ML20245F508
| ML20245F508 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 04/25/1989 |
| From: | Dick G Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8905020400 | |
| Download: ML20245F508 (71) | |
Text
'
April ?5, 1989 i
Docket Nos. 50-498 and 50-499 .
1 LICENSEE: Houston Lighting & Power Company )
FACILITY: South Texas Project, Units 1 and 2 l
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS STATUS OF LICENSEE'S PROBABILISTIC RISK ASSESSNENT (PRA) i 1
1 The subject meeting was held on January 27, 1989. Enclosure 1 contains a list of attendees. A copy of the presentation slides is included as Enclosure 2.
The meeting was held at the request of the licensee to provide the staff with a status of the progress of its PRA, identified as a Probabilistic Safety Study. l The PRA developed with the assistance of Pickard, Lowe, and Garrett was begun in 1984. It is a Level 1 PRA with qualitative containment analysis. A total of about 30 man years has been spent to date to develop the model. The licensee has four people dedicated to the PRA.
After a review of the background and PRA basis, which has been discussed previously with the staff, an overview of preliminary results were presented.
Pointestimateofcoredamagefrequencyjesultingfromtotalinternalinitiating events and exterial events was 1.5 x 10 . The contribution of various initiating event cctegories to the core damage frequency and the results of 20 important core damage event sequences also were presented and discussed with the staff.
The PRA is expected to be submitted to the staff for review in April 1989.
Approximately three months later, the licensee is expected to submit changes '
to the technical specifications in the areas of allowed outage time and surveillance test intervals. Proposed change will be based on the PRA analysis of the three safety train concept cebodied in thp desjgn of..the two units at South Texas. W~
/
Gec!p F. J:: , Jr.
George F. Dick, Jr. , Project Manager Project Directorate - IV Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation As stated nea
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. Docket Nos.50-49E and 50-499 l LICEliSEE: Houston Lighting & Power Company FACIL:TY: South Texas Project, Units I and 2
SUBJECT:
SUf! MARY OF MEETIliG TO DISCUSS STATUS OF !
LICEtiSEE'S PROBABILISTIC RISK ASSESSMENT (PRA) I l 1 1
The subject meeting was held on January 27, 1989. Enclosure 1 contains a list of attendees. A copy of the presentation slides is included as Enclosure 2. ]
l The meeting vers held at the request of the licensee to provide the staff with a status of the progress of its PRA, identified as a Probabilistic Safety Study.
The PRA developed with the E!!istance of Pickard, Lowe, and Garrett was begun j in 1982 It is a Level 1 PRA with qualitative containment analysis. A total I of about 30 man years has been spent to date to develop the model. The licensee has four people dedicated to the PRA.
After a review of the background and PRA basis, which has been discussed l
' previcusly with the staff, an overview of preliminary results were presented.
Point estimate of core damage frequency eventsandexterraleventswas1.5x10gesultingfromtotalinternalinitiating The contribution of various i
initittire event categories to the core damage frequency and the results of 20 important core damage event sequences also trere presented and discussed with the staff.
The PRA is expected to be subritted to the staff for review in April 1989.
Approximately three months later, the licensee is expected te su.mit changes to the technical specifications in the areas of allowed outage time and surveillance test intervals. Prcposed change will be based on the PRA analysis of the three safety train concept erbodied in the design of the two units at South Texas.
Georg F. Dick, J e , Project Manager Project Directorate - IV Division of Reactor Projects - III, l
IV, Y and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
As stated i cc w/ enclosures:
l See next rage
, . l Enclosure 2 l
l l
1 OVERVIEW OF STPEGS I PROBABILISTIC SAFETY STUDY' BY D. J. DENVER R. P. MURPHY R. L. BOYER S. S. RODGERS D. W. STILLWELL JANUARY 27, 1989
- Mr. J. H. Goldbero South Texas Project Houston Lighting and Power Company cc:
Brian Berwick, Esq. - Resident Inspector / South Texas Assistant Attorney General Project Environmental Protection Division c/o U.S. Nuclear Regulatory Connission P. O. Box 12548 P. O. Box 910 Capitol Station Bay City, . Texas 77414 Austin, Texas 78711 Mr. Jonathan' Davis Mr. J. T. Westermeier Assistant City Attorney General Manager, South Texas Project City of Austin Houston Lighting and Power Company P. O. Box 1088 P. O. Box 289 Austin, Texas 78767 Houstcn, Texas 77483 Ms. Pat Coy Mr. R. J. Miner Citizens Concerned About Nuclear .,.
Chief Operating Officer Power City of Austin Electric Utility 10 Singleton 721 Barton Springs Road Eureka Springs, Arkansas 72632 Austin, Texas 78704 Mr. M. A. McBurnett Mr. R. J. Costello Manager, Operations Support Licensing Mr. M. T. Hardt Houston Lighting and Power Company City Public Service Board P. O. Box 289-P. O. Box 1771 Wadsworth, Texas 77483 San Antonio, Texas 78296 Mr. A. Zaccaria.
Jack R. Newman, Esq. Mr. K. G. Hess Newman & Holtzinger, P. C. Bechtel Corporation
, 1615 L Street, NW P. O. Box 2166 l Washington, D.C. 2003E Houston, Texas 77001 Melbert Schwartz, Jr., Esq. Mr. R. P. Verret Baker & Botts Mr. R. L. Range One Shell Plaza Central Power and Light Company-l Houston, Texas 77002 P. O. Box 2121 Corpus Christi, Texas 78403 Mrs. Peggy Buchorn Executive Director Doub, Muntzing and Glasgow Citizens for Equitable Utilitiesc Inc. Attorneys at Law Route 1, Box 1684 Suite 400 Brazoria, Texas 77422 808 Seventeenth Street, N.W.
Washington, D.C. 20006 Mr. S. L. Rosen General Manager, Operations Support Houston Lighting and Power Company
.P. O. Box 289 Wadsworth, Texas 77483
Pr. J. H. Goldberg South Texas Project Ecuston Lighting I, Power Ccrpeny I
cc-Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 I Arlington, Texas 76011 '
Mr. Lanny Sinkin, Counsel for Intervenor Citizens Concerned about Nuclear Power, Inc.
Christic Institute 1324 North Capitol Street ;
Washington, D.C. 20002 Licensing Representative l Houston Lighting and Power Corrpany '
Suite 610 Three Metro Center Bethesda, Maryland 20814 Rufus S. Scott Associate General Ccunsel Houston Lighting & Power Company P. O. Box 1700 Houston, Texas 77001 INP0 Records Center 1100 Circle 75 Parkway Atlanta, Georgia 30339-3064 .
i Joseph M. Hendrie 50 Bellport Lane Bellport, New York 11713 Gerald E. Vaughn, Vice President Nuclear Operations Houston Lighting & Power Company, P. O. Box 289 Wadsworth, Texas 77483 R. W. Chewning, Chairman Nuclear Safety Review Board Houston Lighting & Power Company P. O. Box 289 Wadsworth, Texas 77483
. _ _ _ _ - .__-_-_-_-____A
Enclosure 1.
l South Texas Project PRA Meeting Janua ry 27, 1989 Name Organization George Dick NRC ;
Wayne Harrison HL&P-Daniel Stillwell HL&P l Richard P. Murphy HL&P 1 Richard J. Barrett NRC E. Chelliah NRC Claudia Abbate NRC Roger Boyer HL&P Shawn Rodgers HL&P Jose A. Calvo NRC D. Denver HL&P' l
I I
I I
i
AGENDA I. INTRODUCTION D. J. DENVER II. METHODOLOGY OVERVIEW S. S. RODGERS III. OVERVIEW OF PLANT MODEL R. L. BOYER i
IV. PRELIMINARY RESULTS D. W. STILLWELL V. PSS SCHEDULE FOR COMPLETION ;,.
AND TECH SPEC APPLICATIONS R. P. MURPHY VI.
SUMMARY
D. J. DENVER q
I- i INTRODUCTION O- THE SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION (STPEGS)
INCORPORATES A THREE TRAIN DESIGN PHILOSOPHY.
l HAS BEEN DESIGNED TO POST-TMI REGULATORY CRITERIA )
DESIGN HAS ~ BEEN EXCHANGED WITH THE USSR TO REPRESENT- I THE STATE-OF-THE-ART OF-U.S.' NUCLEAR FACILITIES
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STPEGS FEATURES O STPEGS INCORPORATES SEVERAL SIGNIFICANT IMPROVEMENTS IN SAFETY AND OPERATIONAL DESIGN THREE INDEPENDENT ESSENTIAL SAFETY FEATURE TRAINS, EACH j OF WHICH IS CAPABLE OF SHUTTING THE PLANT DOWN SAFELY l
A SIMILAR DESIGN PHILOSOPHY FOR SUPPORT SYSTEMS, E.G.,
ESSENTIAL AC POWER INCLUDING EMERGENCY DIESEL GENERATORS A QUALIFIED THREE-TRAIN RESIDUAL HEAT REMOVAL SYSTEM ENTIRELY WITHIN CONTAINMENT. THIS DESIGN ELIMINATES THE CLASSIC "V-SEQUENCE" CONCERN FOR CONTAINMENT ~ BYPASS i SCENARIOS FOUR DEDICATED TRAINS OF SECONDARY SIDE HEAT REMOVAL, )
INCLUDING ONE TURBINE DRIVEN AUXILIARY FEEDWATER PUMP.
EACH TRAIN IS CAPABLE OF PROVIDING ADEQUATE SECONDARY SIDE HEAT REMOVAL.
4 A " RAPID REFUELING" DESIGN RESULTING IN IMPROVED ALARA l<
AND MAINTENANCE CONSIDERATIONS )
OTHER UNIQUE DESIGN FEATURES, E.G., DEAEREATING FEEDWATER HEATERS AND OTHER FEEDWATER MODIFICATIONS WHICH SUBSTANTIALLY IMPROVE SECONDARY SIDE WATER QUALITY AND RELIABILITY.
O HOWEVER, LIMITED CREDIT HAS BEEN GIVEN IN THE CURRENT STPEGS TECH SPECS FOR THE ADDITIONAL SAFETY AFFORDED BY THE ADDITIONAL EQUIPMENT. THE PLANT CAN BE BROUGHT TO A STABLE CONDITION WITH ONE TRAIN EXCEPT IN ONE LOW PROBABILITY EVENT.
STPEGS PROBABILISTIC SAFETY STUDY l
0 HOUSTON LIGHTING & POWER COMMENCED A PROGRAM TO COMPLETE A PROBABILISTIC RISK ASSESSMENT OF THE SOUTH TEXAS PROJECT IN l 1984.
UTILITY INITIATIVE, NOT REGUIATORY REQUIREMENT NOW BEING COMPLETED EXECUTIVE
SUMMARY
REPORT COMPLETE ABOUT 3/31/89 O PSS ENCOMPASSES A LEVEL 1 PRA WITH A LIMITED (I.E.,
QUALITATIVE) CONTAINMENT ANALYSIS. INCLUDES EXTERNAL EVENTS AND AN UNCERTAINTY ANALYSIS.
l 0 PRA WILL BE MAINTAINED AS A LIVING TOOL TO ASSESS
- l CHANGES IN PROCEDURES AND EQUIPMENT MODIFICATIONS TO TECH SPECS ACCIDENT MANAGEMENT PROGRAM GENERIC AND OTHER SAFETY ISSUES INDIVIDUAL PLANT EXAMINATION (GENERIC LETTER 88-20)
O THE FOLLOWING PRESENTATIONS PROVIDE AN OVERVIEW OF THE METHODOLOGY, FEATURES AND PRELIMINARY RESULTS OF THE STPEGS PROBABILISTIC SAFETY STUDY.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - -- - _i
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NOTES ON AUXILIARY SYSTEM TO AUXILIARY SYSTEM DEPENDENCIES A. Given a loss of the offsite grid, or a loss of power to a l Class 1L 4.16kv bus, the standby DGs receive automatic start signals.
B. Given a loss of offsite power, these auxiliary.sysceas receive automatic start signals from the load seg aencer once the standby DG is supplying power to the ass ociated bus.
C. Given a loss of offsite power, the control rod drive motor l generator sets lose power, which causes a control rod insertion by gravity. In this case successfu1 reactor trip does not require actuation of the reactor trip breakers.
D. If there is no AC power to the battery chargers, they will eventually fail due to depletion. They are sized for a minimum of two hours.
E. Vital AC fails on loss of DC and AC power supply to the inverters. If there is no AC power to a vital AC channel, it can function if DC power is available. Note that loss of a 4.16kv bus will also fail the associated DC bus due to l
battery depletion.
F. Loss of one HVAC train.
G. Equipment in these systems require DC power to start, but do not require DC power to run.
H. The EAB main area HVAC fans are dependent on DC power to start and to run. On loss cf DC power 11,e dampers fail i
such that smoke purge operation is not available.
I. Loss of DC power to a CRE HVAC train results in damper ;
positioning such that the train is in an " emergency" mode, with isolation from normal outside air, recirculation through the clean up filters, and drawing some outside air through the makeup air units. Smoke purge is unavailable and the makeup air units pass only 1000 cfm versus 17400 for smoke purge.
1 L.
i NOTES ON AUXILIARY SYSTEM TO AUXILIARY SYSTEM DEPENDENCIES I J. Loss of DC power results in the failure of the shunt trip coils of the reactor trip breakers.
K. Each logic train, R and S, is served by two vital AC channels. Loss of both vital AC channels to a logic train will fail the logic train, but trip the reactor.
L. Loss of vital AC power channel results in signal inputs to all parameters of that channel except Containment Pressure HI-3.
M. Loss of a vital AC channel results in the closing of the ECW valve that cools the condenser of the associated chillers, thus failing the chiller. Not recoverable, except locally with hand pumps. I N. Loss of a single logic train of SSPS results in a loss of one of two inputs to ESFAS and to the reactor trip breakers.
O. Loss of an ESFAS train results in no automatic start i signals to the associated ECW train, CCW train, ECH, EAB I l
HVAC train, and CRE HVAC train for those initiating events l that require safety injection actuation (with no LOSP).
Note that trains already running will continue to run. If th'ere is a concurrent LOSP, failure of an ESFAS train does not affect the operation of any of the above system trains since the load sequencer for that train will provide start signals.
P. Failure of the sequen7er results in failure of all the 4.16kv and 480v equipment on that train on loss of offsite power. Failure of the sequencer on SI results in failure of the standby AC powered equipment on that train.
Q. ECH trains A and B are operating with condenser water supplied by ECW train A. With loss of ECW train A, ECW j train B can supply chiller condenser water to ECH trains A j and B without valve changes. ECH train C.is assumed to be aligned to receive chiller condenser water only from ECW train C.
NOTES ON AUXILIARY SYSTEM TO AUXILIARY SYSTEM DEPENDENCIES P. Loss of ECH train A, B, or C results in loss of chilled water to the air herdling units in the EAB HVAC and CRE j HVAC trains A, B, and C, respectively. I S. Loss of EAB HVAC is assumed to result in elevated l temperatures in the EAB such that electrical failures occur.
T. Loss of CRE HVAC is assumed to result in elevated temperatures in the CRE such that' electronic equipment fails.
X. Denotes a guaranteed failure of the system tr'"ain if the associated train fails.
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. NOTES ON THE AUXILIARY SYSTEM TO FRONTLINE SYSTEM DEPENDENCIES A. Offsite grid failure as a result of an' initiating event other than loss of offsite power is assumed to guarantee successful turbine trip.
B. The charging pumps will start when power is available.
C. With a loss of the offsite grid, the RCFCs receive automatic start signal from the the load sequencer when the associated standby DG is supplying power to the bus.
D. The SG PORV's are hydraulically actuated. The hydraulic j system stores energy in an accumulator which is pressurized by a pump powered from Class 1E 480v power. The stroking capacity on loss of AC power is dependent upon accumulator pressure.
E. Failure of the indicated auxiliary system train fails certain containment isolation valves. '
.. Power is lost to the associated block valves.
G. Loss of DC bus E1A11 (Channel I) or DC bus ElB11 (Channel III) de-energizes solenoids such that all the MSIVs close.
H. DC power is required to start but not to run (AFW Train D I requires DC power to open inclation valve).
I.
Loss of DC bus E1A11 (Channe'. I) or DC bus ElB11 (Channel i III) de-energizes solenoids such that turbine bypass is inoperable.
J. Failure of both SSPS trains results in loss of turbine trip signals from the SSPS.
K. Failure of both GSPS trains results in the failure of the main steam isolation function.
L. Failure of both SSPS trains results in the failure of the automatic signal from SSPS to the indicated equipment.
M. Failure of both SSPS trains er the indicated ESFAS train results in the failure of the automatic start signal and the failure of the automatic open signal to the CCW isolation
! valves for the associated RCFCs.
N. Failure of ESFAS trains A and B results in the failure of the main steam isolation function.
. . 1 NOTES ON THE AUXILIARY SYSTEM TO FRONTLINE SYST.EM DEPENDENCIES
pressure relief. Loss of both SSPS trains or loss of the indicated ESFAS train will be disable this function. Manual control'is available. l F. Failure of both SSPS trains, the indicated ESFAS train, or the indicated load sequencer under loss of offsite power conditions results in the loss of an automatic start of the charging pumps.
Q. Loss of all CCW pumps fails the centrifugal charging pumpe due to loss of lubrication and cooling.from CCW . ;
R. Failure of the indicated ECW or CCW train fails cooling to-the essocisted RHR heat exchanger.
S. Failure of all ECW or CCW trains results in-loss of RCP motor and thermal barrier cooling. With charging pump failure, a RCP seal LOCA may occur. The operator can restore RCP cooling using the CVCS positive displacement pump.
T.
Loss of EAB HVAC is assumed to result in elevated temperatures in the EAB such that electrical equipment failures result. 'The indicated equipment fails due to loss of AC power, DC power, or vital AC.
U.
Loss of a QDPS train results in loss of flow control of the corresponding ATW train. The flow control valves fail open on loss of signal from QDPS.
V. Loss of a QDPS train results in loss of the asacciated SG PORV. The PORV fails closed on loss of signal from QDPS.
INITIATING EVENT CATEGORIES SELECTED i FOR QUANTIFICATION m LOSS OF COOLANT INVENTORY (LOCA)
-EXCESSIVE LOCA
-LARGE LOCA l
-MEDIUM LOCA l
-SMALL LOCA A) NONISOLABLE B) ISOLABLE
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-INTERFACING SYSTEMS LOCA l
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-STEAM GENERATOR TUBE RUPTURE 1:
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INITIATING EVENT CATEGORIES SELECTED ,
FOR QUANTIFICATION (CONTINUED) un TRANSIENTS
--REACTOR TRIP
-TURBINE TRIP
-LOSS OF CONDENSER VACUUM l
l
-CLOSURE OF ALL MSIVS
--STEAM LINE BREAK UPSTREAM OF MSIVS A) INSIDE CONTAINMENT B) STUCK OPEN PORV OR SAFETIES
-STEAM LINE BREAK DOWNSTREAM OF MSIVS l
-INADVERTENT SAFETY INJECTION
-GENERAL TRANSIENTS ,
A) TOTAL LOSS OF MAIN FEEDWATER B) PARTIAL LOSS OF MAIN FEEDWATER C) EXCESSIVE FEEDWATER D) CLOSURE OF ONE MSIV E) CORE POWER EXCURSION F) LOSS OF PRIMARY FLOW
i INITIATING EVENT CATEGORIES SELECTED i FOR QUANTIFICATION (CONTINUED) o COMMON CAUSE INITIATING EVENTS i SUPPORT SYSTEMS FAULTS A) LOSS OF 0FFSITE POWER >
B) LOSS OF ONE DC PUS (E1A11 OR ElB11)
C) TOTAL LOSS OF ESSENTIAL COOLING WATER j D) TOTAL LOSS OF COMPONENT COOLING WATER i E) LOSS OF INSTRUMENT AIR l
F) LOSS OF EAB HVAC i G) LOSS OF CONTROL ROOM HVAC 1
--SEISMIC EVENTS l
--FIRES
--EXTERNAL FLOOD I
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ELECTRIC POWER MGDEL o ELECTRIC POWER SYSTEMS
-345 KV 0FFSITE POWER
-UNIT AUXILIARY TRANSFORMER
-STANDBY TRANSFORMER
-13.8 KV BUSES 1F,1G, AND 1H
-4.6 KV BUSES E1A, ElB, AND E1C
-EMERGENCY DIESELS 11, 12 AND 13
-125 V DC POWER E1A11, E1811, E1C11 AND E1D11
-480 V AC BUSES, MCCS, AND LOAD CENTERS o ENGINEERED SAFEGUARDS ACTUATION SYSTEM
--SSPS LOGIC R AND S
-ESFAS TRAINS A, B AND C
-120 V VITAL AC CHANNELS I, II, III AND IV o QUALIFIED DISPLAY PROCESSING SYSTEM o REACTOR PROTECTION SYSTEM
MECHANICAL SUPPORT MODEL a ESSENTIAL COOLING WATER SYSTEM l m ESSENTIAL CHILLED WATER SYSTEM m EAB HVAC SYSTEM EAB HVAC FANS SM0KE PURGE DAMPERS m COMPONENT COOLING WATER SYSTEM i
EARLY RESPONSE MODEL er TURBINE TRIP AND MSIVS um AUXILIARY FEEDWATER SYSTEM su CHEMICAL AND VOLUME CONTROL SYSTEM ,
CENTRIFUGAL CHARGING PUMPS POSITIVE DISPLACEMENT CHARGING PUMP AND TECH CENTER DIESEL LETDOWN AND SEAL RETURN LINES RCP THERMAL BARRIER COOLING 1
mi EMERGENCY CORE COOLING SYSTEM
--HIGH HEAD INJECTION
--LOW HEAD INJECTION
LATE RESPONSE MODEL L
EJ CONTAINMENT SUMP RECIRCULATION EJ CONTAINMENT SPRAY SYSTEM c3 REACTOR CONTAINMENT FAN COOLERS E3 RESIDUAL HEAT REMOVAL SYSTEM EJ CONTAINMENT SUPPLEMENTARY PURGE ISOLATION EJ CONTAINMENT ISOLATION E) STEAM GENERATOR ISOLATION 1
. l
INTERFACE WITH LEVEL 2 PRA ;
o MODEL HAS DEFINED 112 UNIQUE PLANT DAMAGE STATES I
o EACH EVENT SEQUENCE IS MAPPED TO ONE OF THESE STATES o CAPABILITY TO DIRECTLY ADDRESS NUREG-1150 '
ISSUES: ;
-HIGH PRESSURE MELT SCENARIOS !
--DIRECT CONTAINMENT HEATING l
-INDUCED STEAM GENERATOR TUBE RUPTURE o CAPABILITY TO APPLY REC 0VERY MODELS FOR POST-MELT ACCIDENT MANAGEMENT
- _____-_-_m.-_______-_ _-_m __
PLANT DAMAGE STATE GROUPS GROUP I -
CONTAINMENT INTACT, N0 SIGNIFICANT RELEASE GROUP II -
CONTAINMENT ISOLATED AT TIME OF MELT, LATE CONTAINMENT FAILURE GROUP III - SMALL UNIS0 LATED RELEASE PATH AT TIME OF MELT GROUP IV -
LARGE UNIS0 LATED RELEASE PATH AT TIME OF MELT
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4 SCHEDULE FOR COMPLETION OF ACTIVITIES AE.D APPLICATIONS OF THE STPEGS PRA i
'1 RICHARD MURPHY I I
i
STATUS OF STPEGS PSS AND TECH SPEC STUDY O THE STPEGS PROBABILISTIC SAFETY STUDY (PSS) AND THE TECH SPEC ANALYSIS ARE EXPECTED TO BE COMPLETED ON THE FOLLOWING SCHEDULE:
MEETING WITH THE NRC TO DISCUSS THE PRA MODEL 1/27/89 PSS EXECUTIVE
SUMMARY
REPORT COMPLETED FOR REVIEW 3/31/89 TECH SPEC ANALYSIS REPORT COMPLETED FOR REVIEW 6/30/89 O THE TECH SPEC ANALYSIS WILL UTILIZE THE PSS PLANT COMPUTER MODEL.
I
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RISF-BASED ANALYSIS OF STPEGS TECHNICAL SPECIFICATIONS O THE STPEGS PRA COMPUTER MODEL WILL BE UTILIZED TO ANALYZE TECHNICAL SPECIFICATION
-ALLOWED OUTAGE TIMES (AOT's)
-SURVEILLANCE TEST INTERVALS (STI's).
O STI'c AS WELL AS AOT'S ON SELECTED SYSTEMS ARE BEING INCLUDED IN THE STUDY O ALLOWED OUTAGE TIMES WHICH WILL BE ANALYZED ARE SUMMARIZED ON THE NEXT SLIDE
]
O SURVEILLANCE TEST INTERVALS WHICH WILL BE ANALYZED ARE SUMMARIZED ON THE FOLLOWING SLIDE
.l l
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TECH SPECS UNDER EVALUATION l
TECH SPEC SYSTEM __,2QI (DAYS)
CURRENT PROPOSED 3.5.2 EMERGENCY CORE CCOLING SYSTEM (ECCS) 3 10 3.5.6 RESIDUAL HEAT REMOVAL (RHR) 3 10 3.6.2.1 CONTAINMENT SPRAY (CS) 3 10 3.6.2.3 REACTOR CONTAINMENT FAN COOLERS (RCFC) 3 10 3.7.1.2 AUXILIARY FEEDWATER (AFW) 3 10 3.7.3 COMPONENT COOLING WATER (CCW) 3 10 3.7.4 ESSENTIAL COOLING WATER (ECW) 3 10 3.7.14 ESSENTIAL CHILLED WATER (ECH) 3 20 3.8.1.1 STANDBY DIESEL GENERATORS (SDG) 3 10 3.1.2.4 CENTRIFUGAL CHARGING PUMPS (CCP) 3 10 3.5.1 SAFETY INJECTION ACCUMULATORS 1 HOUR 24 HOURS
l TECH SPECS UNDER EVALUATION Surveillance Tcst Interval Chances:
Tech Snec: System / Chance in Interval 4.7.1.2 Auxiliary Feedwater -
Increase STI from 31 days to 92 days.
4.8.1.1 Standby Diesel Diesel Generators Increase STI from 31 to 92 days.
4.3.2 Essential Safety Features Actuation System Increase from 31 to 92 days.
4.3.3 Accident Monitoring System Increase from 31 days to * .
5 3.6.1.7 Containment Ventilation Increase from 31 days to * .
3.7.7 Control Room HVAC Increase from 31 to 92 days.
3.8.3 Onsite Powerl, Distribution Increase STI from 7 days to ___.
o "To date" to be determined.
l
Other ACT's Which Will Be Addressed In The Future:
I_ech Sper* System / Chance in Allowed Outaae Time 3.3.1 Reactor Protection System Increase breaker AOT from 48 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3.4.2 Pressurizer Safety Relief Valves 15 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
3.4.4 Pressurizer PORV Increase from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3.6.1.7 Containment Ventilation Increase from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to * .
3.6.3 Containment Isolation Increase from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to * .
3.7.1.1 Main Steam Safety Relief Valves Increase from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to * .
3.7.1.5 Main Steam Isolation Valves Increase from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to * .
3.7.7 Control Room HVAC Increase from 7 days to
- for 1 operable.
3.8.1 AC Power Sources Increase AOT of Offsite Power Source from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to *.
3.8.2 DC Power Sources Increase from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to * .
3.8.3 Onsite Power Distribution ESF AC Bus AOT from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to * .
Vital AC Panel from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to * .
DC Bus from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to * .
"To date" to be determined.
I
. u.
SUMMARY
O THE SOUTH TEXAS PROJECT UTILIZES A THREE-TRAIN ESSENTIAL SAFETY FEATURE DESIGN WHICH IS UNIQUE IN THE UNITED STATES.
O STPEGS HAS SUBSTANTIALLY ENHANCED SAFETY FEATURES WHICH HAVE RESULTED IN ADDED EQUIPMENT REQUIRING MORE FREQUENT TESTING AND MAINTENANCE.
O SUBSTANTIALLY HIGHER INVESTMENT AND O&M COSTS ARE ASSOCIATED WITH THE IMPROVED SAFETY DESIGN.
O LITTLE CREDIT HAS BEEN GIVEN IN THE CURRENT TECH SPECS FOR THE ADDITIONAL SAFETY AFFORDED BY THE ADDITIONAL EQUIPMENT.
O WITH THE ADDED SAFETY EQUIPMENT, AN AGGRESSIVE MAINTENANCE AND TESTING PROGRAM, AND ThE APPLICATION OF THE STANDARD WESTINGHOUSE "TWO-TRAIN" TECH SPECS, STPEGS IS IN A POSITION TO INCREASE PLANT AVAILABILITY THROUGH THE IMPLEMENTATION OF TECH SPECS THAT PROPERLY REFLECT THE STPEGS DESIGN.
O THE STPEGS PSS WILL PROVIDE THE BASIS FOR IMPLEMENTING A CONTINUING RISK MANAGEMENT PROGRAM.
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