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Initial Exam 2015-301 Final RO Written Exam
	ML15068A031
Person / Time
Site:	Turkey Point
Issue date:	03/09/2015
From:	; NRC/RGN-II
To:	; Florida Power & Light Co
References
	50-250/15-301, 50-251/15-301
	Download: ML15068A031 (102)
	v • d • e

PTN L-15-1 NRC EXAM This information is controlled by P"I N's 2015 Lorr (L-15-1) NRC EXAMINAT ION SECURITY AGREE MEN I'.

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

Date: Facility/Unit: Turkey Point Units 3 and 4 Region: II Reactor Type: w Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. You have 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to complete the examination.

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

- - I - -I- - Points Applicant's Score

--I --I -- Points Applicant's Grade --I --I -- Percent This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 1 Given the following conditions:

Unit 3 is operating at 100% power with all systems in automatic.

Which ONE of the following completes the statement below?

Raising ____ will result in a lower final steady state RCP #1 Seal leakoff flow.

A. Reactor Coolant Drain Tank pressure B. Volume Control Tank pressure C. Reactor Coolant Pump Standpipe level D. Letdown flow This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 2 Given the following conditions:

Unit 4 is in MODE 5.

The Pressurizer is solid with OMS in low pressure operation.

4B RHR loop is in operation.

RCS pressure control is in manual.

Subsequently:

Operator manually lowers demand on TC-4-144A, Letdown Temperature Controller.

Which ONE of the following identifies the plant response with no additional operator action?

A. NRHX CCW flow decreases.

B. Shutdown Margin increases.

C. RCS pressure decreases.

D. RCS temperature increases.

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QUESTION 3 Given the following conditions:

Letdown flow is 105 gpm.

Which ONE of the following completes the statement below?

In accordance with 3-0P-047, CVCS-Charging and Letdown, before removing a letdown orifice from service, PCV-3-145, Low Pressure Letdown Controller demand is (1) to maintain pressure at (2) on Pl-3-145.

A. (1) raised (2) 150 psig B. (1) raised (2) 300 psig

c. (1) lowered (2) 150 psig D. (1) lowered (2) 300 psig This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 4 Given the following conditions:

Unit 3 is in MODE 4.

RCS pressure is 370 psig.

Crew prepares to place RHR in service.

MOV-3-750, Loop 3C RHR Pump Suction Stop Valve, is open.

MOV-3-8628, RHR Suction from RWST, is open.

Subsequently:

RCS pressure rises to 475 psig.

MOV-3-751, Loop 3C RHR Pump Suction Stop Valve, will not open.

Which ONE of the following completes the statements below?

MOV-3-751 (1) prevented from opening by RCS pressure.

MOV-3-751 (2) prevented from opening by MOV-3-8628.

A. (1) is (2) is not

8. (1) is not (2) is not

c. (1) is (2) is D. (1) is not (2) is This information is controlled by PTN' s 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 5 Given the following conditions:

A large break LOCA has occurred on Unit 3.

The 38 4kV Bus is faulted and locked-out.

The crew is performing actions of 3-EOP-E-1, Loss of Reactor or Secondary Coolant.

Conditions are met to isolate the SI Accumulators.

Which ONE of the following completes the statement below and identifies the Accumulator Isolation Valves capable of being closed?

MOV-3-865A and (1) , Accumulator Discharge Isolation valves, can be closed from m_.

A. (1) MOV-3-8658 (2) VPB B. (1) MOV-3-865C (2) VPB

c. (1) MOV-3-8658 (2) MCC using pushbutton D. (1) MOV-3-865C (2) MCC using pushbutton This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 6 Which ONE of the following completes the statements below regarding the operation of the Pressurizer Relief Tank?

PCV-3-473, Nitrogen Regulator, is set to maintain PRT pressure at (1) psig.

PCV-3-473 (2) automatically isolate on a Containment Isolation Phase A.

A. (1) 6 - 8 (2) will B. (1)10-12 (2) will C. (1) 6 - 8 (2) will NOT D. (1)10-12 (2) will NOT This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 7 Given the following conditions:

Unit 3 is in MODE 4.

TCV-3-143, LID Demineralizer Divert Valve, is aligned to AUTO.

RHR is in service.

Subsequently:

Annunciator A3/5, LTDN DEMIN HI TEMP/FLOW DIVERTED, alarms.

Fl-3-620, local NRHX CCW flow is reading 90 gpm.

Which ONE of the following completes the following statements below?

The LTDN DEMIN HI TEMP alarm setpoint is (1}

In accordance with 3-ARP-097.CR.A, A3/5, the RCO must manually adjust (2) to restore Letdown Temperature.

A. (1) 135°F (2) TC-3-144A, UD Temp Controller B. (1) 135°F (2) HCV-3-142, RHR Ltdn to eves C. (1)125°F (2) TC-3-144A, UD Temp Controller D. (1)125°F (2) HCV-3-142, RHR Ltdn to eves This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 8 Which ONE of the following identifies the power supply to the Spray Valve Controller for PCV-3-4558, Pressurizer Spray Valve, Loop B?

REFERENCE PROVIDED A. 3001 B. 3023 C. 3P06 D. 3P08 This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 9 Given the following conditions:

Unit 4 is raising power with Main Generator output at 200 MW.

l&C is performing 4-SMl-086.01, Turbine Emergency Trip Header Pressure Channel Calibration of PS-4-3629.

After discovering PS-4-3629 is calibrated to 45 psig, l&C suspends 4-SMl-086.01 for lunch and leaves work as is.

Subsequently:

Prior to restart l&C verifies PS-4-3630, Turbine Emergency Trip Header Pressure Switch, is tripped prior to recommencing work.

Which of the following completes the sentence below?

The Turbine Emergency Trip Header Pressure Switches are expected to be set at (1)

Continue with (2)

A. (1) 45 psig (2) 4-EOP-E-O, Reactor Trip or Safety Injection B. (1) 45 psig (2) 4-GOP-301, Hot Standby to Power Operation C. (1) 1000 psig (2) 4-GOP-301, Hot Standby to Power Operation D. (1) 1000 psig (2) 4-EOP-E-O, Reactor Trip or Safety Injection This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 10 Given the following conditions:

Unit 4 is at 7% power.

Pressurizer Pressure Channel PT-4-457 fails high.

PT-4-457 is removed from service in accordance with 4-0NOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels.

Which ONE of the choices below completes the follow statements?

At Power Trips Blocked light on VPA is (1) for current plant conditions.

With PT-4-457 removed from service, a Safety Injection (2) occur with a subsequent loss of 4P06.

A. (1) NOT lit (2) will NOT B. (1) NOT lit (2) will C. (1) lit (2) will NOT D. (1) lit (2) will This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 11 Given the following conditions:

Unit 4 is at 50% power.

PT-4-494, C SIG Pressure Transmitter Channel II, fails high.

The bi-stables for the failed channel are tripped in accordance with 4-0NOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels.

Subsequently:

PT-4-495, C SIG Pressure Transmitter Channel Ill, fails low.

Which ONE of the following describes ~ow the plant will respond and why?

An automatic Safety Injection will (1) Loop C Lo Stm Pressure inputs to the actuation logic are indicating a header pressure 100 psig (2) SIG pressure.

A. ( 1) occur because 2/3 (2) greater than

8. (1) occur because 2/3 (2) less than C. (1) NOT occur because only 113 (2) greater than D. (1) NOT occur because only 113 (2) less than This information is controlled by PTN' s 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 12 Given the following plant conditions:

Unit 3 is operating at 100% power.

The breaker to 3A Vital MCC has tripped OPEN.

Which ONE of the following components has lost power?

A. 38 Auxiliary Building Exhaust Fan B. 3A Main Steam Penetration Cooling Fan C. 3A Normal Containment Cooler Fan D. 38 CROM Cooler Fan This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 13 Given the following conditions:

Unit 4 is at 100% power.

Subsequently:

A fault occurs on the 4C SG inside containment.

A Reactor Trip and Safety injection occur.

4C SG completely depressurizes.

The Pressurizer is empty.

Containment temperature is 197°F and lowering.

Containment pressure is 15 psig and lowering.

The plant conditions are stabilized.

Which ONE of the following identifies the instrumentation which is NOT designed to operate under the given conditions?

A. Particulate/Gas Monitors, R-11 /R-12

8. 4C SG Pressure Transmitters, PT-4-494/495/496 C. Pressurizer Level Transmitters, LT-4-459/460/461 D. 4C SG Level Transmitters - Narrow range , LT-4-494/495/496 This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 14 Given the following conditions:

A Unit 3 Reactor Trip and Safety Injection occurs.

RCS pressure is 1000 psig and stable.

RCS subcooling is 45°F.

Containment pressure rises to 16 psig and continues to slowly rise.

Annunciator H5/2, CNTMT ISOLATION ACTIVATED, is in alarm.

The crew is performing 3-EOP-E-O, Reactor Trip or Safety Injection.

The RCO is checking Containment Spray requirements IAW Attachment 3.

Which ONE of the following describes the NEXT required operator action(s) in accordance with ?

A. Manually initiate Containment Spray using ONLY the Containment Spray pushbuttons.

B. Manually start the Containment Spray Pumps and open Containment Spray Isolation Valves.

C. Verify SI is reset. Verify SI Amber Lights on VPB are ALL BRIGHT.

D. Verify SI is reset and stop ALL RCPs.

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QUESTION 15 Given the following conditions:

Unit 4 is at 100% power.

Subsequently:

A Reactor trip occurs.

The crew is performing Immediate Operator Actions of 4-EOP-E-O, Reactor Trip or Safety Injection.

The BOP attempts to manually close MSR Main Steam Stop Supply Valves, but the position indications are NOT lit.

Which ONE of the following is required NEXT in accordance with 4-EOP-E-O?

A. Manually trip the turbine.

B. Close Main Steamline Isolation and Bypass valves.

C. Continue verification of MSR Purge valve closure.

D. Continue verification of Reheater Timing valve closure.

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QUESTION 16 Given the following conditions:

Unit 3 Reactor power is stable at 8%.

The RCO is manually controlling SG levels at 50%.

Subsequently:

The Unit Supervisor directs raising power to 30%.

During Turbine load increase, the Turbine Control valves opened rapidly.

Reactor power increases to 14%.

All SG level deviations are in alarm.

Which ONE of the following completes the following statements?

The initial SG level deviation alarms occur because SG narrow range levels indicate 5% _ill than programmed level. The RCO (2) required to throttle the Main Feedwater Control valves to maintain SIG levels.

A. (1) less (2) is B. (1) greater (2) is C. (1) less (2) is NOT D. (1) greater (2) is NOT This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 17 Given the following:

Unit 4 is at 60% power.

Subsequently:

A Condensate header rupture occurs.

Both Main Feedwater Pumps trip.

After 10 seconds, SG levels are 30% narrow range and lowering.

Which ONE of the following completes the following statements?

An automatic Reactor Trip setpoint (1) currently exceeded. AFW Pumps (2) running.

A. (1) is (2) are B. (1) is (2) are NOT C. (1) is NOT (2) are.

0. (1) is NOT (2) are NOT.

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QUESTION 18 Given the following conditions:

Unit 3 is increasing power following a refueling outage.

Unit 3 experiences a Reactor trip from 25% power.

The crew completes 3-EOP-E-O, Reactor Trip and Safety Injection, and enters 3-EOP-ES-0.1, Reactor Trip Response.

Pressurizer level is 14% and slowly decreasing.

All Steam Generator Narrow Range levels are between 12% and 15% and slowly rising.

Steam Generators pressures are approximately 990 psig and slowly decreasing.

Tavg is 544°F and slowly decreasing.

RCS pressure is 2125 psig and slowly decreasing.

Which ONE of the following identifies the crew's initial required response in accordance with 3-EOP-ES-0.1 to address the conditions above?

A. Establish Emergency Boration.

B. Reduce Auxiliary Feedwater Flow.

C. Close MSIVs and bypass valves.

D. Initiate a Safety Injection and return to 3-EOP-E-O.

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QUESTION 19 Given the following conditions:

The Auxiliary Feedwater (AFW) system receives an auto-start signal.

A AFW Pump governor malfunctions and AFW Turbine speed rises.

Which ONE of the following describes the effect of this event?

The A AFW Pump trips on over-speed at (1)

The Train 1 AFW Flow Control Valves will (2)

A. (1) 5900 rpm (2) remain throttled B. (1) 5900 rpm (2) fully open

c. (1) 6500 rpm (2) remain throttled D. (1) 6500 rpm (2) fully open This information is controlled by PTN ' s 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 20 Given the following plant conditions:

Unit 3 is in MODE 1.

Unit 4 is in MODE 1.

A lightning strike damages the Unit 3 Startup Transformer.

Which ONE of the following statements correctly describes the required response in accordance with TS 3.8.1.1?

Within one hour, the Unit 3 crew (1) required to demonstrate the OPERABILITY of the Unit 4 Startup Transformer and (2) required to demonstrate the OPERABILITY of the Unit 3 EDGs.

A. (1) is (2) is B. (1) is not (2) is

c. (1) is (2) is not D. (1) is not (2) is not This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 21 Which ONE of the following describes the effect of placing the yellow NORMAUISOLATE switch to ISOLATE on the 4B HHSI Pump breaker cubicle?

A Enables the pump's control switch on the 4KV Breaker cubicle door.

B. Aligns backup fuses into the 4B HHSI Pump control circuit.

C. Allows local starts at the SI Pump Room.

D. Disables all 4B HHSI Pump Breaker protective trip signals.

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QUESTION 22 Given the following conditions:

The 3A RHR Pump is started.

After the pump starts, DC control power is lost.

Which ONE of the following completes the following sentence?

The 3A RHR Pump Breaker 3AA15 _ _ __

A. can NOT be tripped remotely.

The breaker's blue light is extinguished.

B. can be tripped remotely.

The breaker's blue light is extinguished.

C. can NOT be tripped remotely.

The breaker's blue light is on.

D. can be tripped remotely.

The breaker's blue light is on.

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QUESTION 23 Given the following conditions:

The 4A EOG is running in parallel with 4A 4KV bus.

The 4A EOG is running at 2800 KW

The Varmeter on the 4A EOG Generator Control Panel 4C12A is reading 500 KVAR out.

Subsequently:

The operator locally positions the Voltage Regulator Switch at the 4A EOG Generator Control Panel 4C12A to LOWER.

Which ONE of the following predicts the initial behavior of EOG parameters?

A. ( 1 ) VARs increase (2) current and temperature lower B. ( 1) VARs decrease (2) current and temperature lower C. ( 1 ) VARs increase (2) current and temperature rise

0. ( 1) VARs decrease (2) current and temperature rise This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 24 Given the following conditions:

A bus fault has caused 125 VDC Electrical Distribution Bus 3D01 to de-energize.

Which ONE of the following plant components will be directly affected by this loss of DC power?

A. Loss of Train 2 Feedwater Isolation capability B. Loss of all 3D Switchgear control power C. Loss C AFW Pump control and protection D. Loss of 3A EDG DC power This information is controlled by PTN's 2015 LOJT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 25 Given the following conditions:

Unit 3 and 4 are at 100% power

Liquid release is in progress IAW O-NOP-061.11C, Controlled Liquid Release From Monitor Tank A.

Subsequently:

Annunciator H1/4, PRMS HI RADIATION, alarms.

Annunciator H1/6, PRMS CHANNEL FAILURE, alarms.

Unit 3 RCO reports no indications are available for Process Radiation Monitor R-18 due to no power.

Which ONE of the following describes the initial required response?

A. Enter the applicable TS Action Statement.

B. Enter 3-0NOP-067, Radioactive Effluent Release.

C. Direct Chemistry to sample effluent in accordance with ODCM.

D. Direct Chemistry to sample effluent in accordance with 3-ARP-097.CR.H, H1/6.

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QUESTION 26 Given the following conditions:

Units 3 and 4 are at 100% power with a normal electrical alignment.

3A and 38 ICW Pumps are running.

Subsequently:

Annunciator 13/4, TRAVELING SCREEN GENERAL TROUBLE, alarms

Annunciator 14/2, ICWP A/B/C TRIP, alarms

Annunciator 14/4, ICW HEADER A/8 LO PRESS, alarms

Ultimate Heat Sink Temperature on Tl-3/4-3605 is 96°F.

3A1 and 3A2 Traveling Screen Wash Pumps are ON.

3A ICW Pump RED light is lit.

Which ONE of the following completes the statement below?

In accordance with 3-ARP-097.CR.I, the RCO (1) start the 3C ICW pump and refer to (2) to address the given conditions?

A. (1)will (2) 3-0NOP-011 .1, Intake Canal Low Level or High Temperature

8. (1) will NOT (2) 3-0NOP-011.1, Intake Canal Low Level or High Temperature C. (1) will (2) 3-0NOP-019, Intake Cooling Water Malfunction D. (1) will NOT (2) 3-0NOP-019, Intake Cooling Water Malfunction This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 27 Given the following conditions:

Units 3 and 4 are at 100% power.

The 4CM Instrument Air compressor is out for maintenance.

3CM is running in LEAD, 3CD is in LAG, 4CD is in STANDBY-LAG.

The Instrument Air systems are cross-tied.

Subsequently:

Unit 4 Annunciators 16/1, INSTR AIR SYSTEM HI TEMP/LO PRESS, alarms.

Which ONE of the following identifies (1) the setpoint at which CV-4-1605, UNIT 4 Instrument Air Crosstie Isolation Control Valve, closes to protect Unit 4 and (2) the minimum pressure on Pl-4-1444, INST AIR PRESS, if it cannot be maintained, when a Unit 4 Reactor Trip is required per O-ONOP-013, Loss of Instrument Air?

A. (1) 80 psig (2) 60 psig B. (1) 80 psig (2) 65 psig C. (1) 90 psig (2) 60 psig D. (1) 90 psig (2) 65 psig This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 28 Given the following conditions:

Unit 3 is in MODE 1.

An emergent Containment Entry is in progress to investigate RCS leakage.

O-ADM-009, Containment Entries When Containment Integrity Is Established, is in effect.

NOTES

TS LCO 3.6.1.3 is Containment Air Locks.

Which ONE of the following completes the following statements?

TS LCO 3.6.1.3 (1) satisfied with a single Personnel Hatch door open for transit.

In accordance with O-ADM-009, the RCO (2) required to log the status of Personnel Hatch Inner and Outer Door indication from VPB.

A (1) is not (2) is not B. (1) is (2) is not C. (1) is (2) is D. (1) is not (2) is This information is controlled by PTN's 2015 LOIT (L 1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 29 Given the following conditions:

Unit 3 is in a refueling outage.

The Upper Internals Assembly is being lifted out of the Reactor Vessel to set into the lower cavity.

Subsequently:

84/1, SOURCE RANGE HI FLUX AT SHUTDOWN, alarms.

The Source Range high flux alarm (1) initiate a containment evacuation alarm.

If flux continues to increase, 84/1 (2) require the operators to initiate Emergency 8oration.

A. (1) will (2) does

8. (1)will (2) does NOT

c. (1) will NOT (2) does D. (1) will NOT (2) does NOT This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 30 Given the following conditions:

Unit 4 is in MODE 5.

Letdown is in service.

The RCS is solid.

A plant cooldown and depressurization is in progress.

Which ONE of the following describes the action to lower RCS pressure?

A. Raise the setpoint of HCV-4-142, RHR LID to eves.

B. Raise the setpoint of PCV-4-145, Low Pressure LTDN Controller.

C. Lower the setpoint of HCV-4-142, RHR LID to eves.

D. Lower the setpoint of PCV-4-145, Low Pressure LTDN Controller.

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QUESTION 31 Given the following conditions:

Unit 3 trips due to Loss of Offsite Power.

3A and 38 EDGs energize their respective 4KV buses.

3-EOP-ES-0.2, Natural Circulation Cooldown, is being implemented.

Automatic SI is blocked in accordance with 3-EOP-ES-0.2.

Subsequently during the cooldown:

Average CET temperature is 586°F and rising.

RCS pressure is 1335 psig and rapidly lowering.

Pressurizer level is 11 % and lowering.

Which ONE of the following actions is required in accordance with 3-EOP-ES-0.2?

A. Verify PORVs are closed.

B. Initiate SI and Phase A.

C. Start all available Charging Pumps.

D. Energize all Pressurizer heaters.

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QUESTION 32 Given the following conditions:

A Unit 3 Reactor Startup is in progress.

Power level is 1 x 1o-8 amps.

Subsequently:

120V Vital Instrument Panel 3P07 is lost.

Which ONE of the following completes the following statement?

A Reactor Trip occurs due to the loss of (1) , and as power lowers below 10E-10 amps_

fil__ will energize.

A. (1) IR Nl-35 (2) SR Nl-31 B. (1) IR Nl-35 (2) SR Nl-32 C. (1) IR Nl-36 (2) SR Nl-31 D. (1) IR Nl-36 (2) SR Nl-32 This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 33 Given the following conditions:

Unit 4 is in MODE 6.

The crew has commenced rod unlatching .

Subsequently:

Annunciator 14/6, CNTMT SUMP HI LEVEL, alarms.

Reactor Cavity water level is 56 feet and lowering.

Bubbles are rising around the outside of the Reactor Vessel.

R-11, Containment Air Particulate Monitor, is in alarm Which ONE of the following completes the following statements based on the given conditions?

Containment Purge Isolation valves (1) required to be closed in accordance with 4-0NOP-033.2, Refueling Cavity Seal Failure.

Normal Containment Coolers are -"""'2""")__ secured.

A. (1) are (2) automatically B. (1) are NOT (2) manually

c. (1) are (2) manually D. (1) are NOT (2) automatically This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 34 Given the following conditions:

A Unit 3 startup is in progress.

The crew prepares to synch the Main Generator to the grid.

Rod Control is in Manual.

Steam Dumps to Condenser Mode Selector Switch is in Manual.

Tavg is 550°F.

Tref is 548°F.

Power Level is 7%.

Subsequently:

The Turbine trips.

Which ONE of the following completes the following sentence for the initial plant response to the Turbine trip?

The Condenser Steam Dumps will _ __

A. modulate open on the Turbine trip program B. quick open on the Turbine trip program C. quick open on the Steam Header pressure D. modulate open on the Steam Header pressure This information is controlled by PTN's 2015 LOIT (L 1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 35 Given the following conditions:

Unit 3 is at 100% power Subsequently:

BOP manually trips the Turbine due to high vibration.

Which ONE of the following identifies how secondary system parameters respond?

PT-3-447, Turbine Inlet Pressure, ( 1) and Condenser Steam Dumps will reduce RCS temperature (2)

A. (1) lowers (2) to no load Tavg B. (1) rises (2) to no load Tavg

c. (1) lowers (2) to 5°F of program D. (1) rises (2) to 5°F of program This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 36 Given the following conditions:

Unit 4 is at 100% power.

Main Condenser vacuum is lowering.

The crew enters 4-0NOP-014, Main Condenser Loss of Vacuum.

Which ONE of the following describes (1) the initial required action and (2) radiation monitoring capabilities?

A. (1) Place the Steam Jet Air Ejector Hogging Jet in service.

(2) SJAE SPING is not OPERABLE.

B. (1) Place the Steam Jet Air Ejector Hogging Jet in service.

(2) SJAE SPING is OPERABLE.

C. (1) Place the Standby Steam Jet Air Ejector in service.

(2) SJAE SPING is not OPERABLE.

0. (1) Place the Standby Steam Jet Air Ejector in service.

(2) SJAE SPING is OPERABLE.

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QUESTION 37 Which ONE of the following is a condition which causes X6/2, RADWASTE BLDG PANEL C46 TROUBLE, to alarm in the Control Room?

A. Spent Resin Storage Tank High Level B. Waste Monitor Tank A High Level C. Reactor Coolant Drain Tank Unit 3 High Level D. Waste Liquid High Radiation This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 38 Given the following conditions:

Unit 3 startup is in progress per 3-GOP-301, Hot Standby to Power Operation.

Unit 3 is at 25% power and stable.

RCS Boron Concentration is 1274 ppm.

Turbine is in MW Control with the following Control Valve positions:

CV-UL (#1) 3% CV-UR (#2) 3%

CV-LL (#3) 14.5% CV-LR (#4) 14.5%

Subsequently:

CV-3-1606, A SIG Atmospheric Steam Dump fails open.

Which ONE of the following identifies the INITIAL response of the plant as result of this failure, with no operator action?

A. Main Generator electrical output will remain constant at 25% with the Turbine in MW control.

B. Reactor Coolant Hot Leg temperature, as read on TR-3-413, rises and results in a decrease in subcooling.

C. The Turbine Control valves, CV-LL and CV-LR, move to maintain constant Turbine Inlet Pressure.

D. Reactor Coolant Cold Leg temperature, as read on TR-3-410, lowers and results in a Reactor power rise.

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QUESTION 39 Given the following plant conditions:

Unit 3 trips from 100% power on a spurious SI actuation.

Crew enters 3-EOP-E-O, Reactor Trip or Safety Injection.

RCS pressure is 2150 psig and recovering.

RCS Tavg stabilizes at 535°F.

Which ONE of the following correctly completes the following statement?

To ensure adequate Shutdown Margin, a minimum boration rate of (1) is required until Shutdown Margin is verified.

From the time of the trip, with no operator action, Shutdown Margin will be (2) after eighteen (18) hours.

A. (1) 20 gpm (2) higher B. (1) 45 gpm (2) higher

c. (1) 20 gpm (2) lower D. (1) 45 gpm (2) lower This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 40 Given the following conditions:

A loss of all AC power occurs on Unit 3.

The crew enters 3-EOP-ECA-0.0, Loss of All AC Power.

Pressurizer level is 5% and lowering.

Containment Sump level is rising.

The crew prepares to depressurize the SGs.

Which ONE of the following actions (1) ensures natural circulation or reflux boiling cooling is sufficient and (2) lists the applicable recovery procedure after power is restored?

A. (1) Verify total AFW flow between 400 and 450 gpm.

(2) 3-EOP-ECA-0.1, Loss of All Power Recovery Without SI Required B. (1) Maintain >7% Narrow Range Level in at least one SG.

(2) 3-EOP-ECA-0.1, Loss of All Power Recovery Without SI Required C. (1) Verify total AFW flow between 400 and 450 gpm.

(2) 3-EOP-ECA-0.2, Loss of All Power Recovery With SI Required D. (1) Maintain >7% Narrow Range Level in at least one SG.

(2) 3-EOP-ECA-0.2, Loss of All Power Recovery With SI Required This information is controlled by PTN' s 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 41 Given the following conditions:

Unit 3 Reactor trips and Safety Injection actuates.

3A 4KV Bus is de-energized due to an overcurrent condition.

All RCPs are tripped in accordance with the Fold-Out page.

The crew is performing Attachment 3 of 3-EOP-E-O, Reactor Trip or Safety Injection.

All other safeguards equipment functions normally.

Which ONE of the following completes the following statement?

In accordance with Attachment 3 of 3-EOP-E-O, the RCO _.....,l.1. . ).......__manually start the Normal Containment Coolers.

The 3B ECG (2) automatically start.

A. (1)will (2) will B. (1)will (2) will NOT C. (1) will NOT (2) will NOT D. (1) will NOT (2) will This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 42 Given the following conditions:

Unit 4 is at 38% power and stable.

4C Loop RCS Flow Meters are: 86%, 89%, and 91 %.

4C RCP motor frame vibration is 3 mils.

4C RCP Oil Reservoir level W' below normal.

Which ONE of the following identifies the crew response in accordance with 4-0NOP-041.1, Reactor Coolant Pump Off-Normal, to the given conditions?

A. No actions required to trip 4C .RCP or Reactor or reduce load.

B. A Reactor trip and 4C RCP trip are required.

C. A Reactor trip is required and 4C RCP trip is NOT required.

D. A Fast Load Reduction is required.

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QUESTION 43 Given the following conditions:

Unit 3 is at 100% power with 3A Charging Pump running .

HCV-3-121, Charging Flow To Regen Heat Exchanger, is throttled.

Subsequently:

Annunciator A6/5, RCP LABYRINTH SEAL LO ~P. alarms.

Annunciator A5/1, CHARGING PUMP A TRIP, alarms.

3C Charging Pump is started in accordance with 3-0NOP-047.1, Loss of Charging Flow.

The RCO balances Charging and Letdown flows, and establishes a stable VCT level.

RCP Seal Injection flows are 4 to 5 gpm.

Which ONE of the following completes the following sentence?

IAW 3-0NOP-041.01, Reactor Coolant Pump Off-Normal, the first action to address the given conditions is _ __

A. locally throttle open 3-297A/B/C RCP Seal Injection Valves B. manually raise 3C Charging pump speed C. manually throttle closed HCV-3-121 D. manually throttle open HCV-3-121 This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 44 Given the following conditions:

Unit 3 is in MODE 4 in 3-GOP-305, Hot Standby to Cold Shutdown.

Unit 3 is cooling down with the 3B RHR loop in accordance with 3-0P-050, Residual Heat Removal System.

3A CCW Heat Exchanger is out of service for maintenance with required pumps in pull-to-lock.

Subsequently:

The 3B CCW pump breaker trips on motor overload, causing an electrical transient resulting in a momentary loss of the 3B 4KV Bus.

3B 4KV Bus power is restored on the 3B EOG.

Which ONE of the following completes the statement below?

Shutdown cooling is restored when 3B RHR pump (1) and 3C CCW Pump (2)

A. ( 1) auto starts from the sequencer (2) auto starts from the sequencer B. (1) is manually started (2) auto starts from the sequencer C. (1) auto starts from the sequencer (2) is manually started D. (1) is manually started (2) is manually started This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGRE EMENT.

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QUESTION 45 Given the following conditions:

Unit 4 is at 80% power.

48 Steam Generator Feed Pump's breaker trips.

Subsequently

PC-4-444J, Pressurizer Pressure Controller, does not respond in automatic.

Which ONE of the following completes the sentences below?

PC-4-444J's output is required to be initially ( 1) in accordance with 4-0NOP-041.5 Pressurizer Pressure Control Malfunction. PORV (2) may be operated by controller PC-4-444J.

A. (1) raised (2) PCV-4-455C B. (1) raised (2) PCV-4-456

c. (1) lowered (2) PCV-4-456 D. (1) lowered (2) PCV-4-455C This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 46 Given the following conditions:

Unit 3 experiences an Anticipated Transient without Scram (ATWS).

3-EOP-FR-S.1, Response To Nuclear Power Generation/ATWS is in progress.

The RCO is initiating Emergency Boration.

Neither 3A nor 3B Boric Acid Pumps starts.

Which ONE of the following describes the required response to initiate Emergency Boration?

A. Open MOV-3-350, Emergency Boration Valve.

B. Close LCV-3-115C, VCT Outlet to Charging Pump Suction.

C. Close FCV-3-113B, Blender to Charging Pump.

D. Open 3-356, Manual Emergency Boration Valve.

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QUESTION 47 Given the following conditions:

A SGTR occurs on 3B S/G.

3-EOP-E-3, Steam Generator Tube Rupture is in progress.

A cooldown is commenced to target temperature.

Subsequently:

RCS Subcooling lowers to 14°F with no other accidents in progress.

Which ONE of the following identifies whether or not the RCPs should be secured and why?

A. Trip the Reactor Coolant Pumps to minimize the potential for RCP damage when an RCS depressurization is initiated.

B. Trip the Reactor Coolant Pump on the affected loop to minimize RCS inventory loss.

C. Keep the Reactor Coolant Pumps running to prevent the automatic opening of the SOTA.

D. Keep the Reactor Coolant Pumps running because a controlled RCS cooldown is in progress.

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QUESTION 48 Given the following conditions:

Unit 3 is at 100% power.

3A EOG is out of service for maintenance.

A fault on the Unit 3 Startup Transformer generates a Transformer Lockout and a 3B 4KV Bus Lockout which do not reset.

3-EOP-ECA-0.0, Loss of All AC Power, is in progress.

Which ONE of the following statements identifies (1) the preferred source of power restoration in accordance with 3-EOP-ECA-0.0 and (2) the design basis battery duration?

A. (1) 3C 4KV Bus (2) 30 minutes B. (1) 3C 4KV Bus (2) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

c. (1) Unit 4 Startup Transformer (2) 30 minutes D. (1) Unit 4 Startup Transformer (2) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 49 Given the following conditions:

Unit 3 is at 3% power while performing a reactor startup.

Vital Instrument Panel 3P06 loses power.

3-0NOP-003.6, Loss of 120V Vital Instrument Panel 3P06, is in progress.

Which ONE of the following completes the statements below?

The Pressurizer Control Heaters _(1 ) _ . 3-0NOP-003.6 directs the reduction of Charging flow to reduce the PRZ fill rate to prevent _(2)_.

A. ( 1) remain on (2) lifting a PRZ PORV B. (1) de-energize (2) lifting a PRZ PORV C. (1) remain on (2) a High PRZ level trip D. ( 1) de-energize (2) a High PRZ level trip This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 50 Given the following conditions:

Unit 3 is in MODE 3.

Vital 480V MCC 3B is out of service.

Subsequently:

Vital DC Bus 3D23 loses power due to a fault on the bus.

The crew is restoring power to DC Bus 3D23 in accordance with 3-0NOP-003.5, Loss of DC Bus 3D23 and 3D23A (3B).

The fault has been isolated.

Which ONE of the following identifies (1) the battery charger that is still OPERABLE and (2) the expected battery charger voltage?

A. (1) 3B1 (2) 120 volts B. (1) 3B1 (2) 135 volts

c. (1) 3B2 (2) 135 volts D. (1) 3B2 (2) 120 volts This information is controlled by PTN' s 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 51 Given the following conditions:

Unit 3 is operating at 95% power.

ICW Header pressure is 8 psig on Pl-3-1619 and Pl-3-1620.

3A 1 and 3A2 Traveling Screen Differential pressures are 12. 7 inches water on R-3-2300 TWS DP Recorder.

TPCW Discharge pressure is 98 psig Pl-3-1468.

CCW Header Supply pressure is 105 psig on Pl-3-640.

Which ONE of the following procedures is required to be entered?

A. 3-0NOP-019, Intake Cooling Water Malfunction B. 3-0NOP-030, Component Cooling Water Malfunction C. 3-0NOP-008, Turbine Plant Cooling Water Malfunction D. 3-0NOP-011, Screen Wash System/Intake Malfunction This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT .

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QUESTION 52 Given the following conditions:

Unit 3 is at 95% power.

The TCS MVARs are oscillating.

The Switchyard voltage is stable.

Voltage Regulator Selector Switch is ON.

Subsequently:

The crew enters 3-0NOP-090, Abnormal Generator MW/MVAR Oscillation.

The U3 Turbine Operator reports the Minimum Excitation Module #5 light is lit.

Which ONE of the following describes the operator's response to stabilize the Main Generator?

A. Raise voltage using the Main Generator AC Voltage Regulator.

B. Raise voltage using the Main Generator DC Voltage Regulator.

C. Lower voltage using the Main Generator AC Voltage Regulator.

D. Lower voltage using the Main Generator DC Voltage Regulator.

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QUESTION 53 Given the following conditions:

3-EOP-ECA-1 .2, LOCA Outside Containment, has been entered.

The crew closed MOV-3-744A & B, RHR Discharge to Cold Leg Isolation Valves.

The leak is between the 3B RHR Heat Exchanger and 3-HCV-758, RHR HX Outlet Flow Control Valve.

Which ONE of the following completes the following statement?

In accordance with 3-EOP-ECA-1.2, isolation of the LOCA outside containment can be verified based on (1)

Local operator actions (2) for Alternate RHR to be available for plant cooldown.

A. (1) increasing RCS pressure (2) are required B. (1) increasing RCS pressure (2) are NOT required C. (1) decreasing Auxiliary Building radiation (2) are required D. (1) decreasing Auxiliary Building radiation (2) are NOT required This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 54 Given the following conditions:

The Reactor trips due to a loss of offsite power.

A loss of all feedwater occurs.

3-EOP-FR-H.1, Loss of Secondary Heat Sink, is in progress.

A source of feedwater is NOT restored.

The crew is establishing RCS bleed and feed when only one Pressurizer PORV opens.

Which ONE of the following correctly completes the statement below?

Based on these plant conditions, the RCS bleed path is (1) and the crew should (2) , while continuing efforts to re-establish a source of feedwater to the SGs.

A. ( 1) adequate (2) depressurize SGs to less than 360 psig B. (1) adequate (2) open all RCS Vents C. (1) inadequate (2) depressurize SGs to less than 360 psig D. (1) inadequate (2) open all RCS Vents This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 55 Given the following conditions:

Unit 3 trips from 100% power due to a 3A 4KV Bus lockout.

Post trip, a large break LOCA develops.

While performing 3-EOP-ES-1.3, Transfer To Cold Leg Recirculation, 38 RHR Pump trips and cannot be restarted.

Containment pressure peaked at 22 psig and is now 18 psig.

2 Emergency Containment Coolers are running.

RWST level is at 55,000 gallons.

Which ONE of the following completes the statements below?

The Containment Spray Pumps must be operated in accordance with {1)

{2) Containment Spray Pump(s) is/are required to be operating.

NOTE 3-EOP-FR-Z.1, Response to High Containment Pressure 3-EOP-ECA-1.1, Loss of Emergency Coolant Recirculation A. (1) 3-EOP-FR-Z.1 (2) One B. (1) 3-EOP- FR-Z.1 (2) Zero C. (1) 3-EOP-ECA-1.1 (2) One D. (1) 3-EOP-ECA-1.1 (2) Zero This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 56 Given the following conditions:

3-EOP-ECA-2.1, Uncontrolled Depressurization of All Steam Generators, is in progress.

RCS temperature decreases from 547°F to 422°F in the last hour.

The crew adjusts AFW flow.

SG NR levels are all off-scale low.

3A SG Safety Valve reseats.

Which ONE of the following describes the AFW flow requirement and the action for the next procedure transition?

A

AFW flow is at a minimum of 50 gpm per SG.

Immediately transition to 3-EOP-E-2, Faulted SG Isolation.

B.

AFW flow is at a minimum of 400 gpm per SG.

Immediately transition to 3-EOP-E-2, Faulted SG Isolation.

C.

AFW flow is at a minimum of 50 gpm per SG.

When 3A SG pressure rises, transition to 3-EOP-E-2, Faulted SG Isolation.

D.

AFW flow is at a minimum of 400 gpm per SG.

When 3A SG pressure rises, transition to 3-EOP-E-2, Faulted SG Isolation.

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QUESTION 57 Given the following conditions:

Unit 4 is at 50% steady state power.

Tavg is matched with Tref at 562PF.

Automatic VCT makeup is in progress.

TCS is in MW control at 410 MW.

The Rod Motion Control Selector Switch is placed in AUTO after moving Control Bank D to 161 steps.

The Axial Flux Difference is -3 when the Rod Motion Control Selector Switch is placed in AUTO.

Subsequently, a few minutes later:

Unit 4 is at 51 % and increasing.

Tavg is 3°F higher than Tref.

TCS is in MW control at 412 MW with Turbine Control Valves closing.

The Axial Flux Difference is -0.5 and trending more positive.

Which ONE of the following completes the following statement?

The conditions identified above are due to a(n) (1) and the applicable ONOP will subsequently (2) to restore Tavg.

A. (1) inadvertent dilution (2) adjust turbine load B. (1) continuous rod withdrawal (2) adjust turbine load C. (1) inadvertent dilution (2) insert control rods D. (1) continuous rod withdrawal (2) insert control rods This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 58 Given the following conditions:

Loss of Offsite Power occurs on Unit 3.

3A and 38 EDGs are powering the 3A and 38 4KV Buses.

3-EOP-ES-0.4, Natural Circulation Cooldown with Steam Void in Vessel (Without RVLMS) is in progress.

RCS pressure is 1635 psig.

Pressurizer level is 30%.

Prior to depressurizing the RCS, Pressurizer PORV PCV-3-456 fails open.

Which ONE of the following identifies the reason for the initial rapidly increasing Pressurizer level during this event?

A. The steam space in the Pressurizer collapses allowing more makeup to be injected immediately into the RCS by the HHSI Pumps.

B. Pressurizer level reference legs flash which results in an increase in indicated level.

C. Safety Injection Accumulators inject into the RCS which increases Pressurizer level.

D. Reactor upper head region voiding occurs which results in mass transfer from the Reactor Head to the Pressurizer.

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QUESTION 59 Given the following conditions:

Unit 3 is at 100% power.

Subsequently:

ES/3, CONDENSER LO VACUUM, alarms.

Crew enters 3-0NOP-014, Main Condenser Loss of Vacuum.

The Crew commences a Fast Load Reduction IAW 3-GOP-100, Fast Load Reduction.

Operators are evaluating Annunciator BB/1, ROD BANK LO LIMIT which has alarmed.

Which ONE of the following describes NEXT action required by 3-GOP-100?

A. Place Control Rods in Manual.

B. Borate the RCS at least 16 gpm.

C. Slow or stop the load reduction.

D. Trip the Reactor and enter 3-EOP-E-O.

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QUESTION 60 Given the following conditions:

Waste Gas Decay Tank D contains high-activity gas.

Waste Gas Decay Tank D Relief Valve develops a flange leak that is slowly dispersing into the Aux Bldg.

R-14, Plant Vent Gas Monitor, alarms.

Crew enters 3-0NOP-067, Radioactive Effluent Release.

Which ONE of the following identifies the plant response to an R-14 alarm, if any occurs?

A. Aux Bldg Exhaust Fans trip.

B. Aux Bldg Supply Fans trip.

C. No effect on Aux Bldg or Control Room Ventilation.

D. Control Room Ventilation shifts to recirculation mode.

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QUESTION 61 Given the following conditions:

A fire was confirmed in the Cable Spreading Room that was affecting plant equipment.

The crew is implementing O-ONOP-105, Control Room Evacuation.

The site has a loss of offsite power.

All emergency safeguards equipment operates as required.

Which ONE of the following identifies the EDG operation in accordance with O-ONOP-105?

At the point when Control of Shutdown Systems is established, the _ _ _ _ __

A. 3A EDG will remain loaded

8. 48 EDG will remain loaded C. 3A EDG will need to be shutdown D. 48 EDG will need to be shutdown This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 62 Given the following conditions:

Unit 3 is currently at 100% preparing for a shutdown due to high RCS Activity levels.

Subsequently:

A RCS leak of 50 gpm continues to increase inside containment.

Which ONE of the following identifies which process radiation monitors will show a continued elevated trend after an automatic safety injection actuation?

NOTE

R-3-11 Containment Air Particulate

R-3-12 Containment Air Gas

R-14 Plant Stack Gas A. R-3-11 Yes, R-3-12 Yes, R-14 Yes B. R-3-11 Yes, R-3-12 Yes, R-14 No C. R-3-11 No, R-3-12 No, R-14 Yes D. R-3-11 No, R-3-12 No, R-14 No This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 63 Given the following conditions:

Unit 3 experiences a Safety Injection.

Total AFW flow is throttled to 450 gpm.

The crew transitions from 3-EOP-E-O, Reactor Trip or Safety Injection, to 3-EOP-E-1, Loss of Reactor or Secondary Coolant.

The crew is determining SI Termination criteria with the following:

Containment temperature is 165°F and slowly decreasing.

Pressurizer level is 17% and rising.

RCS subcooling is 58°F and stable.

RCS pressure is 1550 psig and stable.

SG Levels are 5% and rising.

Which ONE of the following identifies the correct operator response and the reason?

A. Terminate SI since all criteria is satisfied.

B. Do NOT terminate SI since RCS pressure is too low.

C. Do NOT terminate SI since SG levels are too low.

D. Do NOT terminate SI since PZR level is too low.

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QUESTION 64 Given the following conditions:

Unit 3 experiences a major Steam Line Break inside containment concurrent with a Loss of Off-Site Power.

3A and 3B 4KV Buses are powered from the Emergency Diesel Generators.

Containment Pressure Hi signal is actuated.

Containment Temperature is 193°F.

The crew is performing 3-EOP-FR-P.1, Response to Imminent Pressurized Thermal Shock.

Both HHSI and RHR Pumps have been stopped and placed in AUTO.

3A Charging Pump is running with 40 gpm Charging flow.

Letdown is unavailable.

Which ONE of the following is (1) the preferred method to depressurize the plant and (2) the earliest allowable CET subcooling temperature to terminate the depressurization?

A. ( 1) Use Auxiliary Spray (2) 28°F B. (1) Use Auxiliary Spray (2) 82°F C. (1) Open one Pressurizer PORV (2) 28°F D. ( 1) Open one Pressurizer PORV (2) 82°F This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 65 Given the following conditions:

Unit 4 Reactor trips due to a loss of offsite power.

The crew performs 4-EOP-ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (With RVLMS).

Subsequently:

Train 4A RVLMS fails.

Pressurizer level is 6% and lowering rapidly.

Which ONE of the following identifies the required procedural response?

A Remain in 4-EOP-ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (With RVLMS).

B. Transition to 4-EOP-E-1, Loss of Reactor or Secondary Coolant.

C. Transition to 4-EOP-ES-0.0, Rediagnosis.

D. Transition to 4-EOP-E-O, Reactor Trip or Safety Injection.

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QUESTION 66 Which ONE of the following identifies the required reviews prior to assuming the Unit 3 RCO responsibility in accordance with O-ADM-202, Shift Relief and Turnover?

A. Check LMS for quals, review eSOMs for clearances, and review Schedule of Plant Checks and Surveillances (Red Book).

B. Review TSA log book (TCC Index), review Schedule of Plant Checks and Surveillances (Red Book), and perform a Minimum Equipment List check.

C. Check Watchstander Out of Service Book, review eSOMs for clearances, and perform a Minimum Equipment List check.

D. Check Watchstander Out of Service Book, review TSA log book (TCC Index), and review Annunciator Status Log.

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QUESTION 67 Which ONE of the following completes the following statements?

The Tech Spec minimum required Spent Fuel Pool water level is (1)

While raising the fuel assembly in the Spent Fuel Pool, the Bridge Crane hoist stops when the load limit of (2) lbs. has been reached.

A. (1) 57' (2) 2075 B. (1) 56'10" (2) 2075 C. (1) 57' (2) 4000 D. (1) 56'10" (2) 4000 This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 68 Which ONE of the following identifies the individual(s) who may enter a Guarded Area per OP-AA-102-1003, Guarded Equipment, without requesting permission from the Control Room and first completing OP-AA-102-1003-F01, Protected/Guarded Equipment Work Approval Form?

A Security Officer performing official rounds ( 1) required to obtain permission.

A Maintenance Supervisor performing a walkdown of a jobsite for an upcoming high risk activity (2) required to obtain permission.

A. (1) is not (2) is not B. (1) is not (2) is

c. (1) is (2) is not D. (1) is (2) is This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 69 Given the following plant conditions:

Unit 4 is in MODE 5.

4A RHR loop is in operation.

Which ONE of the following is a criterion for the RCS Loop Filled requirement of O-ADM-051, Outage Risk Assessment and Control?

A. 48 RCP is operating.

B. Reactor Coolant System pressure is 50 psig.

C. Unit 4 Steam Generator Wide Range levels are greater than 10%.

D. Unit 4 Steam Generator Narrow Range levels are greater than 10%.

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PTN L-15-1 NRC EXAM This information is controlled by PTN's 2015 LOIT (L-15-1) NRC FXAMIN ATION ~ ECURI I Y AGREE MEN I .

QUESTION 70 Given the following conditions:

Unit 3 is in MODE 6 with a core onload and containment purge in progress.

The refueling crew is lowering an irradiated fuel assembly into the core, when the assembly is inadvertently dropped.

Annunciators X4/1, ARMS HI RADIATION, and H1/4, PRMS HI RADIATION, are lit.

The crew verifies that R-3-12, Containment Gas Monitor, and Rl-3-14028, Unit 3 Containment Operating Floor, are in alarm.

No other annunciators have been received.

As a result of this event:

(1) The Containment Evacuation Alarm will be -1.1L initiated.

(2) A Control Ventilation Isolation will be -1.f.L initiated.

A. ( 1) automatically (2) automatically B. (1) manually (2) automatically

c. ( 1) automatically (2) manually D. (1) manually (2) manually This information is controlled by PTN' s 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 71 Given the following conditions:

Unit 3 has just entered MODE 5 for a refueling outage.

All equipment is in a normal alignment for plant conditions.

A containment purge has been initiated using the Unit 4 Purge Supply and Exhaust Fans.

Subsequently:

Containment Radiation Monitor R-3-11, Containment Air Particulate Monitor, alarms.

The crew enters 3-0NOP-067, Radioactive Effluent Release Which ONE of the following identifies the required actions in accordance with 3-0NOP-067?

A. (1) Check that Unit 4 Purge Supply and Exhaust Fans automatically trip (2) Verify that TS-002, TSC EMER VENT AUTO INITIATE Switch is in INHIBIT B. (1) Manually stop the U4 Purge Supply and Exhaust Fans (2) Verify that TS-002, TSC EMER VENT AUTO INITIATE Switch is in ENABLE C. (1) Check that Unit 4 Purge Supply and Exhaust Fans automatically trip (2) Verify that TS-002, TSC EMER VENT AUTO INITIATE Switch is in ENABLE D. (1) Manually stop the U4 Purge Supply and Exhaust Fans (2) Verify that TS-002, TSC EMER VENT AUTO INITIATE Switch is in INHIBIT This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 72 Given the following conditions:

Unit 3 RCS temperature is 150°F.

Time to boil in the RCS of 35 minutes.

38 RHR pump is Inoperable.

Subsequently,

The 3A RHR pump trips and cannot be restarted.

The crew enters 3-0NOP-050, Loss of RHR.

Which ONE of the following completes the statements below?

Containment Closure is required to be completed in (1) minutes, in accordance with 3-0NOP-050.

The equipment hatch is required to be closed with (2) bolts, in accordance with O-ADM-051, Outage Risk Assessment and Control.

A. (1) 30 (2) four B. (1) 30 (2) eight

c. (1) 35 (2) four D. (1) 35 (2) eight This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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QUESTION 73 Given the following conditions:

Alignment is complete for Unit 3 hot leg recirculation in accordance with 3-EOP-ES-1.4, Transfer to Hot Leg Recirculation.

Plant conditions are stabilized for the accident in progress.

The highest Critical Safety Function Status Tree (CSFST) is a yellow path on Inventory.

Which ONE of the following identifies the CSFST monitoring requirement in accordance with 3-EOP-F-O, Critical Safety Function Status Trees?

A. Monitor continuously.

B. Monitored every 10 to 20 minutes.

C. Suspend monitoring since conditions are stable.

D. Suspend monitoring immediately after completing 3-EOP-ES-1.4.

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QUESTION 74 Given the following conditions:

Unit 3 is at 100% power.

Work is ongoing in the Unit 3 480V Load Center Rooms.

The foreman requests the fire door between A & B and C & D Load Centers be opened to allow better air circulation for the worker's comfort.

A Fire Protection Impairment has NOT been issued.

In response to the foreman's request, which ONE of the following identifies the policy concerning the propping open of fire doors in accordance with O-ADM-016, Fire Protection Program?

A. May be open without compensatory actions as long as work is ongoing.

B. May be open if an hourly roving fire watch is provided.

C. Cannot be opened solely for comfort of personnel.

D. Cannot be opened for more than 30 minutes.

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QUESTION 75 Given the following conditions:

Unit 3 is at 100% power.

Subsequently:

Unit 3 experiences an accident with the following conditions:

Pressure AFW Flow S/GA 950 psig and lowering 260 gpm S/GB 290 psig and lowering 340 gpm S/GC 940 psig and lowering 230 gpm NR Level WR Level S/GA 21 .5 % and lowering 55.8% and lowering S/GB 80.3 % and rising 55.9% and rising S/GC 18.6 % and lowering 54.5% and lowering Containment pressure: 21.5 psig and rising Containment Sump level: 36.2 inches and rising Pressurizer pressure: 1274 psig and lowering Pressurizer level: 1 % and lowering All safeguards equipment is operating as expected.

Based on the indicated parameters, which ONE of the following identifies the initial transition from 3-EOP-E-O,Reactor Trip or Safety Injection?

A. 3-EOP-FR-Z.1, Response To High Containment Pressure

8. 3-EOP-E-1, Loss of Reactor or Secondary Coolant C. 3-EOP-E-2, Faulted Steam Generator D. 3-EOP-E-3, Steam Generator Tube Rupture This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

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L-15-1 NRC EXAM rhii. information is controlled b) PT'I:'~ 2015 LOl"I (1. 1) \l~ C E\.A \II:\ \TIO:\ SECI IUn \GIU. E\IE:-.T.

Site-Specific Written Examination STUDENT REFERENCES U.S. Nuclear Regulatory Commission Site-Specific Written Examination This information is controlled by PT'l:'s 2015 LOIT (L-15-1) ~ RC EXA \-11:\ATIO;\ SECl lRIT\ ,\GREEME:-.T.

PRESSURIZER SPRAY. LOOP B 0

CLOSE

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Atmosphere Gaseous or Particulate Radioactivity Monitoring System, and

b. A Containment Sump Level Monitoring System.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION :

a. With both the Particulate and Gaseous Radioactivity Monitoring Systems inoperable, operation may continue for up to 7 days provided:

1) A Containment Sump Level Monitoring System is OPERABLE;

2) Appropriate grab samples are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;

3) A Reactor Coolant System water inventory balance is performed at least once per s*

hours except when operating in shutdown cooling mode; and

4) Containment Purge, Exhaust and Instrument Air Bleed valves are maintained closed.**

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With no Containment Sump Level Monitoring System operable, restore at least one Containment Sump Level Monitoring System to OPERABLE status within 7 days, or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection System shall be demonstrated OPERABLE by:

a. Containment Atmosphere Gaseous and Particulate Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and

b. Containment Sump Level Monitoring System-performance of CHANNEL CALIBRATION at least once per 18 months.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

- Instrument Air Bleed valves may be opened intermittently under administrative controls.

TURKEY POINT - UNITS 3 & 4 3/4 4-13 AMENDMENT NOS. 260 AND 255 I

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATING 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary-to-secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4-1 up to a maximum of 5 GPM at a Reactor Coolant System pressure of 2235 +/- 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 .*

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or with primary-to-secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding primary-to-secondary leakage, PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than allowed by 3.4.6.2.e above operation may continue provided:

1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months verify that at least two valves in each high pressure line having a non-functional valve are in, and remain in that mode corresponding to the isolated condition, i.e., manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supplies deenergized. Follow applicable ACTION statement for the affected system, and

Test pressure less than 2235 psig are allowed. Minimum differential test pressure shall not be less than 150 psid. Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one-half power.

TURKEY POINT - UNITS 3 & 4 3/4 4-14 AMENDMENT NOS. 260 AND 255 I

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION (Continued)

2. The leakage* from the remaining isolating valves in each high pressure line having a valve not meeting the criteria of Table 3.4-1, as listed in Table 3.4-1, shall be determined and recorded daily. The positions of the other valves located in the high pressure line having the leaking valve shall be recorded daily unless they are manual valves located inside containment.

Otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With any Reactor Coolant System Pressure Isolation Valve leakage greater than 5 gpm, reduce leakage to below 5 gpm within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous or particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the containment sump level at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.** Performance of a Reactor Coolant System water inventory balance at least once per 72***

hours; and

d. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; and

e. Verifying primary-to-secondary leakage is ~ 150 gallons per day through any one SG at least once per 72*** hours.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage* to be within its limit:

a. At least once per 18 months.

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months, and

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

- Not applicable to primary-to-secondary leakage.

- - Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

TURKEY POINT - UNITS 3 & 4 3/4 4-15 AMENDMENT NOS. 260 AND 255 I

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION <Continued)

d. Following valve actuation due to automatic or manual action or flow through the valve :

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying valve closure, and

2. Prior to entering Mode 2 by verifying leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

TURKEY POINT - UNITS 3 & 4 3/4 4-16 AMENDMENT NOS. 260 AND 255 I

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION High-Head Safety Unit 3 Unit 4 Injection Check Valves 3-874A 4-874A Loop A, hot leg 3-875A 4-875A cold leg 3-873A 4-873A cold leg 3-874B 4-874B Loop B, hot leg 3-875B 4-875B cold leg 3-873B 4-873B cold leg 3-875C 4-875C Loop C,cold leg 3-873C 4-873C cold leg Residual Heat Removal Line Check Valves 3-876A 4-876A Loop A, cold leg 4-876E 3-876B 4-876B Loop B, cold leg 3-8760 4-8760 3-876C 4-876C Loop C, cold leg 3-876E MOV4-750 Loop A, MOV4-751 hot leg to RHR MOV3-750 Loop C, MOV3-751 hot leg to RHR ACCEPTABLE LEAKAGE LIMITS

1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable provided that the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between previously measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between previously measured leakage rate and the maximum permissible rate of 5.0 gpm by 50%

or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

TURKEY POINT- UNITS 3 & 4 3/4 4-17 AMENDMENT NOS. 260 AND 255 I

L-15-1 NRC EXAM This information is controlled by PTN's 2015 LOIT (L-15-1) NRC EXAMINATION SECl' RITY AGREEME NT.

P~MARYJ PTN UNIT3 RCP DETAILED DATA

SUMMARY

I ffi V

THERMAL BARRIER RELATED DIBPLAYB AP Bt1111 RCPA

. .INH20 RCPB

. .INH20

~

RCPC l l l l llNH20 SECONDARY I BEAL INJECTION FLOW NUMB&R ONE BEAL LIAKOFF

" ' J " GPM F T W GPM F " "I GPM FLOW -

l - GPM -

l. . . GPM - . l l ! i - GPM POWER I THERMAL BARRIER COOLING WAT1iR LO FLOW ALARM PUMP BEARING

- - l'*'* ***l'1'* ***l.Tll*

T1iMP&RATUREB . . . DEGF l l l l lDEGF . . .DEGF TURBINE I NUMBER ONE BEAL L&AKOFF T1iMP&RATURE . .DEGF . .D&GF llll1DEOF YCT TEMPERATURE . .DEGF . .DEGF . .DEGF ESF I UPPER OIL RESERVOIR HI/LO LEY&LALARM

-~[o];:l.U!1* -~[*l;:UiT!1- ~l~

LOWER OIL RESERVOIR HI/LO -~~~T!1- ~i'iT!1- ~i'iT!1-LEY&LALARM SUPPORT SYS UPPER THRUST BEARING TEMP - DEGF . .DEGF . . .DEGF LOWER THRUST BEARING TEMP l l l l DEGF . .DEGF l l f l l 'oEGF UPPER GUIDE BEARING TEMP . . .DEGF l l l lDEGF - D&GF LOWER GU10E BEARING TEMP . . . DIGF . . .DEGF D l lDEGF UTILITIES MOTOR BEARING COOLING l~[*l~i'iT!1" IOO;:Ti'i1!1" lmiilliT!1" WATER HI TEMP ALARM

,-~!] MOTOR BEARING COOUNC WATER FLOW I E l l GPM l ! l l l OPM . . . OPM STATOR WINDING T1iMP l l l l DEGF llllDEOF - DEOF I P~NT I l~[*lmiT!1" IM*l;:lM!1" ~[o];:li'.T!1" I P~O~S I

Procedure No.: Procedure Ti lie: Page:

10 Approval Date:

O-ADM-115 Notification of Plant Events 8/1/12 ENCLOSURE I (Page 1 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit NOTE: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements.

Deviation from Tech Specs allowed by l 0 One Hour Perform actions immediately necessary to protect public health and safety.

CFR 50.54(x) and 10 CFR 72.32(d) [NUREG-1022] 10 CFR 50.72(b)(l)

Initiation of any nuclear plant shutdown Four Hours "Initiation of any Nuclear Plant Shutdown" is the performance of any action to start reducing required by Tech Specs. reactor power to achieve an operational condition or mode that requires the reactor to be subcritical as a result of a Tech Spec requirement. This includes any means of power reduction, such as control rod insertion, boration, or turbine load reduction [NUREG-1022].

Examples are exceeding an LCO action statement, Tech Spec 3.0.3, Safety Limit violation

[10 CFR 50.36].

10 CFR 50.72(b)(2)(i)

An event that results or should have Four Hours Actuation from a pre-planned sequence during testing or reactor operation is not reportable.

resulted in Emergency Core Cooling [NUREG-1022] 10 CFR 50.72(b)(2)(iv)(A)

System discharge into the RCS as a result of a valid signal.

An event that results in actuation of the Four Hours Manual or automatic RPS actuation not part of a pre-planned sequence is reportable. Actuation Reactor Protection System (RPS) when the from a pre-planned sequence during testing or reactor operation is not reportable. Turbine reactor is critical. runbacks are not part of RPS and therefore not reportable.

[NUREG-1022] 10 CFR 50.72(b)(2)(iv)(B)

One startup transformer and one of the Four Hours required Emergency Diesel Generators Tech Spec 3.8. l.l.

inoperable.

Both startup transformers inoperable. Four Hours Tech Spec 3.8. l.l.

Both standby feedwater pumps inoperable Four Hours Tech Spec 3.7.1.6.

for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W2002/AWT/cls/mr/mr

Procedure No.: Procedure Ti tie: Page:

11 Approval Date:

O-ADM-115 Notification of Plant Events 4/23/13 ENCLOSURE!

(Page 2 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit Note: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements, and O-ADM-117, Equipment Important to Emergency Preparedness (Commitment Step 2.3.5).

Loss of Emergency Assessment Capability I Eight Hours ILoss of Emergency Assessment Capability: A major loss of emergency assessments capability includes those events that significantly impair the safety assessment capability of the licensee. Some engineering judgment is needed to determine the significance of the loss of particular equipment; e.g., loss of only the SPDS for a short period of time may not be reportable, but concurrent loss of SPDS and other assessment equipment may be reportable. Loss of significant portion of control room indication including annunciators or monitor or loss of all plant vent stack radiation monitors are examples of a major loss of emergency assessment capability which should be evaluated for reportability.

Emergency response facilities (ERFs) include the Technical Support Center (TSC) and Emergency Operations Facility (EOF) only. The Operational Support Center (OSC) is not included because it requires no special design criteria that would limit the use of an alternate location.

Examples:

Reportable A. Loss of a significant portion of control room safety indication (loss as delineated in the EALS) including annunciators or monitors. An 8-hour notification is required.

B. Loss of all plant vent stack radiation monitors. An 8-hour notification is required.

C. Loss of Technical Support Center ventilation capability to support habitability [if not restorable]

within the time required for activation. An 8-hour notification is required. (Note: Licensing shall be contacted promptly upon discovery for notification guidance.)

D. Loss of all power to the Technical Support Center if not restorable within the time required for activation. An 8-hour notification is required.

W2002/AWT/cls/mr/mr

Procedure No.: Procedure

Title:

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12 Approval Date:

O-ADM-115 Notification of Plant Events 4/2/13 ENCLOSUREl (Page 3 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit Note: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements, and O-ADM-117, Equipment Important to Emergency Preparedness (Commitment Step 2.3.5).

Loss of Emergency Assessment Capability IEight Hours I E. The complete loss of ERDADS/DCS data concurrent with the loss of other major accident assessment indication (indicators, annunciators, monitors, etc.).

F. The complete loss of ERDADS/DCS for greater than one hour.

Not Reportable A. Unavailability of one redundant component or train such as a radiation monitor for a period of time as permitted by the Technical Specifications generally is not reportable. If a redundant means to assess a key plant parameter for emergency classification is available in the event a monitor becomes unavailable, then the condition should not be reportable.

B. Loss of backup or normal power to the Technical Support Center but not both.

C. If an emergency assessment system or component is governed by Technical Specifications, ODCM, or UFSAR requirement, and operation is pennitted within the LCO Action Statement, then there is NOT a significant loss of assessment capability (e.g. - accident monitoring equipment, radiation monitors, meteorological equipment, etc.) If the governing document requires NRC notification or plant shutdown due to unavailable assessment instrumentation, then a condition would be considered a significant loss of assessment capability and would be reportable.

D. The loss of all meteorological tower data points, where the National Weather Service can provide the required information, would NOT constitute a significant loss of assessment capability.

[NUREG-1022] 10 CFR 50.72(b){3)(xiii)

W2002/AWT/cls/mr/mr

Procedure No.: Procedure Ti tie: Page:

13 Approval Date:

O-ADM-115 Notification of Plant Events 8/1/12 ENCLOSUREl (Page 4 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit Note: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements, and O-ADM-117, Equipment Important to Emergency Prepar(!dness (C_o_mrnitment S~ep 2.3.5).

Loss of Offsite Response Capability I Eight Hours I A major loss of offsite response capability includes those events that would significantly impair the fulfillment of the Emergency Plan. Loss of offsite response capability may typically include the loss of plant access, emergency offsite response facilities, or public prompt notification system (a loss of more than 25% of the station's total sirens, other alerting systems (e.g., tone alert radios,), or more importantly, the lost capability to alert a large segment of the population (for more than l hour) would be considered a major loss and, therefore, reportable per this section).

Note that performing maintenance on an offsite emergency response facility is not reportable if the facility can be returned to service promptly in the event of an accident.

Examples:

Reportable A. Loss of more than 25% of the station's sirens for any reason. Power Systems will notify the Control Room and EP for any siren system degradation that is reportable. An 8-hour notification is required.

B. Offsite authorities have notified the station that plant access is severely limited impacting the ability to timely augment the emergency response facilities in the required timeframe. The station had 2 full shift crews on site to support plant operations and no emergency declaration was made. An ENS call is required because the road closing may prevent the plant staff from adequately staffing the TSC, or from fully responding to some emergencies.

C. Loss of all power to the Emergency Operations Facility if not restorable within the time required for activation. An 8-hour notification is required.

W2002JAWT/cls/mr/mr

Procedure No.: Procedure

Title:

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14 Approval Date:

O-ADM-115 Notification of Plant Events 4/2/13 ENCLOSURE I (Page 5 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit Note: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements, and O-ADM-117, Equipment Important to Emergency Preparedness (Commitment Step 2.3.5).

Loss of Offsite Response Capability I Eight Hours I Not Reportable A. It was observed during siren testing that 5 of 52 alert sirens around the EPZ failed to function .

This was not considered to be a major loss of the offsite response capability since it was significantly less than 25% of the alert sirens.

B. Miami-Dade County was performing a scheduled quarterly full cycle siren test and as they were performing the procedure there was a step requiring the turning of a key in order to make the sirens sound. The sirens did not sound, however, within minutes the individual realized an improper arming configuration, rearmed the siren, turned the key, and the sirens functioned properly. This event would not be reportable because the sirens were functional the entire time because the sirens were still able to alert the public.

C. Loss of backup or normal power to the Emergency Operations Facility but not both.

D. Loss of the NRC ERDS Link is not reportable if the ENS phone is available.

[NUREG-1022] IO CFR 50.72(b)(3)(xiii)

W2002/AWT/cls/mr/mr

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15 Approval Date:

O-ADM-115 Notification of Plant Events 4/2/13 ENCLOSURE 1 (Page 6 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit Note: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements, and O-ADM-117, Equipment Important to Emergency Preparedness (Commitment Step 2.3.5).

Loss of Communications Capability I Eight Hours I A major loss of communications capability would include the loss of ENS and commercial telephone lines. Refer to O-ADM-117 for additional information.

Examples:

Reportable A. ENS and commercial phone lines were discovered to have been cut such that all offsite notification was impacted while crews were digging. This would require an 8-hour notification.

B. Significant degradation of the Alert and Notification System (siren availability less than 90 percent for 12 consecutive months, or less than 75 percent for 2 weeks).

[Commitment - Steps 2.3.l and 2.3.2.J C. Loss of primary and all backup State communications channels.

W2002/AWT/cls/mr/mr

Procedure No.: Procedure

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16 Approval Date:

O-ADM-115 Notification of Plant Events 4/2/13 ENCLOSURE 1 (Page 7 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit Note: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements, and O-ADM-117, Equipment Important to Emergency Preparedness (Commitment Step 2.3 .5).

Loss of Communications Capability I Eight Hours I Not Reportable A. The Hot Ring Down phone in the control room was discovered to be unable to contact the State warning point. Other communication channels were verified to be available.

B. A loss of the EROS function/link while still maintaining ENS does NOT mandate an eight hour NRC notification for major loss of offsite communication capability.

C. If an EROS failure is determined to be in NRC maintained equipment, then the EROS help desk should be notified of the outage so that the NRC can arrange for repair.

D. If either or both of the emergency communications subsystems (ENS and HPN) fail, the NRC Operations Center should be informed of the failure over normal commercial telephone lines.

When notifying the NRC Operations Center, use the backup commercial telephone numbers provided. If the NRC Operations Center notifies Turkey Point that an ENS line is inoperable, there is no need for a subsequent notification by Turkey Point. Loss of either ENS or HPN does not necessitate an event report. The NRC Operations Center contacts the appropriate repair organization.

[NUREG-1022) 10 CFR 50.72(b)(3)(xiii)

Steam Generator Tube Inspection Eight Hours Results in Category C-3 Tech Spec 4.4.5.5 .c W2002/AWT/cls/mr/mr

Procedure No.: Procedure Ti tie: Page:

17 Approval Date:

O-ADM-115 Notification of Plant Events 8/1/12 ENCLOSURE!

(Page 8 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit NOTE: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, IO CFR Regulatory Reportaoility-Requirements.

Any event or condition that at the time of I Eight Hours Systems with operable redundant trains do not meet this criteria.

discovery could have prevented the fulfillment of the safety function of structures or systems necessary to:

Shut down the reactor and maintain it in a safe shutdown condition

Remove residual heat

Control the release of radioactive [NUREG-1022] IO CFR 50.72(b)(3)(v)(A-B-C-D) material

Mitigate the consequences of an accident Any event or condition that results in valid I Eight Hours Valid ESF actuations are those that result from valid signals or from intentional manual actuation of any of the systems listed initiation. Actuation from a pre-planned sequence during testing or reactor operation is not below: reportable.

Reactor Protection System

General Containment Isolation Valid signals are those signals, both automatic and manual, that are initiated in response to actual signals affecting containment plant conditions or parameters satisfying the requirements for initiation of the safety function of isolation valves in more than one the system. This includes those initiated prior to criticality.

system or multiple MSIVs

ECCS including HHSI and RHR RPS actuations that occur when the reactor is not critical should not be referred to as reactor trips

AFW in event notifications.

Containment heat removal and depressurization systems, including containment spray and fan cooler [NUREG-1022] 10 CFR 50.72(b)(3)(iv)(A) systems (Ctmt Spray, ECCs, ECFs)

EDGs W2002/AWT/cls/mr/mr

Procedure No. : Procedure

Title:

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18 Approval Date:

O-ADM-115 Notification of Plant Events 8/1/12 ENCLOSURE 1 (Page 9 of 13)

NRC NOTIFICATION TABLE Notification Plant Condition or Event Notes Time Limit NOTE: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements.

An event or condition, during operating or I Eight Hours Examples are:

shutdown, that results in:

Fuel cladding failures in the reactor or in the storage pool that are unique or wide spread,

The nuclear plant, including its or caused by unexpected factors, and cause a significant release of fission products.

principal safety barriers, being

Cracks or breaks in reactor coolant piping, reactor vessel, or major components in the seriously degraded. primary coolant circuit that have safety relevance.

The nuclear power plant being in an

Significant welding or material defects in the RCS.

unanalyzed condition that

Low temperature over pressure transients where the pressure-temperature relationship significantly degrades plant safety. violates the Tech Spec pressure-temperature curve. Consider heat-up and cool-down rates or OMS actuation.

Loss of containment function or integrity, including leak-rate test results that exceed Tech Spec limits. Reference Containment Leakage Rate Testing Program procedure and associated surveillance procedures.

Loss of pressurizer safety valves function during operation

[NUREG-1022] 10 CFR 50.72(b)(3)(ii)(A&B)

Spent fuel pool water level below 56 ft 10 I 24 Hours Reference Tech Spec 3.9.11 in for 7 days W2002/AWT/cls/mr/mr

Procedure No.: Procedure Ti tie: Page:

19 Approval Date:

O-ADM-115 Notification of Plant Events 8/1/12 ENCLOSURE 1 (Page 10 of 13)

NRC NOTIFICATION TABLE Notification Radioactive Materials Event Notes Time Limit NOTE: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, IO CFR Regulatory Reportability Requirements.

Any incident involving radioactive materials which may have caused or threatens to cause: Immediate

An individual to receive a total effective dose equivalent of 25 rems or more, OR an eye dose equivalent of 75 rems or more, OR a shallow-dose equivalent to the skin or extremities of 250 rads or more.

The release of radioactive material such that, had an individual been present for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the individual could have received an intake 5 times the occupational limit on intake (not applicable to spaces not normally occupied). 10 CFR 20.2202(a)(1&2)

Receipt of any package that is found to contain: Immediate

Removable radioactive surface contamination in excess of 22 dpm/cm2 ; OR

In excess of 200 millirem/hour on the accessible external surface [or 1000 millirem/hour under certain conditions specified in IO CFR 71.47(a)], OR

In excess of 200 millirem/hour at any point on the outer surface of the vehicle; OR IOCFR 71.87, IOCFR 71.47,

In excess of 10 millirem/hour at any point two meters from the vehicle; OR IO CFR 20.1906(d)

Two millirem/hour in any normally occupied positions of the vehicle .

Any lost, stolen, or missing licensed material in an aggregate quantity equal to or greater than Immediate 1,000 times the quantity specified in Appendix C to 10 CFR 20. IO CFR 20.2201(a)(l)(i)

Any event requiring the transport of a radioactively contaminated person to an offsite medical Eight Hours facility for treatment. [NUREG-1022] IO CFR 50.72(b)(3)(xii)

Any incident involving radioactive materials which may have caused or threatens to cause: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

An individual to receive, in a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a total effective dose equivalent of greater than 5 rems, OR an eye dose equivalent of greater than 15 rems, OR a shallow dose equivalent to the skin or extremities of greater than 50 rems.

The release of radioactive material such that, had an individual been present for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the individual could have received an intake greater than the occupational annual limit on intake (not aoolicable to spaces not normally occupied). 10 CFR 20.2202(b)(1&2)

Lost, stolen, or missing licensed material in an aggregate quantity greater than I 0 times the 30 days quantity specified in Appendix C to l 0 CFR 20, that is still missing. 10 CFR 20.220l(a)(l)(ii)

W2002/AWT/cls/mr/mr

Procedure No.: Procedure Ti tie: Page:

20 Approval Date:

O-ADM-115 Notification of Plant Events 8/1/12 ENCLOSURE I (Page 11 of 13)

NRC NOTIFICATION TABLE Security I Noteworthy I Employee Notification Notes Events Time Limit NOTE: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, l 0 CFR Regulatory Reportability Requirements.

Security related events requiring one I One hour The Security Manager or designee will inform the Shift Manager of notifications made and actions hour reporting to the NRCOC are taken. The Shift Manager will notify the NRCOC of security related events requiring one hour contained in SY-AA-102-1017, notifications.

Safeguards Event Reporting. 10 CFR 73.7l(a and b)

Any event or situation, related to the I Four hours Examples of events reportable under this criterion include:

health and safety of the public or on-site

Release of radioactively contaminated tools or equipment to public areas.

personnel, or protection of the

Unusual or abnormal releases of radioactive effluents.

environment, for which a news release is

On-site fatality or injury to three or more individuals all requiring in-patient hospitalization.

planned or notification to other OSHA regulations (29 CFR 1904.39) require reporting onsite fatalities or the death of an government agencies has been or will be employee from a work-related incident, or the in-patient hospitalization of 3 or more employees made (excludes contaminated as a result of a work-related incident, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The OSHA phone number is listed in the groundwater incidents as described in Emergency Response Directory (Government Organizations Section).

Attachment 5). Such an event may include an onsite fatality or inadvertent

Assure the Duty Safety Supervisor is aware of this OSHA notification requirement.

release radioactively contaminated

Assure the Shift Manager follows up with an NRC notification .

materials.

Notification to a government agency.

If the notification is for an event or situation not related to the health and safety of the public or on-site personnel or not related to protection of the environment, it should be reported to the NRC, but classified on the Event Notification Worksheet, Form 443, as Information Only, not as a 10 CFR 50.72(b)(2)(xi) report.

[NUREG-1022] 10 CFR 50.72(b)(2)(xi)

W2002/AWT/cls/mr/mr

Procedure No.: Procedure Tille: Page:

21 Approval Date:

O-ADM-115 Notification of Plant Events 8/1/12 ENCLOSURE 1 (Page 12 of 13)

NRC NOTIFICATION TABLE Security I Noteworthy I Employee Notification Notes Events Time Limit NOTE: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements.

Fitness For Duty Event 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Examples are:

The use, sale, distribution, possession, or presence of illegal drugs, or the consumption or presence of alcohol within a protected area;

Any acts by any person licensed under I 0 CFR parts 52 and/or 55 to operate a power reactor, as well as any acts by SSNM transporters, FFD program personnel, or any supervisory personnel who are authorized under this part, if such acts -

(i) Involve the use, sale, or possession of a controlled substance; (ii) Result in a determination that the individual has violated the licensee's or other entity's FFD policy (including subversion as defined in § 26.5); or (iii) Involve the consumption of alcohol within a protected area or while performing the duties that require the individual to be subject to the FFD program;

Any intentional act that casts doubt on the integrity of the FFD program; and

Any programmatic failure, degradation, or discovered vulnerability of the FFD program that may permit undetected drug or alcohol use or abuse by individuals within a protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program.

If a false positive error occurs on a blind performance test sample submitted to an HHS-certified laboratory, the* licensee or other entity shall notify the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after discovery of the error.

If a false negative error occurs on a quality assurance check of validity screening tests, as required in§ 26.137(b), the licensee or other entity shall notify the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after discovery of the error.

10 CFR26.719, FFD-7 W2002/AWT/cls/mr/mr

Procedure No.: Procedure Ti tie: Page:

22 Approval Date:

O-ADM-115 Notification of Plant Events 8/1/12 ENCLOSUREl (Page 13 of 13)

NRC NOTIFICATION TABLE Followup Notification Notification Notes Time Limit NOTE: For additional details/guidance on NRC reporting requirements refer to O-ADM-560, 10 CFR Regulatory Reportability Requirements.

In addition to the initial notification, the Shift I One hour The NRC may request an open channel for followup notifications.

Manager shall make followup notifications for the following conditions: The time frame for followup reports will be based on NRC recommendations made during

Further degradation in the level of safety of the initial report and on the judgment of the Shift Manager as to the seriousness of the the plant or other worsening plant condition. event. All followup reports shall be made within I hour.

The results of ensuing evaluations or assessments of plant conditions.

The effectiveness of response or protective 10 CFR 50.72(c)(l)(i), 50.72(c)(2)(i),

measures taken. [NUREG-1022] 50.72(c) (2)(ii), 50.72(c) (2)(iii), 50.72(c)(3).

Any information related to plant behavior that is not understood.

W2002/AWT/cls/mr/mr

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Two independent auxiliary feedwater trains including 3 pumps as specified in Table 3.7-3 and associated flowpaths shall be OPERABLE.

APPLICA81LITY: MODES 1, 2 and 3 ACTION:

1) With one of the two required independent auxiliary feedwater trains inoperable, either restore the inoperable train to an OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or place the affected unit(s) in at least HOT STANDBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s* and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2) With both required auxiliary feedwater trains inoperable, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either restore both trains to an OPERABLE status, or restore one train to an OPERABLE status and follow ACTION statement 1 above for the other train. If neither train can be restored to an OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , verify the OPERABILITY of both standby feed-water pumps and place the affected unit(s) in at least HOT STANDBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s* and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Otherwise, initiate corrective action to restore at least one auxiliary feedwater train to an OPERABLE status as soon as possible and follow ACTION statement 1 above for the other train.

3) With a single auxiliary feedwater pump inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , verify OPERABILITY of two independent auxiliary feedwater trains, or follow ACTION statements 1 or 2 above as applicable.

Upon verification of the OPERABILITY of two independent auxiliary feedwater trains, restore the inoperable auxiliary feedwater pump to an OPERABLE status within 30 days, or place the operating unit(s) in at least HOT STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s* and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The provisions of Specification 3.0.4 are not applicable during the 30 day period for the inoperable auxiliary feedwater pump.

SURVEILLANCE REQUIREMENTS 4.7.1 .2.1 The required independent auxiliary feedwater trains shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1) Verifying by control panel indication and visual observation of equipment that each steam turbine-driven pump operates for 15 minutes or greater and develops a flow of greater than or

If this ACTION applies to both units simultaneously, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

TURKEY POINT - UNITS 3 & 4 3/4 7-3 AMENDMENT NOS. 137 AND 132

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS <Continued) equal to 373 gpm to the entrance of the steam generators. The provisions of Specification 4 .0 .4 are not applicable for entry into MODES 2 and 3;

2) Verifying by control panel indication and visual observation of equipment that the auxiliary feedwater discharge valves and the steam supply and turbine pressure valves operate as required to deliver the required flow during the pump performance test above;

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; and

4) Verifying that power is available to those components which require power for flow path operability.

b. At least once per 18 months by:

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of each Auxiliary Feedwater Actuation test signal , and

2) Verifying that each auxiliary feedwater pump receives a start signal as designed automatically upon receipt of each Auxiliary Feedwater Actuation test signal.

4 .7.1 .2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 1 by verifying normal flow to each steam generator.

TURKEY POINT - UNITS 3 & 4 3/4 7-4 AMENDMENT NOS. 137 AND 132

TABLE 3.7-3 AUXILIARY FEEDWATER SYSTEM OPERABILITY UNIT TRAIN STEAM SUPPLY FLOWPATH(3 l PUMP DISCHARGE WATER FLOWPATH(3l SG 3A via CV-3-2816 SG 3C via MOV-3-1405 A or C( 2l 3 1 SG 3B via CV-3-2817 or SG 3B via MOV-3-1404< l SG 3C via CV-3-2818 SG 3A via CV-3-2831 SG 3A via MOV-3-1403 B or C( 2l 3 2 1 SG 3B via CV-3-2832 or SG 3B via MOV-3-1404( )

SG 3C via CV-3-2833 SG 4A via CV-4-2816 SG 4C via MOV-4-1405 A or C(2l 4 1 1 SG 4B via CV-4-2817 or SG 4B via MOV-4-1404( >

SG 4C via CV-4-2818 SG 4A via CV-4-2831 SG 4A via MOV-4-1403 B or C( 2 >

4 2 1 SG 4B via CV-4-2832 or SG 4B via MOV-4-1404( >

SG 4C via CV-4-2833 NOTES:

(11Steam admission valves MOV-3-1404 and MOV-4-1404 can be aligned to either train (but not both) to restore OPERABILITY in the event MOV-3-1403 or MOV-3-1405, or MOV-4-1403 or MOV-4-1405 are inoperable.

2

( lDuring single and two unit operation, one pump shall be OPERABLE in each train and the third auxiliary feedwater pump shall be OPERABLE and capable of being powered from, and supplying water to either train, except as noted in ACTION 3 of Technical Specification 3. 7.1.2. The third auxiliary feedwater pump (normally the "C" pump) can be aligned to either train to restore OPERABILITY in the event one of the required pumps is inoperable.

3

( l1f any local manual realignment of valves is required when operating the auxiliary feedwater pumps, a dedicated individual, who is in communication with the control room , shall be stationed at the auxiliary feedwater pump area. Upon instructions from the control room, this operator would realign the valves in the AFW system train to its normal operational alignment.

TURKEY POINT - UNITS 3 & 4 3/4 7-5 AMENDMENT NOS. 137 AND 132

Turkey Point 2015-301 Reference Package EALs were also part of the reference package as indicated on the exam question documentation. They are only referenced here due to potentially being considered SUNSI.

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