L-2003-317, SRO Written License Examination Comments

SRO Written License Examination Comments

ML033650387
Person / Time
Site:	Turkey Point
Issue date:	12/18/2003
From:	Jones T Florida Power & Light Co
To:	Ernstes M, Reyes L Region 2 Administrator
References
L-2003-317
Download: ML033650387 (75)

Text

FPL DEC 1 8 2003 L-2003-317 Mr. Luis A. Reyes Regional Administrator, Region II Attn: Michael E. Ernstes U. S. Nuclear Regulatory Commission Sam Nunn Atlanta Federal Center 61 Forsyth Street, S. W., Suite 23T85 Atlanta, GA 30303 Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 SRO Written License Examination Comments The provisions of NUREG-1021, Operator Licensing Examiner Standards, Examiner Standards ES-402, Administering Initial Written Examinations, allow the opportunity for submittal of comments on the written portion of the SRO License Examination to the NRC.

This letter documents that Florida Power and Light Co. is submitting comments and recommendations for your review and approval for questions #1 and # 46 related to the site-specific written examination administered at Turkey Point on December 15, 2003. Additionally, question #33 is being submitted for consideration for applicant Timothy Scott. Mr. Scott inadvertently bubbled in the wrong item on his scantron answer sheet. The supporting information for these questions is enclosed.

Should there be any questions, please contact Gregory Laughlin at (305) 246-6274.

Very truly yours, Terry Jones Vice President Turkey Point Nuclear Plant SM cc: Michael E. Ernstes, Chief, Operator Licensing and Human Performance Branch, Region II, USNRC Chief Examiner, Region I, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant Document Control Desk, USNRC, Washington, D.C.

an FPL Group company

Turkey Point 2003 NRC Written SRO Exam Post-Exam Review Recommendations December 18, 2003 Q# 1) Accept either "B" or "C" response as correct.

"C" response is correct per the answer key. It is supported by its reference: BD-ONOP-003.6, Page 6.

Post-exam review revealed that "B" response is also correct. Refer to 3-ONOP-003.6, CAUTION on Page 22 which states " Pressurizerlevel should be monitoredclosely on the operable instrumentationduringperformance of the following steps to avoid uncoveringthe pressurizerheaters or causing a high level trip."

The "following steps" referred to in the CAUTION are steps associated with maintaining pressurizer level and pressure control.

This CAUTION clearly states that operators must use care when controlling pressurizer level (via chaging pump speed control) to prevent a reactor trip on high pressurizer level. This concern forms the basis for the "B" response, making the "B" response also correct.

Note that the distractor analysis says that "pressurizer level is not a concern becausepressurewill increase to the POR V setpoint priorto pressurizerlevel trip criteriabeing reached " The distractor analysis is correct in that pressure will likely increase to 2335 psig as level increases. PORV-456 will auto-open and relieve pressure down to 2315 psig at which time it will close. The PORV will then cycle open and closed between these pressure values while level continues to rise. Level will increase to its auto-trip setpoint of 92% or until operators manually trip the reactor at 80% level as directed by their Operations Department Instruction (ODI) #23. Therefore the distractor analysis is flawed because pressurizer level remains a concern even while the PORV is cycling.

Finally the issue of identifying the ultimate basis for minimizing charging flow must be addressed. With letdown isolated as a result of the loss of 3P06, pressurizer level and pressure will continuously rise at a rate dictated by the charging flow rate. An ONOP-003.6 objective is to maintain stable conditions with near normal parameter values until 3P06 can be restored. Either event (PORV cycling or reactor trip on high level) is undesirable and the procedure gives guidance to minimize charging flow to preclude both. Therefore precluding both events becomes the ultimate basis of reducing charging flow.

References Provided:

3-ONOP-003.6, "Loss of 120V Vital Instrument Panel 3P06" 3-ONOP-003.6, Basis Document ODI-CO-023, "Manual Reactor Trip Guidelines" Simulator generated curve: PRZ. Pressure & Level following loss of 3P06

Turkey Point Nuclear Plant 2003-301 -

SRO Inital Exam 1.'-.

Which ONE of the following is the basis for reducing charging flow to the minimum required to maintain RCP Seal Injection following a loss of 120V Vital Instrument Panel 3P06, as required by Step 3.a, of 3-ONOP-003.6, "Loss of 120V Vital Instrument Panel 3P06?"

A. Reducing charging flow assures proper back pressure on the RCP # 2 seal and ensures the # 2 seal is not cocked.

B. Reducing charging flow extends the time for recovery without tripping the Reactor on high pressurizer level.

C. Minimizing the fill rate of the pressurizer extends the time for recovery without lifting a pressurizer PORV due to compressing the bubble.

D. Minimizing charging pump speed ensures that a loss of charging does not occur due to low oil pressure to ensure that RCP Seal Injection is maintained.

1. 004A1.04 1 SRO Which ONE of the following is the design basis for reducing charging flow to the minimum required to maintain RCP Seal Injection using the "3B" or "3C" charging pump in-manuai speed following a loss of 120V Vital Instrument Panel 3P06, as required by Step 3.a, of 3-ONOP-003.6, "Loss of 120V Vital Instrument Panel 3P06?"

A. Reducing charging flow manually would assures proper back pressure on the RCP

1. 2 seal and ensures the # 2 RGP seal is not cocked.

B. Reducing charging flow manually-will extends the time frame for recovering without tripping the Reactor on high pressurizer level.

C. Minimizing the fill rate of the pressurizer manualy wil extends the time frame for recovery without lifting a the pressurizer PORV due to compressing the bubble.

D. Minimizing charging pump speed manually will ensures that a loss of charging does not occur due to low oil pressure to ensure that RCP Seal Injection is maintained not lost.

Distractor Analysis:

A. Incorrect. This is the basis for having RCS pressure greater than 325 psig.

B. Incorrect. In this case pressurizer level is not a concern because pressure will increase to the PORV setpoint prior to pressurizer level trip criteria being reached.

C. Correct. The loss of 3P06 directly affects the normal control of the pressurizer pressure and level. Operator attention is necessary to maintain the pressurizer in normal level and pressure.

D. Incorrect. Operational experience at PTN has shown that Charging pumps are prone to auto trip on low oil pressure when the speed of the charging pump is reduced to low values of less than 20% demand .

Answer: C Reason for Revision: Eliminated unnecessary words. Minor wordsmithing.

Florida Power & Light Company Turkey Point Nuclear Plant Unit 3

Title:

Loss of 120V Vital Instrument Panel 3P06 Safety Related Procedure Responsible Department: Operations Revision Approval Date: 7/31/03 RTSs 93-0728P, 93-0099P, 93-1422P, 95-0463P, 96-0087P, 96-0602P, 97-0752P, 97-1334P, 97-1417P, 98-0851P, 98-1272P, 00-0439P, 02-0294P, 02-0593P, 03-0467P OTSC 0547-00 PC/Ms93-005, 94-034,95-102, 97-036,98-025 W97:bclswlsw/mrg

LIST OF EFFECTIVE PAGES Revision Page Date 1 07/31/03 2 07/31/03 3 10/07/02 4 10/07/02 5 10/07/02 6 10/07/02 7 10/07/02 8 10/07/02 9 10/07/02 10 10/07/02 11 10/07/02 12 10/07/02 13 10/07/02 14 10/07/02 15 07/31/03 16 10/07/02 17 10/07/02 18 10/07/02 19 10/07/02 20 10/07/02 21 10/07/02 22 10/07/02 23 10/07/02 Foldout Page 10/07/02 W97:bc/swlswlmrg

1.0 PURPOSE This procedure provides instructions to be followed upon receipt of Loss of 120V Vital Instrument Panel 3P06.

2.0 SYMPTOMS OR ENTRY CONDITIONS 2.1 Indications 2.1.1 Power Range N41 Failure (NIS Racks Channel I Lights Out) 2.1.2 Loss of Channel I Vital Instrumentation/Indications 2.1.3 Transfer of Feedwater Control from Automatic to Manual for Steam Generator A 2.1.4 Loss of Power to Pressurizer pressure control Auto/Manual Station (auto lockup) 2.1.5 Loss of Power to the Pressurizer Spray Valve Auto/Manual Station (auto lockup) 2.1.6 Loss of Pressurizer Heaters (Control and Backup) 2.1.7 Isolation of CVCS Letdown Flow 2.1.8 Loss of Power to Pressurizer Level Auto/Manual Station (auto lockup) 2.1.9 Loss of Power to 3A Charging Pump Auto/Manual Station (auto lockup) 2.1.10 RCP Thermal Barrier Cooling Water Valve, MOV-3-626, closes 2.1.11 PORV-456 Auto Open Disabled (if in OMS LOW PRESSURE OPS) 2.1.12 Loss of Power to Steam Generator C Auto/Manual Station (auto lockup) 2.2 Alarms 2.2.1 F 1/2, VITAL AC BUS INVERTER TROUBLE 2.2.2 B 6/5, POWER RANGE LOSS OF DETECTOR VOLTAGE 2.2.3 B 7/1, NIS/RPI ROD DROP ROD STOP 2.2.4 C 6/1, SG A LEVEL DEVIATION 2.2.5 A 1/5 RCP SEAL LEAKOFF HI FLOW 2.2.6 A 6/4, RCP SEAL WATER LO DP 2.2.7 A 7/6, RCP C SEAL WATER BYPASS LO FLOW (if CV-3-307 Open)

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2.2.8 H 7/5, CSP A/B COOLING WATER LO FLOW 2.2.9 A 1/1, RCP THERMAL BARR COOLING WATER HI FLOW 2.2.10 H 6/2, RHR HX HI/LO FLOW 2.2.11 X 4/1, ARMS HI RADIATION 2.2.12 H 1/2, SFP HI TEMP 2.3 General 2.3.1 Loss of the 120V Vital Instrument Panel 3P06 results in a loss of automatic feedwater control, and a loss of power to all channel I instrumentation.

ENCLOSURE 1 of this procedure contains a list of instrumentation lost in the Control Room due to the loss of Vital Instrument Panel 3P06.

2.3.2 As with any loss of a vital AC panel, early diagnosis and recovery is of greatest assistance toward unit restoration.

3.0 REFERENCES

/RECORDS REOUIRED/COMMITMENT DOCUMENTS 3.1 References 3.1.1 Technical Specifications

1. Section 3.3.2, ESFAS Instrumentation

2. Section 3.8.1.1, Diesel Generators

3. Section 3.8.3.1, Onsite Power Distribution

4. Section 3.4.9.3, Overpressure Mitigating Systems 3.1.2 FSAR

1. Section 8.2-7, Electrical 3.1.3 Plant Drawings

1. 5610-E-855, Breaker List W97:bclswlswlmrg

3.1.4 Procedures

1. 3-GOP-103, Power Operation to Hot Standby

2. 3-EOP-E-0, Reactor Trip or Safety Injection

3. -OP-003.3, 120V Vital Instrument System 3.1.5 Plant Change/Modifications

1. PC/M 93-005, Elimination of Turbine Runback Dropped Rod

2. PC/M 94-035, RTDP Related RPS/ESFAS Setpoint Changes

3. PC/M 95-102, Abandonment of the C07 Panel and Sample Train Reconfiguration

4. PCIM 97-036, Quarterly MEP

5. PC/M 98-025, Repowering Of RHR Pressure Interlock (PC-3/4-600X) 3.2 Records Reouired 3.2.1 The date and time procedure completed shall be logged in the Reactor Control Operator (RCO) logbook(s). Also, any problems encountered while performing the procedure should be logged (i.e., malfunctioning equipment, delays due to changes in plant conditions, etc.).

3.3 Commitment Documents 3.3.1 CTRAC 90-0248 W97:bclsw/sw/mrg

STEP llACTIONIEXPECTED RESPONSE I RESPONSE NOT OBTAINED CAUTION If the pressurizer spray valves were open prior to the loss of 3P06, a Reactor Trip may occur due to OTLT or low pressurizer pressure.

_ -_ -_ _ _ _ _ - NOE_ - - - - _

NOTESl

• Step I is an immediate action step.

• All 3P06 (RED) channel indication/controls are affected by failure of 3P06. Enclosure 1 provides a listing of lost functions, indications, and controls.

Check If A Reactor Trip Has Occurred Perform the following:

a. IFa reactor trip is required, THEN manually trip the reactor AND perform 3-EOP-E-0, REACTOR TRIP OR SAFETY INJECTION, while continuing with this procedure.

b. IFreactor trip is NOT required, THEN go to Step 2.

2 Check Unit Operating In Modes I Through 3 Prior To Loss Of 3P06 Perform the following:

a. IF RCS solid, THEN perform the following to prevent RCP damage AND maintain RCS pressure:

1) Stop All RCPs

2) Stop and start charging pumps as necessary to maintain RCS pressure.

b. IFRHR cooling is in service, AND MOV-3-750 is closed or stroking closed, THEN stop the operating RHR pump(s)

AND go to 3-ONOP-050, LOSS OF RHR, while continuing with this procedure.

c. IFOMS is in LOW PRESSURE OPS AND PORV-3-456 isrequired to be open for pressure control, THEN manual action shall be taken to control RCS pressure.

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STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED I

3 Control Pressurizer Pressure As Follows:

a. Reduce charging flow to minimum required to maintain RCP seal Injection using the 3B OR 3C charging pumps in MANUAL speed control

b. Check Pressurizer PORVs - CLOSED b. IF PRZ pressure less than setpoint, THEN manually close PORVs. IF any PRZ PORV can NOT be closed, THEN manually close its block valve.

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I Procedure No.: Procedure

Title:

3-ONOP-003.6 Loss of 120V Vital Instrument Panel 3P06 I lJSTEP I ACTION/EXPECTED RESPONSE l I RESPONSE NOT OBTAINED E NOTES E

• VCT Temperature indication, T-3-116, should be monitored in lieu of Excess Letdown, TI-3-139, which is de-energized.

I Excess letdown flow must be established slowly to minimize thermal I I stresses on the Excess LTDN Heat Exchanger (5 to 10 minutes). I 4 Maintain Pressurizer Level As Follows:

a. Place Pressurizer Level control switch in Position 3 (Ch 2 & 3)

b. Place Excess Letdown in service as follows:

1) Verify Excess LTDN HX CCW Outlet, CV-3-739, open

2) Verify Excess LTDN Divert to WDS, CV-3-389, is aligned to the VCT (switch to NORMAL)

3) Open Excess LTDN Stop Valve, CV-3-387

4) Open Excess LTDN Flow Controller, HCV-3-137 AND adjust flow to control Pressurizer Level W97:bclswlswlmrg

Procedure No.: I Procedure

Title:

3-ONOP-003.6 Lo ss of 120V Vital Instrument Panel 3P06 J STEP I ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED l CAUTIONS

• Reducing feed flow to less than steam flow by 655,000 lbs/hr will I result in a reactor trip due to low level trip logic on Channel I of I I

each steam generator.

• Steam Generator 3A level controls are in MANUAL and 3A FW Bypass Valve fails closed.

• 3A Steam Generator Level Recorder is DE-ENERGIZED

• Steam Generator 3C level controls are in AUTO LOCKUP I - a - a - a - a - -aNOTE 3B Steam Generator Level Controller should remain in AUTOMATIC

- - - w- - - - - - - - - - -- - -

5 Control Steam Generator Water Levels As Follows:

• 3A Steam Generator by manual control of Feedwater flow

• 3C Steam Generator by adjusting the following parameters:

• Blowdown flow

• Feed flow

• Turbine load

• Steam Flow 6 Maintain The Following Plant Parameters -

STABLE:

IF any reactor trip setpoint is approached or exceeded, THEN manually trip the reactor AND perform 3-EOP-E-0, REACTOR TRIP

• Tavg OR SAFETY INJECTION, while continuing

• Reactor power with this procedure.

• Pressurizer Pressure

• Pressurizer Water level

• Steam Generator Water level W97:bc/swlswlmrg

STEP l l ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 7 Check Power Restored To 3P06 Perform the following:

a. Continue efforts to restore power to 3P06.

b. IF power can NOT be restored to 3P06 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, THEN perform the actions required by Technical Specifications as directed by the NPS.

c. Return to Step 1.

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No.:

Procedure No.:

IPrcr Tte Procedure

Title:

IU 3-ONOP-003.6 J ACTIONIEXPECTED RESPONSE lISTEP RESPONSE NOT OBTAINED I CAUTION Auto/Manual controllers should NOT be returned to AUTO until vital power has been completely restored

_ _ _ _ M - - - ~ - - -'_

I NOTE I I When power is restored to a Manual/Auto station, the AUTO light should I i tuam on, after approximately 15 seconds the MANUAL light should turn on.I When the MANUAL light turns on, manual control of the process is available.

L . . _ -. ._ . . ._ .- . . . -- - _ .

8 Restore Equipment To AUTOMATIC Controls As Follows:

a. Pressurizer Pressure Control using section 2 of ATTACHMENT 4

b. Steam Generator Level control as follows:

1) Manually control feed flow to return steam generator to required band for plant operating mode

2) Manually adjust feed flow to match steam flow

3) Place the steam generator level controls to AUTO

4) Repeat Steps 8.b.1) through 8.b.3) until all steam generator controls are in AUTO

c. Direct the Operators to return all controls c. IF AUTOMATIC control is NOT I

listed on ENCLOSURE 1 to AUTOMATIC available OR desired, THEN maintain using appropriate plant procedures controls in MANUAL.

d. Verify all annunciators indicate correctly for d. Perform the actions of the appropriate the current plant status Annunciator Response procedure for the affected alarms.

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lSEPE lI ACTIONIEXPECTED RESPONSE I l RESPONSE NOT OBTAINED l 9 Go To Appropriate Procedure As Determined By The Nuclear Plant Supervisor END OF TEXT W97:bclswlswlmrg

ENCLOSURE I (Page 1 of 4)

CONTROL ROOM FUNCTIONS AND INDICATIONS LOST ON LOSS OF 3P06 FUNCTIONS. Operating Lock up of Pressurizer Pressure Controllers causing spray valves to stay as is Lose Auto and Manual Control of C Feedwater Control Valve, FCV-3-498 Lose Auto Control of A Feedwater Control Valve, FCV-3-478 Lose RCP Thermal Barrier Cooling Water, MOV-3-626 closes Lose Auto and Manual 3A Charging Pump Control causing Auto Lock-up Lose Auto Speed Control of 3B and 3C Charging Pumps Lose the Auto Makeup Control to the Volume Control Tank Lose power to Control Relay from MOV-3-l1 5C which opens LCV-3-11 5B Letdown Isolation Pressurizer heaters de-energize Lose Auto and Manual control of PCV-3-145, Letdown Pressure Controller Loss of 3B Diesel Load Sequencer, 3C23B-1 deenergized Lose AMSAC A Processor Lose the Ability to Block the Source Range Trip Lose Feedwater Isolation signal (Reactor Trip with Tavg <5540 F)

NOTES The following conditions exist which affect Pressurizer Pressure control:

• Pressurizer Pressure Controller PC-444J - AUTO LOCKUP

• PZR Spray Valve Controllers - AUTO LOCKUP l . PZR heaters deenergized I l

• Letdown isolation l l e 3A charging pump - AUTO LOCKUP 3B AND 3C Charging pump loss of auto speed control l

• Minimum charging flow for seal injection should be maintained due to l loss of thermal barrier cooling water caused by closure of MOV-3-626. l W97:bcdswlsw/mrg

ENCLOSURE 1 (Page 2 of 4)

CONTROL ROOM FUNCTIONS AND INDICATIONS LOST ON LOSS OF 3P06 F- - - - - - -- _ - _ _OTES _ - - - - - _ - -

I I

• With vital panel 3P06 deenergized, 3B bus sequencer is out of service I resulting in the following Tech Spec implications: I I I

1. AFW actuation from bus stripping on 3B 4KV bus will NOT be I generated, placing the unit in a shutdown action statement (Tech I I Spec 3.3.2, Table 3.3-2, functional unit 6.d action 23 invokes Tech I I Spec 3.0.3.) I I

2. Loss of Power signals are lost via the 3B bus sequencer, placing the I I unit in a shutdown action statement (Tech Spec 3.3.2, Table 3.3-2, l Functional Unit 7a, b and c)

I I 3. Bus stripping will NOT automatically occur, 3B EDG will NOT automatically close in on the bus and is out service; actions of TechI I

I Spec 3.8.1. 1 apply.

L.. _ - - - _ _- - - - - - _- - - - _- - - -

INDICATORS TI-3-401 RX Vessel Leak of Temp TI-3-133 Seal Water Return Temp TI-3-139 Excess LTDN HX Temp PI-3-121 Charging Pumps Disch Press TI-3-123 Regen Hx Outlet Temp TI-3-141 LTDN Relief To PRT Temp TI-3-143 Non-Regen HX LTDN Temp FI-3-150 Low Pressure Letdown Flow Indication FR-3-154B #1 Seal Leakoff Recorder Low Range (Fails As Is)

FR-3-154A #1 Seal Leakoff Recorder High Range (Fails As Is)

PI-3-154 C RCP Seal AP PI-3-128A B RCP Thermal Barrier AP PI-3-402 RCS Press NR PI-3-403 RCS Press WR TI-3-465 Pzr Safety Valve Temp TI-3-467 Pzr Safety Valve Temp TI-3-469 Pzr Safety Valve Temp TI-3-463 PZR Relief Temp TI-3-452 PZR Spray Loop B Temp TI-3-451 PZR Spray Loop C Temp TI-3-412B A Loop Ovpwr AT TI-3-412A A Loop AT TI-3-412C A Loop Ovtemp AT

_ Ir resswm

ENCLOSURE 1 (Page 3 of 4)

CONTROL ROOM FUNCTIONS AND INDICATIONS LOST ON LOSS OF 3P06 INDICATORS TI-3-412D A Loop Temp Avg PI-3455 PZR Press Ch I LI-3-459A PZR Level Prot/Cont.

FI-3-414 RCS Flow Loop A FI-3-424 RCS Flow Loop B FI-3-434 RCS Flow Loop C TR-3412 Delta-T Recorder NR-346 NIS Overpower Recorder LI-3-474 A Stm Gen Level LI-3-484 B Stm Gen Level LI-3-494 C Stm Gen Level LR-3-477 Stm Gen Wide Range Level (Fails As Is)

FR-3-478 3A Steam Generator Recorder LI-3470 Pzr Relief Tk Level TI-3-471 Pzr Relief Tk Temp PI-3-472 PZR Relief Tank Pressure PC-3444H Auto Manual Station for Pzr Spray Valve PCV-3-455B PC-3444G Auto Manual Station for Pzr Spray Valve PCV-3-455A PC-3-444J Auto Manual Station for Pressurizer Pressure Controller SC-3-151A Auto Manual Station Charging Pump A Control PC-3-145B Auto Manual Station Low Pressure Letdown Pressure FC-3-113A Auto Manual Station Boric Acid to Blend System NI-3-6649B1/B2 Gammametrics Backup NIS N-3-31 Source Range Counts and Source Range Startup Rate N-3-35 Inter Range Current Current and Startup Rate N-3-41 Power Range and Axial Flux Difference LC-3-478A Stm Gen A Control Valve Controller, Lose Inst, MAN Control Only FCV-3-479 3A FW Bypass Valve U-3 Pressurizer Safety Valve Acoustic Monitoring System NIS Rack Ch I (N-31, N-35, N-41)

RI-3-631 1B Containment High Radiation TI-3-610B CCW Pump Inlet Temp TI-3-607B B CCW HX Outlet Temp FI-3-613B B CCW HX HDR Flow RI-3-631 lB Containment High Radiation (VPC)

I ALARMS A 1/1, RCP THERMAL BARR COOLING WATER HI FLOW A 1/5, RCP SEAL LEAKOFF HI FLOW (C RCP only)

A 6/4, RCP SEAL WATER LO DP (C RCP)

C 6/1, SG A LEVEL DEVIATION H 1/2, SFP HI TEMP H 7/3, RHR PP A COOLING WATER LO FLOW H 7/5, CSP A/B COOLING WATER LO FLOW X 3/6, SI PP COOLING WATER LO FLOW W97:bcdswlswlmrg

ENCLOSURE 1 (Page 4 of 4)

CONTROL ROOM FUNCTIONS AND INDICATIONS LOST ON LOSS OF 3P06 r---- - - -- - - - - - -

- - - - - - - - - - -NOTE l I The following listed are shutdown mode concerned failures, and are in l E addition to power mode failures.

L _ - - _ -. _ _ - _ - _- - _ _ _ J FUNCTIONS. Shutdown Lose RHR Suction, MOV-3-750 closes from Loss of PT-3-403 Lose Auto/Manual Control of FCV-3-605 Lose Pressure Control by HCV-3-142, fails closed.

PORV-3-456 Auto Open signal from OMS is defeated, consult Tech Specs if OMS is required to be operable.

Lose ability to open MOV-3-862B/863B due to a loss of power to PC-3-600X Lose B Gammnametrics INDICATORS PT-3-403 RCS pressure HIC-3-142 RHR LTDN To CVCS FC-3-605C Auto Manual Station RHR Ht Exchanger Bypass Flow Control Amber Safety Injection Lights for the following valves:

MOV-3-744B MOV-3-843B MOV-3-862B MOV-3-863B MOV-878A MOV-3-865B MOV-3-860B MOV-3-861 B MOV-3-864B MOV-3-866B ALARMS H 1/2, SFP HI TEMP H 6/2, RHR HX HI/LO FLOW H 7/3, RHR PP A COOLING WATER LO FLOW X 3/6, SI PP COOLING WATER LO FLOW W97:bc/swlsw/mrg

ATTACHMENT 1 (Page 1 of 2)

RESTORATION OF 3P06 VITAL INSTRUMENT AC BUS In the Inverter Room, perform the following:

a. Proceed to the 3C inverter.

b. Open the 3C inverter System Output breaker, CB6.

2. In the Cable Spreading Room, perform the following:

a. At Vital Instrument Panel 3P06, place all breakers to OFF.

b. At Subpanel 3P21, place all breakers to OFF.

3. Check 4P06 being powered by CS Inverter at 4P06A Vital Instrument AC Selector Switch in the Cable Spreading Room.

4. IF 4P06 is powered by the CS Inverter, THEN notify the Nuclear Plant Supervisor.

CAUTION DO NOT proceed with this procedure ff 4P06 is powered by the CS Inverter.

5. IF 4P06 is NOT powered from CS Inverter, THEN place SPARE inverter CS in service to supply 3P06 Vital Instrument AC Bus load as follows:

a. At Vital Instrument Panel 3P06A in the Cable Spreading Room, place Vital Instrument AC Selector Switch 3P06A to the ALTERNATE SUPPLY STANDBY STATIC INVERTER CS position.

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ATTACHMENT 1 (Page 2 of 2)

RESTORATION OF 3P96 VITAL INSTRUMENT AC BUS CAUTION If System Output Breaker, CB6, has tripped, this would indicate an overcurrent condition and the amps should be monitored when each breaker on the Vital and Subpanel is closed. Amps should stabilize at less than 63. This will require a second operator at the CS inverter or at ERDADS to monitor amperage.

6. Notify the Control Room that circuits on 3P06 are about to be energized.

7. At Vital Instrument Panel 3P06, place the following breakers in the ON position:

a. 3P06 - Main

b. 3P06-4, (energizes LC460CX).

c. 3P06-8, (energizes AUTO/MANUAL station for Steam Generator C).

8. At Panel 3P06, place the remaining breakers in the ON position using Attachment 2, I AND allowing five (5) seconds between each breaker. I

9. At Subpanel 3P21, place breakers in the ON position using Attachment 3 AND allowing five (5) seconds between each breaker. I

10. In the Inverter Room, at the (locked) Alternate Source Transfer Switch 3Y05B, perform the following:

a. Unlock Alternate Source Transfer Switch AND place in the BACKUP TO SPARE INVERTER CS position.

I1. At Spare Inverter CS (3Y06), place the Synch Selector Switch inside the inverter panel in the NORMAL (down) position.

12. Notify the Control Room when all breakers are closed.

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ATTACHMENT 2 (Page 1 of 2) 120V AC VITAL INSTRUMENT PANEL 3P06 NORMAL ALIGNMENT Component CoetDcitio Normal Number omponen escripion Position 3P06A Instrument AC Selector Switch ALTERNATE 3P06-Main Feed Bkr from Selector Switch 3P06A ON 3P06-1 3QR1 - Process Protection Rack 1 ON 3P06-2 3QR2 - Process Protection Rack 2 ON 3P06-3 3QR3 - Process Protection Rack 3 ON 3P064 3QR5 - Process Protection Rack 5 ON 3P06-5 3QR6 - Process Protection Rack 6 ON 3P06-6 3QRIO - Process Protection Rack 10 ON 3P06-7 C25 - Boron Recycle and WDS Panel ON 3P06-8 3C01 - Control Console, Left Section ON 3P06-9 SPARE *ON 3P06-10 3QR47 - Misc. Relay Rack 47 and TB3200 ON Gamma Metrics Hi Flux at Shutdown 3P06-11 3QR32 & 3QR37 - Reactor Protection Rack 32 ON (Train A) and Rack 37 (Train B) 3P06-12 TB3134 - CCW From RCP Hi Flow FC-3-626 ON and TIC-3-651 SFP Cooling Hi Temp 3P06-13 3QR59 - Nuc Inst Rack 59 (Ch 1 Source & Inter ON Range Cont Pwr) 3P06-14 3X12; Nuclear Inst. Rack Solatron A ON

• Note: All breakers shall be in the ON position to reduce confusion and have conformity.

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ATTACHMENT 2 (Page 2 of 2) 120V AC VITAL INSTRUMENT PANEL 3P06 NORMAL ALIGNMENT Component Component Description Normal Number Position 3P06-15 3QR46 Misc. Relay Rack 46 ON 3P06-16 3QR30 - Area Radiation Monitor Rack 30 ON 3P06-17 3C03 - Vertical Board, Section A Right ON 3P06-18 3QR59 - Nuc Inst Rack 59 (Ch 1 Power Range Cont ON Pwr) 3P06-19 3C41 - Lighting Remote Control Relay Panel ON 3P06-20 3E03 - No. 3 Gen Excitation Swgr Cabinet Ground ON Detection 3P06-21 2R20 - Power Supply DDPS AC Monitor ON 3P06-22 Emergency Load Sequencer Panel - 3B (3C23B-1) ON 3P06-23 Safety Valve Pos Ind Power Supply - 3QR66 ON 3P06-24 Emergency Load Sequencer Panel - 3B (3C23B-1) ON W97:bc/swlsw/mrg

ATTACHMENT 3 (Page 1 of 1) 120V VITAL INSTRUMENT AC SUBPANEL 3P21 NORMAL BREAKER ALIGNMENT Component Component Description Normal Number Position 3P21-1 RAT-3-631 lB Ctmt Hi Range Monitor ON 3P21-2 PR-3-6306B Ctmt Press Recorder ON 3P21-3 Transducer Cabinet SPDS/SAS C300B ON 3P21-4 RR-3-6311B Ctmt Radiation & Hydrogen ON Recorder 3P21-5 Emergency Diesel Generator 3B Current Xducer ON 3P21-6 Wide Range Rad Monitor C212 ON 3P21-7 SPARE *ON 3P21-8 SPARE *ON 3P21-9 Relay Rack 3QR47 Power - RHR Interlock ON PC-3-600X 3P21-10 RHR & HHSI Valve Position Indication ON 3P21-11 SPARE *ON 3P21-12 Charcoal Filter Dousing Valve Actuation Flow ON Switches at C281B

• Note: All breakers shall be in the ON position to reduce confusion and have conformity.

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ATTACHMENT 4 (Page 1 of 2)

PRESSURIZER LEVEL AND PRESSURE CONTROL WITH 3P06 DE-ENERGIZED SECTION 1 CAUTION Pressurizer level should be monitored closely on the operable instrumentation during performance of the following steps to avoid uncovering the pressurizer heaters or causing a hi level trip.

NOTE PCV-3-145 is in AUTO-LOCKUP. The letdown orifice which was in service phor to the loss of 3P06 should be used when restoring letdown.

Perform the following:

a. Verify Pressurizer PORVs are closed.

b. Verify Pressurizer Level control selector switch in Position 3 (CH 2, & 3).

c. Proceed to Rack 46 (Front) AND manually hold in Relay LC 460 CX.

d. Operate heaters as necessary to return pressure to normal.

e. Restore letdown as follows:

1) Verify Letdown orifice isolation valves - CLOSED

2) Open Letdown From Regen Heat Exchanger Isolation CV-3-204

3) Open High Pressure Letdown Isolation From Loop B Cold Leg, LCV-3-460

4) Open letdown orifice isolation valve to establish desired flow.

2. IF pressure is NOT increasing with heaters energized, THEN proceed to Rack 20 front AND remove the power fuse from the front of PC-444 C&D to close the Pressurizer Spray Valves.

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ATTACHMENT 4 (Page 2 of 2)

PRESSURIZER LEVEL AND PRESSURE CONTROL WITH 3P06 DE-ENERGIZED

3. IF the above preferred method of energizing pressurizer heaters AND restoring letdown flow is NOT successful, THEN proceed as follows:

a. Proceed to the Unit 3 West electrical penetration room AND perform the following:

1) Select LOCAL control of 3A Backup Group Pressurizer heaters.

2) Push START/STOP pushbuttons as necessary to control heater operation.

b. IF necessary, THEN restore Letdown flow by holding valve handswitches in the OPEN position to initiate normal letdown, SECTION 2 I. WHEN power to the Vital AC bus is restored, THEN perform the following:

a. IF relay LC460CX in Rack 46 is being held in, THEN release hold on relay.

b. IF the power fuses for Pressurizer Spray valves were removed in Section 1, Step 2, THEN replace the power fuses for PC-444C and PC-444D in Rack 20.

c. Restore pressure control using 3-OP-041.2, PRESSURIZER SYSTEM.

FINAL PAGE W97:bclsw/swlmrg

FOLDOUTPAGE FOR PROCEDURE3-ONOP-003.6

1. Dispatch an operator to restore power to 3P06 using Attachment 1. I

2. Dispatch an operator to restore pressurizer pressure AND level controls using Attachment 4. 1

3. IF a Reactor Trip has occurred, THEN perform the following: I

a. Close MOV-3-1407 I

b. Close MOV-3-1408 I

b. Close MOV-3-1409 I W97:bcIsw/swlmrg

Florida Power & Light Company Turkey Point Nuclear Plant Unit 3 & 4 (Master)

BASIS DOCUMENT FOR:

• -ONOP-003.6 DATED:

10/7/02 TITLE:

Loss of 120V Vital Instrument Panel *P06 W97:Ibcsw

Page 2 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 l ~BASIS DOCUMENT LIST OF EFFECTIVE PAGES Revision Page Date 1 10/07/02 2 10/07/02 3 10/07/02 4 10/07/02 5 10/07/02 6 10/07/02 7 10/07/02 8 10/07/02 9 10/07/02 10 10/07/02 11 10/07/02 12 10/07/02 13 10/07/02 14 10/07/02 15 10/07/02

[ -- NOTE

- - - - - I I I The intent of this document is to provide justification and explanations for selected I procedural information. The Basis Document does not contain the procedure content as a I I whole and will not be re-dated unless the procedure change affects the content of this I I document. I

- a_ - - _ _a - - _ __ a - a - a _ - -_ -

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Page 3 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 BASIS DOCUMENT 1.0 PURPOSE 1.1 Self-explanatory 2.0 SYMPTOMS OR ENTRY CONDITIONS 2.1 Indications NOTE I

• The indications listed are plant parameters and system responses caused by a loss of I 120V Vital Instrument Panel *P06. Many indications will be received, however only those I indications requiring immediate operator attention are referenced.

l 2.1.1 Instrument bus *P06 supplies power to the NIS rack PR-N41 module.

2.1.2 Loss of instrument bus *P06 results in the loss of all Channel I Vital Instrumentation/Indications (See Enclosure 1).

2.1.3 Instrument bus P06 supplies power to Steam Generator *A Feedwater Controller (FC-478), located in Process Control Rack 6. Upon loss of *P06, Steam Generator *A feedwater flow control automatically transfers to manual. This transfer is indicated by lights on Steam Generator *A Auto Manual Station (FC-478F), on the console. Specifically, the Auto light goes out and the Man light comes on.

2.1.4 Instrument bus *P06 supplies power to the Pressurizer Spray Valve Auto Manual Station (PC-444J), located on the console. All lights at PC-444J go out, while the associated automatic controller (PC-444A), located in Process Control Rack 20, transfers to auto lockup.

2.1.5 Instrument bus *P06 supplies power to the Pressurizer Pressure Controls Auto Manual Stations (PC-444G and PC-444H), located on the console. All lights at PC-444G and PC-444H go out while associated automatic controllers (PC-444C and PC-444D), located in Process Control Rack 20, transfer to auto lockup. As a consequence, pressurizer spray valves PCV-455A and PCV-455B remain as is. Closure of pressurizer spray valves upon loss of *P06 is covered in a later section of this document.

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Page 4 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 ll BASIS DOCUMENT l 2.1.6 Instrument Bus *P06 supplies power to the Pressurizer Level Comparator, LC-3-460C. During normal operations, LC-460C de-energizes relay LC-460X at 14 percent pressurizer level to trip all Control and Backup pressurizer heaters should they become uncovered. On a loss of *P06, LC-460X is de-energized, tripping all pressurizer heaters regardless of pressurizer level.

2.1.7 De-energizing LC-460X on a loss of *P06 also isolates CVCS Letdown flow. Letdown flow is isolated by the closure of orifice isolation valves 200A, 200B and 200C and letdown isolation valve, LCV-460. During normal operations, isolation would occur at 14 percent pressurizer level to prevent loss of reactor coolant inventory.

2.1.8 Instrument Bus *P06 supplies power to Pressurizer Level Auto Manual Station, LC-459G. Loss of *P06 results in all lights at LC-459G going out and Pressurizer Level Controller LC-459F, transferring to auto lockup.

Manual operation of charging pumps may be necessary.

2.1.9 Instrument Bus *P06 supplies power to *A Charging Pump Auto Manual Station, SC-151A. Loss of *P06 results in all lights at SC-151A going out and *A Charging Pump controller transferring to auto lockup.

2.1.10 Instrument Bus P06 supplies power to FIC-*-626, which fails high when it loses power, resulting in MOV-*-626 going closed.

2.1.11 Loss of power to PT-*-403 prevents the automatic functions of PORV-*-456 from opening the valve.

2.1.12 (Unit 3) Instrument Bus 3P06 supplies power to Steam Generator 3C Auto Manual Station (FC-3-498F), located on the console. Loss of 3P06 results in all lights at FC-3-498F going out and Steam Generator 3C Feedwater Flow Controller (FC-3-498) in the process control racks transferring to auto lockup.

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Page 5 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 lr BASIS DOCUMENT l 2.2 Alarms 2.2.1 Panel F, Window 1/2 is a common annunciator indicating the receipt of any of several local inverter alarms.

2.2.2 Panel B, Window 6/5 reflects a loss of voltage to a NIS power range module.

2.2.3 Panel B, Window 7/1 reflects the actuation of an NIS Rod Drop Relay which in turn generates an auto rod withdrawal stop signal. (NC-41KX).

2.2.4-2.2.12 Alarms are the result of a loss of power to the relays allowing them to makeup the alarm circuits.

2.3 General 2.3.1 Self Explanatory 2.3.2 Self Explanatory

3.0 REFERENCES

/RECORDS REOUIREDICOMMITMENT DOCUMENTS Self explanatory W97:Jbcsw

Page 6 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 BASIS DOCUMENT l PROCEDURE STEPS This caution alerts the operator to a potentialautomatic reactor tip.

r _ _ _ _~ _ - _ _ _ - - _ - -

- - - - - - - - - - NOTES K

• Immediate actions are those actions which the operator should be able to perform l before opening and reading the emergency procedures. Although the operator should memorize immediate actions, they need not be memorized verbatim. The operator should know them well enough to complete the intent of each step. l Provides reminders that *P06 is the RED channel and that Enclosure 1 is available to determine what functions, indications, and controls are lost. -

This step checks if a reactor trip has occurred. If a reactor trip has not occurred the operator is directed to check if a reactor trip is required. If a reactor trip is required, the operator is directed to manually trip the reactor and perform EOP-E-0 concurrently. If a trip is not required, the operator is directed to the applicable procedure step.

2. This step is written to direct the operator to check for loss of RHR if the unit is not operating in MODES 1-3. Automatic RHR flow control is lost. MOV-*-750 fails closed when PC-*-403 loses power and PCV-*-142 fails closed which may lead to an RCS overpressure condition.

3. The loss of *P06 directly affects the normal control of pressurizer pressure and level.

Operator attention to the pressurizer is necessary to maintain pressure and level in normal ranges.

a. Minimizing the fill rate of the pressurizer will extend the time frame for recovery without lifting a PZR PORV due to compressing the bubble.

b. Power operated relief valve, PCV-455C, receives its control signal from pressure comparator, PC-444A. Upon loss of *P06, PC-444A output signal locks up as is, with the possibility of maintaining PCV-455C in the open position. For this reason, pressurizer PORVs should be verified shut to prevent inadvertent depressurization of the RCS.

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Page 7 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 BASIS DOCUMENT PROCEDURE STEPS

- - - - - - - - - - NOTES -

Provides the operator with an alternate temperature indication for the VCT. l Reminds the operator to use caution when changing Excess Letdown Heat Exchanger temperatures. -

4. Gives actions that must be taken to restore manual control of PRZ pressure.

a. The function of the Pressurizer Level control switch is to transfer the output of pressurizer level channels I, II and III to level comparators LC-459C and LC-460C.

This allows flexibility of operation for testing and loss of channel conditions. On loss of *P06, both pressurizer level channel I and LC-460C are de-energized. Placing the switch in CH3, CH2 (position 3) selects out de-energized channel I level transmitter, LT-459, although LC-460C remains de-energized.

b. Excess letdown is available and placing it in service will assist in maintaining pressurizer level.

CAUTIONS

• Channel I level protection circuits for A, B, and C Steam Generators introduce a low-low level sgnal (10 percent) to the trip matrix upon loss of power.

Reducing feed flow less than steam flow by 20 percent of full flow (0.665 x 10 lbmlhr) would complete the reactor trip logic, even though steam generator levels may be normal. Caution must be used when reducing feed flow less than steam flow to prevent a reactor trip.

• Self explanatory

• Self explanatory

• Self explanatory (Unit 3 ONLY)

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Page 8 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 1017/02 IL PROCEDURE STEPS BASIS DOCUMENT l

-l N OTE -l

• Self explanatory N Self explanatory I

5. Monitoring steam generator levels is important since *A Feedwater Flow Controller is in manual. The operator must maintain manual control of *A Steam Generator level to prevent a reactor trip.

(Unit 3) The operator must maintain control of 3C Steam Generator level to prevent a reactor trip. Monitoring steam generator levels is important since 3C Feedwater Flow Controller is locked up at signal prior to vital bus loss. However, since no turbine runback has occurred, feed/steam mismatch is slight or non-existent and any change in steam generator level will be slow. Steam Generator 3C level adjustments may be made using careful control of all steam generator parameters.

6. The operator is directed to maintain key plant parameters stable.

7. The operator is directed to check if power has been restored to *P06. If conditions stabilize and *P06 is still not recovered, the operator is directed to perform the following:

a. Continue efforts to restore *P06.

b. If power cannot be restored within the prescribed time frame a unit shutdown is commenced to ensure compliance with Tech Specs. If the NPS determines the unit cannot be shut down safely with the manual controls, the unit may be tripped and stabilized using the EOP network.

c. If one hour has not elapsed and power is not restored to *P06, the operator is looped back to Step I to maintain the unit stable.

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Page 9 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 BASIS DOCUMENT PROCEDURE STEPS CAUTION It Is possible to shift the controller to AUTO approximately 15 seconds after the MANUAL light comes on; however, the controller should remain In MANUAL until vital bus power has been completely restored. This prevents operating the controller In AUTO with deenergized controller input signals and also allows Input signals to stabilize.

I NOTES I I When power is restored to a ManuallAuto station, the AUTO light comes on. In I I approximately 15 seconds, the MANUAL light comes on and the AUTO light goes off. I I Manual control is available at the time the MANUAL Light comes on.

I

• During Low Pressure OPS (Shutdown) PORV-456 has dual power feeds. Loss of l *,~.P065 will cause PT-403 to de-energize and disable the auto open feature of this PORV. This loss will not result in an indication in the Control Room for the PORV.

Refer to Technical Specification for PORV operability.

8. When the Vital AC bus is restored, all altered controls should be returned to normal alignments before continuing power operations.

a. Manual control of pressurizer level is restored using Attachment 4.

b. Steam generator level control is restored by matching steam flow and feed flow on each steam generator and then placing the feedwater regulating valve controls to automatic.

c. The operators are directed to use Enclosure 1 to restore the remaining affected controls to automatic. Caution should be used to ensure the process signals are stable and ready to be returned to automatic.

d. Self explanatory

9. The Nuclear Plant Supervisor will determine the appropriate plant procedure to be used for continued operation.

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Page 10 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7102 t ~BASIS DOCUMENTl ENCLOSURE 1 Provides a comprehensive list of lost functions, indications, and controls not specifically addressed in the procedure. Knowledge of these functions, indications, and controls is important if vital instrument power is not restored promptly.

ATTACHMENT 1

1. This step directs the operator to attempt to restore *P06.

a. Inverter SC is located in the Inverter Room adjacent to the Control Room.

b. Opening inverter *C output breaker isolates the failed inverter from *P06.

2. This step prepares *P06 for restoration.

a. Opening all breakers on *P06 prevents current surges when *P06 is reenergized.

b. Opening *P21 breakers also prevents current surges upon reenergizing the vital bus.

3. Ensuring the &P06 is not being powered by the CS inverter prevents parallel powering of 3P06 and 4P06. If CS inverter is powering &P06, discontinue this procedure.

4. If &P06 is being powered by CS inverter, the Nuclear Plant Supervisor should be notified.

The course of action at this time will depend on the status of Unit & and the &C inverter.

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Page 11 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 1017/02 ll BASIS DOCUMENT l ATTACHMENT 1 (Cont'd)

- CAUTION It is possible to align the output of the spare Inverter CS to vital buses 3P06 and 4P06 simultaneously. Although Inverter CS load capacity Is sufficient to power both buses, parallel operation of two vital buses on one Inverter Is prohibited. This prevents a single failure (Inverter CS) from Impacting both units (3P06 and 4P06) simultaneously.

5. If available, the step transfers *P06 to the spare inverter by providing instructions for placing the spare inverter in service.

a. Designated switch operation is performed at Vital Instrument Panel P06A. Placing the Vital Instrument AC Selector Switch *PO6A to the ALTERNATE SUPPLY STANDBY INVERTER CS position, places the CS spare inverter in service.

b. Alternate source Transfer Switch *YO5B is located in the Inverter Room.

(1) Placing the Alternate Source Transfer Switch in the BACKUP TO SPARE INVERTER CS position, ensures that the CVT will be available to carry Vital Instrument AC loads in the event Spare Inverter CS fails.

c. Placing the Synch Selector Switch in the NORMAL position, enables the inverter to adjust its frequency to match that of the *YO5B CVT.

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Page 12 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 ll BASIS DOCUMENT l ATTACHMENT 1 (Cont'd)

CAUTION Warns the operator about amperage limitations.

6. Keeps the Control Room aware of what actions are being taken or are about to be taken in the field.

7. With CS inverter supplying power to *P06, vital loads may be energized by placing the following breakers to the ON position:

a. Placing P06-Main to ON energizes the supply side of all *P06 load breakers.

b. *PO6-4 energizes relay LC-460CX. Manual operation of LC-460CS may now be discontinued.

c. 3P06-8 energizes Steam Generator 3C Auto Manual feedwater controller. Steam Generator 3C may then be controlled in manual. (Unit 3 ONLY)

8. Attachment 3 lists loads on Vital Panel *P06.

9. Attachment 2 lists loads on Sub-panel *P2 1.

10. This notification permits the Control Room to return to normal indication and control of reactor plant parameters.

ATTACHMENT 2 Provides the required breaker alignment for panel *P06.

ATTACHMENT 3 Provides the required breaker alignment for panel *P2 1.

ATTACHMENT 4. Section 1 This attachment gives necessary instructions for manually controlling pressurizer pressure and level when *P06 is de-energized, and for restoring pressurizer level and pressure controls when

• P06 is restored.

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Page 13 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 BASIS DOCUMENT ATTACHMENT 4. Section 1 (Cont'd)

CAUTION Self explanatory NOTES I Self explanatory

1. The loss of *P06 directly affects the normal control of pressurizer pressure and level.

Operator attention to the pressurizer is necessary to maintain pressure and level in normal ranges.

a. Power operated relief valve, PCV-455C, receives its control signal from pressure comparator, PC-444A. PC-444A output signal locks up as is on loss of power with the possibility of maintaining PCV-455C in the open position. For this reason, pressurizer PORVs should be verified shut to prevent inadvertent depressurization of the RCS.

b. On loss of *P06, both pressurizer level channel I and LC-460C are deenergized.

Placing the switch in CH3, CH2 (position 3) selects out de-energized channel I level transmitter, LT-459, although LC-460C remains de-energized.

c. *P06 supplies power to the pressurizer level comparator, LC-460C. On loss of *P06 and LC-460C, pressurizer level comparator relay LC-460CX de-energizes and operates contacts in the control circuits for letdown isolation valve LCV-460, orifice isolation valves, 200A, 200B and 200C, control group heaters and backup group heaters. As a result, all pressurizer heaters are de-energized and letdown flow is isolated. To regain normal control of pressurizer heaters and letdown flow, LC460CX must be operated manually. LC-460CX is a type BF relay located adjacent to LC-459CX in the front of Aux. Relay Rack 46. The relay is manually operated by depressing the pushbutton located on the relay face. This method of regaining pressure control was chosen because it does not defeat the low level protection feature. If an actual low level signal is present, relay LC-459CX will de-energize pressurizer heaters, preventing heater damage. It is important here to note that only relay LC-460CX should be held in. Do not attempt to hold in both relays.

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Page 14 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 1017/02 8 ~BASIS DOCUMENT l ATTACHMENT 4, Section 1 (Cont'd)

d. Normal control of pressurizer heaters is now available and may be operated if necessary to return pressure to normal.

e. Actions required to restore letdown.

2. If operating pressurizer heaters does not increase RCS pressure, Pressurizer Spray Valves, PCV-455A and B may be partially open. On loss of *P06, Spray Valve Auto Manual stations (PC-444G and H) are de-energized, causing their associated automatic controllers (PC-444C and D) to lock up as is. The spray valves then remain in the same position as they were when *P06 was lost. To shut PCV-455A and B, proceed to the front of Rack 20 and de-energize PC-444C and D by removing the power fuses located on the face of each controller. Spray Valve proportional controllers now sense a zero volt input and shut the spray valves.

3. If the preferred method of restoring pressurizer heaters and letdown flow are not successful, the operator is given alternate methods below.

a. Provisions have been made for local control of pressurizer heater backup group A in the Unit 3 West (Unit 4 North) electrical penetration room. A local/remote selector switch and start/stop pushbuttons are located near heater group A breaker panel. To gain local control, place the selector switch in the local position. Operate the heater group with the start/stop pushbuttons as needed to control pressurizer pressure. Local status lights indicate breaker position. An annunciator at Panel F in the Control Room indicates local control of backup group A. This method of heater control bypasses low pressurizer level heater protection. The heaters must be manually de-energized if pressurizer level drops below 14 percent.

b. Holding the control switches in open will bypass LC-460CX contacts in the valve control circuits, allowing valves to open.

ATTACHMENT 4 Section 2 I1. Provides actions to restore pressurizer controls.

a. With power restored LC-460C and pressurizer level is greater than 14 percent, normal pressurizer heater and letdown control should be available.

b. This action will restore normal operation of the pressurizer spray valves.

c. Self explanatory W97:/bc/sw

Page 15 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 ll BASIS DOCUMENT l FOLDOUT PAGE

1. Self explanatory I

2. Self explanatory

3. Feedwater isolation will not occur following a reactor trip with Tavg less than 554 0F because power is lost to the feedwater isolation circuit. Therefore, if a reactor trip has occurred, feedwater isolation is provided by closing the feedwater isolation MOVs.

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4 1 of 3 ODI-CO-023 MANUAL REACTOR TRIP GUIDELV4ES 04/01/02 Florida Power & Light Company Plant Turkey Point - Nuclear 1.0 REVIEW AND APPROVAL Operations Supervisor: .X T-7N J-'Usot Date Operations Manager: - -- ,  ;: 4qot.

Date THIS INSTRUCTION EXPIRES FEBRUARY 1, 2004 2.1 To provide the Control Room Operators Manual Reactor Trip Guidelines to be used during operation of the plant. Guidelines which will allow the operator to manually trip the reactor and turbine prior to reaching an automatic trip setpoint.

3.0 SCOPE 3.1 This instruction applies when either unit is in Modes 1 or 2.

3.2 This instruction does not conflict or supersede any approved plant procedures or Quality Instructions.

3.3 On-Shift personnel shall adhere to the following guidelines during operation of the plant to allow precluding an Automatic Trip.

A 2 of 3 ODI-CO-023 MANUAL REACTOR TRIP GUIDELINES 04/01/02 4.0 RESPONSIBILITIES 4.1 The Nuclear Plant Supervisor is responsible for:

4.1.1 Ensuring that all On-Shift personnel diligently monitors the Control Room and Plant instrumentation for adverse trends.

4.1.2 Ensuring that if one of these setpoints is reached or if control of the plant is in question that a preemptive trip is called for. Rapidly changing parameters need not require reaching one of these setpoints prior to ordering a Manual Reactor Trip.

4.2 All On-Shift Licensed Operators are responsible for:

4.2.1 Taking timely and proper actions to ensure safe operation of the facility.

4.2.2 Initiating Engineered Safeguards Features or Reactor Trip Actuation if indicators exceed automatic setpoints.

4.2.3 Informing the NPS/ANPS, and taking action should any Manual Trip Guideline be exceeded.

5.0 INSTRUCTIONS 5.1 IF one of these setpoints is reached or a preemptive trip is called for THEN, Trip the Reactor and Turbine AND go to E-0 Reactor Trip or Safety Injection.

TRIP MANUAL SETPOINT Source Range Hi Flux Any unexplained increase in OR Reactor Power, IR amps, or Intermediate Range Hi Flux Source Counts OR Power Range Hi Flux Lo Setpoint Power Range Hi Flux Hi Setpoint Power at 103% and increasing Pressurizer Lo Pressure 2000 psig decreasing Pressurizer Hi Pressure 2350 psig Lo Steam Generator Level 15% Narrow Range PFressurizer Hfigh Level At 80% on 2/3 channels

3 of 3 ODI-CO-023 MANUAL REACTOR TREP GUIDELINES 04/01/02 5.2 IF any of these guidelines are exceeded and Reactor Power is ABOVE 10%

THEN, Trip the Reactor AND Turbine and go to E-0, Reactor Trip or Safety Injection:

TRIP MANUAL SETPOINT Lo Vacuum MWe 24.5 Inches of Vacuum Decreasing< 531 22 Inches oFVacuum Decreasing >

531 MWe S/G Hi Level 75% Narrow Range 5.3 IF any of these guidelines are exceeded AND Reactor Power is BELOW 10%

TCHEN, Trip the Turbine and Maintain Stable plant conditions TRIP MANUAL SETOINT Lo Vacuum 24.5 Inches of Vacuum Decreasing SIG Hi Level 75% Narrow Range

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PRESSURIZER PRESSURE PSIA F 14.7 (2200.00- 2500.00)

H1-LT-460 TR.ANSMIrR OUTPUTSI1G.  %~ F10.6 (50.00- 100.00)

Turkey Point Nuclear Plant 2003-301 SRO Inital Exam 33.

The crew is responding in accordance with E-1, Loss of Reactor or Secondary Coolant.

Pressurizer level has risen continuously even though the RCS pressure has been dropping steadily. All Reactor Coolant pumps are in operation.

Which ONE of the following leak locations is consistent with the plant conditions just described?

Weld break on:

the pressurizer sample line.

t51one of the CRDM nozzle penetrations.

( the line to Pressurizer PORV Block valve MOV-536.

/ the Charging header connection to the RCS.

C.c

Florida Power & Light Company Turkey Point Nuclear Plant Unit 3 a.

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tN t d

Title:

I Loss of RHR Safety Related Procedure Responsible Department: Operations Revision Approval Date: 10/16/98C1 RTSs 93-1422P, 94-0349P, 94-1102P, 95-0957P, 95-1012P, 96-1534P 96-1456P, 98-0588P, 98-0929P PCIM 89-332,96-081 OTSC 0412-95 W97/DHldaj

LIST OF EFFECTIVE PAGES Revision Page Date 1 10/16/98C1 2 10/16/98C1 3 10/16/98 4 10/16/98 5 10/16/98 6 10/16/98 7 10/16/98C 8 10/16/98 9 10/16/98 10 10/16/98 11 10/16/98 12 10/16/98C 13 10/16/98 14 10/16/98 15 10/16/98C 16 10/16/98C 17 10/16/98 18 10/16/98 19 10/16/98 20 10/16/98 21 10/16/98 22 10/16/98 23 10/16/98 24 10/16/98 25 10/16/98C 26 10/16/98 27 10/16/98 W97/DHldaj

1.0 PURPOSE 1.1 This procedure provides instructions to be followed in the event of a loss of flow in the RHR system.

2.0 SYMPTOMS OR ENTRY CONDITIONS 2.1 Annunciators 2.1.1 H 6/2, RHR HX HI/LO FLOW 2.1.2 H 6/4, RHR PP A/B TRIP 2.1.3 I 7/6, RHR SUMP PUMP ROOM A HI LEVEL 2.1.4 I 8/6, RHR SUMP PUMP ROOM B HI LEVEL 2.1.5 I 3/6, RHR SUMP HX ROOM HI LEVEL 2.1.6 I 7/3, RX VESSEL DRAINDOWN LO-LO-LEVEL 2.1.7 A 7/1, PRT HI/LO LEVEL HI PRESS/TEMP 2.1.8 A 9/6, RHR MOV-750/751 LETDOWN ISOLATION 2.2 Indications 2.2.1 Neither RHR pump is operating when required for decay heat removal 2.2.2 Loop 3C RHR Suction Stop Valve(s), MOV-3-750 or MOV-3-751, indicate closed when RHR is required for decay heat removal 2.2.3 Rapid increase in RCS pressure and OMS actuation when the RCS is solid 2.2.4 Low flow indicated on FI-3-605

3.0 REFERENCES

/RECORDS REOUIRED/COMMITMENT DOCUMENTS 3.1 References 3.1.1 Technical Specifications for Turkey Point Unit 3 and Unit 4 3.1.2 Turkey Point Unit 3 and Unit 4 Final Safety Analysis Report W97VDH/daj

3.1.3 Operating Diagrams

1. 5613-M-3050, Residual Heat Removal System

2. 5613-M-3062, Safety Injection System 3.1.4 Procedures

1. 3-ONOP-030, Loss of Component Cooling Water

2. 3-ONOP-041.3, Excessive Reactor Coolant System Leakage

3. 3-ONOP-041.8, Shutdown LOCA [Mode 5 or 6]

4. 3-OP-041.1, Reactor Coolant Pump

5. 3-OP-050, Residual Heat Removal System

6. 3-OP-073, Condensate System

7. 0-OP-074.1, Standby S/G Feedwater System 3.1.5 Plant Change/Modifications

1. PC/M 89-332, Generic Letter 88-17, Loss of Decay Heat Removal Programmed Enhancement - RCS Redundant Level Monitors

2. PC/M 96-081, Setpoint Change for RCP Seal Leakoff Low Flow 3.1.6 Miscellaneous Documents

1. JPN-PTN-SEMJ-89-094, Adequacy of Core Cooling

2. JPN-PTN-SENP-92-009, Substantial Safety Hazards Evaluation Related to Pressurizer Vents at Cold Shutdown

3. Westinghouse Technical Bulletin ESBU-TB-93-01, Revision 1

4. Westinghouse EOP Rev C Changes 3.2 Records Required 3.2.1 None W97/DH/dal

3.3 Commitment Documents 3.3.1 NRC Inspection Report 89-053, March 14, 1990 3.3.2 NRC EIN No.86-101, Loss of Decay Heat Removal Due to Loss of Fluid Levels In Reactor Coolant System 3.3.3 NRC Generic Letter 88-17, Loss of Decay Heat Removal 3.3.4 NRC IN-92-16, Loss of Flow from the Residual Heat Removal Pump During Refueling Cavity Draindown 3.3.5 INPO SOER 85-4, Loss or degradation of Residual Heat Removal Capability in PWRs (CTRAC No. 85-1178-34) 3.3.6 INPO SER 17-86, Loss of Shutdown Cooling Flow (CTRAC No. 87-0823) 3.3.7 INPO SER 23-86, Loss of Decay Heat Removal Flow (CTRAC No. 86-0982) 3.3.8 INPO OE-1744, RHR Gas Binding Due to Erroneous Half Loop Indication (CTRAC No. 85-1178-34) 3.3.9 INPO SER 9-92, Loss of Residual Heat Removal with Reduced Reactor Vessel Water Level 3.3.10 JPN-PTN-SENP-95-026, CCW Flow Balance and Post-Accident Alignment Requirements to Support Thermal Up-Rate (LER 250/95-006)

W97/DH/dai

STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION If leakage from the RHR system Is discovered, the leak should be Isolated using 3-ONOP-041.3, EXCESSIVE REACTOR COOLANT SYSTEM LEAKAGE.

1-----.-

NOTES e Steps I through 3 are IMMEDIATE ACTION steps if loss of RHR is due to a loss of off-site power capability, power and RHR flow should be restored utilizing 3-ONOP-004, LOSS OF OFFSITE POWER or 3-EOP-ECA-0.0, 1 LOSS OF ALL AC. During a loss of power, this procedure should be used to establish l containment closure and alternate cooling if RHR flow remains unavailable. I L . . . - . . . - . . . . . . . . . . . . . . .

1 Dispatch An Operator To Monitor RHR Pumps As Follows:

a. Obtain radio

b. Monitor RHR pump locally

c. Maintain communication with control room

d. Stay near RHR pump until normal RHR flow is restored W97/DH/daj

ISTEPACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED

~

_~~~~

~

NOTE

- - - - - _u I

• RCS heatup rate is required to be monitored by the Shift Engineer or any available operator while efforts continue to restore RHR cooling.

2 Monitor RCS Heatup Rate As Follows:

a. Plot core exit temperature every minute a. IFcore exit temperatures are NOT for 5 minutes available, THEN perform the following:

1) Assume a 120F per minute heatup rate unless the refueling cavity is flooded. IFthe refueling cavity is flooded, THEN use 40F per minute.

2) Observe NOTE prior to Step 3 AND go to Step 3

b. Calculate RCS heatup rate

c. Determine time required to reach saturation in RCS

d. Report results to unit RCO and NPS

e. Repeat this step every 15 minutes until RHR cooling is Restored W97IDH/daj

STEP I ACTION/EXPECTED RESPONSE l l RESPONSE NOT OBTAINED l NOTE Interrupt feature for MOV-3-750 and MOV-3-751 is only functional with OMS in LO PRESS I

• OPS. I L _ _ _ _- _ _ _ _ _- _ _- _ --- _ _

3 Check Loop 3C RHR Pump Suction Stop Perform the following Valves - OPEN

a. Stop RHR pumps.

• MOV-3-750

• MOV-3-751 b. IF a momentary pressure spike has caused either or both valves to start closing, THEN perform the following at the Pushbutton Interrupt switches:

1) Determine affected valve(s).

• Yellow light -ON

2) Verify over pressure signal NOT present:

• Blue light -ON

3) Push Interrupt Pushbutton for affected valve(s).

4) Verify yellow light - DE-ENERGIZES.

5) WHEN blue light DE-ENERGIZES, THEN verify affected valve(s) - OPEN.

6) Go to Step 4.

c. IF RCS pressure GREATER THAN 525 psig, THEN perform the following:

1) Stop the charging pump(s).

2) Reduce RCS pressure to 450 psig.

d. IF Loop 3C RHR Pump Suction Stop Valve(s) were NOT closed to isolate system leakage, THEN reopen RHR Loop Suction Stop Valve(s). IF either valve can NOT be opened, THEN direct an operator to locally reopen Loop 3C RHR Pump Suction Stop Valve(s).

e. IF either valve can NOT be reopened, THEN observe NOTE prior to Step 20 AND go to Step 20.

W97/DH/daj

STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 4 Verify RHR Discharge To Cold Leg Isolation IF RHR Discharge To Cold Leg Isolation Valves - OPEN valve(s) were NOT closed to isolate system leakage, THEN perform the following:

• MOV-3-744A

• MOV-3-744B a. Reopen RHR discharge valve(s).

b. IF at least one valve can NOT be opened, THEN perform the following:

1) Stop RHR pump(s).

2) Direct operators to locally reopen RHR Discharge To Cold Leg Isolation Valve(s).

3) Observe NOTE prior to Step 20 AND go to Step 20.

W97/DH/daj

Procedure No.: Procedure

Title:

Page: 1 3-ONOP-05 Loss of R 3-ONOP-050 Loss of RIIR Approval Date:1 10116198 I I ACTIONIEXPECTED RESPONSE 1 I RESPONSE NOT OBTAINED I 5 Check RHR Pumps- ANY RUNNING Perform the following:

a. Close RHR Heat Exchanger Outlet Flow valve, HCV-3-758.

b. Close RHR Heat exchanger Bypass Flow valve, FCV-3-605.

c. Verify MOV-3-750 and MOV-3-751 -

OPEN

d. Attempt to restart previously running RHR pump.

e. IF previously running RHR pump can NOT be started, THEN start alternate RHR pump.

f. IF neither RHR pump can be started, THEN perform the following:

1) Direct appropriate personnel to restore at least one RHR pump to operable status.

2) Observe NOTE prior to Step 20 AND go to Step 20.

g. Return RHR Heat Exchanger Bypass Flow valve, FCV-3-605, to AUTOMATIC operation at desired flow.

h. Open RHR Heat Exchanger Outlet Flow valve, HCV-3-758, as necessary to maintain desired RCS temperature.

6 Verify RHR Pump NOT Cavitating Go to Step 12.

• Running RHR pump amps - STABLE

• RHR flow - STABLE

• RHR pump noise level - NORMAL 7 Check For RHR Flow Control Valve Failure

a. Verify RHR Heat Exchanger Bypass Flow, a. Go to Step 8.

FCV-3-605 - MAINTAINING DESIRED FLOW IN AUTOMATIC

b. Go to Step 18 W97/D-ldaj

_ACTIONIEXPECTED ISTEP RESPONSE I RESPONSE NOT OBTAINED 8 Control RHR Bypass Flow

a. Manually control RHR Heat Exchanger a. Perform the following at the 10 foot Bypass flow, FCV-3-605, to establish elevation platform in the RHR Heat desired flow Exchanger room to locally control RHR flow:

1) Remove seal and place Safe shutdown FCV-3-605 Manual Control Air Isolation Valve, 3-40-1895, in MANUAL.

2) Verify Safe Shutdown FCV-3-605 Manual Control Air Vent Valve, 3-40-1896, In NORMAL.

3) Adjust Safe Shutdown FCV-3-605 Manual Controller, PCV-3-605, to establish desired flow.

4) Observe CAUTION prior to Step 9 AND go to Step 9.

b. GotoStep`10 W97IDH~daj

lISTEP ACTIONIEXPECTED RESPONSE l RESPONSE NOT OBTAINED CAUTION If only two CCW Heat Exchangers are In service and MOV-3-749A and MOV-3-749B are open, two CCW Pumps are required to be maintained In PULL-TO-LOCK 9 Check Desired RHR Heat Exchanger Perform the following:

Bypass Flow, FCV-3-605 - BEING MAINTAINED USING SAFE SHUTDOWN a. Close RHR Heat Exchanger Bypass Flow MANUAL CONTROL valve, FCV-3-605.

b. Open RHR Heat Exchanger Outlet Flow Valve, HCV-3-758, to establish between 3500 and 3700 gpm.

c. Locally open breakers for RHR Heat Exchanger Outlet Component Cooling Water valves.

• 30617 for MOV-3-749B

• 30721 for MOV-3-749A

d. Control the cooldown by locally throttling both of the RHR Heat Exchanger outlet Component Cooling Water MOVs while maintaining CCW flow through both RHR Heat Exchangers.

• MOV-3-749A

• MOV-3-749B 10 Notify Appropriate Personnel To Correct RHR Flow Control Failure 11 Verify RHR Pump NOT Cavitating Go to Step 12.

a. Running RHR Pump

• Amps - STABLE

• Flow - STABLE

• Noise level - NORMAL

b. Go to Step 14 W97/DH/daj

l~.J SElI ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I 12 Reduce RHR Flow To Stop Cavitation

a. Adjust automatic setpoint for RHR Heat a. Perform the following:

Exchanger Bypass Flow, FCV-3-605, to establish RHR flow between 1000 and 1) Take manual control of RHR Heat 1500 gpm Exchanger Bypass Flow, FCV-3-605, to establish between 1000 and 1500 gpm.

2) IF manual control can NOT be established, THEN perform the following:

a) Remove seal and place SAFE SHUTDOWN FCV-3-605 MANUAL AIR ISOLATION VALVE, 3-40-1895, in MANUAL.

b) Verify SAFE SHUTDOWN FCV-3-605 MANUAL CONTROL AIR VENT VALVE, 340-1896, in NORMAL.

c) Adjust SAFE SHUTDOWN FCV-3-605 MANUAL CONTROLLE R, PCV-3-605, to establish between 1000 to 1500 gpm.

13 Verify RHR Pump Cavitation Has Stopped Perform the following:

• Running RHR pump amps - STABLE a. Stop RHR pumps

• RHR flow - STABLE

• RHR pump noise level - NORMAL b. Observe NOTE prior to Step 20 AND go to Step 20.

W97/DHldaj

lSEPj I ACTIONIEXPECTED RESPONSE 1 l RESPONSE NOT OBTAINED l

.a A 14 Establish RHR Flow

a. Check RHR pumps - ALL STOPPED a. Go to Step 15.

b. Verify RCS temperature - LESS THAN b. Observe NOTE prior to Step 20 AND go 3500 F to Step 20.

c. Close RHR Heat Exchanger Outlet Flow valve, HCV-3-758

d. Close RHR Heat Exchanger Bypass Flow valve, FCV-3-605

e. Start previously running RHR pump e. Start alternate RHR pump. IF neither RHR

• pump can be started, THEN observe NOTE prior to Step 20 AND go to Step 20.

f. Establish 1500 gpm RHR flow as follows:

• Slowly open RHR Heat Exchanger Outlet Flow valve, HCV-3-758

g. Check RHR Pump NOT Cavitating g. Return to Step 12.

• Check RHR pump amps - STABLE

• Check RHR flow - STABLE

• RHR pump noise level - NORMAL W97IDH/daj

STUEP llIACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 15 Increase RHR Flow

a. IF RHR flow is less than desired flow, a. IF FCV-3-605 is fully open, THEN open THEN increase RHR Flow 500 gpm from RHR Heat Exchanger Outlet Flow valve, current value as follows: HCV-3-758, to establish desired flow.

• Adjust RHR Heat Exchanger Bypass Flow valve, FCV-3-605

b. Verify RHR pump NOT Cavitating b. Perform the following:

• Running RHR pump amps - STABLE 1) Decrease RHR flow 500 gpm from

• Check RHR flow - STABLE current value.

• RHR pump noise level - NORMAL

2) Return to Step 12.

c. Check RHR flow - GREATER THAN OR c. Perform the following:

EQUAL TO 3000 GPM

• IF RHR flow is less than 3000 gpm BUT increasing, THEN return to Step 15a.

OR

• IF RHR flow can NOT be restored, THEN go to Step 16.

d. Go to Step 18 16 Check If RCS - OPERATING DRAINED Observe NOTE prior to Step 20 AND go to DOWN (LESS THAN 10% COLD CAL PZR Step 20.

LEVEL) 17 Go To 3-ONOP-041.8, SHUTDOWN LOCA

[MODE 5 OR 61 18 Maintain Stable Plant Conditions

a. Verify RCS temperature - STABLE OR a. Perform the following:

DECREASING a Adjust HCV-3-758 to obtain desired cooldown rate.

• Adjust FCV-3-605 to maintain desired RHR flow rate.

b. Verify RCS temperature - LESS THAN b. Observe NOTE prior to Step 20 AND go 200OF OR trending to NPS DESIRED to Step 20.

TEMPERATURE 19 Go to Step 34 W97/DH/daj

STEP l [ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

• NOTE u I RCS heatup rate is required to be monitored by the Shift Engineer or any other available l I operator while efforts continue to restore RHR cooling.

L . . . -. -. . . - - - - - - - - -

20 Continue To Monitor RCS Heatup Rate As Follows:

a. Plot core exit temperature every minute a. IF core exit temperatures are NOT for 5 minutes available, THEN perform the following:

1) Assume a 120 F per minute heatup rate unless the refueling cavity is flooded. IF the refueling cavity is flooded, THEN use 40 F per minute.

2) Go to Step 21.

b. Calculate RCS heatup rate

c. Determine time required to reach saturation in RCS

d. Report results to unit RCO and NPS

e. Repeat this step every 15 minutes until RHR cooling is restored W97/DH/daj

Procedure No.: Procedure

Title:

Page:

Approval Date:

3-ONOP-050 Loss of RHR 10/16/98 Lj I .

ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 21 Isolate Containment If Required

a. Direct appropriate personnel to close any open containment penetrations:

• Equipment hatch

• Airlocks

• Refueling transfer tube

• Any other openings

b. Direct personnel to stop work on all RCS openings

c. Check RCS temperature - GREATER c. WHEN RCS temperature is greater than THAN 200OF 2000 F, THEN do Steps 21d, 21e, 21f, 21g and 21h. Continue with Step 22.

d. Announce over the plant PA system: d. Request NPS pass supervisory announcement over radio to order

• Attention all personnel Inside personnel out of containment.

Unit 3 Containment Evacuate Unit 3 Containment

e. Actuate Containment Evacuation Alarm

f. Announce over the plant PA system: f. Request NPS pass supervisory announcement over radio to order

• Attention all personnel Inside personnel out of containment.

Unit 3 Containment Evacuate Unit 3 Containment

g. Actuate Containment Isolation Phase A:

1) Manually actuate containment isolation phase A

2) IF any containment isolation phase A

2) Containment isolation phase A valve valve is NOT closed, THEN manually white lights on VPB - ALL BRIGHT close valve. IF valve(s) can NOT be manually closed, THEN manually or locally isolate affected containment penetration.

h. Reset Phase A Containment Isolation W97/DH/daj

STEP lACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 22 Establish Secondary Heat Sink

a. Verify at least two S/Gs available: a. Perform the following:

• Secondary side manways - 1) IF RCS temperature is decreasing, INSTALLED THEN continue efforts to restore RHR

• S/G hot leg manway - INSTALLED cooling AND go to Step 32.

• LI-3-462 -GREATER THAN 10%

• RCS Loops - FILLED 2) IF RCS temperature is increasing, THEN go to 3-ONOP-041.8, SHUTDOWN LOCA [MODE 5 OR 6].

b. Establish SIG makeup to the available b. Perform the following:

S/Gs using one of the following methods:

1) IF RCS temperature is decreasing,

• Start a standby feedwater pump using THEN continue efforts to restore RHR O-OP-074.1, STANDBY STEAM cooling AND go to Step 32.

GENERATOR FEEDWATER SYSTEM 2) IF RCS temperature is increasing, THEN go to 3-ONOP-041.8, OR SHUTDOWN LOCA [MODE 5 OR 6].

• Start a condensate pump using 3-OP-073, CONDENSATE SYSTEM OR

• Start a condensate transfer pump aligned to S/G fill line

c. Open available S/G Steam dump to atmosphere valves as necessary to maintain desired RCS temperatures W97IDH/daj

Procedure No.: Procedure

Title:

Page: l Approval Date:

3-ONOP-050 Loss of RHR 10/16/98 I STEP I ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED MMM I

• NOTE l The effectiveness of steaming the available SIGs may NOT be readily apparent during l

• natural circulation. Plant conditions should be allowed to stabilize prior to performing

• Step 23.

23 Determine If Blowdown Should Be Established

a. Core exit temperatures - INCREASING a. Go to Step 26

b. Available SIG steam dump to atmosphere b. Open available SIG Steam dump to valves - FULL OPEN atmosphere valves as necessary to maintain desired RCS temperatures. IF RCS temperature can be controlled using steam dump to atmosphere valves, THEN go to Step 26.

W97/DH/daj

STEP ACTIONIEXPECTED RESPONSE l RESPONSE NOT OBTAINED 24 Align Blowdown From Avallable SIG(s)

a. Verify RE-19, S/G Blowdown Radiation a. Direct Nuclear Chemistry to sample Monitor, - INSERVICE available S/G(s) for activity.

b. Prepare for blowdown: b. Go to Step 26.

1) Place blowdown keylock switch(s) for available S/G(s) in DRAIN/FILL position:

• HS-3-1427X for SIG A

• HS-3-1426X for S/G B

• HS-3-1425X for S/G C

2) Verify S/G Liquid Sample valve(s) on available S/G(s) - OPEN:

• MOV-3-1427 for S/G A

• MOV-3-1426 for S/G B

• MOV-3-1425 for S/G C

3) Verify Blowdown Flow valves -

CLOSED:

• FCV-3-6278A

• FCV-3-6278B

• FCV-3-6278C

4) Locally close S/G blowdown Manual Containment Isolation valve(s) on available S/G(s):

• SGB-3-007 for S/G A

• SGB-3-008 for S/G B

• SGB-3-009 for S/G C

5) Open Blowdown Containment Isolation valve(s) on available S/G(s):

• CV-3-6275A for S/G A

• CV-3-6275B for S/G B

• CV-3-6275C for SIG C

6) Locally open S/G Blowdown Manual Containment Isolation valve(s) on available S/G(s):

• SGB-3-007 for S/G A

• SGB-3-008 for S/G B

• SGB-3-009 for S/G C W971DHldaj

STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED OeR __

Lo Establish Blowdown From Available SIG(s)

a. Align blowdown to discharge canal:

1) Open Blowdown Tank Vent To Atmosphere, CV-3-6267A

2) Close Blowdown Tank Vent To Feedwater Heaters, CV-3-6267B

3) Close Blowdown Tank To Condenser, HIS-3-6265A

4) Open Blowdown Tank to Canal, HIS-3-6265B

b. Locally throttle open Blowdown Flow Valve on available S/G(s) to obtain maximum flow

• FCV-3-6278A for S/G A

• FCV-3-6278B for S/G B

• FCV-3-6278C for S/G C 26 Maintain Level In Available SIG(s)

a. Check narrow range levels - GREATER a. Increase S/G makeup to available S/G(s).

THAN 6%

b. Continue SIG makeup to maintain narrow range level between 6% and 50%

W97/DHLdaj

STp1I ACTION/EXPECTED RESPONSE l~j I I RESPONSE NOT OBTAINEDl 27 Determine If One RCP Should Be Started

a. RCS - INTACT a. Go to Step 32

b. Verify RCS Cold Leg Temperature - b. Perform the following:

GREATER THAN 2750 F

1) Locally obtain S/G secondary temperature measurements. Refer to 3-OP-041.1, Reactor Coolant Pump, for methods of obtaining S/G temperatures.

2) IF any S/G secondary water temperature is greater than 10OF above any RCS cold leg temperature, THEN go to Step 32.

28 Check Plant Conditions For Starting Perform the following:

Desired RCP

1. Verify natural circulation using

a. A or B 4KV bus - ENERGIZED FROM ATTACHMENT 1. IF natural circulation STARTUP TRANSFORMER can NOT be verified, THEN increase dumping steam.

b. Number one seal AP - GREATER THAN 200 PSID 2. Go to Step 32.

c. Thermal barrier AP - GREATER THAN 0 INCHES OF WATER

d. Verify proper number one seal leak-off flow - GREATER THAN 0.8 GPM

e. RCP number one seal leak-off temperature - LESS THAN 225 0F W97/DH/daj

I STEP ACTION/EXPECTED RESPONSE IRESPONSE NOTOBTAINED CAUTION CCW System load requirements of 3-OP-030, COMPONENT COOLING WATER SYSTEM, shall NOT be exceeded.

29 Maintain Proper CCW System Alignment For RCP Operation

a. CCW Heat Exchangers - THREE IN a. Perform the following:

SERVICE

1) Start or stop CCW pumps as necessary to establish ONLY ONE RUNNING CCW PUMP.

2) IF MOV-3-749A and MOV-3-749B are open, THEN stop and place in PULL-TO-LOCK all except one running CCW pump.

3) Go to Step 29c.

b. CCW pumps - ONLY TWO RUNNING b. Start or stop CCW pumps as necessary to establish ONLY TWO RUNNING CCW PUMPS.

c. Check CCW from RHR Heat Exchangers c. Perform the following:

- AT LEAST ONE CLOSED

1) Isolate one Emergency Containment

• MOV-3-749A Cooler by placing one ECC Control

• MOV-3-749B Switch in STOP, then go to Step 29d.

2) IF unable to isolate one ECC, THEN stop all RCP's AND verify natural circulation using ATTACHMENT 1.

Go to Step 32.

d. Verify B CCW header flow - NORMAL d. Perform the following:

1) Verify natural circulation using ATTACHMENT 1. IF natural circulation can NOT be verified, THEN increase dumping steam.

2) Go to Step 32.

W97/DHldaj

ACTION/EXPECTED ISTEP RESPONSE RESPONSE NOT OBTAINED 30 Establish Proper CCW Valve Alignment For RCP Operation

a. RCP Thermal Barrier CCW Outlet, a. IF containment isolation phase B NOT MOV-3-626 - OPEN actuated, CCW radiation levels are normal, and RCP number one seal leak-off temperature is less than 2251F, THEN manually open MOV-3-626. IF MOV-3-626 can NOT be manually opened, THEN direct the operator to locally open MOV-3-626.

b. Verify the following valves - OPEN b. IF containment isolation phase B NOT actuated, THEN manually open MOV(s).

• MOV-3-716A, RCP CCW Inlet IF MOV(s) can NOT be manually opened,

• MOV-3-716B, RCP CCW Inlet THEN direct operator to locally open

• MOV-3-730, RCP Bearing CCW MOV(s).

Outlet

c. Open CCW To Normal Containment Cooler valves

• MOV-3-1417

• MOV-3-1418

d. Reset and start normal containment coolers W97/DHldaj

Procedure No.: Procedure

Title:

Page: 25 I OOApproval Date: 1 3-ONOP-050 Loss of RHR 10/16198C j STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

• NOTE fIf possible, RCPs B or C should be run to provide normal PZR spray. RCP A does NOT l

• provide adequate spray flow.

L . . . . . . -- - . . . - . . . . . . .

31 Try To Start One RCP

a. Start oil lift pump

b. Check that the oil lift pump has been b. WHEN 2 minute oil lift pressure time delay running - AT LEAST 2 MINUTES is satisfied, THEN verify Permissive To Start light ON AND perform Steps 31c, 31d and 31e. Continue with Step 32.

c. Start one RCP c. Perform the following:

1) Verify natural circulation using ATTACHMENT 1. IF natural circulation can NOT be verified, THEN increase dumping steam.

2) Stop oil lift pumps.

3) Go to Step 32.

d. Check that the RCP has been running - d. WHEN RCP has been running greater GREATER THAN 1 MINUTE than 1 minute, THEN stop oil lift pump AND continue at Step 32.

e. Stop the oil lift pump 32 Maintain Stable Plant Conditions:

a. Maintain PZR pressure - STABLE

b. Maintain PZR level - STABLE b. IF PZR level can NOT be maintained, FHEN perform 3-ONOP-041.3, EXCESSIVE REACTOR COOLANT SYSTEM LEAKAGE while continuing with this procedure.

c. Maintain intact S/G narrow range levels -

STABLE

d. Maintain RCS average temperature -

STABLE AT DESIRED TEMPERATURE W97IDH/daj

I STEP I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I 33 Verify RHR Flow Restored Return to Step 14.

34 Go To Appropriate Plant Procedure As Determined By The Nuclear Plant Supervisor END OF TEXT W97/DHldaj

ATTACHMENT I (Page 1 of 1)

NATURAL CIRCULATION INDICATIONS The following conditions support or indicate natural circulation flow:

• RCS subcooling based on core exit TCs - GREATER THAN 300 F[210 0F]

• S/G pressures - STABLE OR DECREASING

• RCS hot leg temperatures - STABLE OR DECREASING

• Core exit TCs - STABLE OR DECREASING

• RCS cold leg temperatures - WITHIN 350 F OF SATURATION TEMPERATURE FOR S/G PRESSURE FINAL PAGE W97/D1ldaj