ML16341F484
| ML16341F484 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 12/22/1989 |
| From: | Rood H Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.02, TASK-TM TAC-68346, TAC-68347, NUDOCS 9001030271 | |
| Download: ML16341F484 (86) | |
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50-275 and'50-323 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 December 22, 1989
.LICENSEE:
FACILITY:
PACII.IC GAS AND ELECTRIC COMPANY (PGKE)
DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1
AND 2
SUBJECT:
SUMMARY
OF DECEMBER 5, 1989 PUBLIC MEtTING ON GAGGING ONE OF THE THREE PRESSURIZER SAFETY RELIEF VALVES (SRVs)
AND THE CREATION OF TECHNICAL SPECIFICATIONS COYERING THE STEAM GENERATOR ATMOSPHERIC DUMP VALVES (ADVs) (TAC NUMBERS 68346 AND 68347)
On December 5, 1989 the NRC staff met with PGSE in Rockville, Maryland to
.discuss the above subjects.
Attendees at the meeting are given in Enclosure 1.
The meeting was divided into two parts.
The first part addresse a potential technical specification change to allow plant operation while one of the three pressurizer safety valves is disabled (gagged).
If such a technical specification were approved, the licensee would be allowed to~gag a leaking bRY and continue to operate for the rest of the cycle, rather than being required to shut down and repair or replace the leaking SRV.
The viewgraphs presented by the licensee during this part of the meeting are given 'in Enclosure 2.
At
.tne conclusion of this part of tne meeting, the NRC staff stated that, based on the material presented at the meeting, it had a number of safety concerns about safety valves, including:
(1) what are the uncertainties affecting SKY performance, including the recent setpoint shift reported by Westinghouse?
(2) what is the best way to set SRYs?
(3) how should the overpressure analysis deal with clearing of the loop seals?
(o) how can the loss of loop seal be detected (high tail pipe temperature occurs after the loop seal is lost)?
(5) what is the impact of loop seal loss on the LOCA analysis required to satisfy TMI item II.K.3.2?
the NRC staff stated that until these concerns are resolved, it is premature to consider operation with a gagged SRY, or the relaxation of the SRV setpoint tolerance from plus-or-minus 1 percent to plus-or-minus 3 percent, as proposed by the licensee in another recent technical specification request (I.AR 89-11).
It was noted that a meeting on SRVs between the NRC staff and the Westinghouse Owner's Group would be held on December 8, 1989, and many of the staff's concerns would be discussed at that meeting.
The licensee indicated that its probab'i listic risk assessment (PRA) of Diablo Canyon's compliance with the ATWS rule with a gagged SRY would be submitted to the staff in the near future.
'900i03027i 85'i222
, PDR ADQCK 05000275 P
The second part of the meeting addressed the NRC staff's review of the licensee's steam generator tube rupture (SGTR) analysis, and the need for a technical specsfication to assure ADV operability.
The staff'ad previously requested that the licensee submit such a technical specif)cation, because the ADVs are needed to cope with the SGTR accident.
The viewgraphs presented by the licensee during this part of the meeting are given in Enclosure 3.
The licensee described the ADV system enhancements that will be installed at the Cycle 4 refueling outage for both units.
For Unit 1, this wi 11 be in February
- 1991, and for Unit 2, September 1991.
At the conclusion of the meeting, the licensee stated that it would submit a proposed ADV technical specification by
- February, 1990.
The licensee stated that until an ADV technical specification is approved, ADV operation wi 11 be administratively controlled by procedures similar to the mar ked-up Catawba technical specification presented in the handout.
Enclosures:
1.
Meeting Attendees 2.
Licensee's Viewgraphs, Part I 3.
Licensee's Yiewgraphs, Part II cc: w/enclosures - see next page Harry Roo
, Senior Project Manager project Directorate Y
Division of Reactor Projects - III, IV, V and Special Projects
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J December 22, 19 The second part of the meeting addressed the NRC staff's review of the licensee's steam generator tube rupture (SGTR) analysis, and the need for a technical specification to assure ADV operability.
The staff had previously requested that the licensee submit such a technical specification, because the ADVs are needed zo cope with the SGTR accident.
The viewgraphs presented by the licensee during this part of the meeting are given in Enclosure 3.
The licensee described the ADV system enhancements that will be installed at the Cycle 4 refueling outage for both units.
For Unst I, this will be in February
- 1991, and for Unit 2, September 1991.
At the conclusion of the meeting, the licensee stated that it would submit a proposed ADV technical specification by
- February, 1990.
The licensee stated that until an ADV technical specification is approved, ADV oper ation w111 be administratively controlled by procedures similar to the marked-up Catawba technical specification presented in the handout.
cc: w/enclosures
- see next page DISTRIBUTION NRC and LPDR J. Sniezek PDV Reading G. Knighton H.
Rood OGC E. Jordan R. Jones G. Kelly N. ChoKshi K. Dempsey K. Desai E. Sull>van L. Marsh G.
Hammer ACHS (10)
H. B. Clayton, 17G-21 M. mendonca, Region V
P. Narbut, Region V
DRSP/PD5~
HRood:pml 12/g.i/89 DR
- Pu5 GK ton 12/yy/89
Enclosures:
1.
Meeting Attendees 2.
Licensee's Viewgraphs, Part I 3.
Licensee's Viewgraphs, Part II original signed by Harry Rood Harry Rood, Senior Project Manager Project uirectorate V
Division of Reactor Projects - III, IV, Y and Special Projects
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Mr'. J.
D. Shiffer Pacific Gas and Electric Company Diab lo Canyon CC:
Richard F,. Locke, Esq.
Pacific Gas 8 Electric Company Post Office Box 7442 San Francisco, California 94120 Ms. Sandra A. Silver 660 Granite Creek Road Santa Cruz, California 9b065 Mr. Peter H. Kaufman Deputy Attorney General State of California 110 West A Street, Suite 700 San Diego,.California 92101 Managing Editor The Count Tele ram Tribune P.
Box 112 C airman San Luis Obispo, California 93406 Ms. Nancy Culver 192 Luneta Street San Luis Obispo, California 93401 Regional Administrator, Region Y
U.S. Nuclear Regulatory Commission 1450 Maria Lane Suite 210 Walnut Cree~, California 94596 NRC Resident Inspector Diablo Canyon Nuclear Power Plant c/o U.s. Nuclear Regulatory Commission I. 0.
Box 369 Avila Beach, California 93424 Bruce Norton, Esq.
c/o Richard F. Locke, Esq.
Pacific Gas and Electric Company Post Office Box 7442 San Francisco, California 94120 Or. R. B. Ferguson Sierra Club - Santa Lucia Chapter Roc'anyon Star Route Creston, California 93432 San Luis Obispo County Board of Supervisors Room 270 County Government Center San Luis Obispo, California 93408 Michael M. Strumwasser, Esq.
Special Assistant Attorney General State of California Department of Justice 3580 Wilshire Boulevard, Room BUO Los Angeles, California 90010 Dr. Gerard C. Mong, Chief Radi o log ica 1 Materi a 1 s Control Section State Department of Health Services 714 P Street, Office Building 88 Sacramento, Califor nia 95814
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v' ENCLOSURE I ATTENDEES Public Meeting on Diablo Canyon SRVs and ADVs
- Tuesday, December 5, 1989 Nilesh Chokshi K. Daschke Ken DemPsey Kulin D. Desai John M. Gisclon T. L. Grebel John B. Hoch Doug Holderbaum R.
C. Jones Glenn Kelly George Knignton Tien P.
Lee Barclay S.
Lew Y. Justin Liu Raliegh M. Nakao Meiita Osborne Harry Rood Anthony M. Sicari David t;. Tateosian J.
E. Tomksns Ken Vavrek Robert C.
Webb ORGANIZATION NRC/RES/PRAB Westinghouse (PG&E)
NRC/NRR/EMTB NRC/NRR/SRXB PG&E PG&E PG&E Westinghouse (PG&E)
NRC/NRR/SRXB NRC/NRR/PRAB NRC/NRR/PDV PG&E PG&E PG&E Bechtel (PG&E)
NRC/NRR/PDV Westinghouse (PG&E)
PG&E PG&E Westinghouse (PG&E)
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ENCLOSURE 2
VIEWGRAPHS PRESENTED BY PG&E, PART I Public Meeting on Diablo Canyon Pressurizer Safety Relief Valves
- Tuesday, December 5,
1989
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SAFETY VALVE MEETING DacaMsaR 5,
1989 Room 13B9 file: SV Pocffh Oao end Eloclrh Company a
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AGENDA FOR SAFETY VALVE MEETING 1 ~
STATUS FROM LAST NRC MEKTXNG AND OVKRVXEM OF CURRENT PROPOSAL 2.
STATUS OF IMPROVED VALVE PERFORMANCE EFFORTS PGRE ROOT CAUSE ANALYSXS IMPROVEMENTS 3.
OvERvxEw DF DCPP SPEcxFxe ATWS ANALYSXS PGSE/M 4.
PROPOSED TECH SPEC 5.
RECENT UNXT 1 BEAUMONT TEST RESULTS PGSE 6.
NRC FEEDBACK AND FUTURE ACTXONS NRC/PGLE pecgto Oae an@ Kteelcle Cocnpeny
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0 STATUS FROM LAST MEETING AND OVERVIEW OF CURRENT PROPOSAL 0
AGENDA O
JULY 20, 1989 MEETING WITH THE NRC INITXAL PG&E PROPOSAL FOR DXSABLXNG A SAFETY VALVE PG&E/W SAFETY EVALUATXON VALVE PERFORMANCE IMPROVEMENTS NRC COMMENTS REGARDING ATWS 0
VALVE PERFORMANCE IMPROVEMENTS O
DCPP SPECXFXC ATWS ANALYSIS AND EVALUATION OF DETERMINISTXC SAFETY ANALYSIS REQUIREMENTS O
PROPOSED TECH SPECS WHXCH'OMBXNE OPERABXLXTY OF PORVS AND SAFETY VALVES 0
DISCUSS RECENT UNIT 1 BEAUMONT TEST RESULTS O
0 PacNo Oaa and ElacUto Company
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STATUS OF IMPROVE0 VALVE PERFORMANCE EFFORTS PacNo Oaa and Klaetrh Company
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STATUS OF ROOT CAUSE ANALYSIS 0
PREVIOUS MEETING CITED SEVERAL POSSIBLE ROOT CAUSES FOR THE UNIT 2 MARCH 1989 LEAKAGE FOREIGN MATERIAL QN SEAT THE VALVE AND UPSTREAM PIPING WAS INSPECTED AND FOUND TO BE FREE OF FURTHER FOREIGN MATERIAL NOZZLE LOADS A REVIEW OF THE NOZZLE LOADS FOUND THEM TO BE HIGHER THAN THE CURRENT CROSBY RECOMMENDED MAXIMUMS HYDROGEN BUBBLE WESTINGHOUSE IS CURRENTLY INVESTIGATING THE POSSIBILITY AND EFFECT OF A HYDROGEN BUBBLE FORMING AGAINST THE DISC DUE TO THE PARTICULAR LOOP SEAL GEOMETRY 0
PG&E MILL MONITOR PERFORMANCE OF THE SAFETY VALVES DURING PLANT HEATUP MONITORING TAILPIPE TEMPERATURE MONITOR TAlLPIPE ACOUSTICS MONITOR PRT LEVEL AND TEMPERATURE Paclflo Oaa and Elaetrlc Company
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'ESULTANT CORRECTIVE ACTIONS o
IMPROVE CALIBRATION METHODS ALL VALVES ARE NOW SENT TO THE BEAUMONT TESTING FACILITY TO BE CALIBRATED USING LIVE STEAM WITH WATER FILLED LOOP SEALS OF IDENTICAL GEOMETRY IN AN IDENTICAL VALVE ENVIRONMENT (RECENT UNIT I RESULTS ARE DISCUSSED LATER)
PGRE WILL CONTINUE TO REVISE AND IMPROVE CURRENT SETPOINT TESTING METHODS AS EXPERIENCE AND KNOWLEDGE IS GAINED o
DISABLE VALVE IN CLOSED POSITION o
REVIEW OF POSSIBLE CONFIGURATION CHANGES AND THEIR EFFECTIVENESS IN ELIMINATING PROBLEMS OF LEAKAGE AND SETPOINT DRIFT 0
CONTINUE TO MONITOR WOG EFFORT paclfla Oaa and Etaelrlc Company
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POSSIBLE CONFIGURATION CHANGES UNDER INVESTIGATION o
CROSBY FLEXI-DISC MODIFICATION CHANGE OF DISC DESIGN PERFORMANCE SHOWN ACCEPTABLE IN OVER HALF A DOZEN UPS.
PLANTS o
FRENCH OVER-PRESSURE PROTECTION SYSTEM ELIMINATION OF SEPARATE PORVS VALVES ARE PILOT OPERATED PILOT OPERATED BACKUP STOP VALVE NO SET POINT VARIATION OR VALVE INSTABILITYWHETHER DISCHARGING
- WATER, STEAM OR TWO PHASE FLOW LEAK TIGHT SEAL UP TO OPENING PRESSURE PERFORMANCE SHOWN ACCEPTABLE IN EXTENSIVE EUROPEAN PWR OPERATIONAL EXPERIENCE IMPLEMENTATION REQUIRES ASME CODE CASE, EXTENSIVE DCPP PIPING CHANGES, AND NRC APPROVAL OF SYSTEM DESIGN Paclth Oaa and Khctrlc Company
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OVERVIEW OF DCPP SPECIFIC ATWS ANALYSIS pacl0c Oaa and Electric Company files SV
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ASSESSMENT OF COMPLIANCE WITH ATWS RULE BASIS PRESSURIZER SAFETY VALVE UNAVAILABLE O
ASSESSMENT USED WOG MODEL OF MCAP-11993 0
PR NODE MODXFXED TO REFLECT COMBXNATXONS OF PORVS AND SAFETY VALVES UETS CALCULATED USXNG MOG FEEDBACK MODEL MOG FEEDBACK MODEL BOUNDS DXABLO CANYON 0
DXABLO CANYON "EXTRA" PORV MODELLED MXTH 904 STEAN. DUMP 0
OTHER UNAVAXLABXLXTXESBASED ON DXABLO CANYON PRA taeHto Oaa and Saotrh Company
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ATWS RULE COMPLIANCE RESULTS (AVAILABILITYFOR.1004 OF CYCLE LENGTH)
PORV AVAILABLE SV AVAILABLE 0/3 1/3 3/3 1.20E-05 9.95E-06 5.1E-06.
3/3 7.9E-06 1.8E-06
- ANALOGOUS TO MCAP-11993 VALUES COMPARED TO THE TARGET OF L
10 X 10 5
CDF flee! SV tecNo Oee end Stelrh Compeny
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ATMS RULE COMPLIANCE CONCLUSIONS 0
TARGET IS TO HAVE ATMS CDF NEAR OR BELOW 1 X 1O
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0 CDF CALCULATED WITH 2 OF 3 SAFETY VALVES AVAXLABLE (NO CREDXT FOR ANY PORYS)
MEETS THIS. GOAL 0
BASES FOR ATMS RULE XS SATXSFXED O
PGSE IS COMPLETING THXS ANALYSXS USXNG THE DCPP PRA AND BELIEVES THXS ANALYSIS MILL RESULT XN LOWER'ISK THAN THE MOG MODEL L
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PROPOSED TECHNICAL SPECIFICATION PacNo eaa and Electtk Company
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REACTOR COOLNT SYSTEM OPERATIHG LIHITING COXOITIOX FOR OPERATIOX
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3.4.2.2 All pressurizer.'d'de safety valves shall be OPERABLE fifth a lift setting of 2485 psfg t C.+
'APPLICABILITY:
HOOES l, 2 and 3.
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ACTIN:
.- ' -::':'uo 5(.'th one p,pressurfzer Code safety valveSfnoperable, o44er restore
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inoperable valve to OPERABLE" status vfthfn 15 afnutes er be fn at least HOT STANDBY vfthfn 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and fn at least HOT SHUT vfthfn the
. follerfng 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.;...
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The provfsfohs of Specfffcatfon 3.0.4 say be suspended for up to 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> per valve for entry into and durfng operations fn NDE 3 for the purpose of setting thy pressurizer Code safety valves under ambfent (hot)'condftfons provided a prelfafnary cold setting, was sade prior to heatup.
SURVEILLANCE RE UIREHEXTS 1.4.2.2 No addftfonal requirements other than those required by Specifica-tion i.0.5.
gj+ JIM pre55Qf(zer fggg $tgEQ vo44 lhopiJJ)l5>0~~ p~v CanHnye pnwided W NcIper4SL MW ls I/xLpAICof~
Ona Q l~t one Pesarher Power-cPerahd reltef Ala h NE t0~~ ~ value, is qadi. 5hcmise, reAxe )he incpe~te~
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"The liftsetting pressure shall correspond to ambient conditions of the valve at nomfnal operating temperature and pressur'e.
OIABLO CANYON " UNITS 1 4 2 3j4 4-S
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, RECENT UNIT 1 BEAUMONT TEST RESUl.TS 0
UNXT 1 SAFETY VALVES TESTED WXTH LOOP SEAL AT MESTXNGHOUSE IN BEAUMONT BETWEEN OCTOBER 23 AND 25, 1989 O
TOLERANCE DEVXATXONS WERE +7.324,
+8.84, 3.864 0
No DAMAGE WAS OBSERVED 0
DEVXATXONS RESULT FROM DXFFERENT TEST METHOD (X.E.
l.OOP SEAL) 0 TOLERANCE DEVXATXONS OF THIS MAGNITUDE ARE NOT EXPECTED XN FUTURE f)lea SV pacHlo Oaa and Efactrlo Company
TREVXTEST O
TREVXTEST APPARATUS XS HYDRAULXC ASSIST MECHANISM ATTACHED TO THE STEM OF THE VALVE 0
LOOP SEAL MAS DRAXNED MHEN USXNG TREVXTEST ftto SV tecNo Oee end Soctrlo Com peny a
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BEAUMONT 0
BEAUMONT TEST USXNG LXVE STEAM MXTH LOOP SEAL 0
IDENTXCAL GEOMETRY TO PLANT DESXGN 0
CONTROLLED VALVE AMBXENT TEMPERATURE ftte: SV PaeNc Oaa and Seell@ Company
REASON FOR THE SETPOINT DEVXATXON 0
THE VALVES MERE PREVIOUSLY TESTED WITHOUT LOOP SEAL USXNG TREVXTEST O
THE BEAUMONT TEST USED LXVE STEAM MXTH LOOP SEAL 0
PAST TESTING AT BEAUMONT HAS SHOWN 3-84 DXFFERENCE XN LXFT POXNT WXTH PRESENCE OF LOOP SEAL ff'tee SV lecNo Oee end Nectrto Compeny
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1MPACT OF OUT-OF-TOLERANCE SETPOINTS O
PG&E HAS'OMPLETED PRKLXMXNARY ANALYSXS 0
RCS PEAK PRESSURE MAY EXCEED THE 1104 OF THE DESIGN LXMXT FOR FSAR DESXGN BASIS ACCXDENT 0
RCS PEAK PRESSURE XS BELOW THE 1104 OF THE DESXGN LXMXT XF ONK PORV AVAILABLE ftte: SV 1y taoNe Oae end Sectrh Company
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ENCLOSURE 3
VIEWGRAPHS PRESENTED BY PGSE, PART II Public Meeting on Diablo Canyon Steam Generator Atmospheric uump Valves
- Tuesday, December 5,
1989
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ATMOSPHERIC DUMP YALVE MEETING Dacawsaa 5,
1989 RooM 13B9 tacNc Oaa and Etectrtc Company
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AGENDA FOR ADV MEETING 1.
INTR0DUGTxoN AND PURPosE 2.
OVERVXEM OF SGTR ANALYSXS PG&E PG&E 3.
DxseussxoN DF CURRENT DESXGN AND ENHANCEMENTS PG&E 4.
ADMXNXSTRATXVE CONTROLS AND TECHNXCAL SPECXFXCATXONS NRC/PG&E taclflc Oea and Electric Coeyany a
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AGENnA o
RESP ONn To NRC REvzEw oF DCPP'GTR ANALvsxs (APRXI 29, 1988 PGRE SusMXTTAL To THE NRC) o CLARZFxcATXQN 0F ExxsTxNG DEsxGN 0F ADVs 0
REvxEM ENHANGEMENTs To ExxsTxNG DEsxGN o
IOENTxFv AOMXNXSTRATxvE CoNTRoLS ON ADVs AnMXNXSTRATzvE CoNTRoLS oN ADVs ADV TEcH SPEc REauEST pictttc GIO littd Ktectrtc OompanY
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OVERVIEW OF SGTR ANALYSIS tacmo Oaa and Sectrh Company Err~~ anv
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BACKGROUND TO THE SGTR ANALYSIS 0
SGTR AT GXNNA (1982)
CALLED INTO QUESTXON TRADXTXONAL FSAR SGTR ASSUMPTXONS 0
THREE XSSUES RESULTED:
1)
OPERATOR ACTXON TXMES 2)
EQUXPMENT QUALXFXCATXON 3)
WORST CASK SXNGLK FAXLURE O
PG&E PERFORMED A NKW SGTR ANALYSXS TO SATISFY A UNIT 2 LXCENSXNG COMMXTMKNT
- 1) ANALYSIS DONE BY M USXNG MOG METHODOLOGY 2)
4 4
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r PGRE STEAM GENERATOR TUBE RUPTURE (SGTR)
ANALYSIS TWO STAGE ANALYSIS COMPRISED OF TWO DIFFERENT TRANSIENTS:
(1)
DEMoNBTRATE THAT sTEAN GENERATQR (SG)
OVERFXLL DOES NOT OCCUR ASSUMXNG:
(A)'OUBLE ENDED GUXLLOTXNE BREAK AT TUBE SHEET (B)
Loss oF oFFsxTE FowER COXNCXDENT MXTH REACTOR TRXP (c)
FAxLURE QF AUx FEED LEvEL coNTROL vALvE (AFW LCV) To ClOSE UPON DEMAND Faclfto Oae and Ettctrk Company
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DEMONSTRATE ACCEPTABLE OFFSXTE DOSE CONSKaUENCES AFTER A SGTR ASSUMING:
(A)
DOUBLE ENDED GUILLOTINE BREAK AT TUBE SHEET (B)
LOSS OF OFFSXTE POWER COIN-CXDENT MXTH REACTOR TRXP (C) 104 ATMOSPHERIC DUMP VALVE (ADV) ON FAULTED LOOP FAXLS OPEN 10 MINUTES AFTER SGTR (D)
FAXLED OPEN ADV LOCALLY CLOSED 30 MINUTES AFTER FAXLXNG OPEN (E)
RCS COOLDOWN XNXTXATED 5 MXNUTKS AFTER ADV LOCALLY CLOSEO PocUle 4II and Eleetrle Company
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C ANALYSIS RESULTS OVERFXLL ANALYSXS (A)
QVERFXLL AVQXDKD (S)
ALL THREE ADV S ON XNTACT LOOPS USED TO COOLDOMN RCS MXTH SXNGLE FAXLURE (C)
BREAK FLOW TERMXNATKD 45 MXNS.
AFTER SGTR DOSE ANALVSXS (A)'FFSXTE DOSE CONSEQUENCES ACCEPTABLE (S)
BREAK FLOW TERMXNATED 1 HR.
AND 18 MXNS. AFTER SGTR Pac}fh Oaa and Ktaetclo Company a
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DISCUSSION OF CURRENT ADV SYSTEM DESIGN AND PLANNED ENHANCEMENTS pacific Oaa and Elactrlo Company ii%1
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ADY SYSTEM DESCRIPTION CURRENT CONFIGURATION 0
MAINTAXN PRESSURE BOUNDARY 0
AXR-TO-OPEN / FAXL.CLOSE ACTUATOR 0
LOCAL MANUAL OPERATXQN 0
BACKUP AXR / NXTROGEN NITROGEN - PASSXYE CLASS II AXR - SEISMICALLY QUALIFIED CLASSXFXED AS 1A FOR SEXSMXC CONFIGURATXON CONTROL PacNo Oaa an4 Santo Company
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v CURRENT CONFlaVRAYXON HC SDCS SDCS
>>>>>> <<>>>>>>>> va STEAN PRESSURE CONTROL AIR/M 1/P C
TP VO CONTROL A1R/N 0
[sSocs EHERG12E FOR TAVG CONTROL
[sSocs --
NE SY C
NE sv D
Hl C A1R/N EHERGlZE T TRlP OPEN HD
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EHERG1ZE FOR 5ACKLIP AlR R
ll POS I
POS GL CLOSED IL"-----"-/ERGO RL OPEN AlR 2250 PSlG ATOIFC AT}ISPHERE PacHlc Oaa and Electric Company
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ADV DESIGN AND OPERATION FUNCTION 0
PROVIDE ARTXFXCXAL LOAD DURING PRXMARYlSECONDARY LOAD MXSMATCH 0
ANTXCXPATORY STEAM RELXEF TO PREVENT UNNECESSARY SAFETY VALVE CHALLENGES PROVXDE COOLDOWN CAPABILITY XF CONDENSER Xs UNAVAXLABLK MOTIVE POWER 0
CLASS II INSTRUMENT AXR -
LOST DUE TO LOOP
.iO CLASS II NITROGEN - HIGHLY RELXAGLE CLASS I HIGH PRESSURE AXR DOTTLE-SEXSMICALLY QUALIFIED LOSS OF SOTH CLASS II AXR AND NITROGEN ALXGNS CLASS I AXR
4 C
CONTROL POWER 0
CLASS II - HXGHLY RELIABLE FROM VITAL BUS 0
0 CLASS I - HXGHLY RELXABLE FROM VXTAL BUS TWO VALVES ARE POWERED FROM EACH BUS CONTROLS 0
NORMALLY ALXGNED TO STEAM DUMP CONTROLS 0
HAND CONTROL'LER XN BOTH CONTROL ROOM'ND HOT SHUTDOMN PANEL 0
BACKUP AIR CONTROLS LOCATED XN CONTROL ROOM ffte! AQY tacNo Oss end Etectrlo Company
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ENHANCEMENTS Q
MODXFY EXXSTXNG BACKUP AXR CONTROLS To ALLOW OPERATOR To MANUALLYSELEGT THxs OPTxoN FROM CONTROL ROOM 0
MODXFY VALVES So THAT THE CLASS I CONTROLS ARE POWERED FROM A DXFFERENT BUS O
SCHEDULED FOR 1R4 8
2R4 tacSo Oee and Efeetrlo Company
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'CCEPTABILITY OF CURRENT CONFIGURATION CURRENT CONFXGURATXON IS ACCEPTABLE MHXLE ENHANCEMENTS ARE IMPLEMENTED AS DXSCUSSED BELOW.
Q ADMXNXSTRATXVE
- OPERATxNG PRocEDUREs TRAXNXNG SURYEXLLANcE 5 FUNGTxoNAL TEsTxNG 0
DESXGN CLASS 2 FEATURES ARK RELXABLK POWER SUPPLXES ARK RELXABLE
- DCPP STEAM GENERATOR EXPERXENCK XS GOOD 4
SGRR ANALYSXS CONTAXNS CONSERYATXYE ASSUMPTXONS PGSE MxLL CLARxFY IT's APRXL 29, 1988 LETTER, DCL"88-114 IN THK NEAR FUTURE taalOe Caa and Electric Company a
ADMINISTRATIVECONTROLS AND TECHNICAL SPECIFICATIONS pacNo Oaa and Sootrfo Company
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ADMINISTRATIVE CONTROLS FOR ATMOSPHERIC DUMP VALVES ACTIONS TAKEN 0
OPERATXONS SHIFT ORDER XSSUED 11/17/89 0
STP V-3R1 TEsTs STROKE TIME EACH CQLD SHUTDOWN 0
OPERATOR ROUND SHEETS ON AXR PRESSURE VERIFICATXON ACTIONS TO 8E TAKEN PLANT PROCEDURE To IMPLEMENT ADMXNXSTRATIVE CONTROLS 0
REVISE SURVEXLLANCE TEST PROCEDURES (STPS)
FULLFXLL ADDXTXONAL OPERATOR TRAXNXNG ON THESE PROCEDURAL REQUIREMENTS 0
COMPLETE AOT ANALYSIS F 1990 FacNc Cia and Electric Company SUBMIT LAR IN THE SPRXNG 0
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~'~lk'TRYST HS STOPS GLHKRATOR nsPH8 c EvbtP VALVES 0
l cvHf +d'a assoc kl, b/n< 'whys L?MIYIHG COHO T ON F R OPERATION l4~
p,~y $+Acric. of~
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APPL ABILITY:
HOOKS ls 8, 3, and 4.'n>'.
a.
M]th on>> less than the requir>>d NA restore the inoperable steam generator to OPER within 7 days; or be in at least HOT STANDBY with)n and in HOT SHUTOON with the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and ranulradpisldualpsat saovalv@ep ln oparat$ on for reaoval.
Ofe) Vssus-b.
M)th two )ass than the requ)red restore at least one of the inoperable steam generat
'OPEAASLK status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least-HO the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTONN within the foll and place the required in decay heat removal, E4C gee~
OPERA&LEs I.E status ho next S hours place the decay heat ADA OP
- 8LE, r
to STANOSY within
<ng 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />'s operation for SURVEILLANCE RE U
T assoc~~
block a1
$.1.S.S Each stais SsnaratorJuh, nd assocTats Encludfng the she l be demon Clog 4 kacA,up a4 We 4VC eA a.
At l>>est once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by ver$ fyin that each has a pre than or equal to 98$ psig, and SQae y c,
~
At least once per 18 months by ver)fy)ng that all steam gener operate thr'ough on>> cyc)e of.full travel using reiot controls. and
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Sf d ys S/re~ ga.~a~Q>>
1SDt/ block Yal~> aM n
'ls treted SLS:
yCC4'~m sure greater
~
AbV'er
~tv"ill aanuel "A>V>
c~~ g~
"Nen steam g>>narrators are be)ng used for decay heat removal.
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