ML17311A067

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Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17311A067
Person / Time
Site: Ginna Constellation icon.png
Issue date: 01/14/1999
From: Widay J
ROCHESTER GAS & ELECTRIC CORP.
To: Vissing G
NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17265A517 List:
References
NUDOCS 9901220051
Download: ML17311A067 (33)
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Text

CATEGORY 1 REGULA'RY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9901220051 DOC.DATE: 99/01/14 NOTARIZED: NO DOCKET I FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME . AUTHOR AFFILIATION WIDAY,J.A.

RECXP.NAME Rochester Gas S Electric Corp.

RECIPIENT AFFILIATION pm~

VISSING,G.S.

SUBJECT:

Forwards revised Emergency Operating Procedures for plant.

DISTRIBUTION CODE: A002D COPIES RECEIVED:LTR ENCL SXZE:

TITLE: OR Submittal:Inadequate Core Cooling (Item II.F.2) GL 82-2 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72) . 05000244~

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 1 VISSING,G. 1 1 INTERNA LE CE T R 0 1 1 NRR/DSSA/SRXB 1 1 RES DE 1 1 RES/DET/EIB 1 1 RES/DSIR/RPSB 1 1 EXTERNAL: NOAC NRC PDR 1 1 0

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 9 ENCL 9

ROCHESTERGASANDELECTRICCORPORATION o 89EASTAVENUE ROCHESTER, N.K 14649<001 8F:

JOSEPH A. WIDAY TELEPHONE Plant Managor AAEACODE 716 546.2700 Ghna Nucloar Plant January 14, 1999 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555

Subject:

Emergency Operating Procedures R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Vissing:

As requested, enclosed are Ginna Station Emergency Operating Procedures.

Very truly yours, a.d oseph A. Wida JAW/jdw XC: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Ginna USNRC Senior Resident Inspector Enclosure(s):

Zyoo82 AP Index ATT-5.2, Rev 3 ATT- Index ATT-8.0, Rev 5 E Index ATT-14.6, Rev 1 ECA Index E-1, Rev 17.

ES Index ECA-1.1, Rev 16 FR Index ES-1.3, Rev 26 AP-SW.1, Rev 14 FR-Z.2, Rev 4 9'F01220051 9'PQ1.14

,PDR F - " '",

ADOCK 05000244 PDR

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REPORT NO. 01 GINNA NUCLEAR POWER PLANT 12/14/98 PAGE: 1 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRAP ABNORMAL PROCEDURE P ERS: DOC TYPES - PRAP PRE PRECA PRER PRES STATUS: EF QU 5 YEARS ONLY:

PROCEDUR EFFECT LAST NEXT NUYBER PROCEDURE TITLE DATE REVIEW REVIEW ST AP-CCW.1 ~E IÃIO THE COMPONENT COOLING LOOP 013 10/30/98 05/01/98 05/01/03 EF AP-CCW.2 SS OF CCW DURING POWER OPERATION 012 02/24/96 08/30/94 08/30/99 EF AP-CCW.3 LOSS 0 PLANT SHUTDOWN 010 03/29/96 08/30/94 08/30/99 EF AP-CR.1 CONTROL ROOM SIBILITY 015 01/26/98 11/17/94 11/17/99 EF AP-CVCS.1 CVCS LEAK 012 05/01/98 05/01/98 05/01/03+F AP-CW.1 LOSS OF A CIRC WATER PUMP 010 07/16/98 05/01/98 05/0 03 EF AP-ELEC.1 LOSS OF 12A AND/OR 128 BUSSES 016 05/01/98 05/01/98 5/01/03 EF AP-ELEC.2 SAFEGUARD BUSSES LOW VOLTAGE OR SYSTEM LOW FR UENCY 008 01/26/98 02/11/ci4 02/11/99 EF AP-ELEC.3 LOSS OF 12A A%)/OR 12B TRANSFORMER (BELOW 350 F) 005 05/01/98 05/ /98 05/01/03 EF AP-ELEC. 14/16 LOSS OF SAFEGUARDS BUS li/16 000 06/09/97 6/09/97 06/09/02 EF AP-ELEC.17/18 LOSS OF SAFEGUARDS BUS 11/18 001 02/27/ 8 06/09/97 06/09/02 EP AP-FW. 1 PARTIAL OR COMPLETE LOSS OF PAIN FEEDWATER 011 02 7/98 02/27/98 02/27/03 EF AP- IA. 1 LOSS OF INSTRUMENT AIR 01 05/01/98 05/01/98 05/01/03 EF AP-PRZR.1 ABNORMAL PRESSURIZER PRESSURE 09 0 3/96 09/29/94 09/29/99 EF AP-RCC.1 CONTINUOUS CONTROL ROD WITHDRAWAL/INSERTION 006 02/24/9 05/14/98 05/14/03 EF AP-RCC.2 RCC/RPI MALFUNCTION 008 11/16/98 02/0 97 02/06/02 EF AP RCC.3 DROPPED ROD RECOVERY 004 11/16/98 02/27/98 /27/03 EF AP-RCP.1 RCP SEAL MALFUNCTION 012 05/01/98 05/01/98 05/0 03 EF AP-RCS.1 REACTOR COOLANT LEAK 013 05/01/98 05/01/98 05/01/03 AP-RCS.2 LOSS OF REACTOR COOLANT FLOW 010 12/14/98 05/01/98 05/01/03 EF AP-RCS.3 HIGH REACTOR COOLANT ACTIVITY 007 08/05/97 08/05/97 08/05/02 EF AP-RCS.4 SHUTDOWN LOCA 009 05/01/98 05/01/98 05/01/03 EF AP-RHR.1 LOSS OF RHR 012 05/01/98 05/01/98 05/01/03 EF AP-RHR.2 LOSS OF RHR WHILE OPERATIN T RCS REDUCED INVENTORY CONDITIONS 007 05/15/97 03/21/95 03/21/00 EF

REPORT NO. 01 GINNA NUCLEAR POWER PLANT 12/14/98 PAGE: 2 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRAP ABNORMAL PROCEDURE PARAMETERS: DOC TYPES - PRAP PRE PRECA PRER PRES STATUS: EF QU 5 YEARS ONLY:

PROCEDURE EFFECT LAST NEXT NUMBER PROCEDURE TITLE DATE REVIEW REVIEW ST AP SW. 1 SERVICE WATER LEAK 013 08/24/98 06/03/98 06/03/03 EF AP TURB.1 TURBINE TRIP WITWOlTT RX TRIP REQUIRED 009 10/10/97 10/10/97 10/10/02 EF AP-TURB. 2 TURBINE LOAD REJECTION 016 02/27/98 05/13/98 05/13/03 EF AP-TURB. 3 TURBINE VIBRATION 008 12/04/96 02/10/98 02/10/03 EF AP-TURB. 4 LOSS OF CONDENSER VACUUM 014 05/01/98 05/01/98 05/01/03 EF AP-TURB.5 RAPID LOAD REDUCTION 003 02/27/98 07/10/95 07/10/00 EF TOTAL FOR PRAP 30

REPORT NO. 01 GINNA NUCLEAR POWER PLANT 10/30/98 PAGB: 1 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRATT EOP ATTACHMENTS PARAMETERS: DOC TYPES - PRATT 5 YEARS ONLY:

PROCEDURE LAST EFFBCI'ATE NUMBER PROCEDURE TITLB REVI EW ATT-1. 0 ATTACHMENT AT POWER CCW ALIGNMENT 001 07/26/94 02/10/98 02/10/03 EF ATT-2. 0 ATTACHMENT AUX BLDG SW 06/26/98 ATT-2. 1 ATI'ACHMENT MIN SW 004 06/26/98 02/10/98 02/10/03 BF ATT-2. 2 ATTACHMENT SW ISOLATION 005 10/30/98 08/11/98 08/11/03 EP ATT-2. 3 ATTACHMENT SW LOADS IN CNMT 003 01/25/95 01/25/95 01/25/00 EP ATT-3. 0 ATTACHMENT CI/CV I 004 02/27/98 02/11/94 02/11/99 BP ATT-3. 1 ATTACHMENT CNMT CLOSURE 002 07/26/94 02/ll/94 02/11/99 EP ATT-4. 0 ATTACHMENT CNMT RECIRC PANS 003 07/26/94 05/13/98 05/13/03 EF ATT-5. 0 ATTACHMENT COND TO S/G 004 01/25/95 01/25/95 01/25/00 EP ATI'-5. 1 ATTACHMENT SAFW 006 07/07/98 11/08/94 11/08/99 EF ATT-5. 2 ATTACHMENT PIRE WATER COOLING TO TDAFW PUMP 002 10/26/95 03/04/94 03/04/99 BF ATT-6. 0 ATTACHMENT COND VACUUM 003 12/18/96 02/10/98 02/10/03 EF ATT-7. 0 ATTACHMENT CR EVAC 004 05/04/98 02/10/98 02/10/03 EF ATT-8. 0 ATTACHMENT DC LOADS 004 06/03/98 01/26/94 01/26/99 EF ATT-8. 1 ATTACHMENT D/G STOP 004 11/03/95 02/10/98 02/10/03 EF ATT-8. 2 ATTACHMENT GEN DEGAS 005 07/26/94 02/11/94 02/11/99 EF ATT-8. 3 ATTACHMENT NONVITAL 003 07/26/94 02/10/98 02/10/03 EF ATT-8. 4 ATI'ACHMENT S I/UV 004 04/24/97 02/10/98 02/10/03 EF ATT-9. 0 ATTACI94ENT LETDOWN 006 04/07/97 01/26/94 01/26/99 EF ATT-9. 1 ATTACHMENT BXCESS L/D 002 07/26/94 02/10/98 02/10/03 EF ATT-10.0 ATTACHMENT PAULTED S/G 005 10/03/96 05/13/98 05/13/03 EF ATT-11. 0 ATTACHMENT IA CONCERNS 002 04/07/97 08/11/98 08/11/03 EF ATT-11. 1 ATTACHMENT IA SUPPLY 002 04/07/97 08/11/98 08/11/03 EF ATT-11.2 ATTACHMENT DIESEL AIR COMPRESSOR 000 04/03/98 04/03/98 04/03/03 EF

I REPORT NO. 01 GINNA NUCLEAR POWER PLANT 10/30/98 PAGE: 2 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRATT EOP ATTACHMENTS PARAMETERS: DOC TYPES - PRATT 5 YEARS ONLY:

PROCEDURE EFFECT . LAST NUMBER PROCEDURE TITLE DATE REVIEW ATT-12.0 ATTACHMENT N2 PORVS 003 03/24/97 02/10/98 02/10/03 EP ATT-13.0 ATTACHMENT NC 002 07/26/94 02/10/98 02/10/03 EF ATT-14.0 ATTACHMENT NORMAL RHR COOLING 002 04/07/97 10/19/94 10/19/99 EF ATT-14. 1 ATTACHMENT RHR COOL 004 05/01/98 05/01/98 05/01/03 EP ATT-14.2 ATTACHMENT RHR ISOL 001 07/26/94 02/10/98 02/10/03 EP ATT-14.3 ATTACHMENT RHR NPSH 002 08/01/97 01/26/94 01/26/99 EF ATT-14.4 ATTACHMENT RHR SAMPLE 001 07/26/94 01/26/94 01/26/99 EF ATT-14.5 ATTACHMENT RHR SYSTEM 002 07/26/94 02/10/98 02/10/03 EF ATT-14. 6 ATTACHMENT RHR PRESS REDUCTION 000 04/07/94 04/07/94 04/07/99 EP ATT-15.0 ATI'ACHMENT RCP START 005 05/22/97 04/20/95 04/20/00 EF ATT-15.1 ATI'ACHMENT RCP DIAGNOSTICS 003 04/24/97 02/10/98 02/10/03 EF ATT-15.2 ATTACHMENT SEAL COOLING 003 05/22/97 02/10/98 02/10/03 EP ATT-16 ' ATTACHMENT RUPTURED S/G 008 03/17/98 11/08/94 11/08/99 EF ATT-17.0 ATTACHMENT SD-1 006 11/03/95 02/03/95 02/03/00 EP ATT-17. 1 ATTACK48NT SD-2 005 09/26/96 01/26/94 01/26/99 EF ATT-18. 0 ATTACHMENT SFP - RWST 004 10/08/97 02/10/98 02/10/03 EF ATT-19.0 ATTACHMENT SI FLUSH XX 01/25/95 DE ATT-20.0 ATTACHMENT VENT TIME 003 07/26/94 02/10/98 02/10/03 EP ATT-21.0 ATTACHMENT RCS ISOLATION 001 07/26/94 02/10/98 02/10/03 EP ATT-22.0 ATTACHMENT RESTORING FEED FLOW 000 03/24/97 03/24/97 03/24/02 EF TOTAL FOR PRATT 44

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0

REPORT NO. 01 GINNA NUCLEAR POWER PLANT 12/14/98 PAGE: 3 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRE EMERGENCY PROCEDURE PARAMEIERS: DOC TYPES - PRAP PRE PRECA PRER PRES STATUS: EF QU 5 YEARS ONLY:

PROCEDURE EFFECT LAST NEXT NUMBER PROCEDURB TITLE DATE REVIEW REVIEW ST E-O REACIOR TRIP OR SAFETY INJECTION 024 05/01/98 05/01/98 05/01/03 EF E-1 LOSS OF REACTOR OR SECONDARY COOLANT 016 05/01/98 05/01/98 05/01/03 EF E 2 FAULTED STEAM GENERATOR ISOLATION 008 05/01/98 05/01/98 05/01/03 EF E-3 STEAM GENERATOR TUBE RUPTURE 023 12/14/98 05/01/98 05/01/03 EF TOTAL FOR PRE

REPORT NO. 01 GINNA NUCLEAR POWER PLANT 12/14/98 PAGE: 4 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRECA EMERGENCY CONTINGENCY ACTIONS PROC PARAMETERS: DOC TYPES - PRAP PRE PRECA PRER PRES STATUS: EF 5 YEARS ONLY:

QU PROCEDURE EFFECI'EV LAST NEXT NUMBER PROCEDURE TITLE DATE REVIEW REVIEW ST ECA-0. 0 LOSS OF ALL AC POWER 019 12/14/98 05/01/98 05/01/03 EF ECA-0. 1 LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED 06/26/98 015 05/01/98 05/01/03 EF ECA 0 ' LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED 011 12/14/98 05/01/98 05/01/03 EF ECA-1. 1 LOSS OF EMERGENCY COOLANT RECIRCULATION 015 05/01/98 05/01/98 05/01/03 EF ECA-1.2 LOCA OUTSIDE CONTAINMENT 05/01/98 005 05/01/98 05/01/03 EF ECA 2.1 UNCONTROLLED DEPRESSURIZATION OF BOTH STEAM GENERATORS 011 06/26/98 05/01/98 05/01/03 EF ECA 3.1 SGTR WITH LOSS OF REACfOR COOLANT-SUBCOOLED RECOVERY DESIRED 016 06/26/98 05/01/98 05/01/03 EF ECA 3.2 SGTR 'WITH LOSS OF RFACfOR COOLANT. SATURATED RECOVERY DESIRED 020 06/26/98 05/01/98 05/01/03 EF ECA-3.3 SGTR WITHOUT PRESSURIZER PRESSURF. CONTROL 020 06/26/98 05/01/98 05/01/03 EF TOTAL FOR PRECA

'I II

REPORT NO. 01 GINNA NUCLEAR POWER PLANT 12/14/98 PAGE: 7 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRES EQUIPMENT SUB- PROCEDURE ~

PARAMETERS: DOC TYPES PRAP PRE PRECA PRER STATUS: EF QU 5 YEARS ONLY:

PROCEDURE NEXT EFFECI'ATE NUMBER PROCEDURE TITLE REV REVIEW ST ES-0.0 REDIAGNOSIS 010 05/01/98 05/01/98 05/01/03 EF ES-0. 1 REACIOR TRIP RESPONSE 015 12/14/98 05/01/98 05/01/03 EF ES-0.2 NATURAL CIRCULATION COOLDOWN 012 05/01/98 05/01/98 05/01/03 EF ES 0.3 NATURAL CIRCUIATION COOLDOWN WITH STEAM VOID IN VESSEL 008 05/01/98 05/01/98 05/01/03 EF ES-1.1 SI TERMINATION 015 06/26/98 05/01/98 05/01/03 EF ES-1.2 POST LOCA COOLDOWN AND DEPRFSSUR IZATION 018 06/26/98 05/01/98 05/01/03 EF ES-1.3 TRANSFER TO COLD LEG RECIRCUIJLTION 025 06/26/98 05/01/98 05/01/03 EF ES- 3. I POST.SGTR COOLDOwN USING BACKFILL 013 05/01/98 05/01/98 05/01/03 EF I

ES-3.2 POST-54'%OO!I~ USIN> BI~MWN 014 05/01/98 05/01/98 05/01/03 EF ES-3.3 POST ~ SGTR COOIXKWN USIN'I STFW'4 O'WP 014 05/01/98 05/01/98 05/01/03 EF TOTAL FOR PRES 10

REPORT NO. 01 GlNNA NUCLEAR POWER PLANT 12/14/98 PAGE: 1 REPORT: NPSP0200 PROCEDURES INDEX DOC TYPE: PRFR FUNCI'IONAL RESTORATION GUIDELINE PROC PARAMETERS: DOC TYPES - PRFR STATUS: EF QU 5 YEARS ONLY:

PROCEDURE EFFECT LAST NEXT NUMBER PROCEDURE TITLE REV DATE REVIEW REVIEW ST FR-C.1 RESPONSE TO INADEQUATE CORE COOLING 016 12/14/98 05/01/98 05/01/03 EF FR-C.2 RESPONSE TO DEGRADED CORE COOI ING 014 05/01/98 05/01/98 05/01/03 EF PR-C.3 RESPONSE TO SATURATED CORE COOI ING 008 05/01/98 05/01/98 05/01/03 EF FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK 020 06/26/98 05/01/98 05/01/03 EF FR-H.2 RESPONSE TO STEAM GENERATOR OVERPRESSURE 004 05/01/98 05/01/98 05/01/03 EF FR-H. 3 RESPONSE TO STEAM GENERATOR HIGH LEVEL 005 05/01/98 05/01/98 05/01/03 EF FR H.4 RESPONSE TO LOSS OF NORMAL STEAM REI EASE CAPABII ITIES 004 05/01/98 05/01/98 05/01/03 EF FR-H. 5 RESPONSE TO STEAM GENERATOR LOW LEVEL 004 05/01/98 05/01/98 05/01/03 EF FR-I. 1 RESPONSE TO HIGH PRESSURIZER l,EVEi. 009 06/26/98 05/01/98 05/01/03 EF FR-I.2 RESPONSE TO LOW PRESSURIZER LEVEl. 006 11/16/98 05/01/98 05/01/0) EF FR-I.3 RESPONSE TO VOIDS IN REACTOR VESSEl 011 06/26/98 05/01/98 05/01/03 EF FR-P. 1 RESPONSE TO IMM!NENI'RESSURIZED THERMAI, SHOCK CONDITION 019 12/14/98 05/01/98 05/OI/OI EF FR-P. 2 RESPONSE TO ANTICIPATED PRESSUR!ZED THERMAL Sl(OCK CONDITION 007 05/01/98 05/OI/98 05/01/OI EF FR-S. 1 RESPONSE TO REACTOR RESTART/ATWS 013 12/14/98 05/OI/98 05/01/03 EF FR-S.2 RESPONSE TO ASS OF CORE SHUIDOWN 008 05/01/98 05/01/98 05/01/03 EF FR-Z. 1 RESPONSE TO HIGH CONTAINMENI'RESSURE 005 12/14/98 05/01/98 05/01/03 EF FR-Z. 2 RESPONSE TO CONTAINMENT FIDODING 003 05/01/98 05/01/98 05/01/03 EF FR-Z. 3 RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL 004 05/01/98 05/01/98 05/01/03 EF TOTAL FOR PRFR 18

EOP TITLE:

REV: 13 AP-SW.1 SERVICE WATER LEAK PAGE 1 of 10 ROCHESTER GAS AND ELECTRIC CORPORATION GINNA STATION CONTROLLED COPY NUMBER kL~

RESPONSIBLE NAGER F-EFFECTIVE DATE CATEGORY 1.0 REVIEWED BY:

EOP TITLE:

REV: 13 AP-SW.1 SERVICE WATER LEAK PAGE 2 of 10 A. PURPOSE This procedure provides the necessary instructions to respond to a service water system leak.

B. ENTRY CONDITIONS/SYMPTOMS

1. SYMPTOMS The symptoms of SERVICE WATER LEAK are:
a. Service water header pressure low alarms on comput: er, or
b. Sump pump activity increases in containment, the AUX BLDG, or INT BLDG, OR
c. Unexplained increase in the waste hold-up tank, or
d. Visual observation of a SW leak, or
e. Annunciator C-2, CONTAINMENT RECIRC CLRS WATER OUTLET HI TEMP 217', lit, or
f. Annunciator C-10, CONTAINMENT RECIRC CLRS WATER OUTLET LO FLOW 920 GPM, lit, or
g. Annunciator E-31, CONTAINMENT RECIRC FAN CONDENSATE HI-HI LEVEL alarm, exhibits an unexplained increase in frequency, or
h. Annunciator H-G, CCW SERVICE WATER LO FLOW 1000 GPM, lit.

fOP: TITLE:

REV: 13 AP-SW.1 SERVICE WATER LEAK PAGE 3 of 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

  • 4 * *
  • 1 *
  • 0 4
  • 1 1 ~ 1 t
  • 1
  • 1 I 4 ~ t 4 ~ ~ 4 1 4 1 1 t
  • 4 4 t 1 I
  • CAUTION o IF. AT ANY TIME DURING THIS PROCEDURE. A REACTOR TRIP OR SI OCCURS. E-O, REACTOR TRIP OR SAFETY INJECTION, SHALL BE PERFORMED.

o IF EITHER D/G RUNNING WITHOUT SW COOLING AVAILABLE, THEN STOP THE AFFECTED D/G TO PREVENT OVERHEATING.

1 Verify 480V AC Emergency Ensure associated D/G(s) running Busses 17 and 18 ENERGIZED and attempt to manually load busses 17 and/or 18 onto the D/G(s) if necessary.

2 Verify At, Least One SW Pump IF a SW pump has tripped. THEN Running In Each Loop: ensure other pump in the affected loop is running.

~ A or B pump in loop A

~ C or D pump in loop B

EOP: TITLE:

REV 13 AP-SW. 1 SERVICE WATER LEAK PAGE 4 of 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: Abnormally low pressure in either SW loop may indicate that the idle pump check valve is open. This may be corrected by restarting or isolating the idle pump.

3 Check SW System Status:

a. Check SW loop header pressures: a. IF three SW pumps operating and either loop pressure less than o Pressure in both loops 40 psig, THEN trip the reactor APPROXIMATELY EQUAL and go to E-O. REACTOR TRIP OR SAFETY INJECTION.

o PPCS SW low pressure alarm status - NOT LOW IF only two SW pumps operating and either loop pressure less o Pressure in both loops- than 45 psig. THEN start one STABLE OR INCREASING additional SW pump (243 kw each pump).

b. Check SW loop header pressures b. IF either SW loop pressure is GREATER THAN 55 PSIG less than 55 PSIG with three SW pumps running AND cause can NOT be corrected, THEN initiate a controlled shutdown while continuing with this procedure (Refer to 0-2.1, NORMAL SHUTDOWN TO HOT SHUTDOWN).

EOP TITLE:

REV: 13 AP-SW.1 SERVICE WATER LEAK PAGE 5 of 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: o If SW is lost to any safeguards equipment, the affected component should be declared inoperable and appropriate actions taken as required by ITS, Section 3.

o CNMT sump A level of 10 feet is approximately 6 feet 6 inches below the bottom of the reactor vessel.

4 Check For SW Leakage In CNMT:

a. Check Sump A indication a. IF the SW leak is NOT in the CNMT, THEN go to Step 6.

o Sump A level - INCREASING

-OR-o Sump A pump start frequency-INCREASING (Refer to RCS Daily Leakage Log)

b. Evaluate Sump A conditions: b. Plant shutdown should be considered, consult plant staff.
1) Verify Leakage within capacity of one Sump A pump (50 gpm)
2) Check Sump A level - LESS THAN 10 FEET
c. Direct RP to establish conditions for CNMT entry

EOP: TITLE:

REV: 13 AP-SW.1 SERVICE WATER LEAK PAGE 6 of 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

~ ~ 1 I 1 4 * ~ t 0 ~ I ~ 4 I ~ ~ ~ I 1 4 ~ 0 t ~ ~ 4 1 ~ 4 1 0 I 1 0 t 0 0 0 ~ ~

CAUTION BEFORE ISOLATING SW TO CNMT RECIRC FANS, REFER TO ITS SECTION 3.6.6 FOR OPERABILITY REQUIREMENTS.

NOTE: o One Reactor Compartment cooling fan should be running whenever RCS temperature is greater than 135'F.

o CNMT recirc fan condensate collector level indicators may be helpful in identifying a leaking fan cooler.

5 Check CNMT fan indications: Dispatch AO to perform Attachment SW LOADS IN CNMT as necessary.

o CNMT recirc fan collector dump WHEN CNMT SW leak location frequency - NORMAL (Refer to RCS identified, THEN go to Step 9.

Daily Leakage Log) o CNMT recirc fan SW flows APPROXIMATELY EQUAL (INTER BLDG basement by IBELIP) o Reactor compartment cooler SW outlet pressures - APPROXIMATELY EQUAL (INTER BLDG SAMPLE HOOD AREA)

~ Cooler A - PI 2232

~ Cooler B - PI 2141

EOP: TITLE:

REV: 13 AP-SW 1 ~ SERVICE WATER LEAK PAGE 7 of 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 6 Dispatch AO To Screenhouse To Perform The Following:

a. Verify idle SW pump check valve a. Notify Control Room of any closed indication of check valve failure.

o Idle pump shaft stopped o Idle pump discharge pressure

- ZERO (unisolate and check local pressure indicator)

b. Investigate for SW leak in b. Perform the following:

Screenhouse - NO EXCESSIVE LEAKAGE INDICATED 1) Identify leak location.

IF increase in leakage from underground header indicated, THEN isolation of header should be considered (Refer to Attachment SW ISOLATION)

2) Notify Control Room of leak location.

NOTE: Refer to Attachment SW ISOLATION for a list of the major non-safeguards loads supplied by each service water header.

7 Check Indications For Leak Dispatch AO to the specific area to Location: investigate for leakage.

o AUX BLDG sump pump start frequency - NORMAL (Refer to RCS Daily Leakage Log) o Annunciator L-9, AUX BLDG SUMP HI LEVEL - EXTINGUISHED o Annunciator L-17. INTER BLDG SUMP HI LEVEL - EXTINGUISHED

~'

EOP: TITLE:

REV 13 AP-SW.1 SERVICE WATER LEAK PAGE 8 of 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 8 Dispatch AO To Locally Investigate For SW Leakage And To Monitor Operating Equipment

~ Turbine BLDG

~ SAPW pump room

EOP: TITLE:

REV: 13 AP-SW.1 SERVICE WATER LEAK PAGE 9 of 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: If SW is lost to either D/G. refer to ER-D/G 2, ALTERNATE COOLING FOR if cooling

~

EMERGENCY D/Gs. is required.

9 Evaluate SW Leak Concerns

a. Check SW pump status - AT LEAST a. IF either SW header pressure THREE PUMPS RUNNING less than 45 psig. THEN start third SW pump.
b. Intact SW loop header pressure- b. Dispatch AO to perform the GREATER THAN 45 PSIG following:
1) Split A and B SW headers:

o Close V-4669 OR V-4760 in B D/G room.

o Close V-4611 OR V-4612 in Screenhouse.

o Close V-4625 OR V-4626 in INT BLDG clean side.

o Close V-4639 OR V-4756 in INT BLDG clean side.

2) IF plant at power. THEN initiate a controlled shutdown (Refer to 0-2.1, NORMAL SHUTDOWN TO HOT SHUTDOWN).
3) Go to Step 10.
c. Verify leak location - IDENTIFIED c. Return to Step 3.
d. Verify plant operating at power d. Verify SW system conditions appropriate for plant mode (Refer to ITS Section 3.7.8) and go to Step 10.
e. Leak isolation at power e. IF plant shutdown requir'ed. THEN ACCEPTABLE refer to 0-2.1, NORMAL SHUTDOWN TO HOT SHUTDOWN.

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I REV: 13 AP-SW.1 SERVICE WATER LEAK PAGE 10 of 10 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 10 Dispatch AO(s) To Locally Isolate SW Leak As Necessary ll Verify SW Leak Isolated

a. Monitor SW System Operation a. IF SW leak can NOT be isolated within the affected header, THEN o SW loop header pressure stop SW pumps in the affected RESTORED TO PRE-EVENT VALUE loop and go to Step 12.

Archive PPCS point ID loop A P2160 OR loop B P2161) o Both SW loop header pressures STABLE

b. Verify at least one SW pump b. Refer to ITS Section 3.7.8 for available from each screenhouse limiting conditions for AC Emergency bus operation.

~ Bus 17 SW pumps B or D

~ Bus 18 SW pumps A or C NOTE: o Refer to 0-9.3, NRC IMMEDIATE NOTIFICATION, for reporting requirements.

o An Action Report, per IP-CAP.1. should be submitted for a SW leak in CNMT.

12 Notify Higher Supervision

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REV: 13 AP-SW.1 SERVICE WATER LEAK PAGE 1 of 1 AP-SW.1 APPENDIX LIST TITLE

1) ATTACHMENT SW ISOLATION (ATT-2. 2)
2) ATTACHMENT SW LOADS IN CNMT (ATT-2.3)

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REV: 1 ATT-5. 2 ATTACHMENT FIRE WATER COOLING TO TDAFW PUMP PAGE 1 of 1 Superintendent Date 7-4l -5 To rovide Fire Water Coolin to the TDAFW Pum erform the NOTE: A 12 inch "crescent" wrench will be needed to make the connections.

1. Obtain the hose from the Turbine Bldg. hose locker (top floor KEY 579).
2. CLOSE service water root valve to the TDAFW pump thrust bearing and lube oil cooler V-4087C (located on the south side of the pump below MOV-4013).
3. Check OPEN instrument root valve to PI-2134 SW inlet to TDAFW oil cooler V-4288 (located at the north west corner of the oil sump).
4. Connect the hose between PI-2134 and the Fire Water Main at valve V-9226. (V-9226. is located by the S/G blowdown CNMT Isol Valves)

NOTE: The following action may result in an automatic start of the Diesel Driven Fire Pump.

5. OPEN Containment hose reel supply drain & test connection valve V-9226.
6. VERIFY water i:s being supplied to the TDAFW pump thrust bearing by observing flow past valve V-4089A TDAFW Thrust bearing SW outlet block valve. Verify PI-2134 indicates less than 150 psig.
7. Notify the Control Room the attachment is complete.

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REV 3 ATT-8.0 ATTACHMENT DC LOADS PAGE 1 of 1 Superintendent Date A) The A and B MFP oil pumps should be shed within 15 minutes to reduce battery-load to guarantee design battery life.

The following DC loads could be shed from the A and B DC B) busses, if necessary (obtain DC panel key from SS):

1A MAIN DC DISTRIBUTION PANEL (A battery room):

o Switchgear breaker test cabinet, position 1 o Rod drive MG set control panel, position 4 o Iso phase control, position 5 o Exciter equipment panel, position 6 1B MAIN DC DISTRIBUTION PANEL (B battery room):

o Rod drive MG set control panel, position 3 AUX BLDG DC DIST PANEL 1B (south end bus 16 on column near SI pump recirc valves) ~

o Aux Bldg Heat & Vent cont .panel, position 1 SCREENHOUSE DC DISTRIBUTION PANEL B (screenhouse near MCC G) o Traveling screen control, position 1 TURB BLDG DC DISTRIBUTION PANEL (TURB BLDG basement near fire water storage tank) o Hydrogen panel for gl generator, position 3 (after degassification) o MCC F, position 5 o Water treatment panel, position 6

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