ML16179A271

(Unit 1) 2016-301 SRO As-Given Written Exam

ML16179A271
Person / Time
Site:	Summer
Issue date:	06/24/2016
From:	NRC/RGN-II
To:	South Carolina Electric & Gas Co
References
Download: ML16179A271 (130)

Text

ES-401, Page 48 of 50 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:

Facility / Unit 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 percent overall, with 70 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 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.

Applicant's Signature Results RO/SRO-Only/Total Examination Values

______ / ______ / ______

Points Applicant's Score

______ / ______ / ______

Points Applicant's Grade

______ / ______ / ______

Percent X

X V.C. Summer Unit 1

2016 V.C. Summer h



EZInitial Written Examination Class 14-01 Name _______________________________________ Date ______________________

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NRC Initial Written Examination Class 14-01 Name _______________________________________ Date ______________________

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Name: ________________________________

14-01 NRC written Form: 0 Version: 0

1. Given the following plant conditions:

100% Power.

Reactor trip occurs.

One (1) control rod is stuck at 220 steps.

ALL other control rods indicate 0 steps.

TAVG 557°F and stable.

Which ONE of the choices below completes the following statements?

Ten (10) minutes after the trip, shutdown margin is ____(1)____ ;

For the conditions above, operators __(2)__ be required to borate using MVT-8104, EMERG BORATE in accordance with EOP-1.1, ES-0.1, REACTOR TRIP RESPONSE.

1) increasing

2) will

1) increasing

2) will not

1) decreasing

2) will

1) decreasing

2) will not A.

B.

C.

D.

14-01 NRC written

2. Given the following plant conditions:

• PCV-445A PWR RELIEF has failed partially open.

• Reactor trip and Safety Injection (SI) occurred.

• RCS pressure is 1700 psig and decreasing.

TAVG is 550°F and decreasing.

"A" Charging pump is running.

"B" Charging pump failed to start manually or automatically.

Pressurizer level is 32% and increasing at 7%/minute.

All steam generator levels are 50% and stable.

Which ONE of the choices below completes the following statements?

In accordance with EOP-1.2 ES-1.1 SAFETY INJECTION TERMINATION, operators

___(1)___ terminate SI in the current plant condition.

When operators are restoring normal charging, EOP-1.2 will require ___(2)___ to be closed.

1) cannot

2) either MVG-8801A or MVG-8801B, HI HEAD TO COLD LEG INJ

1) cannot

2) both MVG-8801A and MVG-8801B, HI HEAD TO COLD LEG INJ

1) can

2) either MVG-8801A or MVG-8801B, HI HEAD TO COLD LEG INJ

1) can

2) both MVG-8801A and MVG-8801B, HI HEAD TO COLD LEG INJ A.

B.

C.

D.

14-01 NRC written

3. Given the following plant conditions:

Time 1400:

A Small break LOCA has occurred.

Safety injection has actuated.

"A" and "B" Charging pumps started automatically then tripped.

Time 1425:

EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT is in progress.

"A" and "B" Charging pumps are still tripped.

RB pressure is 12.5 psig and rising.

RCS pressure is 1400 psig and decreasing.

The crew tripped all Reactor Coolant Pumps (RCPs).

Which ONE of the following identifies the reason that operators tripped the RCPs?

Reduces heat input into the RCS.

Prevents damage to RCP motor bearings.

Conserves inventory by reducing the flow of coolant from the break.

Saves RCPs for future starts in degraded core cooling procedures.

A.

B.

C.

D.

14-01 NRC written

4. Initial conditions:

100% Power "A" CCW train is the active loop.

"B" CCW pump inoperable.

"C" CCW pump aligned to "B" train.

Final conditions:

Large break LOCA occurred.

EOP-2.2, TRANSFER TO COLD LEG RECIRCULATION in progress.

Operators are reading EOP-2.2 step 13, "Shift the CCW Train to fast speed in the Active Loop. REFER TO SOP-118, COMPONENT COOLING WATER."

Which ONE of the choices below completes the following statement?

The active loop CCW train will be _________

kept in slow speed to prevent damage to the RHR pump on that train.

kept in slow speed to prevent damage to the Charging pump on that train.

shifted to fast speed to provide cooling for non-essential loads.

shifted to fast speed to provide additional cooling for an RHR heat exchanger.

A.

B.

C.

D.

14-01 NRC written

5. Which ONE of the following identifies a condition that would require the immediate trip of the affected reactor coolant pump?

Pump frame vibration of 4 mils and stable.

Pump shaft vibration of 8 mils and increasing.

Motor stator temperature of 312°F and increasing.

Lower seal water bearing temperature 190°F and increasing.

A.

B.

C.

D.

14-01 NRC written

6. Given the following plant conditions:

Time 0600:

100% power.

TAVG is stable on program.

PZR LEVEL CNTRL is selected to 459 + 460.

PVT-8149A, LTDN ORIFICE A ISOL is open.

PVT-8149B, LTDN ORIFICE B ISOL is open.

PVT-8149C, LTDN ORIFICE C ISOL is closed.

FI-122A, CHG FLOW GPM, reads 90 gpm and stable.

Time 0605:

FI-122A, CHG FLOW GPM, decreases to 15 gpm and stabilizes.

Time 0606:

PVT-8149B, LTDN ORIFICE B fails closed.

Which ONE of the choices below answers both of the following questions:

The flow value, as read on FI-122A, at time 0605 can be caused by pressurizer level transmitter __(1) __ failing high.

At time 0606 pressurizer level is __(2)__.

ASSUME NO OPERATOR ACTIONS

1) LT-459

2) Increasing.

1) LT-459

2) Decreasing.

1) LT-460

2) Increasing.

1) LT-460

2) Decreasing.

A.

B.

C.

D.

14-01 NRC written

7. Given the following plant conditions:

A Large break LOCA has occurred.

RM-G7/18 - CNTMT HI RNG GAMMA are in alarm.

All other radiation monitor readings are normal.

RHR A(B) LEVEL FEET, LI-1969/ 1970 read 413 feet and increasing.

RWST LEVEL %, LI-990/992 read 17% and decreasing.

Which ONE of the following identifies a condition that is present as indicated by the instrument readings above?

Reactor coolant is being lost outside of the RB.

RWST level is above the level at which automatic recirculation realignment occurs.

ECCS Pump suction requirements are not met for recirculation.

Vital equipment required for accident mitigation in the RB is submerged.

A.

B.

C.

D.

14-01 NRC written

8. Given the following plant conditions:

Time 1700:

Mode 4 RHR Train "B" running in the shutdown cooling mode.

"A" CCW loop is active.

"A" CCW pump has tripped.

XCP-601, 1-2, CCP A/C AUTOSTART FAIL.

"A" Charging pump in service.

AOP-118.1, LOSS OF COMPONENT COOLING WATER is in progress.

Component temperatures are being monitored in the Control Room only.

Which ONE of the following identifies the earliest time at which a pump must be stopped in accordance with AOP-118.1?

Stop "A" Charging pump at 1720.

Stop "A" Charging pump at 1830.

Stop "B" RHR pump at 1720.

Stop "B" RHR pump at 1830.

A.

B.

C.

D.

14-01 NRC written

9. Initial conditions:

100% power initially.

Reactor Protection System testing was in progress.

A small break LOCA occurred.

The NROATC reported an ATWS during Immediate Actions.

Current conditions:

Reactor Trip Bypass Breaker "B" is racked-in, closed and will not open locally.

All other reactor trip breakers are OPEN.

EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT in progress.

Operators are attempting to reset Safety Injection (SI).

Which ONE of the choices below completes the following statement, given the conditions above?

__(1)__ and can be reset from the Main Control Board.

Condenser steam dumps __(2)__ armed.

1) only one (1) train of SI

2) are

1) only one (1) train of SI

2) are not

1) both trains of SI

2) are

1) both trains of SI

2) are not A.

B.

C.

D.

14-01 NRC written

10. Initial conditions:

100% power.

"A" EDG is inoperable.

Current conditions:

A tube rupture on "A" Steam Generator occurred.

Reactor Trip and Safety Injection occurred.

230 KV offsite power has been lost.

EOP-4.0, E-3 STEAM GENERATOR TUBE RUPTURE is in progress.

Which ONE of the choices identifies an effect of the loss of 230 KV power as actions are taken in EOP-4.0?

The RCS cooldown rate will be restricted to 50°F/hr.

The RCS depressurization will take longer to complete.

The RCS cooldown rate will not be restricted but will take longer to complete.

The RCS depressurization may have to be stopped prior to terminating SI due to voiding in the reactor core.

A.

B.

C.

D.

14-01 NRC written

11. Given the following plant conditions:

Time 1000:

• 45% power initially.

A steam leak begins on "B" Main Steam line.

"B" NR Steam Generator rose to 70% and then began to decrease.

Time 1001:

• Operators manually tripped the reactor.

Time 1002:

• XCP 615, 1-2, RCS TAVG LO goes into alarm.

Time 1003:

• XCP 615, 1-3, RCS TAVG LO-LO goes into alarm.

Which ONE of the following identifies the earliest time stated above at which a Feedwater Isolation actuation occurred?

1000 1001 1002 1003 A.

B.

C.

D.

14-01 NRC written

12. Given the following plant conditions:

Time 1100:

35% power, plant startup is in progress.

"B", "C" and "D" Feedwater Booster pumps are running.

"A" and "B" Feedwater pumps are operating.

Time 1105:

XCP-639, 2-2, BUS 1C O/C 51BX-1C comes into alarm.

Time 1106:

BOP reports that "B" Feedwater Booster pump has tripped.

Which ONE of the following identifies the time at which a manual Reactor Trip was required?

1105 because an ATWS has occurred.

1105 because there is only one (1) Feedwater Booster pump running.

1106 because there is only one (1) Feedwater Booster pump running.

1106 because there are no Feedwater Booster pumps running.

A.

B.

C.

D.

14-01 NRC written

13. Given the following plant conditions:

A loss of offsite power (115 KV and 230 KV) occurred.

The reactor is tripped.

EOP-1.0, E-0, REACTOR TRIP OR SAFETY INJECTION in progress.

Operators are reading step 9, "Check RCS temperature."

Which ONE of the choices below completes the following statement in accordance with EOP-1.0?

Given the conditions above, EOP-1.0, step 9 will have operators check to ensure that

__(1)__ is trending toward __(2)__ to ensure that automatic controls are controlling properly.

1) TAVG

2) 557°F.

1) TCOLD

2) 557°F.

1) TAVG

2) 564°F.

1) TCOLD

2) 564°F.

A.

B.

C.

D.

14-01 NRC written

14. Given the following plant conditions;

• 100% power

• A failure of internal regulator circuitry on Inverter XIT-5901 occurred.

• The XIT-5901 static switch failed to automatically transfer.

• APN-5901 is deenergized.

• All other equipment is OPERABLE.

Which ONE of the choices completes the following statement?

To restore power to APN-5901 using a path that is not routed through the static switch, the operator will take the _____.

TEST TRANSFER Switch to ALT.

MAN BYP Switch to PREF - ISOLATE.

MAN BYP Switch to BYP to ALT.

MAN BYP Switch to BYP to ALT - ISOLATE.

A.

B.

C.

D.

14-01 NRC written

15. Given the following plant conditions:

Time 0900:

100% power.

"C" Service Water pump breaker was inoperable.

A loss of offsite power (115 KV and 230 KV) occurred.

"A" and "B" Service Water pump breakers tripped and cannot be reclosed.

Operators entered AOP-117.1, LOSS OF SERVICE WATER.

Time 0910:

XCP-636, DGA ENG TEMP TRBL is in alarm.

An AO reports a jacket water temperature of 193°F and rising.

System controller reports that power will not be available for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Which ONE of the choices below answers both of the following questions:

1) Will the diesel cooling system design provide automatic initiation of Fire Service cooling if operators take no action?

2) How will diesel be operated after Fire Service cooling is established in accordance with AOP-117.1?

1) Yes.

2) Operated up to the full rated load capacity because Fire Service cooling will provide adequate cooling at the rated load.

1) Yes.

2) Operated at a reduced load capacity because Fire Service cooling will not provide adequate cooling at the rated load.

1) No.

2) Operated up to the full rated load capacity because Fire Service cooling will provide adequate cooling at the rated load.

1) No.

2) Operated at a reduced load capacity because Fire Service cooling will not provide adequate cooling at the rated load.

A.

B.

C.

D.

14-01 NRC written

16. Given the following plant conditions:

5% power initially.

Plant Startup was in progress.

The following alarms were received in the control room:

- XCP-607, 2-5, INSTR AIR PRESS LO FLOW HI XCP-607, 2-6, SERV AIR PRESS LO Instrument Air pressure is now 75 psig and decreasing.

AOP-220.1, LOSS OF INSTRUMENT AIR has been entered.

An AO has been sent to investigate.

Which ONE of the choices below answers both of the following questions?

1) What is the event that caused the indications above?

2) What is a condition that requires a reactor trip, in accordance with AOP-220.1?

1) The operating Instrument Air Compressor has tripped.

2) Instrument Air header pressure decreases to 60 psig.

1) The operating Instrument Air Compressor has tripped.

2) An MSIV indicates off it's open seat.

1) A large leak exists in the Instrument Air system.

2) Instrument Air header pressure decreases to 60 psig.

1) A large leak exists in the Instrument Air System.

2) An MSIV indicates off it's open seat.

A.

B.

C.

D.

14-01 NRC written

17. Given the following plant conditions:

• 100% power.

• HMI screen indicates "LIMITING" for turbine control.

• Weather reports indicate a massive thunderstorm is approaching.

• Operators are reviewing AOP-301.1, RESPONSE TO ELECTRICAL GRID ISSUES.

Which ONE of the choices below completes the following statements in accordance with AOP-301.1?

An increase in V.C. Summer reactor power could be caused by __(1)__, as identified in a CAUTION statement.

The value specified on the Reference Page of AOP-301.1 above which a Reactor Trip is required is ___(1)___

1) Turbine control valves opening due to a decreased grid frequency.

2) 100%

1) Turbine control valves opening due to a decreased grid frequency.

2) 101%

1) a RCS flow increase due to an increase in grid frequency.

2) 100%

1) a RCS flow increase due to an increase in grid frequency.

2) 101%

A.

B.

C.

D.

14-01 NRC written

18. Given the following plant conditions:

100% power initially.

A loss of all Feedwater pumps occurred.

Emergency Feedwater flow (EFW) cannot be established.

EOP-15.0, FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK has been entered.

Steam Generator levels indicate 10% wide range level.

Which ONE of the choices below answer both of the following questions in accordance with EOP-15.0?

1) Is bleed and feed cooling required in the current plant condition?

2) If bleed and feed were required and only two (2) pressurizer PORVs could be opened, would Reactor Vessel Head vents also be opened?

1) Yes.

2) Yes.

1) Yes.

2) No.

1) No.

2) Yes.

1) No.

2) No.

A.

B.

C.

D.

14-01 NRC written

19. Given the following plant conditions:

Time 0800:

45% power initially.

A Stator coolant runback begins.

Time 0801:

XCP-621, 1-2, CRB INSRT LMT LO goes into alarm.

Time 0802:

XCP-621, 1-1, CRB INSRT LMT LO LO goes into alarm.

The runback stops.

Which ONE of the choices below completes the following statements?

___(1)___ is the earliest time that an Emergency Boration is required.

It is required to ensure that ___(2)___.

1) 0801

2) compliance is maintained with Technical Specification Shutdown Margin requirements.

1) 0801

2) Axial Flux Differential remains within the Technical Specification requirement.

1) 0802

2) rods do not go below Technical Specification rod insertion limits.

1) 0802

2) compliance with Technical Specification Shutdown Margin requirements is restored.

A.

B.

C.

D.

14-01 NRC written

20. Given the following plant conditions:

Mode 6.

Core reload in progress.

The personnel hatch is closed.

Reactor Building Purge is in service.

Auxiliary Building ventilation is in service.

Which ONE of the choices below answers both of the following questions:

The highest potential magnitude of release would occur if a __(1)__ fuel assembly is dropped in the Reactor Building.

Under the conditions above, a release of noble gases from a fuel assembly dropped in the Reactor Building __(2)__ be detected by RM-A3, MAIN PLANT VENT.

1) new

2) would

1) new

2) would not

1) spent

2) would

1) spent

2) would not A.

B.

C.

D.

14-01 NRC written

21. Given the following plant conditions:

100% power.

Steam Generator blowdown return is aligned to the condenser.

One tube on Steam Generator "B" has just completely ruptured.

Which ONE of the choices below completes the following statement?

The first alarm to indicate that a primary to secondary break has occurred will be from

__(1)__ and the return of blowdown to the condensor is terminated by automatic action provided by __(2)__.

Assume that a Reactor trip setpoint has NOT been reached.

1) RM-A9, CNDSR EXHAUST GAS ATMOS.

2) RM-L7, NUCLEAR BLOWDOWN WASTE EFFLUENT LIQUID.

1) RM-A9, CNDSR EXHAUST GAS ATMOS.

2) RM-L3, STEAM GENERATOR BLOWDOWN LIQUID.

1) RM-G19B, STMLN HI RNG GAMMA.

2) RM-L7, NUCLEAR BLOWDOWN WASTE EFFLUENT LIQUID.

1) RM-G19B, STMLN HI RNG GAMMA.

2) RM-L3, STEAM GENERATOR BLOWDOWN LIQUID.

A.

B.

C.

D.

14-01 NRC written

22. Which ONE of the choices below answer both of the following questions:

1) What value for a containment parameter would require entry into an action statement in accordance with either of the following V.C. Summer Technical Specifications for CONTAINMENT SYSTEMS?

T.S. 3.6.1.4 - INTERNAL PRESSURE T.S. 3.6.1.5 - AIR TEMPERATURE

2) Is the action required to be completed within one (1) hour?

1) Negative Containment pressure of 0.3 psig (- 0.3 psig).

2) Yes.

1) Negative Containment pressure of 0.3 psig (- 0.3 psig).

2) No.

1) Containment average air temperature of 115°F.

2) Yes.

1) Containment average air temperature of 115°F.

2) No.

A.

B.

C.

D.

14-01 NRC written

23. Given the following plant conditions:

A small break LOCA has occurred.

There are no ECCS pumps available.

All RCPs are stopped.

NR RVLIS level is 32% and decreasing.

Core Exit TCs are 790°F and increasing.

EOP-14.0, FR-C.1 RESPONSE TO INADEQUATE CORE COOLING is in progress.

Attempts to restore an ECCS injection flowpath have failed.

Which ONE of the following identifies the next major action required in an attempt to restore adequate core cooling in accordance with EOP-14.0?

Restart an RCP.

Open pressurizer PORVs.

Depressurize steam generators.

Open the reactor head vent valves.

A.

B.

C.

D.

14-01 NRC written

24. Which ONE of the choices below answers both of the following questions regarding EOP-1.5, REDIAGNOSIS?

1) Is a demand or actuation of Safety Injection required for using EOP-1.5, REDIAGNOSIS?

2) What is a check that is performed in EOP-1.5 to diagnose an event?

1) Yes.

2) Check if PRT conditions are normal.

1) Yes.

2) Verify no SG tubes are ruptured.

1) No.

2) Check if PRT conditions are normal.

1) No.

2) Verify no SG tubes are ruptured.

A.

B.

C.

D.

14-01 NRC written

25. Given the following plant conditions:

A small break LOCA has occurred.

Condenser pressure is 6" Hg.

All Circulating Water pumps are running.

RCS pressure is 1100 psig and stable.

"A" Charging pump is running with lower than normal amperage.

"B" Charging pump is OFF.

Core Exit TCs are 570°F and rising.

The crew is performing actions contained in EOP-2.1, ES-1.2 POST-LOCA COOLDOWN AND DEPRESSURIZATION.

Which ONE of the following describes the method and rate at which the RCS cooldown will occur in accordance with EOP-2.1?

Condenser steam dumps will be used at the maximum achievable rate. RCS cooldown rate limits do not apply for this condition.

Condenser steam dumps will be used at less than 100°F per hour cooldown rate.

S/G PORVs will be used at the maximum achievable rate. RCS cooldown rate limits do not apply for this condition.

S/G PORVs will be used at less than 100°F per hour cooldown rate.

A.

B.

C.

D.

14-01 NRC written

26. Given the following plant conditions:

100% power initially.

A reactor trip occurred.

Steam dumps failed to operate.

Steam Generator Safety valves failed to operate.

All Steam Generator Pressures are at 1235 psig.

A YELLOW status is indicated on IPCS for HEAT SINK.

EOP-1.1, ES-0.1 REACTOR TRIP RESPONSE, step 1 is in progress.

Which ONE of the choices completes the following statements in accordance with OAP-103.4 EOP/FSP/AOP USER'S GUIDE?

EOP-15.1, STEAM GENERATOR OVERPRESSURE __(1)__ allowed to be used in the current condition.

The CRS __(2)__ required to enter EOP-15.1 when the use of YELLOW path procedures is allowed by OAP-103.4.

1) is

2) is

1) is

2) is not

1) is not

2) is

1) is not

2) is not A.

B.

C.

D.

14-01 NRC written

27. Initial conditions:

A small break LOCA occurred.

Charging pumps failed to start manually or automatically.

Indications of inadequate core cooling were present.

Current conditions:

Charging pump capability has been restored.

EOP-2.1, ES-1.2 POST-LOCA COOLDOWN AND DEPRESSURIZATION in progress.

RB pressure is 3 psig and decreasing.

Integrated dose for RB Radiation is 200,000 Rads.

"A" Charging pump has just been stopped in accordance with EOP-2.1.

Which ONE of the choices completes the following statement?

The SI Re-initiation Criteria for the current conditions, as specifically stated on the EOP-2.1 Reference Page, is either a subcooling value less than __(1)__ or a PZR level less than __(2)__.

1) 52.5°F

2) 10%.

1) 52.5°F

2) 28%.

1) 67.5°F

2) 10%.

1) 67.5°F

2) 28%.

A.

B.

C.

D.

14-01 NRC written

28. Initial conditions:

100% power.

Alternate Seal Injection is out of service.

The meter deflection on the HCV-186 INJ FLOW control is at 100%.

Current condition:

The meter deflection on the HCV-186 INJ FLOW control is at 0%.

Which ONE of the choices below answers both of the following questions:

1) How has Seal Injection flow changed due to the change in the position of HCV-186?

2) What is the minimum adjustment that will restore Seal Injection flow to within the allowed range in accordance with SOP-101, REACTOR COOLANT SYSTEM.

1) Flow has increased.

2) Lower to less than the maximum allowed value of 13 gpm.

1) Flow has increased.

2) Lower to less than the maximum allowed value of 10 gpm.

1) Flow has decreased.

2) Raise to more than the minimum allowed value of 8 gpm.

1) Flow has decreased.

2) Raise to more than the minimum allowed value of 6 gpm.

A.

B.

C.

D.

14-01 NRC written

29. Given the following plant conditions:

100% power.

"A" CCW train is the active loop.

All CCW pumps are OPERABLE.

Which ONE of the following events would require tripping of all Reactor Coolant Pumps the soonest?

Closure of MVB-9524B/9526B, LP B NON-ESSEN LOAD ISOL.

Closure of MVG-9600, TO THERM BARR ISOL.

Closure of MVG-9568, TO RB LOAD.

Trip of "A" CCW pump.

A.

B.

C.

D.

14-01 NRC written

30. Initial conditions:

100% power.

"A" Charging pump is running.

Current conditions:

"A" Charging pump tripped.

Which ONE of the choices below completes the following statements?

For the first 30 seconds after the Charging Pump trip, RCP seal cooling __(1)__ be provided.

After 30 seconds, RCP seal injection will automatically be reestablished using water from the __(2)__.

ASSUME NO OPERATOR ACTIONS

1) will not

2) RWST.

1) will not

2) VCT.

1) will

2) RWST.

1) will

2) VCT.

A.

B.

C.

D.

14-01 NRC written

31. Given the following plant conditions:

Plant is shut down with a cooldown in progress.

"A" RHR train is in service.

"B" RHR pump switch is in PULL TO LK NON-A.

"A" RCP is running.

All Steam Generator NR levels are at 60% and stable.

XCP-610, 2-5, RCS TEMP LO AND RHR SUCT VLV NOT OPEN is in alarm.

Operators verified that "A" RHR train valves are in the normal alignment.

Which ONE of the choices below completes the following statements?

This alarm could be caused by ___(1)___ being partially closed.

Technical Specifications that protect against ___(2)___ are not fully met.

1) MVG-8809B, RWST TO RHR PP B

2) RCS failures near the Nil Ductility Temperature

1) MVG-8809B, RWST TO RHR PP B

2) loss of heat removal capability during Mode 4 conditions

1) MVG-8702B, RCS LP C TO PUMP B

2) RCS failures near the Nil Ductility Temperature

1) MVG-8702B, RCS LP C TO PUMP B

2) loss of heat removal capability during Mode 4 conditions A.

B.

C.

D.

14-01 NRC written

32. Given the following plant conditions:

Time 0100:

A large-break LOCA occurred.

XCP-612, 4-3, RWST LVL LO-LO XFER TO SUMP is in alarm.

Time 0101:

The CRS begins reading step 1 of EOP-2.2, ES-1.3 TRANSFER TO COLD LEG RECIRCULATION.

Which ONE of the choices below completes the following statements?

At time 0101, Charging Pump suctions are supplied from the __(1)__.

When the actions of EOP-2.2 are complete, Charging Pump miniflow line valve MVG-8106, CHG PP will be __(2)__.

1) RWST.

2) open.

1) RWST.

2) closed.

1) RHR pump discharge.

2) open.

1) RHR pump discharge.

2) closed.

A.

B.

C.

D.

14-01 NRC written

33. Given the following plant conditions:

100% power initially.

A faulted steam generator occurred.

Both Charging pumps are running in injection mode.

RCS pressure lowered to 1520 psig and then steadily increased to the current value of 2300 psig.

Which ONE of the choices below completes the following statements?

PRT level is increasing because there is coolant flowing from a _____.

Pressurizer Code Safety valve.

relief valve on the RHR pump suction.

relief valve on the normal letdown line.

relief valve on the Seal Return line.

A.

B.

C.

D.

14-01 NRC written

34. Given the following plant conditions:

Time 1700:

100% power.

VCT level is 35% and stable.

• PVT-8149A, LTDN ORIFICE A ISOL and PVT-8149B, LTDN ORIFICE B ISOL are open.

PVT-8149C, LTDN ORIFICE C ISOL is closed.

Time 1715:

Operator realigned the valves above in accordance with SOP-102, CHEMICAL AND VOLUME CONTROL SYSTEM as follows:

- PVT-8149B, LTDN ORIFICE B ISOL and PVT-8149C, LTDN ORIFICE C ISOL are now open.

PVT-8149A, LTDN ORIFICE A ISOL is now closed.

Time now 1730.

Which ONE of the choices below identifies a change in a system parameter that occurred as a result of realigning valves to the current condition?

Assume all parameters are stable at the current conditions.

Cooling flow through the Letdown Heat Exchanger is higher.

Regenerative Heat Exchanger outlet temperature is lower.

Flow through the normal letdown line is lower.

VCT level is lower.

A.

B.

C.

D.

14-01 NRC written

35. Given the following plant conditions:

• Time 1300.

Plant startup in progress using GOP-2, PLANT STARTUP AND HEATUP (MODE 5 TO MODE 3).

"B" RHR Train is in service.

"A" RCP is running.

A loss of Component Cooling Water occurs Which ONE of the following identifies the earliest time at which the trip of a component is required in accordance with AOP-118.1, LOSS OF COMPONENT COOLING WATER?

Assume CCW is not restored.

"A" RCP at 1310.

"A" RCP at 1320.

"B" RHR pump at 1320.

"B" RHR pump at 1330.

A.

B.

C.

D.

14-01 NRC written

36. Given the following plant conditions:

75% power and stable.

The TAVG input to the Master Level Controller fails promptly to a value of 557°F.

Which ONE of the choices below completes the following statements?

XCP-616, 1-5, PZR LCS DEV HI/LO __(1)__ in alarm.

As a result of this malfunction, __(2)__ automatically.

1) is

2) FCV-122, CHG FLOW opened further

1) is.

2) Pressurizer backup heaters energized

1) is not

2) FCV-122, CHG FLOW opened further

1) is not

2) Pressurizer backup heaters energized A.

B.

C.

D.

14-01 NRC written

37. Given the following plant conditions:

100% power.

Power range channel N-43 has failed high.

I&C reports that the channel will take a week to repair due to the lack of a replacement part.

Which ONE of the following identifies a set of actions that will be required in accordance with AOP-401.10, POWER RANGE CHANNEL FAILURE?

The ROD CNTRL BANK SEL Switch will be __(1)__.

Tripping of an Overtemperature T bistable __(2)__ be required.

1) taken to MAN

2) will not

1) taken to MAN

2) will

1) left in AUTO

2) will not

1) left in AUTO

2) will A.

B.

C.

D.

14-01 NRC written

38. Given the following plant conditions:

A plant load reduction is in progress.

Turbine First Stage Pressure Channel, PT-446, sticks at 50% of range.

Reactor power is currently 25%.

Which ONE of the choices below completes the following statements?

RPS permissive __(1)__ for this reactor power.

In this condition, a minimum of __(2)__ RCP(s) tripped will cause a Reactor Trip on low RCS flow.

1) P-7 is in its proper condition

2) one (1)

1) P-7 is in its proper condition

2) two (2)

1) P-7 is not in its proper condition

2) two (2)

1) P-8 is not in its proper condition

2) one (1)

A.

B.

C.

D.

14-01 NRC written

39. Which ONE of the following identifies the power supply to "B" Train Engineered Safety Features Loading Sequencer?

APN-5902 APN-5903 APN01DB2 DPN-1HB1 A.

B.

C.

D.

14-01 NRC written

40. Given the following plant conditions:

100% power.

PT-951, Reactor Building Pressure Channel, has failed HIGH.

Which ONE of the choices below completes the following statements?

1) When the PT-951 bistable associated with HIGH-1 is tripped in accordance with SOP-401 REACTOR PROTECTION AND CONTROL SYSTEM it is __(1)__.

2) After the HIGH-1 bistable is tripped, the logic using the remaining operable bistables that will initiate a HIGH-1 RB Pressure actuation is 1 out of __(2)__.

1) energized.

2) 2.

1) energized.

2) 3.

1) deenergized.

2) 2.

1) deenergized.

2) 3.

A.

B.

C.

D.

14-01 NRC written

41. Initial conditions:

100% power.

The following Reactor Building Cooling Units are running in NORM:

XFN0064A-AH XFN0064B-AH XFN0065B-AH RBCU TRAIN A EMERG is in XFN-64A.

RBCU TRAIN B EMERG is in XFN-65B.

Current conditions:

A large break LOCA occurs.

Which ONE of the following describes the operation of the RBCUs for these conditions?

XFN-64B stops; XFN0064A-AH and XFN0065B-AH remain running.

Cooling water swaps to Service Water for all RBCUs.

XFN-64B stops; XFN0064A-AH and XFN0065B-AH remain running.

Cooling water swaps to Service Water for running RBCUs only.

All RBCUs stop; XFN0064A-AH and XFN0065B-AH start.

Cooling water swaps to Service Water for all RBCUs.

All RBCUs stop; XFN0064A-AH and XFN0065B-AH start.

Cooling water swaps to Service Water for running RBCUs only.

A.

B.

C.

D.

14-01 NRC written

42. Given the following plant conditions:

100% power initially.

A LOCA event is in progress.

The HI-3 signal actuation failed to occur automatically at the required setpoint.

The following timeline of actions and events occur:

0700 - XCP-626, 5-1, PZR SI is in alarm.

0701 - RB pressure is 4 psig, increasing.

0705 - RB pressure is 14 psig, increasing.

0706 - RB spray is manually actuated.

0715 - SI is RESET.

0716 - Phase A isolation is RESET.

0717 - Phase B isolation is RESET.

0718 - RB Spray is RESET.

Which ONE of the choices below completes the following statements?

A demand signal for MVG-3003A(B), SPRAY HDR ISOL LOOP A(B) to open first occurred at time __(1)__.

The actuation system circuit logic to allow closure of these valves was first satisfied at time __(2)__.

1) 0700

2) 0716

1) 0701

2) 0716

1) 0706

2) 0717

1) 0706

2) 0718 A.

B.

C.

D.

14-01 NRC written

43. Initial conditions:

A small break LOCA has occurred.

EOP-2.1, ES-1.2 POST-LOCA COOLDOWN AND DEPRESSURIZATION is in progress.

Steam generator pressures are 825 psig and lowering at 1 psig/ second.

Current conditions:

Operators are preparing to begin a RCS cooldown.

Steam generator pressures are 500 psig and dropping at 10 psig/sec.

RB pressure is 15 psig and increasing.

Which ONE of the choices below completes the following statement?

In accordance with EOP-2.1, operators will ________.

transition to EOP-3.0, E-2 FAULTED STEAM GENERATOR ISOLATION.

transition to EOP-3.1, ECA-2.1 UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS.

transition to EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT.

ensure MSIVs are closed and remain in EOP-2.1.

A.

B.

C.

D.

14-01 NRC written

44. Given the following plant conditions:

100% power initially.

The reactor tripped automatically.

EOP-1.0, E-0 REACTOR TRIP OR SAFETY INJECTION is in progress.

RB pressure has steadily increased to the current value of 5.5 psig.

Steam Generator pressures have steadily decreased to the following values:

"A" 650 psig.

"B" 670 psig.

"C" 670 psig.

Which ONE of the following describes the status of the Phase B isolation and Main Steamline Isolation actuation signals?

Neither signal has actuated.

Only Phase B isolation has actuated.

Only Main Steamline isolation has actuated.

Both signals have actuated.

A.

B.

C.

D.

14-01 NRC written

45. Initial conditions:

Plant power escalation was in progress.

"A" S/G Feedwater Flow Control Valve (PVT-478) will not move from it's current position automatically or manually from the Main Control Board.

"A" S/G level is 57% and decreasing.

"B" and "C" S/G levels are 61% and stable.

Power is 35% and stable.

Current conditions:

AOP-210.1, FEEDWATER FLOW CONTROL VALVE FAILURE in progress.

Operators are attempting to stabilize "A" SG level.

"A" S/G level is 52% and decreasing.

All SG pressures are 1000 psig.

Which ONE of the choices below completes the following statements?

The immediate actions of AOP-210.1, will allow operators to raise Feedwater pump speed until a maximum Feedwater pump pressure of __(1)__ is reached and will direct operators to __(2)__ if necessary.

1) 1250 psig

2) open FCV-3321, LOOP A MAIN FW BYPASS from the Control Board.

1) 1250 psig

2) dispatch an AO to locally operate PVT-478, SG A FWF.

1) 1600 psig

2) open FCV-3321, LOOP A MAIN FW BYPASS from the Control Board.

1) 1600 psig

2) dispatch an AO to locally operate PVT-478, SG A FWF.

A.

B.

C.

D.

14-01 NRC written

46. Given the following plant conditions:

Time 10:00 8% power.

"A" Feedwater pump running.

"B" and "C" Feedwater pumps TRIP/RESET switches indicate TRIP.

Turbine-driven EFW pump is OFF.

"A" and "B" MD EFW Pumps are in NORMAL AFTER STOP.

Hand Controllers IFV-3531(3541)(3551), MD EFP TO SG A(B)(C) are at 0% and indicate full closed.

Flow Control Valve Switches FCV-3531(3541)(3551), MD EFP TO SG A(B)(C) are in MANUAL.

Time 10:02 "A" Feedwater pump has tripped.

Steam Generator narrow range levels are at 45%, lowering.

Which ONE of the following describes the condition of components associated with the Motor-driven (MD) EFW pumps?

Assume no operator actions.

"A" and "B" MD EFW Pumps are running; FCV-3531(3541)(3551) are closed.

"A" and "B" MD EFW Pumps are running; FCV-3531(3541)(3551) are 100% open.

"A" and "B" MD EFW Pumps are off.

FCV-3531(3541)(3551) are closed.

"A" and "B" MD EFW Pumps are off.

FCV-3531(3541)(3551) are 100% open.

A.

B.

C.

D.

14-01 NRC written

47. Given the following plant conditions:

A partial loss of off-site power occurred.

"A" EDG is supplying bus 1DA.

No other power source can be restored.

The Shift Manager has directed that a maximum limit of 4676 KW be strictly observed on bus 1DA.

Which ONE of the choices below completes the following statement?

In order to prevent failure of the "A" EDG, bus 1DA can be loaded to a maximum of 4676 KW _____, in accordance with SOP-306, EMERGENCY DIESEL GENERATOR.

continuously without limits on duration.

for one 7-day period only.

for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> out of every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the next year only.

for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> out of every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the next year or for one 7-day period.

A.

B.

C.

D.

14-01 NRC written

48. Given the following plant conditions; 20% power Which ONE of the following would require, if it were isolated, entry into an action statement of T.S. 3.8.1.1. AC SOURCES until the component is restored to service?

Assume no plant modifications or temporary cabling.

XTF005, UNIT 2 ENGINEERED SAFEGUARD TRANSFORMER.

XTF0031, EMERGENCY AUXILIARY TRANSFORMER #1.

XTF5052, ALTERNATE AC SOURCE TRANSFORMER.

230 KV switchyard Bus 2.

A.

B.

C.

D.

14-01 NRC written

49. Given the following plant conditions:

Time 1400:

A loss of all offsite power (115 KV and 230 KV) has occurred.

"A" and "B" EDGs failed to start.

EOP-6.0, ECA-0.0 LOSS OF ALL ESF AC POWER is in progress.

Which ONE of the choices below completes the following statement?

In accordance with V.C. Summer design basis, at time 1800, the minimum terminal voltage that can be expected on batteries XBA1A and XBA1B is __(1)__.

EOP-6.0 __(2)__ contain actions to shed non-safety related DC loads.

1) 108 VDC.

2) does

1) 108 VDC.

2) does not

1) 110 VDC.

2) does

1) 110 VDC.

2) does not A.

B.

C.

D.

14-01 NRC written

50. Given the following plant conditions:

Time 0700:

100% power initially.

MAIN XFMR FEED XCB 8902 opens spuriously.

"A" RCP tripped.

NROATC reports both reactor trip breakers are closed.

Time 0702:

XTF-31 transformer lockout occurs.

51BX-1DA lockout occurs.

"B" EDG fails to start automatically or manually.

Which ONE of the following describes an Immediate Action that is contained in the procedure that will be implemented at time 0702?

Verify the reactor is tripped.

EOP-13.0, FR-S.1 RESPONSE TO ABNORMAL NUCLEAR POWER GENERATION.

Trip the Main turbine.

EOP-13.0, FR-S.1 RESPONSE TO ABNORMAL NUCLEAR POWER GENERATION.

Trip the Main turbine.

EOP-6.0, ECA-0.0 LOSS OF ALL ESF AC POWER.

Restore power to any ESF bus.

EOP-6.0, ECA-0.0 LOSS OF ALL ESF AC POWER.

A.

B.

C.

D.

14-01 NRC written

51. Which ONE of the following sets of valves will re-position as indicated on a sensed Hi Radiation signal generated by detector RM-L3, STEAM GENERATOR BLOWDOWN LIQUID MONITOR?

PVT-524, TO CW will close.

PVG-525, DIVERT TO NB will open.

PVT-524, TO CW will close.

PVG-525, DIVERT TO NB will close.

PVD-6121, NUC BLOWDOWN DISCHARGE will close.

PVG-525, DIVERT TO NB will close.

PVD-6121, NUC BLOWDOWN DISCHARGE will close.

PVG-525, DIVERT TO NB will open.

A.

B.

C.

D.

14-01 NRC written

52. Which ONE of the following describes the buses that directly supply power to the "B" CCW pump and the "B" Component Cooling Water Booster Pump?

"B" CCW "B" CCWBP 1DB 1B2X 1DB 1DB2X 1DB1 1B2X 1DB1 1DB2X A.

B.

C.

D.

14-01 NRC written

53. Given the following plant conditions:

• Refueling is in progress.

• TBCCC has been taken out of service.

• The Supplemental Air Compressor is running supplying breathing air.

Which ONE of the following identifies a supply of Station Instrument Air that can be used in these conditions in accordance with SOP-220, STATION AND BACKUP INSTRUMENT AIR SYSTEMS?

Instrument air compressor "B" using backup cooling supplied from Raw Water.

Diesel air compressor with pressure control selected to "ONBOARD".

Reactor Building air compressors through XVA-2659, RB INSTRUMENT AIR BACK-UP SUP ISOL VLV.

Supplemental instrument air compressor through XVB02633-IA IN BACKUP SYSTEM SUP HDR ISOLATION VLV.

A.

B.

C.

D.

14-01 NRC written

54. Which ONE of the following describes the normal position of SVX-9364B RCS LOOP B SAMPLE HEADER ISOLATION VLV and which ESF function closes this valve?

Normally open and receives a CLOSE signal on phase A Normally open and receives a CLOSE signal on phase B Normally closed and receives a CLOSE signal on phase A Normally closed and receives a CLOSE signal on phase B A.

B.

C.

D.

14-01 NRC written

55. Given the following plant conditions:

A LOCA has occurred.

The operating crew has entered EOP-17.1, RESPONSE TO REACTOR BUILDING FLOODING.

Which ONE of the following describes the method that will be used to evaluate for leakage from Service Water to the Reactor Building?

Compare flow indication from each SW Booster Pump discharge to that pumps return flow to the pond.

Compare "A" SW Booster pump flow and pressure to "B" SW Booster pump flow and pressure.

Isolate Service Water to one (1) RBCU train and monitor the rate of sump increase.

Verify indications for Service Water flow to all individual loads are normal.

A.

B.

C.

D.

14-01 NRC written

56. The following plant conditions exist:

100% power.

The Master Pressure Controller is in AUTO.

Power has been lost to Pressurizer pressure transmitter PT-444.

Which ONE of the following describes the condition that will occur if power to PT-444 is not restored?

Assume no operator actions.

The reactor will trip on high pressurizer pressure.

The pressurizer spray valves will open to maintain RCS pressure.

Two pressurizer PORVs will cycle to maintain RCS pressure.

The reactor will trip on low pressurizer pressure.

A.

B.

C.

D.

14-01 NRC written

57. Given the following plant condition:

Time 0500:

The reactor at 100% power at End of Life and stable.

XCP-620, 5-1, ROD CNTRL SYS FAIL URGENT is in alarm.

AXIAL FLUX DIFFERENCE (AFD) indications are as follows.

N-41 +4 N-42 +4 N-43 +4 N-44 +4 Time 0501:

A Main Turbine runback begins.

Time 0502:

The Main Turbine runback stopped.

Power is at 95% and stable.

AFD indications are now as follows:

N-41 +7 N-42 +7 N-43 +7 N-44 +7 Which ONE of the following completes the following statement?

Over the next 10 minutes, changes in __(1)__ will directly cause AFD to change.

During that time AFD will trend in the __(2)__ direction.

Assume no operator actions

1) Xenon

2) positive

1) Xenon

2) negative

1) Iodine

2) positive

1) Iodine

2) negative A.

B.

C.

D.

14-01 NRC written

58. Given the following plant conditions:

• 100% power.

• Pressurizer Level Control Channel Selector switch is in the "459 + 460" position.

Which ONE of the choices below answers both of the following:

1) Which level channel will eventually cause a reactor trip with no operator action if it fails both high or low?

2) Do LT-459 and LT-460 both serve as Reactor Trip protection channels?

1) LT-459

2) Yes.

1) LT-459

2) No.

1) LT-460

2) Yes.

1) LT-460

2) No.

A.

B.

C.

D.

14-01 NRC written

59. Given the following plant conditions:

Time 0300:

The plant is in a refueling outage.

The core off-load has just been completed.

The spent fuel pool temperature is 102°F.

Time 0305 A loss of all offsite power (115KV and 230 KV) has occurred.

Both EDGs failed to start.

In accordance with a note in AOP-123.4, LOSS OF SPENT FUEL COOLING, the maximum expected heatup rate in the Spent Fuel Pool for the condition above is

__(1)__.

One Spent Fuel Cooling pump can be started from it's normally aligned source of power if either bus __(2)__ is restored,

1) 3°F/hr.

2) 1DA or 1DB

1) 3°F/hr.

2) 1A or 1B

1) 20°F/hr.

2) 1DA or 1DB

1) 20°F/hr.

2) 1A or 1B A.

B.

C.

D.

14-01 NRC written

60. Given the following plant conditions:

Mode 6.

Reactor cavity is at the normal level for refueling.

Fuel Transfer Tube is open.

The equipment hatch is open.

Reactor Building Purge is in progress.

XFN-11A, SPLY FAN A is running.

XFN-11B, SPLY FAN B is OFF.

XFN-13A, EXH FAN A is running.

XFN-13B, EXH FAN B is running.

Which ONE of the following identifies an action that would cause Reactor Cavity level to decrease?

Closing the equipment hatch.

Stopping XFN-11A.

Stopping XFN-13A.

Starting XFN-20, FUEL BLDG SPLY FAN.

A.

B.

C.

D.

14-01 NRC written

61. Given the following plant conditions:

Time 0900:

A spurious safety injection previously occurred.

EOP-1.2, ES-1.1 SAFETY INJECTION TERMINATION is in progress.

Time 0910:

A double-ended rupture of one (1) Steam Generator u-tube occurs.

RCS pressure is 1700 psig and decreasing.

Which ONE of the choices below completes the following statement regarding operation of the ECCS system?

SI Reinitiation Criteria, as contained on the EOP-1.2 REFERENCE PAGE, are first applicable when __(1)__

When Safety Injection is reinitiated per the REFERENCE PAGE, operators __(2)__.

1) MVG-8801A(B), HI HEAD TO COLD LEG INJ are closed.

2) open injection valves and start Charging pumps manually.

1) MVG-8801A(B), HI HEAD TO COLD LEG INJ are closed.

2) actuate Safety Injection using switches SI ACTUATION CS-CR0IB(C).

1) the RHR pumps are stopped.

2) open injection valves and start Charging pumps manually.

1) the RHR pumps are stopped.

2) actuate Safety Injection using switches SI ACTUATION CS-CR0IB(C).

A.

B.

C.

D.

14-01 NRC written

62. Given the following plant conditions:

A waste gas release was in progress.

HCV-014, WASTE GAS DISCHARGE CONTROL VALVE, has tripped shut.

XCP-644, 3-1, PLANT VENT GAS RM-A3 HI RAD in alarm.

XCP-645, 2-3, GAS WST DISCH RM-A10 HI RAD has not alarmed during the release.

Which ONE of the choices below answers both of the following questions:

1) What condition must be satisfied prior to recommencing the release in accordance with HPP-709, SAMPLING AND RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS?

2) What is the minimum control manipulation necessary to enable opening of HCV-014?

1) RM-A3 reading must decrease to pre-release background reading.

2) HCV-014 selector switch must be cycled to CLOSE, then the valve re-opened.

1) A Request for Redundant Analysis must be initiated.

2) HCV-014 controller must be taken to ZERO (0), then the valve re-opened.

1) A Request for Redundant Analysis must be initiated.

2) HCV-014 selector switch must be cycled to CLOSE, then the valve re-opened.

1) RM-A3 reading must decrease to pre-release background reading.

2) HCV-014 controller must be taken to ZERO (0), then the valve re-opened.

A.

B.

C.

D.

14-01 NRC written

63. Given the following plant conditions:

Mode 6.

RB Purge is in progress.

The following alarms come in on XCP-644:

- 1-5, RM-G17A, MANIP CRN RM-G17A HI RAD

- 1-6, RM-G17A, MANIP CRN RM-G17A TRBL Operators have determined that the RM-G17A detector has failed high.

NOTE THE FOLLOWING NUMBERS AND VALVE NAMES:

1 XVB00001A-AH, REACTOR BUILDING PURGE SUPPLY ISOLATION VALVE 2 XVB00001B-AH, RB PURGE SUPPLY ISOL VALVE (IRC).

3 XVB00002A-AH, REACTOR BUILDING PURGE EXHAUST ISOLATION VALVE 4 XVB00002B-AH, RB PURGE EXHAUST ISOL VALVE (IRC).

Which ONE of the choices below completes the following statements?

Valves __(1)__ closed as a result of the failure.

When RM-G17A is removed from service in accordance with SOP-124, PROCESS AND AREA RADIATION MONITORING SYSTEM, the final position of the INTERLOCK switch on RM-G17A will be __(2)__.

1) 1 and 3

2) ON.

1) 1 and 3

2) OFF.

1) 2 and 4

2) ON

1) 2 and 4

2) OFF A.

B.

C.

D.

14-01 NRC written

64. Which ONE of the choices below completes the following statement?

The highest pressure, as read on PI-8386, RB AIR HDR PRESS, at which PVA-2659, INST AIR TO RB AIR SERV will open is __(1)__.

93 psig 90 psig 70 psig 65 psig A.

B.

C.

D.

14-01 NRC written

65. Given the following conditions:

The plant was operating at 100% power.

The Control Room (CR) is being evacuated due to a fire in the Main Control Board.

FEP-4.0, CONTROL ROOM EVACUATION DUE TO A FIRE has been implemented by the Shift Manager.

Operators determine that immediate actions can be performed in the control room as necessary prior to leaving the Control Room.

A spurious Reactor trip occurred.

Which ONE of the following describes an immediate action the NROATC will perform in the Control Room?

Trip RCPs.

Open PORV disconnect switches.

Emergency start "B" EDG.

Start "B" Motor-driven EFW pump.

A.

B.

C.

D.

14-01 NRC written

66. Initial conditions:

100% power initially.

A Steam generator tube rupture occurred on "A" Steam Generator.

Current conditions:

"B" Steam Generator was also determined to be ruptured.

EOP-4.0, E-3 STEAM GENERATOR TUBE RUPTURE is in progress.

Operators are reading step 9 of EOP-4.0.

Steam Generator pressures are as follows:

"A" 1050 psig "B" 950 psig "C"

950 psig Operators have determined that there is a coincident RCS leak to the RB.

Reactor Building pressure is 2.0 psig and rising slowly.

Which ONE of the choices below completes the following statement?

In accordance with EOP-4.0, operators will cool the RCS to less than a maximum temperature of ______.

REFERENCE PROVIDED 482°F 469°F 466°F 453°F A.

B.

C.

D.

Determine the required core exit TC temperature for RCS cooldown from the table below:

9 LOWEST RUPTURED SG PRESS (PSIG)

CORE EXIT TC TEMP

(°F)

CONTROLLER SETPOINT 1101-1200 494 [478]

4.9 1001-1100 482 [466]

4.4 901-1000 469 [453]

3.8 801-900 455 [439]

3.4 701-800 439 [423]

2.8 601-700 421 [405]

2.3 460-600 392 [376]

1.6 Check if any RCP is running.

10 With no RCP running, RCS cooldown and depressurization may cause RUPTURED loop Tcold to falsely indicate a transition to EOP-16.0, FR-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK, is required. Disregard the RUPTURED loop Tcold indication prior to performing Step 40.

10 EOP-4.0 REVISION 24 E3, STEAM GENERATOR TUBE RUPTURE ACTION/EXPECTED RESPONSE ALTERNATIVE ACTION PAGE 8 OF 51

14-01 NRC written

67. Which ONE of the choices below completes the following statement regarding T.S.

3.4.7, CHEMISTRY?

T.S. 3.4.7, CHEMISTRY __(1)__apply in Mode 5.

The steady state limit for oxygen in the Reactor Coolant System is __(2)__.

1) does

2) 0.10 ppm

1) does not

2) 0.10 ppm

1) does

2) 0.15 ppm

1) does not

2) 0.15 ppm A.

B.

C.

D.

14-01 NRC written

68. Given the following plant conditions:

50% power Which ONE of the choices below answers both of the following questions?

1) What is the plant position that maintains OAP-100.6, CONTROL ROOM CONDUCT AND CONTROL OF SHIFT ACTIVITIES, ATTACHMENT IA - REACTIVITY CONTROL PARAMETERS.

2) What is the frequency at which Attachment IA is required to be performed?

1) Reactor Engineer

2) Daily

1) Reactor Engineer

2) Weekly

1) Reactor Operator

2) Daily

1) Reactor Operator

2) Weekly A.

B.

C.

D.

14-01 NRC written

69. Given the following plant conditions:

• The Reactor Vessel head is removed.

• IFV-3531-O-EF, OPER-SG A MTR DR EF PUMP FLOW CONT VLV, has been isolated to replace the actuator.

Which ONE of the choices below completes the following statement in accordance with SAP-0205 STATUS CONTROL AND REMOVAL AND RESTORATION?

An ___(1)___ Removal and Restoration will be used to track the status of IFV-3531-O-EF.

When the valve is restored to the NEUTRAL position a ___(2)____ chain will be installed.

1) Action

2) red

1) Action

2) silver

1) Outage

2) red

1) Outage

2) silver A.

B.

C.

D.

14-01 NRC written

70. Initial conditions:

Reactor startup in progress in accordance with GOP-3 REACTOR STARTUP FROM HOT STANDBY TO STARTUP (MODE 3 TO MODE 2).

The NROATC is withdrawing Control Bank "A" Rods.

Current condition:

Power ascension in progress in accordance with GOP-4A POWER OPERATION (MODE 1 - ASCENDING).

Which ONE of the choices below completes the following statements?

While withdrawing Control Bank "A" under the initial conditions, the ROD CNTRL BANK SEL Switch is required to be in the ___(1)___ position in accordance with GOP-3.

The ROD CNTRL BANK SEL Switch is placed in AUTO when power is greater than a minimum of ___(1)___ in accordance with GOP-4A.

1) CBA

2) 10%

1) CBA

2) 15%

1) MANUAL

2) 10%

1) MANUAL

2) 15%

A.

B.

C.

D.

14-01 NRC written

71. Given the following plant conditions:

Mode 6.

Spent Fuel Pool level decreasing.

AOP-123.1, DECREASING LEVEL IN THE SPENT FUEL POOL OR REFUELING CAVITY DURING REFUELING, in progress.

Which ONE of the choices below completes the following statement(s)?

The radiation threshold that is specifically used in AOP-123.1 to trigger immediate evacuations of all personnel from the RB is _____.

2 R/hr, as read on RM-G5, RB PERSONNEL ACCESS AREA GAMMA.

2 R/hr, as read on RM-G17A(B), RB MANIP CRANE AREA GAMMA.

20 R/hr, as read on RM-G5, RB PERSONNEL ACCESS AREA GAMMA.

20 R/hr, as read on RM-G17A(B), RB MANIP CRANE AREA GAMMA.

A.

B.

C.

D.

14-01 NRC written

72. Given the following plant conditions:

• A Large break LOCA has occurred.

• Major core damage has occurred.

• The Severe Accident Management Guidelines have been entered.

• The Auxiliary Building has been evacuated due to high radiation levels.

• An operator must be dispatched to close a valve to terminate a release from containment to the environment.

Which ONE of the following describes the highest allowed radiation dose to perform this operation in accordance with EPP-020, EMERGENCY PERSONNEL EXPOSURE CONTROL?

No more than 5 Rem.

No more than 10 Rem.

No more than 25 Rem.

Greater than 25 Rem on a voluntary basis.

A.

B.

C.

D.

14-01 NRC written

73. Given the following plant conditions:

Initially at 100% power.

A Reactor trip and Safety Injection occurred.

EOP-1.0, E-0 REACTOR TRIP OR SAFETY INJECTION has been entered.

Operators have just completed the IMMEDIATE ACTIONS of EOP-1.0.

An AO reports that there is a loud rumbling noise at the door to the Intermediate Building on 436' and that the door is hot.

RB pressure is 0.5 psig and stable.

All RB sump levels are normal.

All Steam Generator pressures are 750 psig and decreasing.

"A" and "B" Motor-driven EFW pumps breakers indicate tripped.

Which one of the following manipulations is performed without specific direction from an EOP step under the current plant conditions in accordance with OAP-103.4, EOP/AOP USER'S GUIDE?

Closure of all MSIVs.

Reduction of EFW flow to a total of 450 gpm.

Complete Isolation of EFW flow to all Steam Generators.

Isolation of steam to the Turbine-driven EFW pump.

A.

B.

C.

D.

14-01 NRC written

74. Given the following plant conditions:

Time 0600:

A Small break LOCA occurred.

Operators entered EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT.

RCS pressure was 1300 psig and stable.

Operators stopped "A" and "B" RHR pumps.

Time 0630:

EOP-2.1 ES-1.2 POST-LOCA COOLDOWN AND DEPRESSURIZATION is in progress.

Operators are preparing to commence a operator-controlled cooldown.

Actions have been delayed due to failure of steam dumps to open.

The NROATC notes the following history of RCS pressure indications.

Time 0615 650 psig 0620 600 psig 0625 550 psig 0630 500 psig Which ONE of the following identifies the earliest time at which operators are required to restart the RHR pumps in injection mode in accordance with direction contained in EOP-2.1, if at all?

ASSUME NO ADDITIONAL OPERATOR ACTIONS.

ASSUME A CONSTANT RATE OF RCS PRESSURE DECREASE.

They must be restarted at 0630.

They must be restarted between 0635 and 0640.

They must be restarted between 0640 and 0645.

They must be restarted between 0645 and 0650.

A.

B.

C.

D.

14-01 NRC written

75. Given the following plant conditions:

A Reactor trip and Safety Injection occurred.

Actions of EOP-1.0, E-0 REACTOR TRIP OR SAFETY INJECTION are complete.

Operators transferred to EOP-3.0, E-2 FAULTED STEAM GENERATOR ISOLATION one (1) second ago.

Which ONE of the following identifies a Major Action Category of EOP-3.0 in accordance with the EOP-3.0 (E-2) Background Document?

Terminate Safety Injection.

Prepare for cooldown to COLD SHUTDOWN.

Check for at least one non-faulted Steam Generator.

Identify and isolate ruptured SGs.

A.

B.

C.

D.

14-01 NRC written

76. Initial conditions:

100% power initially.

A reactor trip occurred due to a loss of Feedwater.

EOP-1.0, E-0, REACTOR TRIP OR SAFETY INJECTION in progress.

Eight rods are indicated at the following positions, in steps, on DRPI:

B-8 at 78 F-4 at 228 J-9 at 120 K-6 at 204 K-8 at 12 M-6 at 174 P-6 at 228 P-10 at 12 All other control and shutdown rods indicate 0 steps.

"A" Reactor Trip breaker GREEN indicating light is LIT.

"B" Reactor Trip breaker RED indicating light is LIT.

Reactor power indicates 2% and decreasing.

Which ONE of the choices below completes the following statement?

In accordance with EOP-1.0, the crew is required to transition to:

EOP-1.1, ES-0.1 REACTOR TRIP RESPONSE and borate a minimum of 2500 gallons.

EOP-1.1, ES-0.1 REACTOR TRIP RESPONSE and borate a minimum of 5800 gallons.

EOP-13.0, FR-S.1 RESPONSE TO ABNORMAL NUCLEAR POWER GENERATION and borate a minimum of 2500 gallons.

EOP-13.0, FR-S.1 RESPONSE TO ABNORMAL NUCLEAR POWER GENERATION and borate a minimum of 5800 gallons.

A.

B.

C.

D.

14-01 NRC written

77. Given the following plant conditions:

Time 0500:

100% power initially.

A small break loss of coolant accident occurred.

230 KV power was lost concurrent with reactor trip.

Time 0600:

RCS Loop TCOLD temperatures are as follows "A" is 468°F and decreasing.

"B" is 455°F and decreasing.

"C" is 265°F and decreasing.

Core Exit TCs are 490°F and decreasing.

RCS pressure is 600 psig and decreasing.

Which ONE of the choices below completes the following statements?

At 0600, the CRS will be __(1)__.

In accordance with the bases for the INTEGRITY Status Tree, there __(2)__ a potential for growth of an existing flaw at this RCS temperature and pressure.

REFERENCE PROVIDED

1) required to use EOP-16.0, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK.

2) is

1) required to use EOP-16.0, RESPONSE TO IMMINENT PRESSURIZED".

THERMAL SHOCK.

2) is not

1) allowed to use EOP-16.1 RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK.

2) is

1) allowed to use EOP-16.1 RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK.

2) is not A.

B.

C.

D.

EOP-12.0 EOP-12.0 REVISION 14 REVISION 14 ATTACHMENT 4 ATTACHMENT 4 PAGE 2 of 2 PAGE 2 of 2 INTEGRITY INTEGRITY PLANT OPERATIONAL LIMITS CURVE PLANT OPERATIONAL LIMITS CURVE 100 150 200 250 300 350 400 0

500 1000 1500 2000 2500 3000 RCS Tcold (°F)

RCS PRESSURE (PSIG)

LIMIT T1 250°F LIMIT T2 280°F REGION OF ACCEPTABLE OPERATION 195°F 0 PSIG LIMIT A 230°F 2050 PSIG 2560 PSIG (PZR SAFETY VALVES ASSUMED FULL OPEN)

LIMIT T1 LIMIT T2 REGION OF 195 LIMIT A 230 2560 PSIG PAGE 9 OF 12 PAGE 9 OF 12

14-01 NRC written

78. Given the following plant conditions:

Time 0100 100% power initially.

A large break LOCA occurred.

"B" RHR pump failed to start manually or automatically.

Time 0140:

The crew entered EOP-2.2, ES-1.3 TRANSFER TO COLD LEG RECIRCULATION.

The CRS has just begun reading Step 1.

A lockout occurs on XTF0004, UNIT 1 ENGINEERED SAFEGUARD TRANSFORMER.

The BOP reports that "A" EDG cannot be started from the Main Control Board.

Which one of the following identifies the procedure action, out of the ones listed below, that will be implemented next to maintain core cooling?

Remain in EOP-2.2 and use AOP-304.1A LOSS OF BUS 1DA WITH THE DIESEL NOT AVAILABLE to reenergize bus 1DX to allow a restart of "A" RHR pump.

Remain in EOP-2.2 and use AOP-304.1A LOSS OF BUS 1DA WITH THE DIESEL NOT AVAILABLE to locally start "A" EDG to allow a restart of "A" RHR pump.

Transfer to EOP-2.4, LOSS OF EMERGENCY COOLANT RECIRCULATION and stop a Charging Pump to preserve the RWST.

Transfer to EOP-2.4, LOSS OF EMERGENCY COOLANT RECIRCULATION and add makeup to the RWST to maintain a water source for Charging.

A.

B.

C.

D.

14-01 NRC written

79.

Given the following plant conditions:

The time is now 1300 on 5/20.

The plant is at 557°F.

"B" CCW Train is active.

All rods are inserted.

Engineering reports that "A" and "C" CCW pumps must be declared inoperable due to a common failure mode found during a review of repair paperwork.

The condition is expected to be corrected in approximately 7 days.

Which ONE of the following identifies the latest time by which the plant must be in COLD SHUTDOWN in accordance with Technical Specifications?

REFERENCE PROVIDED 1900 on 5/21 0100 on 5/22 1900 on 5/24 0100 on 5/25 A.

B.

C.

D.

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water loops shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

+

-+

With only one component cooling water loop OPERABLE, rstore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the~ following. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.3 At least two component cooling water loops shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.

SU+ER - UNIT 1

+

3/4 7-11

14-01 NRC written

80. Given the following plant conditions:

6% power with plant startup in progress using GOP-4A, POWER OPERATION (MODE 1 - ASCENDING).

XCP-625, 2-1, FWP A/B/C TRIP comes into alarm.

All Feedwater pumps indicate tripped.

Steam Generator Narrow Range levels indicate 57% and decreasing.

AOP-210.3, FEEDWATER PUMP MALFUNCTION is in progress.

All EFW pumps are running.

Which ONE of the choices below completes the following statement?

In accordance with AOP-210.3, power will be reduced to a minimum of __(1)__ and operators will then be required to __(2)__.

1) 1 to 3%.

2) transition to GOP-4B, POWER OPERATION (MODE 1 - DESCENDING).

1) 1 to 3%.

2) start a Feedwater pump using SOP-210, FEEDWATER SYSTEM.

1) 5%.

2) transition to GOP-4B, POWER OPERATION (MODE 1 - DESCENDING).

1) 5%

2) start a Feedwater pump using SOP-210, FEEDWATER SYSTEM.

A.

B.

C.

D.

14-01 NRC written

81. Given the following plant conditions:

Time 1000, 6/1:

Mode 3.

All offsite power (115 KV and 230 KV) was lost.

"B" EDG failed to start automatically.

The BOP operator started "B" EDG and the diesel successfully loaded.

Time now 1300, 6/1:

230 KV power is restored.

Bus 1DB is being fed by BUS 1DB NORM FEED.

The "B" EDG auto-start circuitry is being evaluated.

Which ONE of the choices below identifies the latest time by which the plant must be placed in COLD SHUTDOWN in accordance with T.S. 3.8.1.1, AC SOURCES -

OPERATING, if it is so required?

REFERENCE PROVIDED Transition to COLD SHUTDOWN is not required with the current conditions.

1600 on 6/2.

0400 on 6/3 1600 on 6/5 A.

B.

C.

D.

314.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. ~ Two physically independent circuits between the offsite transmission network and the onsite Class I E distribution system, and b.

Two separate and independent Emergency Diesel Generators (EDG), each with:

A ~iiate day fuel tank containing a minimum-volume of 360 gallons of

fuel, 2.

A separate fuel storage system containing a minimum volume of 48,500 gallons of fuel, and 3.

A separate fuel transfer pump.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

~

a.

With one ofisite circuit of 3.8.1.1.a inoperable:

1.

Demonstrate the OPERABILITY of the remaining offsiteA.C. sources by S.n~

performing Surveillance Requirement 4.8.1.1.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, and 2.

If either EDG has not been successfully tested within the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, demonstrate its OPERABILITY by performing Surveillance Requirement 4.8.1.1.2.a.3 separately for each such EDG within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the diesel is already operating, and 3.

Restore the offsite circuit to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s-and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With one EDG of 3,8.1.1.b inoperable:

1.

Demonstrate the OPERABILITY of the A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1 within 1 hourand at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, and 2.

If the EDG became inoperable due to any cause other than preplanned preventive maintenance or testing:

a) determine the OPERABLE EDG is not inoperable due to a common cause failure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or b) demonstrate the OPERABILITY of the remaining EDG by performing Surveillance Requirement 4.8.1.1.2.a.3 within 24

hours, and (U,

Completion of Action b.2 is required regardless of when the inoperable EDG is restorecilo OPERABILITY.

SUMMER-UNIT 1 3/4 8-1 Amendment No. 50, 77, 84.03.150. ~54

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

3.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, verify that required systems, subsystems, trains, components and devices that depend on the remaining EDG as a source of emergency power are also OPERABLE and in MODE 1, 2, or 3, that the Turbine Driven Emergency Feed Pump is OPERABLE. If these conditions are not satisfied within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4.

Restore the EDG to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, unless the following condition exists:

a)

The requirement for restoration of the EDG to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be extended to 14 days if the Alternate AC (AAC) power source is or will be available within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, as specified in the Bases, and b)

If at any time the AAC availability cannot be met, either restore the AAC to available status within the remainder of the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in 4.a (not to exceed 14 days from the time the EDG originally became inoperable), or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c.

With one offsite circuit and one EDG inoperable:

1.

Demonstrate the OPERABILITY of the remaining offsite A.C. source by performing Surveillance Requirement 4.8.1.1.1 within one hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, and

2.

• If the EDG became inoperable due to any cause other than preplanned preventative maintenance or testing:

a) determine the OPERABLE EDG is not inoperable due to a common cause failure within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or b) demonstrate the OPERABILITY of the remaining EDG by performing Surveillance Requirement 4.8.1.1.2.a.3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and

3.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, verify that required systems, subsystems, trains, components and devices that depend on the remaining EDG as a source of emergency power are also OPERABLE and in MODE 1, 2, or 3, that the Turbine Driven Emergency Feed Pump is OPERABLE. If these conditions are not satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4.

Restore one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and

5.

Restore the other A.C. power source (offsite circuit or diesel generator) to OPERABLE status in accordance with the provisions of Section 3.8.1.1 Action Statement a. or b.,

as appropriate, with the time requirement of that Action Statement based on the time of initial loss of the remaining inoperable A.C. power source.

• Completion of Action c.2 is required regardless of when the inoperable EDG is restored to OPERABILITY.

SUMMER - UNIT 1 3/4 8-2 Amendment No. 77, 98, 164, 178

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued) d.

- With two of the required offsite A.

C.

circuits inoperable:

1.

Demonstrate the OPERABILITY of the two LOGs by sequentially performing Surveillance Requirement 4.8.LL2.a.3 on both

~wjitiin8hours, unless the EDOs are alreaiiy operating, and 2.

Restore one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.

Following restoration of one offsite source, follow Action Statement a. with the time requirement of that Action Statement based on the time of initial loss of the remaining inoperable offsite A.C. circuit.

e.

With two of the above required EDGs inoperable:

1.

Demonstrate the OPERABILITY of two offsite At. circuits by performing Surveillance Requirement 4.8.1.1.1 within one hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, and 2.

Restore one of the inoperable EDGs to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COU SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.

Following restoration of one EDO, follow Action Statement b.

with the time requirement of that Action Statement based on the time of initial loss of the remaining inoperable diesel generator.

SURVEI LLANCE REQUIREMENTS 4.8.1.1.1 Lath of the above required physically independent circuits between the offsite traflsmission network and the onsite Class 1E distribution system shall be determined OPERABLE at least once per 7 days by verifying correct breaker alignment and indication of power availability for each Class 1E bus and its preferred offsite power source.

S SU*IER

- UNIT 1 3/4 8-2a AJIENDMENT NO.

77, 98 4

14-01 NRC written

82. Given the following plant conditions:

Time 0400:

90% power with a plant startup in progress.

A continuous rod withdrawal occurred.

• Attempts to manually trip the reactor from the Main Control Board failed.

Time 0401:

• An AO was dispatched to locally open Reactor Trip breakers.

• Power is 99% and increasing.

Time 0403:

The AO opened the Reactor Trip breakers locally.

Time 0420:

All EFW pumps were lost.

Narrow Range Steam Generator levels are at 25% and decreasing.

Time 0445:

"A" Motor-Driven EFW pump was restarted.

Which ONE of the choices below answers both of the following questions in accordance with EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN and EPP-002 COMMUNICATION AND NOTIFICATION:

The highest classification for this event was __(1)__.

If declaration occurred when conditions were met, the time by which state and local authorities must have been notified for the highest classification was __(2)__.

Do not consider Emergency Director Judgement as a basis for your emergency classification.

REFERENCE PROVIDED

1) an Alert.

2) 0415.

1) an Alert.

2) 0430.

1) a Site Area Emergency.

2) 0435.

1) a Site Area Emergency.

2) 0450.

A.

B.

C.

D.

14-01 NRC written

83. Given the following plant conditions:

Time 1200:

Chemistry previously reported indications of fuel failure.

Operators were performing local operations to place RHR in service.

TAVG was 320°F and decreasing.

A partial set of RMS monitor HI-RAD alarm setpoints were as follows:

RM-A3 (gas), MAIN PLANT VENT set at 5.0 x 102 c/m.

RM-A13, MAIN PLANT EXHAUST HIGH RANGE set at 1.0 mR/hr.

Time 1205:

A Safe Shutdown Earthquake (SSE) event occurred.

A leak from the in-service Waste Gas Compressor began.

Current time is 1215:

All personnel have been directed to evacuate the Auxiliary Building due to high airborne.

The following RMS monitor readings were noted:

RM-A3(gas), reads 9.0 x 102 c/m and increasing.

RM-A13, reads 3 mR/hr and increasing.

Which ONE of the choices completes the following statement in accordance with EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN?

At time the current time of 1215, the highest emergency classification condition that is met is an __(1)__ and activation of the TSC __(2)__ required.

Do not consider Emergency Director Judgement as a basis for your emergency classification.

REFERENCE PROVIDED

1) UNUSUAL EVENT;

2) is

1) UNUSUAL EVENT;

2) is not

2) ALERT;

2) is

2) ALERT;

2) is not A.

B.

C.

D.

14-01 NRC written

84. Given the following plant conditions:

Previously, at 0500 on 5/1.

Plant is in Mode 3.

Preparations are being made to begin a Reactor startup.

I&C reports that both RM-G7 and RM-G18 REACTOR BUILDING HIGH RANGE monitors are inoperable.

Date today is 5/5.

I&C reports that RM-G7 has been restored to OPERABLE status.

Which ONE of the choices completes the following statements in accordance with T.S.

3.3.3.6, INSTRUMENTATION: ACCIDENT MONITORING INSTRUMENTATION?

A Special Report to the NRC __(1)__ due within the next fourteen days.

Taking the reactor critical in the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> __(2)__ allowed.

REFERENCE PROVIDED

1) is

2) is

1) is

2) is not

1) is not

2) is

1) is not

2) is not A.

B.

C.

D.

iNSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION L1NUTlNG CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY MODES 1,2, and 3.

ACTION:

a.

With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown on Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 30 days or submit a Special Report within the following 14 days from the time the action is required. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the Instrumentation channels to operable status.

b.1 With the number of OPERABLE Reactor Building radiation monitoring channels less than the Minimum Channels Operable requirement of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

i)

Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and ii)

Submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 daysfollowing the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

b.2 Deleted b.3 With the number of OPERABLE accident monitoring channels less than the Minimum Channels Operable requirement of Table 3.3-10, either restore the inoperable channels to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 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 />.

c.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring Instrumentation channel shall be demonstrated OPERABLE by performing a monthly CHANNEL CHECK and a CHANNEL CALIBRATION every refueling outage. The Reactor Building Radiation Level Instrumentation CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for the range decades above I OR/hr and a single point calibration of the detector below 1 OR/hr with an installed or portable gamma source.

SUMMER

- UNIT I 3/4 3-56 Amendment No. 44-8~ 170

TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF CHANNELS INSTRUMENT CHANNELS QP~EAABLE Reactor Building Pressure - Narrow Range 2

1 Instrument Loop/Indicator:

Channel 13 IPT-951/IPI-951

Channel. B IPT-952JlPl-952 2.

Reactor Building Pressure - Wide Range 2

1 Instrument Loop/Indicator:

Channel D IPT-954MP1-954A Channel E IPT-954B/IPI-954B 3.

Reactor Building Radiation Level

- High Range 2

1 Instrument Loop/Indicator:

ChannelA 13MG-lB Channel B RMG-7 4.

Deleted 5.

Reactor Building/RHA Sump Level 2

1 Instrument Loop/Indicator:

Channel A ILT-l 969/ILl-I969 Channel B ILT-1 970/ILl-i 970 6.

Reactor Coolant Outlet Temperature - T~ - Wide Range 2

1 Instrument Loop/Indicator:

Channel A ITE-413/ITI-413 Channel A ITE-423/ITI-423 Channel E ITE-433/ITR-413 7.

Reactor Coolant Inlet Temperature - Td-Wide Range 2

1 Instrument Loop/Indicator:

Channel E ITE-410/ITI-410 Channel E ITE-420/ITI-420 Channel E ITE-430/ITR-410

14-01 NRC written

85. Given the following plant condition:

Initial conditions:

A large-break LOCA occurred.

RBCUs XFN0064A, XFN0064B and XFN0065B failed to start automatically or manually.

RBCUs XFN0065A is running in SLOW speed.

Both RHR pump and RB Spray pump suctions failed to align to the RB Sump automatically or manually.

EOP-2.4, ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION was in progress.

RB Pressure was 27 psig and rising slowly.

Operators stopped the A RB Spray Pump in accordance with EOP-2.4.

Current conditions:

RB Pressure is 43 psig, rising slowly.

The CRS has implemented EOP-17.0, FR-Z.1 RESPONSE TO HIGH REACTOR BUILDING PRESSURE.

RWST Level is 13%, stable.

Which ONE of the choices below completes the following statements?

Based on the current conditions, operators ___(1)___ required to restart the A RB Spray pump in accordance with EOP-17.0. The basis is because ___(2)___.

1) are not

2) the RWST does not provide sufficient net positive suction head at the current level.

1) are not

2) with recirculation not available, conserving the RWST has priority over the containment overpressure.

1) are

2) there is less than a complete train of containment depressurization equipment operating.

1) are

2) RB leakage may exceed the design basis limits if RB pressure exceeds the design value.

A.

B.

C.

D.

14-01 NRC written

86. Given the following plant conditions:

3% power The high voltage power supply for Intermediate Range Channel N-35 fails low.

Given the conditions above, which ONE of the following is the highest value of power, of those listed, that operators can establish and maintain in accordance with Technical Specifications?

REFERENCE PROVIDED 7.4 x 10-6%.

4%.

9%.

100%.

A.

B.

C.

D.

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE 1

FUNCTIONAL UNIT OF CHANNELS TOTRIP OPERABLE MODES ACTION

-I

I.

Manual Reactor Trip 2

1 2

1,2

~

1 2

1 2

3*, 4*~5*

2.

Power Range, Neutron Flux A.

HighSetpoint 4

2 3

1,2 2~

B.

LowSetpoint 4

2 3

1###,2 2

3.

Power Range, Neutron Flux 4

2 3

1,2 2#

High Positive Rate 4.

Deleted

5~

Intermediate Range, Neutron Flux 2

1 2

~

2 3

6.

Source Range, Neutron Flux A.

Startup 2

1 2

2 4

B.

Shutdown 2

0

1 3,4apd5 5

C.

Shutdown 2

1 2

3*,4*,5*

9 1.

Overtemperature AT ThreeLoopOperation 3

2 2

1 2~

ft Two Loop Operation nd,,

~

8.

Overpower AT F

Three Loop Operation 3

2 2

1 2 6

Two-Loop Operation

• 1.**

9.

PressurizerPressure-Low 3

2 2

1 10.

Pressurizer Pressure--High 3

2 2

1,2 6

S S

S

TABLE 3.3-1 (Continued)

TABLE NOTATION

• With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.

1. The provisions of Specification 3.0.4 are not applicable.

1. 1. Below the P-6 (Intermediate Range Neutron Flux Interlock) setpoint.

1. 1. 1. Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

• • • • Values left blank pending NRC approval of 2 loop operation.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.

c.

Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.

SUMMER - UNIT 1 3/4 3-6 Amendment No. 101, 177

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below the P-6 (Intermediate Range Neutron Flux Interlock) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b.

Above the P-6 (Intermediate Range Neutron Flux Interlock) setpoint but below 10 percent of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10 percent of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes.

ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and

b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 - With less than the Minimum Number of Channels OPERABLE, within one hour determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

SUMMER - UNIT 1 3/4 3-7 Amendment No. 101, 177

14-01 NRC written

87. Plant conditions:

• EOP-3.1, ECA-2.1 UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS in progress.

• Operators throttled EFW flow to 150 gpm.

• AOs have been dispatched to attempt to close MSIVs locally.

• "A" Charging pump was stopped and placed in standby 10 seconds ago.

• "B" Charging pump is running in injection mode.

• Operators are checking to determine whether Normal Charging can be established in accordance with EOP-3.1 step 15.

• The STA reports the following:

HEAT SINK status is RED All other Critical Safety Functions are GREEN.

• Steam Generator pressure are as follows:

- "A" 300 psig, decreasing.

- "B" 400 psig, increasing.

- "C" 300 psig, decreasing.

• Steam Generator WR levels are as follows:

- "A" 5%, decreasing.

- "B" 6%, increasing.

- "C" 5%, decreasing.

Which ONE of the following describes how the CRS should proceed?

Remain in EOP-3.1 until SI is terminated at Step 19; THEN Transition to EOP-3.0, E-2 FAULTED STEAM GENERATOR ISOLATION and isolate "A" and "C" SG's.

Transition to EOP-4.0, E-3 STEAM GENERATOR TUBE RUPTURE and isolate "B" SG.

Transition to EOP-3.0, E-2 FAULTED STEAM GENERATOR ISOLATION and isolate "A" and "C" SG's.

Transition to EOP-15.0, FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK and initiate BLEED and FEED.

A.

B.

C.

D.

14-01 NRC written

88. Initial conditions:

"A" CCW is the active loop.

"C" Service Water pump is INOPERABLE.

All offsite power was lost (115 KV and 230 KV).

Current conditions:

Steam Generator (SG) Tube Rupture on "B" SG occurred.

Operators have just completed the last step of EOP-4.0, E-3 STEAM GENERATOR TUBE RUPTURE.

"B" SG NR level is 92%, stable.

Steam Generator blowdown is not available.

"A" Service Water Pump tripped and cannot be restarted.

Which ONE of the choices below completes the following statements?

The correct transfer out of EOP-4.0 is to __(1)__.

The guidance in AOP-117.1, LOSS OF SERVICE WATER, will be used

__(2)__ to restore the active CCW loop.

1) EOP-4.1A, ES-3.1 POST-SGTR COOLDOWN BY BACKFILLING THE REACTOR COOLANT SYSTEM.

2) in parallel with actions of EOP-4.1A

1) EOP-4.1A, ES-3.1 POST-SGTR COOLDOWN BY BACKFILLING THE REACTOR COOLANT SYSTEM.

2) after recovery actions of EOP-4.1A are complete

1) EOP-4.1C, ES-3.3 POST-SGTR COOLDOWN USING STEAM DUMP.

2) in parallel with actions of EOP-4.1C

1) EOP-4.1C, ES-3.3 POST-SGTR COOLDOWN USING STEAM DUMP.

2) after recovery actions of EOP-4.1C are complete A.

B.

C.

D.

14-01 NRC written

89. Given the following plant conditions:

Time 0030 on 5/10:

100% power.

Station Instrument Air Dryer malfunctions caused a loss of Instrument Air Header pressure.

Time 0045 on 5/10:

100% power.

Instrument air header pressure has been restored to normal pressure.

It was determined that the Accumulator for IFV-3556, TD EFP TO SG C depressurized during the event.

No repairs have been performed.

Which ONE of the following describes the latest time by which the plant must be placed in HOT STANDBY in accordance with T.S.3.7.1.2, EMERGENCY FEEDWATER SYSTEM, if applicable?

REFERENCE PROVIDED The plant is not required to be taken to HOT STANDBY.

0645 on 5/10.

0645 on 5/13.

1245 on 5/13.

A.

B.

C.

D.

SUMMER - UNIT 1 3/4 7-4 Amendment No. 112, 114, 173 PLANT SYSTEMS EMERGENCY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator emergency feedwater pumps and flow paths shall be OPERABLE with:

a.

Two motor-driven emergency feedwater pumps, each capable of being powered from separate emergency busses, and

b.

One steam turbine driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.

With one emergency feedwater pump inoperable, restore the required emergency feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With two emergency feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

With three emergency feedwater pumps inoperable, immediately initiate corrective action to restore at least one emergency feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each emergency feedwater pump shall be demonstrated OPERABLE:

a.

At least once per 31 days by:

1.

Verifying that each motor driven pump develops a total head of greater than or equal to 3800 feet at greater than or equal to 90 gpm flow.

2.

Verifying that the steam turbine driven pump develops a total head of greater than or equal to 3140 feet at a flow of greater than or equal to 97 gpm when the secondary steam supply pressure is greater than 865 psig.

The provisions of Specification 4.0.4 are not applicable.

3.

Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

PLANT SYSTEMS SURVEI LLANCE REQUIREMENTS (Continued) 4.

Verifying that each automatic valve in the flow path from the condensate storage tank to the steam generators is in the fully open position whenever the emergency feedwater system is placed in automatic control or when above 10% RATED THERMAL POWER.

5.

Verifying that valves 1010-El and 1007-EF are locked in the

-ppe position.

b.

At least once per 3 months by verifying that the check valve in the instrument air supply line to the six emergency feedwater control valve air accumulators closes when the normal instrument air supply is not available.

c.

At least once per 18 months during shutdown by verifying that:

1.

Each emergency feed pump starts as designed automatically upon receipt of an emergency feedwater actuation test signal.

2.

The six emergency feedwater control valves can be closed and held closed for three hours with air from the accumulators when the normal instrument air supply is not available.

3.

The turbine driven emergency feedwater pump can be manually stopped from the main control board by closing the steam supply valve with air from the accumulator when the normal instrument air supply is not available.

4.

Each automatic valve in the flow patti actuates to its correct position on receipt of an emergency feedwater actuation test signal.

SUMMER - UNIT 1 3/4 7-5

14-01 NRC written

90. Given the following plant conditions:

Time 1000:

Small break LOCA occurred.

All Charging pumps failed to start automatically or manually.

NR RVLIS is 38% and decreasing.

Core Exit TCs are 670°F and increasing.

RB pressure is 14 psig, increasing.

"A" RB Spray pump is running.

"B" RB Spray failed to start automatically or manually.

XFN0064A is running in SLOW.

XFN0065B is running in SLOW.

Time 1018:

NR RVLIS is 36% and decreasing.

Core Exit TCs are 728°F and increasing.

Time 1025:

FI-4496, SWBP B DISCH FLOW GPM reads 500 gpm and stable.

Time 1030:

NR RVLIS is 32% and decreasing.

Core Exit TCs are 790°F and increasing.

Charging is expected to be restored in 20 minutes.

Which ONE of the following identifies the time when conditions were first met for a GENERAL EMERGENCY in accordance with EPP-001, ACTIVATION AND IMPLEMENTATION OF EMERGENCY PLAN?

Do not consider Emergency Director Judgement as a basis for your emergency classification.

REFERENCE PROVIDED 1000.

1018.

1025.

1030.

A.

B.

C.

D.

14-01 NRC written

91. Given the following plant conditions:

Time 0700 on 5/1:

A Large break LOCA occurred.

Time 0700 on 5/3:

EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT in progress.

Hydrogen concentration is 1.9% and rising.

Time 1000 on 5/4:

EOP-2.0 LOSS OF REACTOR OR SECONDARY COOLANT in progress.

Hydrogen concentration is 3.1% and rising.

"A" and "B" Post Accident Hydrogen Recombiners are inoperable.

The TSC has determined that hydrogen must be vented from the Reactor Building to prevent a challenge to building integrity.

Which ONE of the choices below completes the following statements?

At time 0700 on 5/3, EOP-2.0 required operators to start __(1)__ Post Accident Hydrogen Recombiner(s).

At time 1000 on 5/4, operators will use __(2)__ to reduce the hydrogen concentration.

1) 1

2) RB Alternate Purge in accordance with SOP-114, REACTOR BUILDING VENTILATION SYSTEM

1) 1

2) Backup RB Purge in accordance with SOP-122, POST ACCIDENT HYDROGEN REMOVAL SYSTEM

1) 2

2) RB Alternate Purge in accordance with SOP-114, REACTOR BUILDING VENTILATION SYSTEM

1) 2

2) Backup RB Purge in accordance with SOP-122, POST ACCIDENT HYDROGEN REMOVAL SYSTEM A.

B.

C.

D.

14-01 NRC written

92. Given the following plant conditions:

0800 on 5/1:

• The automatic valve isolation on a HI RAD alarm provided by RM-L10, NUCLEAR BLOWDOWN LIQUID RADIATION MONITOR was found to be inoperable.

0900 on 5/2:

Repairs have been completed satisfactorily.

An Analog Channel Operational Test was performed as a retest.

Which ONE of the following identifies the latest date by which the surveillance above can next be performed without requiring action for a missed surveillance, as specified in the ODCM?

REFERENCE PROVIDED 7/31.

8/2.

8/23.

8/25.

A.

B.

C.

D.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-5 Table 1.1-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT CHANNEL CHECK SOURCE CHECK CHANNEL CALIBRA-TION ANALOG CHANNEL OPERATIONAL TEST

1.

GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a.

Liquid Radwaste Effluent Line -

RM-L5 or RM-L9

b.

Nuclear Blowdown Effluent Line RM-L7

c.

Steam Generator Blowdown Effluent Line - RM-L3, RM-L10

d.

Turbine Building Sump Effluent Line - RM-L8

e.

Condensate Demineralizer Backwash Effluent Line RM-L11 D

D D

D D

P P

M M

M R(2)

R(2)

R(2)

R(2)

R(2)

Q(1)

Q(1)

Q(1)

Q(1)

Q(4)

2.

FLOW RATE MEASUREMENT DEVICES

a.

Liquid Radwaste Effluent Line

b.

Penstocks Minimum Flow Interlock

c.

Nuclear Blowdown Effluent Line

d.

Steam Generator Blowdown Effluent Line D(3)

D(3)

D(3)

D(3)

N.A.

N.A.

N.A.

N.A.

R R

R R

Q Q

Q Q

3.

TANK LEVEL INDICATING DEVICES

a.

Condensate Storage Tank D

N.A.

R Q

See Table 1.1-3 for explanation of frequency notation.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 1.0-6 Table 1.1-2 (Continued)

TABLE NOTATION (1)

The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm/trip setpoint.

2.

Loss of Power (alarm only).

3.

Low Flow (alarm only).

4.

Instrument indicates a Downscale Failure (alarm only).

5.

Normal/Bypass switch set in Bypass (alarm only).

6.

Other instrument controls not set in Operate mode.

(2)

The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(3)

CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic or batch releases are made.

(4)

The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and local panel alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm/trip setpoint.

2.

Loss of Power (alarm only).

3.

Low Flow (alarm only).

4.

Instrument indicates a Downscale Failure (alarm only).

5.

Normal/Bypass switch set in Bypass (alarm only).

6.

Other instrument controls not set in Operate mode.

14-01 NRC written

93. Given the following plant conditions:

The plant is being cooled down for a refueling outage.

RCS temperature is 195°F and lowering.

RCS pressure is 375 psig and stable.

Which ONE of the choices below completes the following statements in accordance with T.S. 3.7.6, CONTROL ROOM EMERGENCY FILTRATION SYSTEM (CREFS)?

In the current conditions, CREFS must remain OPERABLE to ensure that control room occupancy does not result in more than a maximum of __(1)__ rem, after the __(2)__,

that is assumed in the bases for T.S. 3.7.6.

1) 5

2) loss of coolant accident.

1) 5

2) rupture of an outside Waste Gas Decay tank.

1) 25

2) loss of coolant accident.

1) 25

2) rupture of an outside Waste Gas Decay tank.

A.

B.

C.

D.

14-01 NRC written

94. Which ONE of the choices below answers both of the following questions in accordance with OAP-114.1 PROTECTED EQUIPMENT PROGRAM?

Work is not normally allowed within a minimum of __(1)__ feet from protected equipment.

The lowest level of authority that can give permission to work within this distance of protected equipment is the __(2)__.

1) 2

2) Shift Manager.

1) 2

2) Work Control Supervisor.

1) 3

2) Shift Manager.

1) 3

2) Work Control Supervisor.

A.

B.

C.

D.

14-01 NRC written

95. Given the following plant conditions:

The plant has been on line for two (2) weeks after a refueling outage.

Power is 100%.

A review of Surveillance testing revealed that verification that RBCU XFN-64A runs for fifteen (15) minutes in slow speed was last performed correctly sixty (60) days ago.

Based on a risk evaluation, it has been determined that the current status poses minimal risk for continued operation.

Which of the following describes the longest time allowed before the surveillance must be performed in accordance with Technical Specifications?

REFERENCE PROVIDED 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

8 days.

31 days.

39 days.

A.

B.

C.

D.

CONTAINMENT SYSTEMS REACTOR BUILDING COOLING SYSTEM LIMITING CONDITIONS FOR OPERATION 3.6.2.3 T~wo independent groups of reactor building cooling units shall be OPERABLE with at least one of two cooling units OPERABLE in slow speed in each group.

~:-

APPLICABILITY: ~t1OOES1; 2, 3 and 4.

ACTION:

a.

With one group of the above required reactor building cooling units inoperable and both reactor building spray systems OPERABLE, restore the inoperablc group ef cooling units to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With two groups of the above required reactor building tooling units inoperable, and both reactor building spray systems OPERABLE, restore at least one group of cooling units to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Restore both above -required groups of cooling units to OPERABLE status within 7 days of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and i:~

COLD SHUTDOWN within tht tollowinq 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c.

With one group of the ar...e required reactor building cooling units inoperable and one reactor building spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Restore the inoperable group of containment cooling units to OPERABLE status within 7 days of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEI LLANCE REQUI REMENTS 4.6.2.3 Each group of reactor building cooling units shall be demonstrated OPERABLE:

a.

At least once per 31 days by:

1.

Starting each cooling unit group from the control room, and verifying that each cooling unit group operates for at least 15 minutes in the slow speed mode.

b.

At least once per 18 months by:

1.

Verifying that each fan group starts automatically on a safety injection test signal.

2.

Verifying a cooling water flow rate of greater than or equal to 2,000 gpm to each cooling unit group.

SUMMER

- UNIT 1 3/4 5-14 Amendment No. 69

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3.

At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a simulated SI test signal or on an ESFLS, as applicable.

4.

Atleast once-per 18--months, by-ver-ify.ing that each service water system booster pump starts automaticallyon a safety injection signal.

SU*IER - UNIT 1 3/4 6-14a Amendment No.

21

14-01 NRC written

96. Given the following plant conditions:

All offsite power (115 KV and 230 KV) was lost.

Neither EDG started.

The crew is implementing EOP-6.0, ECA-0.0 LOSS OF ALL ESF AC POWER.

The crew is at Step 33 to "select the appropriate recovery procedure."

RB Pressure is 2.6 psig and increasing.

RCS Subcooling on TI-499A(B) is 57°F and decreasing.

PZR Level is 5% and decreasing.

Flow on FI-943, CHG LOOP B CLD/HOT LG is 0 gpm Based on the provided indications, operators will first transition to __(1)__ and then operators will __(2)__.

Assume no further system malfunctions or degradations.

1) EOP-6.1, ECA-0.1 LOSS OF ALL ESF AC POWER RECOVERY WITHOUT SI REQUIRED.

2) return to EOP-1.0, E-0 REACTOR TRIP OR SAFETY INJECTION when actions in EOP-6.1 are complete.

1) EOP-6.1, ECA-0.1 LOSS OF ALL ESF AC POWER RECOVERY WITHOUT SI REQUIRED.

2) transition to EOP-1.3, NATURAL CIRCULATION COOLDOWN when actions in EOP-6.1 are complete.

1) EOP-6.2, ECA-0.2 LOSS OF ALL ESF AC POWER RECOVERY WITH SI REQUIRED.

2) remain in that procedure until conditions for placing RHR in service are met.

1) EOP-6.2, ECA-0.2 LOSS OF ALL ESF AC POWER RECOVERY WITH SI REQUIRED.

2) transition to EOP-2.0, LOSS OF REACTOR OR SECONDARY COOLANT when actions in EOP-6.2 are complete.

A.

B.

C.

D.

14-01 NRC written

97. Given the following plant conditions:

A Loss of offsite power has occurred.

"B" Steam Generator is ruptured.

"B" Steam Generator pressure is 1100 psig and stable.

Both Motor-driven EFW pumps failed to start automatically or manually.

Based on the conditions above, which ONE of the choices below completes the following statements with regard to completing EPP-002 Attachment I, NUCLEAR POWER PLANT EMERGENCY NOTIFICATION FORM?

Protective Action Recommendations under 5 B and 5 C of EPP-002, Attachment I

__(2)__ required for this event.

Number 6 on EPP-002, Attachment I will be indicated as __(1)__.

REFERENCE PROVIDED

1) are

2) A-None.

1) are

2) B - is Occurring.

1) are not

2) A-None.

1) are not

2) B - is Occurring.

A.

B.

C.

D.

EPP-002 ATTACHMENT I PAGE 1 of 11 REVISION 37 NUCLEAR POWER PLANT EMERGENCY NOTIFICATION FORM

1. A DRILL B ACTUAL EVENT MESSAGE # _______

2. A INITIAL B FOLLOW-UP NOTIFICATION: TIME___________ DATE_____/_____/_____ AUTHENTICATION # ________

3. SITE: V. C. Summer Confirmation Phone # (____)_________________

4. EMERGENCY CLASSIFICATION:

A UNUSUAL EVENT B ALERT C SITE AREA EMERGENCY D GENERAL EMERGENCY BASED ON EAL #______________ EAL DESCRIPTION: _________________________________________________________________________

5. PROTECTIVE ACTION RECOMMENDATIONS:

A NONE B EVACUATE ___________________________________________________________________________________________________

C SHELTER _____________________________________________________________________________________________________

D CONSIDER THE USE OF KI (POTASSIUM IODIDE) IN ACCORDANCE WITH STATE PLANS AND POLICY.

E OTHER________________________________________________________________________________________________________

6. EMERGENCY RELEASE:

A None B Is Occurring C Has Occurred

7. RELEASE SIGNIFICANCE:

A Not applicable B Within normal operating limits C Above normal operating limits D Under evaluation

8. EVENT PROGNOSIS:

A Improving B Stable C Degrading

9. METEOROLOGICAL DATA:

Wind Direction* from _______ degrees Wind Speed* _______mph

(*May not be available for Initial Notifications)

Precipitation* _______

Stability Class* A B C D E F G

10. A DECLARATION B TERMINATION Time ________________ Date _____/______/_______

11. AFFECTED UNIT(S):

2 3 All

12. UNIT STATUS:

(Unaffected Unit(s) Status Not Required for Initial Notifications)

U1 _____% Power Shutdown at Time _____________ Date ___/_____/____

B U2 _____% Power Shutdown at Time _____________ Date ___/_____/____

C U3 _____% Power Shutdown at Time _____________ Date ___/_____/____

13. REMARKS: ______________________________________________________________________________________________________________

FOLLOW-UP INFORMATION (Lines 14 through 16 Not Required for Initial Notifications)

EMERGENCY RELEASE DATA. NOT REQUIRED IF LINE 6 A IS SELECTED.

14. RELEASE CHARACTERIZATION:

TYPE: A Elevated B Mixed C Ground UNITS: A Ci B Ci/sec C µCi/sec MAGNITUDE: Noble Gases:__________ Iodines:___________ Particulates:__________ Other: ____________

FORM: A Airborne B Liquid Start Time __________ Date ___/_____/____Stop Time _________ Date ___/_____/____

Start Time __________ Date ___/_____/____Stop Time _________ Date ___/_____/____

15. PROJECTION PARAMETERS:

Projection period: ________Hours Estimated Release Duration ________Hours Projection performed:

Time _________ Date ___/_____/____

16. PROJECTED DOSE:

DISTANCE TEDE (mrem)

Adult Thyroid CDE (mrem)

Site boundary 2 Miles 5 Miles 10 Miles

17. APPROVED BY:

____________________________ Title _____________________

Time _________ Date___/_____/____

NOTIFIED BY:___________________________

RECEIVED BY: _____________________________

Time _________ Date ___/_____/____

A 1

14-01 NRC written

98. Given the following plant conditions:

Time 0900, 6/20 Waste Gas Decay Tank "H" sampled.

A Release Permit was approved for a release of Waste Gas Decay Tank (WGDT) "H".

WGDT "H" pressure 18 psig.

Time 0945, 6/20 Release was commenced.

HCV-014, WASTE GAS DISCHARGE CONTROL VALVE went closed due to a leak on the valve operator diaphragm.

Time 1300, 6/20 WGDT "H" returned to normal service.

Gas transferred from WGDT "C" to WGDT "H".

Time 0200, 6/21 HCV-014 has been repaired.

WGDT "H" is aligned for release.

WGDT "H" pressure 22 psig.

RM-A3 reading Normal background.

Wind direction From the West-Southwest.

Which ONE of the choices below answers both of the following questions?

1) Can the release be restarted at 0200, 6/21 with the Release Permit approved on 6/20?

2) When the release is recommenced, what is the minimum frequency for monitoring meteorological conditions?

1) Yes.

2) At least once per hour.

1) Yes.

2) Every eight (8) hours.

1) No.

2) At least once per hour.

1) No.

2) Every eight (8) hours.

A.

B.

C.

D.

14-01 NRC written

99. Given the following plant conditions:

A tube rupture has occurred on "A" Steam Generator.

EOP-1.0 E-0, REACTOR TRIP OR SAFETY INJECTION is in progress.

All RCPs are running.

Operators are checking parameters for a potential transition to EOP-4.0, E-3 STEAM GENERATOR TUBE RUPTURE.

Which ONE of the choices below completes the following statements?

The earliest point in the event at which RCP Trip criteria based on RCS pressure no longer apply is the time at which __(1)__;

The reason that the criteria no longer apply is that __(2)__.

1) Safety Injection is terminated.

2) decay heat load has fallen low enough that tripping RCPs is not required to prevent exceeding peak clad temperatures even if a LOCA exists.

1) Safety Injection is terminated.

2) RCP trip criteria applicable to small break LOCAs are not required when mitigating a Steam Generator Tube Rupture event.

1) the operator controlled cooldown in EOP-4.0 is started.

2) decay heat load has fallen low enough that tripping RCPs is not required to prevent exceeding peak clad temperatures even if a LOCA exists.

1) the operator controlled cooldown in EOP-4.0 is started.

2) RCP trip criteria applicable to small break LOCAs are not required when mitigating a Steam Generator Tube Rupture event.

A.

B.

C.

D.

14-01 NRC written 100.

Given the following plant conditions:

Time 1000:

Mode 1, plant shutdown in progress for a refueling outage.

Operators are moving new fuel assemblies in the Spent Fuel Pool.

XCP-608, 1-2, SFP LVL HI/LO comes into alarm.

Time 1100:

Operators entered the appropriate procedure and corrected the cause of the Spent Fuel Pool level decrease.

The NROATC reports that indications on POOL LEVEL FEET LI-7431 and LI-7433 have decreased one (1) foot since XCP-608, 1-2 alarmed.

Which ONE of the choices below answers both of the following questions in accordance with T.S. 3.7.10 WATER LEVEL - SPENT FUEL POOL?

Actions required by T.S. 3.7.10 __(1)__ require operators to stop movement of new fuel in the Spent Fuel Pool In accordance with the basis for T.S. 3.7.10, if Spent Fuel Pool level remains at the current level, there may not be a sufficient water volume to __(2)__.

1) do

2) remove fission product gases that are released from a ruptured fuel assembly.

1) do

2) provide an adequate heat sink for an off-loaded core during a loss of Spent Fuel Pool cooling.

1) do not

2) remove fission product gases that are released from a ruptured fuel assembly.

1) do not

2) provide an adequate heat sink for an off-loaded core during a loss of Spent Fuel Pool cooling.

A.

B.

C.

D.

ALL CONDITIONS GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT RG1.1 R

Abnormal Rad Levels

/

Rad Effluent 1

Rad Effluent 2

Irradiated Fuel Event H

Hazards 2

Seismic Event 4

Fire 5

Hazardous Gas 1

Security 6

Control Room Evacuation 7

Judgment Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer 2

3 4

5 6

DEF 1

RU1.1 2

3 4

5 6

DEF 1

RU1.2 Reading on any Table R-1 effluent radiation monitor

> column "UE" for 60 min. (Notes 1, 2, 3) 2 3

4 5

6 DEF 1

RA1.1 2

3 4

5 6

DEF 1

RA1.2 Reading on any Table R-1 effluent radiation monitor >

column "ALERT" for 15 min. (Notes 1, 2, 3, 4, 5) 2 3

4 5

6 DEF 1

RS1.1 Reading on any Table R-1 effluent radiation monitor

> column "SAE" for 15 min. (Notes 1, 2, 3, 4, 5) 2 3

4 5

6 DEF 1

RS1.2 Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4, 5) 2 3

4 5

6 DEF 1

RS1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 100 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.

(Notes 1, 2) 2 3

4 5

6 DEF 1

Reading on any Table R-1 effluent radiation monitor

> column "GE" for 15 min. (Notes 1, 2, 3, 4, 5) 2 3

4 5

6 DEF 1

RG1.2 Dose assessment using actual meteorology indicates doses > 1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4, 5) 2 3

4 5

6 DEF 1

RG1.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 1000 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 5000 mrem for 60 min. of inhalation.

(Notes 1, 2)

Significant lowering of water level above, or damage to, irradiated fuel Unplanned loss of water level above irradiated fuel 2

3 4

5 6

DEF 1

RU2.1 UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following:

• Refueling Cavity: LI-7403 MCB annunciator XCP-609 2-6 (REFUEL CAV LVL HI/LO)

• Spent Fuel Pool: LI-7431 or LI-7433 MCB annunciators XCP 608(609)1-2 (SFP LVL HI/LO)

• Fuel Transfer Canal: LI-7405 MCB annunciator XCP-612 1-6 (FUEL XFER CANAL LVL HI/LO)

AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors:

• RM-G6 Rx Bldg Refueling Bridge

• RM-G17A/B Rx Bldg Manipulator Crane (when installed)

• RM-G8 FHB Refueling Bridge Area Gamma Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown 2

3 4

5 6

DEF 1

RA3.1 Dose rate > 15 mR/hr in EITHER of the following areas:

• Control Room (RM-G1)

• Central Alarm Station (by survey)

Seismic event greater than OBE levels 2

3 4

5 6

DEF 1

HU2.1 2

3 4

5 6

DEF 1

HU3.1 A tornado strike within the PROTECTED AREA FIRE potentially degrading the level of safety of the plant 2

3 4

5 6

DEF 1

HU4.1 A FIRE is NOT extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

Report from the field (i.e., visual observation)

Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown 2

3 4

5 6

DEF 1

HA5.1 2

3 4

5 6

DEF 1

HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by Security Team Leader OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat 2

3 4

5 6

DEF 1

HA1.1 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Team Leader OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site HOSTILE ACTION resulting in loss of physical control of the facility HOSTILE ACTION within the PROTECTED AREA 2

3 4

5 6

DEF 1

HS1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Team Leader 2

3 4

5 6

DEF 1

HG1.1 2

3 4

5 6

DEF 1

HA6.1 An event has resulted in plant control being transferred from the Control Room to the Control Room Evacuation Panels (CREP)

Inability to control a key safety function from outside the Control Room 2

3 4

5 6

DEF 1

HS6.1 An event has resulted in plant control being transferred from the Control Room to the Control Room Evacuation Panels (CREP)

AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1):

Reactivity control

Core cooling

RCS heat removal 2

3 4

5 6

DEF 1

HU7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs 2

3 4

5 6

DEF 1

HA7.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

2 3

4 5

6 DEF 1

HS7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public.

Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

2 3

4 5

6 DEF 1

HG7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

Other conditions exist that in the judgment of the Emergency Director warrant declaration of General Emergency None Modes:

1 2

3 4

5 6

DEF Power Operations Startup Hot Standby Hot Shutdown Cold Shutdown Refueling Defueled EPP-001, Attachment I, Rev. 31 EAL Classification Matrix Page 1 of 3 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes Confirmed SECURITY CONDITION or threat Control Room evacuation resulting in transfer of plant control to alternate locations Other conditions exist that in the judgment of the Emergency Director warrant declaration of an Alert Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE Other conditions exist that in the judgment of the Emergency Director warrant declaration of Site Area Emergency Sample analyses for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2)

Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4, 5)

None None A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Team Leader AND EITHER of the following has occurred:

Any of the following safety functions cannot be controlled or maintained Reactivity control Core cooling RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT 2

3 4

5 6

DEF 1

RA2.1 Damage to irradiated fuel resulting in a release of radioactivity as indicated by a Hi-Rad alarm on any of the following radiation monitors:

• RM-G8 FHB Refueling Bridge Area Gamma

• RM-A6 Fuel Handling Bldg Exhaust

• RM-G6 Rx Bldg Refueling Bridge

• RM-G17A/B Rx Bldg Manipulator Crane (when installed) 2 3

4 5

6 DEF 1

RA2.2 Uncovery of irradiated fuel in the REFUELING PATHWAY Seismic event > OBE as indicated by EITHER:

Triaxial Seismic Switch MCB annunciator XCP-638 3-5 (RB FOUND SEIS SWITCH OBE EXCEED)

Any red OBE light on the Triaxial Response Spectrum Recorder Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-3 area AND Entry into the area is prohibited or impeded (Note 6)

None None Prepared for SCE&G by: Operations Support Services, Inc. - 3/7/15 2

3 4

5 6

DEF 1

RA1.4 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 10 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.

(Notes 1, 2) 3 Area Radiation Levels 2

3 4

5 6

DEF 1

RA3.2 An UNPLANNED event results in radiation levels that prohibit or impede access to any Table R-2 area (Note 6)

Hazardous event 2

3 4

5 6

DEF 1

HU3.2 Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode 2

3 4

5 6

DEF 1

HU3.3 Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) 2 3

4 5

6 DEF 1

HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 9) 2 3

4 5

6 DEF 1

HU4.2 Receipt of a single fire alarm (i.e., no other indications of a FIRE)

AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) 2 3

4 5

6 DEF 1

HU4.3 A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) 2 3

4 5

6 DEF 1

HU4.4 A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish 3

Natural or Tech.

Hazard Reactor Building Auxiliary Building Control Building Fuel Handling Building Intermediate Building Diesel Generator Building Turbine Building Service Water Pumphouse Safe Shutdown Yard Areas:

- RWST

- CST

- DG Fuel Oil Storage Table H-1 Fire Areas E

ISFSI 1

Confinement Boundary Damage to a loaded cask CONFINEMENT BOUNDARY 2

3 4

5 6

DEF 1

EU1.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than the following on the surface of the spent fuel cask (overpack):

60 mrem/hr ( + ) on the top of the overpack 600 mrem/hr ( + ) on the side of the overpack None None None 3

3, 4, 5 4, 5 3, 4, 5 1, 2, 3, 4, 5 3, 4, 5 3

4, 5 3, 4, 5 2, 3 3, 4, 5 1, 2 Table R-2 Safe Operation & Shutdown Areas Area Mode Applicability Auxiliary Building 374' Auxiliary Building 388' Auxiliary Building 400' Auxiliary Building 412' Auxiliary Building 436' Auxiliary Building 463' Intermediate Building 412' Intermediate Building 436' Intermediate Building 463' Control Building 412' Control Building 436' Turbine Building (All levels) 3 3, 4, 5 4, 5 3, 4, 5 1, 2, 3, 4, 5 3, 4, 5 3

4, 5 3, 4, 5 2, 3 3, 4, 5 1, 2 Table H-3 Safe Operation & Shutdown Areas Area Mode Applicability Auxiliary Building 374' Auxiliary Building 388' Auxiliary Building 400' Auxiliary Building 412' Auxiliary Building 436' Auxiliary Building 463' Intermediate Building 412' Intermediate Building 436' Intermediate Building 463' Control Building 412' Control Building 436' Turbine Building (All levels) 3 3, 4, 5 4, 5 3, 4, 5 1, 2, 3, 4, 5 3, 4, 5 3

4, 5 3, 4, 5 2, 3 3, 4, 5 1, 2 Note 1:

The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2:

If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3:

If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Notes Note 4:

During a tube rupture with reactor at power RM-G19A/B/C monitor readings are affected by 16N therefore they are not reliable until reactor has tripped and the monitors stable Note 5:

The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Note 7:

If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Note 8:

A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies Note 6:

If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted..

Note 9:

This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents 2

3 4

5 6

DEF 1

RA2.3 Lowering of spent fuel pool level to Level 2 (ele. 455 6) 2 3

4 5

6 DEF 1

RS2.1 Lowering of spent fuel pool level to Level 3 (ele. 437 0)

Spent fuel pool level at the top of the fuel racks 2

3 4

5 6

DEF 1

RG2.1 Spent fuel pool level cannot be restored to at least Level 3 (ele. 437 0) for > 60 min. (Note 1)

Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer 2

3 4

5 6

DEF 1

RA1.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses

> 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

Main Plant Vent Exhaust Gaseous Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE 14 mR/hr 28,000 cpm 2 X Hi-Rad alarm Liquid RB Purge Exhaust 740 mR/hr N/A 74 mR/hr N/A 2 X HI-Rad alarm Main Steam Line (Note 4)

Liquid Waste and Nuclear Blowdown Discharge 535 mR/hr 53.5 mR/hr 5.4 mR/hr N/A N/A 2 X Hi-Rad alarm 280,000 cpm N/A N/A N/A 7.4 mR/hr N/A N/A N/A N/A N/A RM-A3 (gas)

RM-G19 A/B/C RM-A13 RM-A4 (gas)

RM-A14 RM-L9 VCS UNIT 1 None VCS Unit 1 None None None None None None None None None None

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT S

System Malfunct.

1 Loss of ESF AC Power 6

RTS Failure 3

Loss of Control Room Indications 7

Loss of Comm.

4 RCS Activity 8

CMT Isolation Failure Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor SU1.1 Loss of all offsite AC power (Table S-1) capability to 7.2 KV ESF buses 1DA and 1DB for 15 min. (Note 1)

SA1.1 AC power capability to 7.2 KV ESF buses 1DA and 1DB reduced to a single power source (Table S-1) for 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS SA6.1 An automatic or manual trip fails to shut down the reactor AND Manual actions taken at the reactor control console are not successful in shutting down the reactor as indicated by reactor power 5% (Note 8)

SS6.1 An automatic or manual trip fails to shut down the reactor AND All manual actions to shut down the reactor are not successful in shutting down the reactor as indicated by reactor power 5%

AND EITHER of the following conditions exist:

CSFST Core Cooling-RED path conditions met CSFST Heat Sink-RED path conditions met Automatic or manual trip fails to shut down the reactor SU6.1 An automatic trip did not shut down the reactor after any RTS setpoint is exceeded AND A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8)

None Loss of all onsite or offsite communications capabilities SU7.1 Loss of all Table S-3 onsite communication methods OR Loss of all Table S-3 ORO communication methods OR Loss of all Table S-3 NRC communication methods Reactor coolant activity greater than Technical Specification allowable limits SU4.1 RCS leakage for 15 minutes or longer SU5.1 RCS unidentified or pressure boundary leakage > 10 gpm for 15 min.

OR RCS identified leakage > 25 gpm for 15 min.

OR Leakage from the RCS to a location outside containment

> 25 gpm for 15 min.

(Note 1) 1 2

3 4

1 2

3 4

Prolonged loss of all offsite and all onsite AC power to ESF buses Loss of all offsite and all onsite AC power to ESF buses for 15 minutes or longer SS1.1 Loss of all offsite and all onsite AC power (Table S-1) capability to 7.2 KV ESF buses 1DA and 1DB for 15 min.

(Note 1)

SG1.1 Loss of all offsite and all onsite AC power capability to 7.2 KV ESF buses 1DA and 1DB (Table S-1)

AND EITHER of the following:

• Restoration of at least one ESF bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

• CSFST Core Cooling-RED path conditions met Loss of all vital DC power for 15 minutes or longer SS2.1

< 108 VDC on both Train A and Train B vital 125 VDC systems for 15 min. (Note 1) 1 2

3 4

1 2

3 4

1 2

3 4

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress SA3.1 UNPLANNED loss of Control Room indications for 15 minutes or longer SU3.1 1

2 3

4 1

2 3

4 1

2 3

4 1

2 3

4 With letdown in service, RM-L1 high range monitor

> 40,000 cpm 1

2 3

4 FG1.1 1

2 3

4 Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1)

FS1.1 1

2 3

4 Loss or potential loss of any two barriers (Table F-1)

FA1.1 1

2 3

4 Any loss or any potential loss of either Fuel Clad or RCS (Table F-1)

F Fission Product Barrier Degradation Modes:

1 2

3 4

5 6

DEF Power Operations Startup Hot Standby Hot Shutdown Cold Shutdown Refueling Defueled EPP-001, Attachment I, Rev. 31 EAL Classification Matrix Page 2 of 3 HOT CONDITIONS RCS > 200°F)

An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1)

An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1)

AND Any of the following transient events in progress:

Automatic or manual runback greater than 25%

thermal reactor power

Electrical load rejection greater than 25% full electrical load

Reactor trip

ECCS actuation None Loss of all but one AC power source to ESF buses for 15 minutes or longer Loss of all offsite AC power capability to ESF buses for 15 minutes or longer Sample analysis indicates that a primary coolant activity value is > an allowable limit specified in Technical Specifications 3/4.4.8 SU4.2 None None None Prepared for SCE&G by: Operations Support Services, Inc. - 3/7/15 None SG1.2 Loss of all offsite and all onsite AC power (Table S-1) capability to 7.2 KV ESF buses 1DA and 1DB for 15 min.

AND

< 108 VDC on both Train A and Train B vital 125 VDC systems for 15 min. (Note 1) 1 2

3 4

2 Loss of Vital DC Power SU6.2 A manual trip did not shutdown the reactor AND A subsequent automatic trip or manual trip action taken at the reactor control consoles is successful in shutting down the reactor as indicated by reactor power < 5%

(Note 8)

Failure to isolate containment or loss of containment pressure control SU8.1 Containment isolation actuated AND At least one isolation valve in each penetration is not closed within 15 min. of the actuation (Note 1) 1 2

3 4

SU8.2 Containment pressure > 12 psig AND

< one full train of depressurization equipment (Table S-4) is operating per design for 15 min. (Note 1) 1 2

3 4

5 RCS Leakage None None None None None None None None Table S-4 Full Train Depressurization Equipment RBCU Groups Operating Containment Sprays Operating 2

1 0

0 1

2 Table F-1 Fission Product Barrier Matrix Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss A. CSFST Core Cooling-RED path conditions met A. RM-G7 or RM-G18 CNTMT HI RNG Gamma > 2,000 R/hr B. Dose equivalent I-131 coolant activity > 300 µCi/gm A. RM-G7 or RM-G18 CNTMT HI RNG Gamma > 100 R/hr A. An automatic or manual ECCS (SI) actuation required by EITHER:

UNISOLABLE RCS leakage SG tube RUPTURE A. Any condition in the opinion of the ED that indicates loss of the RCS barrier B. CSFST RCS Integrity-RED path conditions met A. Containment isolation is required AND EITHER:

Containment integrity has been lost based on ED judgment UNISOLABLE pathway from containment to the environment exists A. Any condition in the opinion of the ED that indicates loss of the Containment barrier A. CSFST Containment-RED path conditions met A. RM-G7 or RM-G18 CNTMT HI RNG Gamma > 20,000 R/hr A. Any condition in the opinion of the ED that indicates potential loss of the Containment barrier

1. RCS or SG Tube Leakage

2. Inadequate Heat Removal

3. CMT Radiation /

RCS Activity

4. CMT Integrity or Bypass

5. ED Judgment None None None None None A. A leaking or RUPTURED SG is FAULTED outside of containment None None None C. Containment pressure > 12 psig AND

< one full train of depressurization equipment (Table F-2) is operating per design for 15 min. (Note 1)

B. Containment hydrogen concentration > 4%

None None None None A. CSFST Heat Sink-RED path conditions met AND Heat sink required A. Operation of a standby charging pump is required by EITHER:

UNISOLABLE RCS leakage SG tube RUPTURE A. CSFST Core Cooling-ORANGE path conditions met B. CSFST Heat Sink-RED path conditions met AND Heat sink required A. CSFST Core Cooling-RED path conditions met AND Restoration procedures not effective within 15 min. (Note1)

B. Indications of RCS leakage outside of containment A. Any condition in the opinion of the ED that indicates loss of the fuel clad barrier A. Any condition in the opinion of the ED that indicates potential loss of the Fuel Clad barrier A. Any condition in the opinion of the ED that indicates potential loss of the RCS barrier Table S-1 AC Power Supplies Offsite:

115 KV power to XTF-4 and XTF-5

230 KV power to XTF-31

Parr Hydro Plant 13.8 KV power to ESF Bus 1DA or 1DB Onsite:

Diesel Generator A

Diesel Generator B None Table S-2 Safety System Parameters Reactor power Reactor vessel/pressurizer level RCS pressure Core Exit TCs Level in at least one SG EFW/AFW flow 1

1 1

1 Table F-2 Full Train Depressurization Equipment RBCU Groups Operating Containment Sprays Operating 2

1 0

0 1

2 Note 1:

The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Notes Note 8:

A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.

SA9.1 The occurrence of any Table S-5 hazardous event AND EITHER of the following:

Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.

The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.

Table S-5 Hazardous Events

Seismic event (earthquake)

Internal or external flooding event

High winds or tornado strike

FIRE

EXPLOSION

Other events with similar hazard characteristics as determined by the Shift Manager None 9

Hazard Event Affecting Safety System Loss of all AC and vital DC power sources for 15 minutes or longer 1

2 3

Table S-3 Communications Methods Gai-Tronics system X

Radio system X

X Private Branch Network X

Public Switched Telephone Network X

X X

Fiberoptic system X

X X

Satellite phone system X

X Federal Telephone System X

X ORO Dedicated System X

X System Onsite ORO NRC VCS UNIT 1 1

2 3

4 VCS Unit 1 None

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT C

Cold SD/

Refueling System Malfunct.

2 Loss of ESF AC Power 1

RCS Level 3

RCS Temp.

4 Loss of Vital DC Power 5

Loss of Comm.

Loss of reactor vessel/RCS inventory affecting fuel clad integrity with Containment challenged Loss of reactor vessel/RCS inventory affecting core decay heat removal capability Loss of all offsite and all onsite AC power to ESF buses for 15 minutes or longer CU2.1 CA2.1 Loss of all offsite and all onsite AC power (Table C-2) capability to 7.2 KV ESF buses 1DA and 1DB for 15 min.

(Note 1)

CS1.1 CONTAINMENT CLOSURE not established AND Reactor vessel level < 429 elevation (6" below the bottom of the hot leg penetration)

CG1.1 Reactor vessel level < 427 elevation (top of active fuel) for 30 min. (Note 1)

AND Any of the following indications of containment challenge:

CONTAINMENT CLOSURE not established (Note 7)

Containment hydrogen concentration > 4%

UNPLANNED increase in Containment pressure 5

6 DEF Loss of Vital DC power for 15 minutes or longer CU4.1

< 108 VDC on required DC buses for 15 min. (Note 1) 5 6

None Loss of reactor vessel/RCS inventory Unplanned loss of reactor vessel/RCS inventory for 15 minutes or longer CU1.1 UNPLANNED loss of reactor coolant results in reactor vessel/

RCS level less than a required lower limit for 15 min.

(Note 1)

CA1.1 Loss of reactor vessel/RCS inventory as indicated by level

< 429-6 elevation (bottom of hot leg penetration)

CU1.2 Reactor vessel/RCS level cannot be monitored AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of reactor vessel/RCS inventory 5

6 None Inability to maintain the plant in cold shutdown UNPLANNED increase in RCS temperature CU3.1 UNPLANNED increase in RCS temperature to > 200°F CA3.1 UNPLANNED increase in RCS temperature to > 200°F for

> Table C-3 duration (Note 1)

CU3.2 Loss of all RCS temperature and reactor vessel/RCS level indication for 15 min. (Note 1) 5 6

5 6

5 6

Loss of all onsite or offsite communications capabilities CU5.1 Loss of all Table C-4 onsite communication methods OR Loss of all Table C-4 ORO communication methods OR Loss of all Table C-4 NRC communication methods 5

6 COLD CONDITIONS (RCS 200°F)

Modes:

1 2

3 4

5 6

DEF Power Operations Startup Hot Standby Hot Shutdown Cold Shutdown Refueling Defueled EPP-001, Attachment I, Rev. 31 EAL Classification Matrix Page 3 of 3 5

6 CS1.3 Reactor vessel/RCS level cannot be monitored for 30 min.

(Note 1)

AND Core uncovery is indicated by any of the following:

• RM-G6 Rx Bldg Refueling Bridge or RM-G17A/B Rx Bldg Manipulator Crane > 90,000 mR/hr (when installed)

• Erratic source range monitor indication

• UNPLANNED increase in any Table C-1 sump / tank level of sufficient magnitude to indicate core uncovery 5

6 CS1.2 CONTAINMENT CLOSURE established AND Reactor vessel level < 427 elevation (top of active fuel) 5 6

None None AC power capability to 7.2 KV ESF buses 1DA and 1DB reduced to a single power source (Table C-2) for 15 min.

(Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Loss of all but one AC power source to ESF buses for 15 minutes or longer 5

6 DEF Prepared for SCE&G by: Operations Support Services, Inc. - 3/7/15 5

6 5

6 CA1.2 Reactor vessel/RCS level cannot be monitored for 15 min. (Note 1)

AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of reactor vessel/RCS inventory 5

6 CG1.2 Reactor vessel/RCS level cannot be monitored for 30 min.

(Note 1)

AND Core uncovery is indicated by any of the following:

• RM-G6 Rx Bldg Refueling Bridge or RM-G17A/B Rx Bldg Manipulator Crane > 90,000 mR/hr (when installed)

• Erratic source range monitor indication

• UNPLANNED increase in any Table C-1 sump / tank level of sufficient magnitude to indicate core uncovery AND Any of the following indications of containment challenge:

CONTAINMENT CLOSURE not established (Note 7)

Containment hydrogen concentration > 4%

UNPLANNED increase in Containment pressure 5

6 5

6 DEF CA3.2 UNPLANNED RCS pressure increase > 10 psig (This EAL does not apply during water-solid plant conditions) 5 6

None None Table C-2 AC Power Supplies Offsite:

115 KV power to XTF-4 and XTF-5

230 KV power to XTF-31

Parr Hydro Plant 13.8 KV power to ESF Bus 1DA or 1DB Onsite:

Diesel Generator A

Diesel Generator B Table C-1 Sumps/Tanks

• RB sump

• CCW surge tank

• PRT

• RCDT Note 1:

The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Notes Note 7:

If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced the EAL is not applicable Table C-3 RCS Reheat Duration Thresholds RCS Status Containment Closure Status Heat-up Duration N/A established not established 60 min.

• 20 min.

• 0 min.

Intact AND not at REDUCED INVENTORY Not intact OR at REDUCED INVENTORY Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.

CA6.1 The occurrence of any Table C-5 hazardous event AND EITHER of the following:

Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.

The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.

5 6

6 Hazardous Event Affecting Safety Systems None None None Table C-5 Hazardous Events

Seismic event (earthquake)

Internal or external flooding event

High winds or tornado strike

FIRE

EXPLOSION

Other events with similar hazard characteristics as determined by the Shift Manager Table C-4 Communications Methods Gai-Tronics system X

Radio system X

X Private Branch Network X

Public Switched Telephone Network X

X X

Fiberoptic system X

X X

Satellite phone system X

X Federal Telephone System X

X ORO Dedicated System X

X System Onsite ORO NRC VCS UNIT 1 VCS Unit 1 None None

ANSWER KEY REPORT for 14-01 NRC written Test Form: 0 Answers 0

1 B

2 B

3 B

4 B

5 C

6 B

7 C

8 A

9 A

10 C

11 C

12 C

13 B

14 D

15 B

16 D

17 D

18 B

19 D

20 D

21 C

22 A

23 C

24 B

25 D

26 B

27 D

28 D

29 C

30 C

31 C

32 B

33 D

34 A

35 A

36 B

37 D

38 B

39 B

40 C

41 D

42 A

43 A

44 C

45 C

46 A

47 D

Tuesday, June 14, 2016 2:20:17 PM 1

ANSWER KEY REPORT for 14-01 NRC written Test Form: 0 Answers 0

48 B

49 A

50 C

51 A

52 A

53 A

54 C

55 A

56 C

57 A

58 A

59 C

60 C

61 A

62 D

63 A

64 B

65 A

66 B

67 A

68 D

69 D

70 D

71 D

72 D

73 A

74 D

75 C

76 B

77 D

78 D

79 C

80 A

81 C

82 C

83 C

84 A

85 B

86 B

87 A

88 A

89 C

90 D

91 B

92 D

93 B

94 B

Tuesday, June 14, 2016 2:20:17 PM 2

ANSWER KEY REPORT for 14-01 NRC written Test Form: 0 Answers 0

95 C

96 D

97 D

98 C

99 C

100 A

Tuesday, June 14, 2016 2:20:17 PM 3

2016 V.C. Summer NRC Initial Written Examination Class 14-01 Name _______________________________________ Date ______________________

1.

26.

51.

2.

27.

52.

3.

28.

53.

4.

29.

54.

5.

30.

55.

6.

31.

56.

7.

32.

57.

8.

33.

58.

9.

34.

59.

10.

35.

60.

11.

36.

61.

12.

37.

62.

13.

38.

63.

14.

39.

64.

15.

40.

65.

16.

41.

66.

17.

42.

67.

18.

43.

68.

19.

44.

69.

20.

45.

70.

21.

46.

71.

22.

47.

72.

23.

48.

73.

24.

49.

74.

25.

50.

75.

• • • • • • KEY******

• • • • • KEY****

2016 V.C. Summer NRC Initial Written Examination Class 14-01 Name _______________________________________ Date ______________________

76.

77.

78.

79.

80.

81.

82.

83.

84.

85.

86.

87.

88.

89.

90.

91.

92.

93.

94.

95.

96.

97.

98.

99.

100.

• • • • KEY****

• • • • KEY****