ML20211P588
| ML20211P588 | |
| Person / Time | |
|---|---|
| Issue date: | 07/18/1986 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Boxer B HOUSE OF REP. |
| References | |
| NUDOCS 8607230297 | |
| Download: ML20211P588 (2) | |
Text
..
DISTRIBUTION CENTRAL FILES DMossburg RRAB RDG FILE ED0 RDG FILE ENE DSR0 CHRON FILE MBridgers The Honorable Barbara Boxer VStello 001790 U.S. House of Representatives HRDenton OCA Washington, DC 20515 TPSpeis PPAS FJCongel BWSheron
Dear Congresswomen Boxer:
JERosenthal SECY 86-524 SAcharya PDR In your letter to the President dated May 1,1986, you requested information
[-
regarding the safety of U.S. reactors as a result of the Chernobyl reactor f-t-c accident in the Soviet Union.
I expect that the response to your concern about the Hanford N-reactor and the other four reactors in Savannah River, South Carolina cited in your letter will be provided by the DOE. The following discussion provides our response to your concern about the safety of the commercial nuclear power plants in the U.S. licensed by NRC.
All plants licensed to operate were reviewed in accordance with Conmission guidelines, rules, and regulations based on sound engineering and defense-in-depth principles to protect the health and safety of the public.
The engineered safety features (ESFs) and the containment structures of the plants are capable of coping with a spectrum of design basis accidents (DBAs),
such as an accident involving the total rupture of the largest reactor coolant or steam pipe.
Calculated radiological doses to the public due to release of radionuclides from the DBAs are limited by the ESFs and the containment structure to below the 10 CFR Part 100 guideline values (25 rem to the whole body and 300 rem to the thyroid). Moreover, the dose calculations employed for the DBAs utilize substantially conservative assumptions, and realistic doses to the public from these DBAs are expected to be significantly less than the calculated values. We have also been studying accidents more severe than the DBAs for some time. This effort was accelerated following the TMI accident. Based on these studies, we have concluded that all containment structures surrounding nuclear power reactors provide some degree of protection to the public should an accident more severe than the DBAs occur.
Probabilistic risk assessment (PRA) studies have been performed for a number of operating nuclear power plants. These studies assess plant vulnerability for severe core damage accidents.* Some of these plants have made design and procedure improvements based on insights from the PRAs. However, we do not have PRAs for all plants. Therefore, we have not ranked the nuclear power plants in terms of severe core dc.;aage accident vulnerability, or identified the plants with greatest risk.
- The phrase " severe core damage accident" is used to describe an accident in which the reactor fuel integrity is lost due to inadequate core cooling. The extent of the core damage can range from partial core melting without h
penetration of the reactor vessel as exemplified by the TMI-2 accident to an f
1 accident in which the fuel undergoes extensive to complete melting.
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Congresswomen Boxer For operating plants without PRAs, implementation of the NRC's Severe Accident Policy will involve a systematic examination of all operating reactors to identify and remedy outstanding contributors (severe accident vulnerabilities) to severe core damage accident probability and risk. The NRC staff will issue guidance for the individual plant examinations by the end of 1986, and the examinations will be performed by the utilities starting in early 1987.
For new nuclear power plant applications, plant-specific PRA studies are required and will provide for identification of plant-specific vulnerabilities that can be corrected in the design stage.
My May 19, 1986 memorandum to the Commissioners (Attachment 1) provides a discussion on our earlier and current estimates of the industry-wide likelihood of severe core damage accidents. Concern regarding the acceptable level of severe core damage accident risk is being addressed in the Commission's Safety Goal Policy Statement which was recently approved by the Comission. The final form of the policy establishes a goal such that the mortality risk to an individual member of the public from severe core damage accidents be only a very small fraction (one one-thosandth) of similar background risk from other causes. The final text is presently being prepared for publication in the Federal Register. We would be pleased to send you a copy as soon as the statement is available if you so desire.
Sincerely, Original signed bE' yictor Stellop,
Victor Stello, Jr.
Executive Director for Operations
Attachment:
May 19, 1986 memorandum of Victor Stello, Jr. to the Commissioners cc: NRC PDR
- SEE PREVIOUS CONCURRENCE NRR:D EDO jf//
07/01/86 07/fy86 F, l p HRDenton*
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4i Office DSR0:RRAB*.....:DSR0:RRAB*..:DSR0:RRAB*..:DSR0:DD*....:DSR0:D*.....:NRR:DD*....:
Surname SAcharya : sj.... :JERosenthal. : FJ Conge1.... : BWSheron.... :TPSpei s..... : RVollmer.... :
Date 06/21/86.......:06/23/86....:06/23/86....:06/23/86....:06/23/86....:07/01/86....:
OFFICIAL RECORD COPY
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UNITED STATES Cys: Stell0 c'
NUCLEAR REGULATORY COMMISSION Roe y
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I wash NGTON,0. C. 20555 Rehm Sniezek May 19, 1986 Taylor Heltemes i
GCunningham EDO R/F MEMORANDUM FOR:
Chairman Palladino
~
Commissioner Roberts Commissioner Asselstine Commissioner Bernthal Commissioner Zech
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FROM:
Victor Stello, Jr.~
f<J Executive Director f6r Operations
SUBJECT:
FREQUENCY OF SEVERE CORE DAMAGE ACCIDENT Enclosed for your information is the staff's answer to the question "What is the estimated likelihood of a severe core damage accident due to internal initiators for all operating reactors (100 plants) over the next 20 years?"
The enclosed l
information was developed by the staff based on the latest information currently being developed in the PRA rebaselining of six reference plants by the Office of Nuclear Regulatory Research.
Enclosure:
As stated cc:
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CORE MELT LIKELIHOOD OVEST10N.
WHAT IS THE ESTIMATED LIKEllH00D OF A SEVERE CORE DAMAGE ACCIDENT DUE TO INTERNAL INITIATORS FOR ALL OPERATING REACTORS (100 PLANTS OVER THE NEXT 20 YEARS?
ANSWER.
IT IS IMPOSSIBLE TO ESTIMATE WITH PRECISION WHAT THE ACTUAL INDUSTRY-WIDE LIKEL'IH00D OF SEVERE CORE DAMAGE IS TODAY, SINCE Do.As HAVE NOT BEEN DONE ON ALL PLANTS, PLANT-SPECIFIC DESIGN AND PROCEDURE DIFFERENCES CAN MAKE SUBSTANTIAL DIFFERENCES IN THE RESULTS OF PRAs, THERE ARE SUBSTANTIAL UNCERTAINTIES IN THE RESULTS OF PRAs,. PLANTS HAVE MADE DESIGN AND PROCEDURE CHANGES BASED ON INSIGHTS FROM PRAs, AND ALL PLANTS ARE IN VARIOUS STAGES OF IMPLEMENTING REQUIREMENTS RESULTING FROM THE TMI ACCIDENT.
HOWEVER, IT IS INFORMATIVE TO COMPARE ESTIMATES BASED ON PREVIOUS PRA RESULTS WITH INTERIM RESULTS FROM THE RESEARCH CURRENT PERFORMED SUPPORTING THE RISK REBASELINING' 0F REFERENCE PLAN REBASELINING MEANS UPDATING THE PRA TO REFLECT: '(1) CHAN PLANT DESIGN AND PROCEDURES, AND (2) IMPROVEMENTS'IN THE STATE-OF-THE-ART OF PRA.
REFERENCE ENCLOSURE 1 WH'ICH LISTS SOME OF THESE CHANGES.
2 NOTE:
SEVERE CORE DAMAGE IS THE STATE THAT IS QUANTIFIED IN' PRAs, AND IT IS DEFINED AS THE SITUATION WHERE THERE IS INSUFFICIENT CORE COOLING TO MAINTAIN FUEL INTEGRITY.
HOWEVER, SEVERE CORE DAMAGE MIGHT NOT PROCEED TO EXTENSIVE MELTING AND PENETRATION OF THE REACTOR PRESSURE VESSEL, AS EXEMPLIFIED BY THE TMI-2 ACCIDENT.
WE CANNOT AT PRESENT QUANTIFY THE DISTINCTION BETWEEN SEVERE CORE DAMAGE AND A
" CORE MELT" THAT PENETRATES THE VESSEL.
THE STAFF PROVIDED, ON SEPTEMBER 16, 1985, AN INSERT FOR PAGE 39, LINE 827 0F THE RECORD OF THE SUBCOMMITTEE ON ENERGY CONSERVATION AND POWER'S APRIL 17, 1985, HEARING ON THE NRC BUDGET.
IN THIS INSERT, THE STAFF ESTIMATED AN AVERAGE SEVERE CORE DAMAGE FREQUENCY OF 3 X 10-4 PER REACTOR-YE (RY) BASED UPON SIX RECENT PRAs COVERING NINE REACTOR UNITS **.
THESE STUDIES WERE ORIGINALLY PERFORMED BY THE f
OWNER-0PERATORS OF THE PLANTS.
IN THREE CASES, THE NRC STAFF REVISED THE OWNER-CALCULATED ACCIDENT FREQUENCY UPWARD, AND THE AVERAGE OF 3 X 10-4 SEVERE CORE DAMAGE ACCIDENTS PER RY REFLECTS THESE NRC REVISIONS.
IN EACH CASE, CHANGES ACTUALLY MADE IN PLANT DESIGN AND OPERATION TO REFLECT POST-TMI ORDERS WERE CONSIDERED IN THE PRAs.
UNDER THE ASSUMPTION THAT 3 X 10-4 SEVERE CORE DAMAGE ACCIDENTS THIS NUMBER INCLUDESA SMALL CONTRIBUTION DUE TO EXi'ER EVENTS.
3 PER RY WAS THE INDUSTRY AVERAGE, THE STAFF IN 1985 CAlfULATED THAT THE PROBABILITY OF ONE OR MORE SUCH ACCIDENTS IS 0.45 IN A POPULATION OF 100 PLANTS OPERATING OVER THE NEXT 20 YEARS.
THE NRC RECOGNIZES THAT PLANTS WITH PARTICULARLY PROMINENT VULNERABILITIES, OR OUTLIERS, CONTROL THE INDUSTRY AVERAGE ACCIDENT FREQUENCY.
THIS EFFECT IS A STRONG REASON WHY THE NRC FOCUSSED ATTENTION ON THE SEARCH FOR OUTLIERS; THAT IS, THE PLANTS THAT MAY BE VULNERABLE TO A SPECIFIC SEQUENCE THAT CAN LEAD TO SEVERE CORE DAMAGE.
WE BELIEVE OUR REGULATORY PROCESS IS AIMED AT
~
CONTINUALLY REDUCING THE LIKELIHOOD OF SEVERE REACTOR ACCIDENTS BY REQUIRING PLANT-DESIGN IMPROVEMENTS (E.G., THE ATWS IMPROVEMENTS),
IMPROVED OPERATOR TRAINING, IMPROVED MANAGEMENT OF PLANT OPERATIONS, AND IMPLEMENTATION OF THE COMMISSION'S SEVERE ACCIDENT POLICY STATEMENT. 'SUCH ACTIVITIES ARE A CONTINUING PART OF OUR
~
REGULATORY PROCESS.
~
AS STATED IN THE APRIL 1985 DOCUMENT, THERE ARE REASONS TO SUSPECT THAT THE ASSUMED VALUE OF 3 X 10-4 PER RY MIGHT BE CONSERVATIVE INDEED, MORE RECENT ANALYSES OF REACTOR ACCIDENTS APPEAR TO BEAR THIS OUT.
RECENTLY, THE OFFICE OF NUCLEAR REGULATORY RESEARCH HAS BEEN REBASELINING THE RISK FROM FIVE REFERENCE PLANTS.USING UP-TO-DATE PLANT INFORMATION AND PRA l
O n.
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4 TECHNIQUES TO ESTIMATE THE MEAN VALUE OF THE FREQUENCY OF SEVERE CORE DAMAGE ACCIDENTS DUE TO INTERNAL ACCIDENT INITIATORS. THESE RESULTS WILL BE PUBLISHED IN SEPTEMBER AS DRAFT NUREG-1150.
SOME INTERIM RESULTS FROM THESE STUDIES ARE PROVIDED BELOW.
INTERIM SEVERE CORE DAMAGE PLANT VENDOR CONTAIKMENT FREQUENCY (PERRY)
SURRY W
SUBATMOSPHERIC 2.5X10-5 PEACH BOTTOM GE MARK I IX10-5 SEQUOYAH W
ICE CONDENSER 1.1X10-4 GRAND GULF GE MARK III 2.6X10-5 wii W
LARGE DRY 1.5X10-4*
- THIS ESTIMATE IS BASED ON THE PREVIOUS STAFF ANALYSIS OF THE UTILITY'S PRA.
THERE WAS NO ATTEMPT MADE TO REBASELINE THE PRA.
THESE LATEST ANALYSES ARE OF PLANTS WHICH HAVE UNDERGONE PRA
- !A'.YSIS IN THE PAST AND WHICH HAVE INCORPORATED NOT ONLY POST-TMI i'.ANDATED CHANGES BUT OTHER REFINEMENTS IDENTIFIED IN PRA ANALYSIS TO IMPROVE RELIABILITY.
IN SOME CASES IT CAN BE SEEN THAT IMPROVEMENTS HAVE LOWERED THE SEVERE CORE DAMAGE FREQUENCY FROM INTERNAL EVENTS TO LEVELS AS LOW AS 1 x 10-5 PER YEAR.
A'ND EVEN IN THESE PLANTS, NOT ALL OF THE POTENTIALLY PRACTICAL liELIABILITY IMPROVEMENTS HAVE BEEN MADE.
l 1
5 TO UNDERSTAND THE POTENTIAL INDUSTRY WIDE SIGNIFICANCE OF THIS RELIABILITY IMPROVEMENT PROCESS, WHICH IS STILL GOING ON, ONE MIGHT POSTULATE THAT THESE INTERIM VALUES REPRESENT INDUSTRY AVERAGES FOR REACTORS OF THESE' TYPES.
IN THIS CASE THE INDUSTRY AVERAGE SEVERE CORE DAMAGE FREQUENCY WOULD BECOME ABOUT PER RY, AND THE LIKELIHOOD OF A SEVERE CORE DAMAGE ACCIDENT OCCURRING IN THE NEXT 20 YEARS IN A POPULATION OF 100 PLANTS WOULD BE 0.12, OR ONE CHANCE IN 8.
THE NRC STAFF BELIEVES THAT SUCH REDUCTIONS IN SEVERE CORE DAMAGE FREQUENCY CAN INDEED BE ACHIEVED OR EVEN IMPROVED FURTHER, BY FULL IMPLEMENTATION OF THE TMI FIXES AND AGGRESSIVE IMPLEMENTATIN OF THE COMMISSION'S SEVERE ACCIDENT POLICY STATEMENT.
li-CORE MELT IS LIKELY HOW CAN PLANTS BE SAFE EN0 UGH?
IF THERE IS A DISTINCT POSSIBILITY OF A SEVERE CORE-DAMAGE ACCIDENT IN THE NEXT 20 YEARS, EVEN ONLY ONE CHANCE IN 8, ARE REACTORS SAFE EN0 UGH?
OVERALL WE BELIEVE THE ANSWER IS "YES",
~
THAT REACTORS PROVIDE A HIGH LEVEL OF PROTECTION FOR THE PUBLIC FROM THE RADIOLOGICAL RELEASES THAT MIGHT RESULT FROM REACTOR l
ACCIDENTS.
NOT ALL SEVERE CORE DAMAGE ACCIDENTS ARE LIKELY TO PROGRESS TO SUBSTANTIAL CORE MELTING, AND EVEN WITH SUBSTANTIAL CORE MELTING THE CORE MATERIAL MAY NOT BE RELEASED FROM THE REACTOR VESSEL.
THREE MILE ISLAND INVOLVED A SEVERE CORE DAMAGE i
ACCIDENT IN WHICH THE REACTOR CORE BEGAN TO MELT, YET MOST OF THE RADI0 ACTIVE MATERI ALS RELEASED FROM THE FUEL IN THE AC'C'IDE REMAINED WITHIN THE PLANT, THE HEAVIER CORE AND FISSION l
6 PRODUCT MATERIALS STAYED IN THE REACTOR COOLANT SYSTEM AND THE MORE VOLATILE MATERIALS STAYED INSIDE CONTAINMENT. BASED ON THE PRESIDENT'S COMMISSION REPORT ON THE ACCIDENT AT THREE MILE ISLAND, EVEN IF THE TMI CORE HAD MELTED DOWN COMPLETELY, AND 11ELTED THROUGH THE REACTOR VESSEL, THE TMI CONTAINMENT,WOULD HAVE HELD.
ONLY A SMALL FRACTION OF THE LARGE-SCALE CORE-MELT ACCIDENTS ARE EXPECTED TO HAVE MAJOR OFFSITE RELEASES AND THUS SIGNIFICANT OFFSITE CONSEQUENCES, SINCE ONLY THOSE ACCIDENTS THAT INVOLVE THE EARLY FAILURE OR BYPASS OF THE REACTOR CONTAINMENT STRUCTURE COULD RESULT IN SIGNIFICANT RELEASES OF RADIDACTIVITY.
RESEARCH RESULTS
~
ON SEVERE ACCIDENT TECHNOLOGY ARE SHOWING THAT CONTAINMENT STRUCTURES HAVE MORE SAFETY MARGIN THAN PREVIOUSLY ESTIMATED AND THEREFORE LESS RADIDACTIVITY WILL LIKELY BE RELEASED FROM CONTAINMENT STRUCTURES IN THE EVENT OF A CORE-MELT ACCIDENT THAN HAD PREVIOUSLY BEEN POSTULATED.
THIS IS NOT TO SAY THAT THERE ARE NO LOW PROBABILITY ACCIDENT SEQUENCES THAT COULD BREACH CONTAINMENT.
THERE ARE SOME KINDS OF ACCIDENT SEQUENCES THAT COULD CAUSE FAILURE OF ANY CONTAINMENT DESIGN, ALTHOUGH A SUBSTANTIAL FRACTION OF THE RADI0 ACTIVE MATERIAL WOULD BE TRAPPED WITHIN THE PLANT.
OO
ENCLOSURE 1 SUWARY OF !%JOR DIFFERENCES THE FOLLOWING StW ARIZES THE MJOR IDENTIFIED REASONS FOR THE DIFFERENCES BETWEEN THE PREVIOUS PRA RESULTS AND THE REBASELINED RESULTS FOR SURRY, PEACH BOTTOM, SEQUOYAH, AND GRAND GULF. SUCH A COMPARISON CANNOT BE PADE FOR ZION, SINCE THE PREVIOUS STAFF EVALUATION OF THE LICENSEE'S PRA WILL BE USED FOR THE
~
"REBASELINED" RESULT IN NUREG-1150.
THE SURRY PLANT HAS BEEN MODIFIED SIGNIFICANTLY Sil4CE 1975, REFLECTING #4 EFFORT BY THE UTILITY TO RESPOND TO THE LESSONS LEARNED FROM WASH-1400, AS WELL
.I RESPONDING TO POST-TMI FIXES AND OTHER OPERATIONAL E'XPERIENCE. SPECIF LY, THERE ARE NOW CROSS-TIES BETWEEN ThE HPI AND AFW SYSTEPS OF UNITS 1 AND 2, THE BORON INJECTION TANK HAS BEEN REMOVED, THE SHUNT TRIP HAS BEEN ADDED., ALS TRAINING HAS BEEN IMPROVED, AND THE LPI VALVE TESTING FREQUENCY HAS BEEN MODIFIED TO LOWER THE LIKELIHOOD OF " EVENT V."
ADVN4CES IN UNDERSTNOING OF THERf%L-HYDRAULIC BEHAVIOR HAVE LED TO SEVERAL MODELING DIFFERENCES AS WELL. FOR EXAMPLE, IMPROVED ANALYSIS OF REACTOR CAVITY NO CONTAltNENT StkiP LEVELS DURING A St%LL LOCA HAVE INDICAT'ED. SEQUENCE S 2
WILL NOT RESULT IN CORE DAMAGE, THE SUCCESS CRITERION FOR CONTAINMENT SPRAY OPERATION DURING RECIRCULATION HAS BEEN MODIFIED TO REQUIRE 1/4 TRAINS RA W ER THAN 2/4, CREDIT FOR FEED AND BLEED CORE COOLING HAS BEEN ADDED, AND SEQUENCE S G HAS CREN ELIMINATED BECAUSE 8 DAYS ARE AVAILABLE FOR CORRECTIVE ACTION l
2
2 BEFORE CONTAINMENT FAILURE. THESE CHANGES HAVE LED TO A REDUCTION IN THE PREDICTED FREQUENCY OF SEVERE CORE DAMAGE NO SOME CHN4GES IN DOMitW4T SEQUENCES.
THE SEQUOYAH ANALYSIS HAS BEEN SIMILARLY AFFECTED BY PLANT CHANGES AND THE CAPABILITY FOR IMPROVED MODELING. SPECIFICALLY, THE RPS SHUNT TRIP HAS BEEN ADDED AND CREDIT GIVEN FOR IMPROVED AL S TRAINING, SYMPTOM ORIENTED PROCEDURES HAVE BEEN IMPLEMEfGED, LPI CHECK VALVES ARE SUBJECTED TO INCREASED TEST 1NG, 40 ADDITIONAL PROCEDURES ARE NOW AVAILABLE FOR ENSURING THE REMOVAL OF THE REFUELING DRAIN PLUG. N DELING CHANGES INCLUDE CREDIT FOR BLEED N O FEED CORE COOLING, THE SUCCESS STATE FOR HPl HAS BEEN MODIFIED TO 1/4 TRAINS RATHER THAN 2/4, IN CERTAIN. INSTANCES ONLY ONE RHR OR CONTAltNEtn SPRAY HEAT EXCHANGER IS REQUIRED, RATHER THAN TWO, RCP SEAL LOCAS HAVE BEEN ADDED, AND THE STATION BLACKOUT ANALYSIS IS MORE DETAILED. HOWEVER, THE NEW ANALYSES ALSO IDENTIFIED A NEW DOMINANT SEQUENCE-A COPtON-CAUSE FAILURE MODE IN THE SERVICE WATER SYSTEM.
AT PEACH BOTTOM THE PLANT HAS ALSO BEEN MODIFIED SINCE 1975. NST SIGNIF1-CANTLY, CREDIT HAS BEEN GIVEN FOR IMPLEMENTATION OF THE ATWS RULE, CONTAltNENT VENTING PPDCEDUaES ARE AVAILABLE, AND ADS NOW ACTUATES ON LOW WATER LEVEL ALONE, RATHER THAN LOW LEVEL AND HIGH DRY WELL PRESSURE. CREDIT HAS ALSO BEEN GIVEN FOR IMPROVED RELIABILITY OF DIESEL GENERATORS, BASED dN PLANT-SPECIFIC
~
DATA RESULTING FROM A HIGHLY EFFECTIVE RELIABILITY PROGRAM, AND AN EXTENS]VE HWAN RELIABILITY ANALYSIS HAS BEEN PERFORMED ON OPERATOR PERFORMAf4CE l
ALS.
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WE HAVE NOT YET CRITICALLY REVIEWED THE GRAND GULF RESULTS. BASED O PRELIMINARY REVIEW, IT APPEARS CHANGES RESULT FROM PLANT N DIFICATIONS ASSOCIATED WITH THE ATWS RULE, MORE DETAILED MODELING OF STATION BLACKOUT SEQUENCES INCLUDING THE POTENTIAL FOR LONG-TERM LOSS OF HPCS AND R LOSS OF CONTROL POWER OR PLNP SEAL COOLING, AND MORE REALISTIC CONSIDERATION OF RECOVERY FROM ACCIDENT SEQUENCES ItNOLVING LONG-TERM LOSS OF THE RHR 9
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Congresswoman Boxer For operating plants without PRAs, implementation of the NRC's Severe Accident Policy will involve a systematic examination of all operating reactors to identify and remedy outstanding contributors (severe accident vulnerabilities) to severe core damage accident probability and risk. The NRC staff will issue guidance for the individual plant examinations by the end of 1986, and the examinations will be performed by the utilities starting in early 1987.
For new nuclear power plant applications, plant-specific PRA studies are required and will provide for identification of plant-specific vulnerabilities that can be corrected in the design stage.
My May 19, 1986 memorandum to the Commissioners (Attachment 1) provides a discussion on our earlier and current estimates of the industry-wide likelihood of severe core damage accidents. Concern regarding the acceptable level of severe core damage accident risk is being addressed in the Commission's Safety Goal Policy Statement which was recently approved by the Commission. The final form of the policy establishes a goal such that the mortality risk to an individual member of the public from severe core damage accidents be only a very smal.1 fraction (one one-thousandth) of similar background risk from other causes. The final text is presently being prepared for publication in the Federal' Register. We would be pleased to send you a copy as soon as the statement is available if you so desire.
A copy of this letter is being placed in the NRC's Public Document Room for information of the interested members of the public.
Sincerely, 1
\\
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Victor Stello, Jr.
Executive Director for Opd ations
Attachment:
May 19, 1986 memorandum of Victor Stello, Jr. to the Commissioners cc: NRC PDR
- SEE PREVIOUS CONCURRENCE HRDenton VStello Oy/;/86 07/ /86 Office DSR0:RRAB*.....:DSR0:RRAB*..:DSR0:RRAB*..:DSR0:DD*....:DSR0:D*.....:NRR:D Surname SAcharya: sj.... :JERosenthal. : FJCongel.... : BWSheron.... :TPSpeis..... : RVol mer.... :
06/21/86.......:06/23/86....:06/23/86....:06/23/86....:06/23/86....:Op/[/86....:
Date OFFICIAL RECORD COPY
?
Congresswoman Boxer For operating plants without PRAs, implementation of the NRC's Severe Accident Policy will involve a systematic examination of all operating reactors to identify and remedy outstanding contributors (severe accident vulnerabilities) to severe core damage accident probability and risk. The NRC staff will issue guidance for the individual plant exa inations by the end of 1986, and the examinations will be performed by th utilities starting in early 1987.
For new nuclear power plant applidations, plant-specific PRA studies are required and will provide for identifica' tion of plant-specific vulnerabilities that can be corrected in the design stage'
/, my May 19, 1986 memo andum te the Commissioners, provides a discussion on our earlier and curre t estimates of the industry-wide likelihood of severe core damage'acc ents. Concern regarding the acceptable level of severe core damage accident sk is being addressed in the Comission's Safety Goal Policy Statem t which was recently approved by the Commission. The final form of the poli y would require the mortality risk to an individual member of the public from s4 vere core damage accidents to be only a very small fraction (one one-thousandth)\\of similar background risk from other Thepolicywouldrequirestringent\\restrictionsonseverecoredamage causes.
probability for individual plants, and stringent reoairements on containment performance capability to limit the release of radiunuclides should a severe core damage accident occur.
A copy of this letter is being placed in the NRC's Public Document Room for information of the interested members of the public.
A Sincerely,
\\
Victor Stello, Jr.
~
Executive Director for Operations s
Attachment:
May 19, 1986 memorandum of Victor Stello, Jr. to the Commissioners cc: NRC PDR EDO NRR:DD VStello RVollmer 06/ /86 f,
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0FFICIAL RECORD COPY
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UNITED STATES
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~i NUCLEAR REGULATORY COMMISSION
- y WASHINGTON, D. C. 20555
(...../
JUL 181986 The Honorable Barbara Boxer U.S. House of Representatives Washington, DC 20515 Deer Congresswomen Boxer:
In your letter to the President dated May 1,1986, you requested information regarding the safety of U.S. reactors as a result of the Chernobyl reactor accident in the Soviet Union.
I expect that the response to your concern about the Hanford N-reactor and the other four reactors in Savannah River, South Carolina cited in your letter will be provided by the D0E. The following discussion provides our response to your concern about the safety of the commercial nuclear power plants in the U.S. licensed by NRC.
All plants licensed to operate were reviewed in accordance with Coninission guidelines, rules, and regulations based on sound engineering and defense-in-depth principles to protect the health and safety of the public.
The engineered safety features (ESFs) and the containment structures of the plants are capable of coping with a spectrum of design basis accidents (DBAs),
such as an accident involving the total rupture of the largest reactor coolant or steam pipe. Calculated radiological doses to the public due to release of radionuclides from the DBAs are limited by the ESFs and the containment structure to below the 10 CFR Part 100 guideline values (25 rem to the wtole body and 300 rem to the thyroid). Moreover, the dose calculations employed for Ethe DBAs utilize substantially conservative assumptions, and realistic doses to the public from these DBAs are expected to be significantly less than the calculated values. We have also been studying accidents more severe than the DBAs for some time. This effort was accelerated following the TMI accidcnt. Based on these studier, we have concluded that all containment structures surrounding nuclear power rea: tors provide some degree of protection to the public should an accident more severe than the DBAs occur.
Probabilistic risk assessment (PRA) studies have been performed for a number of operating nuclear power plants. These studies assess plant vulnerability for severe core damage accidents.* Some of these plants have made design and procedure improvements based on insights from the PRAs. However, we do not have PRAs for all plants. Therefore, we have not ranked the nuclear power plants in terms of severe core damage accident vulnerability, or identified the plants with greatest risk.
- The phrase " severe core damage accident" is used to describe an accident in which the reactor fuel integrity is lost due to inadequate core cooling. The extent of the core damage can range from partial core melting without penetration of the reactor vessel as exemplified by the TMI-2 accident to an accident in which the fuel undergoes extensive to complett melting.
Congresswomen Boxer For operating plants without PRAs, implementation of the NRC's Severe Accident Policy will involve a systematic examination of all operating reactors to identify and remedy outstanding contributors (severe accident vulnerabilities) to severe core damage accident probability and risk. The NRC staff will issue guidance for the individual plant examinations by the end of 1986, and the examinations will be performed by the utilities starting in early 1987. For new nuclear power plant applications, plant-specific PRA studies &re required and will provide for identification of plant-specific vulnerabilities that can be corrected in the design stage.
My May 19, 1986 memorandum to the Comissioners (Attachment 1) provides a discussion on our earlier and current estimates of the industry-wide likelihood of severe core damage accidents. Concern regarding the acceptable level ~ of severe core damage accident risk is being addressed in the Commission's Safety Goal Policy Statement which was recently approved by the Comission. The final form of the policy establishes a goal such that the mortality risk to an individual member of the public from severe core damage accidents be only a very small fraction (one one-thosandth) of similar background risk from other causes. The final text is presently being prepared for publication in the Federal Register. We would be pleased to send you a copy as soon as the statement is available if you so desire.
Sincerel lf Victor Stdllo, Jr.
Executive Director for Operations
Attachment:
May 19, 1986 memorandum of Victor Stello, Jr. to the Commissioners cc: NRC PDR b
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UNITED STATES Cys: Stello
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p NUCLEAR REGULATORY COMMISSION Roe WASHWGTON, D. C. 20655 Rehm gk
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May 19, 1986 h.
ue Taylor Heltemes GCunningham EDO R/F MEMORANDUM FOR:
Chairman Palladino Commissioner Roberts Commissioner Asselstine Commissioner Bernthal Commissioner Zech
.-/
,' '
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i FROM:
Victor Stello, Jr.'
J f,-
Executive Director f6r Operations
SUBJECT:
FREQUENCY OF SEVERE CORE DAMAGE ACCIDENT Enclosed for your information is the staff's answer to the question "What is the estimated likelihood of a severe core damage accident due to internal initiators for all operating reactors (100 plants) over the next 20 years?"
The enclosed
.information was developed by the staff based on the latest information currently being developed in the PRA rebaselining of six reference plants by the Office of Nuclear Regulatory Research.
Enclosure:
As stated cc:
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CORE MELT LIKELTHOOD QUESTION.
WHAT IS THE ESTIMATED LIKEllH00D OF A SEVERE CORE DAMAGE ACCIDENT DUE TO INTERNAL INITIATORS FOR ALL OPERATING REACTORS (100 PLANTS OVER THE NEXT 20 YEARS?
ANSWER.
IT IS IMPOSSIBLE TO ESTIMATE WITH PRECISION WHAT THE ACTUAL INDUSTRY-WIDE LIKEL'IHOOD OF SEVERE CORE DAMAGE IS TODAY, SINCE DD.As HAVE NOT BEEN DONE ON ALL PLANTS, PLANT-SPECIFIC DESIGN AND PROCEDURE DIFFERENCES CAN MAKE SUBSTANTIAL DIFFERENCES IN THE RESULTS OF PRAs, THERE ARE SUBSTANTIAL UNCERTAINTIES IN THE RESULTS OF PRAs, PLANTS HAVE MADE DESIGN AND PROCEDURE CHANGES BASED ON INSIGHTS FROM PRAs, AND ALL PLANTS ARE IN VARIOUS STAGES OF IMPLEMENTING REQUIREMENTS RESULTING FROM THE TMI ACCIDENT.
HOWEVER, IT IS INFORMATIVE TO COMPARE ESTIMATES BASED ON PREVIOUS PRA RESULTS WITH INTERIM RESULTS FROM THE RESEARCH CURRENTLY BEING PERFORMED SUPPORTING THE RISK REBASELINING' 0F REFERENCE PLANTS.
REBASELINING MEANS UPDATING THE PRA TO REFLdCT: '(1) CHANGES IN PLANT DESIGN AND PROCEDURES, AND (2) IMPROVEMENTS'IN THE STATE-OF-THE-ART OF PRA.
REFERENCE ENCLOSURE 1 WH'ICH LISTS SOME OF THESE CHANGES.
2 NOTE:
SEVERE CORE DAMAGE IS THE STATE THAT IS OUANTIFIED IN PRAs, AND IT IS DEFINED AS THE SITUATION WHERE THERE IS INSUFFICIENT CORE COOLING TO MAINTAIN FUEL INTEGRITY.
i HOWEVER, SEVERE CORE DAMAGE MIGHT NOT PROCEED TO EXTENSIVE MELTING AND PENETRATION OF THE REACTOR PRESSURE VESSEL, AS EXEMPLIFIED BY THE TMI-2 ACCIDENT.
WE CANNOT AT PRESENT QUANTIFY THE DISTINCTION BETWEEN SEVERE CORE DAMAGE AND A
" CORE MELT" THAT PENETRATES THE VESSEL.
THE STAFF PROVIDED, ON SEPTEMBER 16, 1985, AN INSERT FOR PAGE 39, LINE 827 0F THE RECORD OF THE SUBCOMMITTEE ON ENERGY CONSERVATION AND POWER'S APRIL 17, 1985, HEARING ON THE NRC BUDGET.
IN THIS INSERT, THE STAFF ESTIMATED AN AVERAGE SEVERE CORE DAMAGE FREQUENCY OF 3 X 10-4 PER REACTOR-Y (RY) BASED UPON SIX RECENT PRAs COVERING NINE REACTOR UNITS".
THESE STUDIES WERE ORIGINALLY PERFORMED BY THE i
OWNER-OPERATORS OF THE PLANTS.
IN THREE CASES, THE NRC STAFF REVISED THE OWNER-CALCULATED ACCIDENT FREQUENCY UPWARD, AND THE AVERAGE OF 3 X 10-4 SEVERE CORE DAMAGE ACCIDENTS PER RY REFLECTS THESE NRC REVISIONS.
IN EACH CASE, CHANGES ACTUALLY MADE IN PLANT DESIGN AND OPERATION TO REFLECT POST-TM1 ORDE WERE CONSIDERED IN THE PRAs.
UNDER THE ASSUMPTION THAT 3 X 10-4 SEVERE CORE DAMAGE ACCIDENTS THIS NUMBER INCLUDESA SMALL CONTRIBUTION DUE TO EXT'E
~
EVENTS.
3 PER RY WAS THE INDUSTRY AVERAGE, THE STAFF IN 1985 CALCULATED THAT THE PROBABILITY OF ONE OR MORE SUCH ACCIDENTS IS 0.45 IN A POPULATION OF 100 PLANTS OPERATING OVER THE NEXT 20 YEARS.
4-THE NRC RECOGNIZES THAT PLANTS WITH PARTICULARLY PROMINENT VULNERABILITIES, OR OUTLIERS, CONTROL THE INDUSTRY AVERAGE ACCIDENT FREQUENCY.
THIS EFFECT IS A STRONG REASON WHY THE NRC FOCUSSED ATTENTION ON THE SEARCH FOR OUTLIERS; THAT IS, THE PLANTS THAT MAY BE VULNERABLE TO A SPECIFIC SEQUENCE THAT CAN LEAD T0 SEVERE CORE DAMAGE.
WE BEllEVE OUR REGULATORY PROCESS IS AIMED AT t
CONTINUALLY REDUCING THE LIKEllH00D OF SEVERE REACTOR ACCIDENTS BY REQUIRING PLANT-DESIGN IMPROVEMENTS (E.G., THE ATWS IMPROVEMENTS),
IMPROVED OPERATOR TRAINING, IMPROVED MANAGEMENT OF PLANT OPERATIONS, AND IMPLEMENTATION OF THE COMMISSION'S SEVERE ACCIDENT POLICY STATEMENT.
SUCH ACTIVITIES ARE A CONTINUING PART OF OUR REGULATORY PROCESS.
ASSTATEDkNTHEAPRIL1985 DOCUMENT,THEREAREREASONSTOSUSPECT THAT THE ASSUMED VALUE OF 3 X 10-4 PER RY MIGHT BE CONSERVATIVE.
INDEED, MORE RECENT ANALYSES OF REACTOR ACCIDENTS APPEAR TO BEAR THIS OUT.
RECENTLY, THE OFFICE OF NUCLEAR REGULATORY RESEARCH HAS BEEN REBASELINING THE RISK FROM FIVE REFERENCE PLANTS.USING UP-TO-DATE PLANT INFORMATION AND PRA e@
O a
4 TECHNIQUES TO ESTIMATE THE MEAN VALUE OF THE FREQUENCY OF SEVERE CORE DAMAGE ACCIDENTS DUE TO INTERNAL ACCIDENT INITIATORS.
THESE RESULTS WILL BE PUBLISHED IN SEPTEMBER AS DRAFT NUREG-1150.SOME INTERIM RESULTS FROM THESE STUDIES ARE PROVIDED BELOW.
INTERIN SEVERE CORE DAMAGE PLANT VENDOR CONTAINMENT FREQUENCY (PERRY)
SURRY W
SUBATMOSPHERIC 2.6X10-5 PEACH BOTTOM GE MARK I IX10-5 SEQUOYAH W
ICE CONDENSER 1.1X10-4 GRAND GULF GE MARK 111, 2.6X10-5 u vii W
LARGE DRY 1.5X10-4*
- THIS ESTIMATE IS BASED ON THE PREVIOUS STAFF ANALYSIS OF THE UTILITY'S PRA.
THERE WAS NO ATTEMPT MADE TO REBASELINE THE PRA.
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THESE LATEST ANALYSES ARE OF PLANTS WHICH HAVE UNDERGONE PRA
"!aLYSIS IN THE PAST AND WHICH HAVE INCORPORATED NOT ONLY POST-TMI F.ANDATED CHANGES BUT OTHER REFINEMENTS IDENTIFIED IN PRA ANALYSIS T0 IMPROVE RELIABILITY.
IN SOME CASES IT CAN BE SEEN THAT IMPROVEMENTS HAVE LOWERED THE SEVERE CORE DAMAGE FREQUENCY FROM INTERNAL ~ EVENTS TO LEVELS AS LOW AS 1 x 10-5 PER YEAR.
A'ND EVEN IN THESE PLANTS, h0T ALL OF THE POTENTIALLY PRACTICAL iiELIABILITY IMPROVEMENTS HAVE BEEN MADE.
5 TO UNDERSTAND THE POTENTIAL INDUSTRY WIDE SIGNIFICANCE OF THIS RELIABILITY IMPROVEMENT PROCESS, WHICH IS STILL GOING ON, ONE MIGHT POSTULATE THAT THESE INTERIM VALUES REPRESENT INDUSTRY AVERAGES FOR REACTORS OF THESE TYPES.
IN THIS CASE THE INDUSTRY AVERAGE SEVERE CORE DAMAGE FRE00ENCY WOULD BECOME ABOUT 6 PER RY, AND THE LIKELIHOOD OF A SEVERE CORE DAMAGE ACCIDENT OCCURRING IN THE NEXT 20 YEARS IN A POPULATION OF 100 PLANTS WOULD BE 0.12, OR ONE CHANCE IN 8.
THE NRC STAFF BELIEVES THAT SUCH REDUCTIONS IN SEVERE CORE DAMAGE FREQUENCY CAN INDEED BE ACHIEVED OR EVEN IMPROVED FURTHER, BY FULL IMPLEMENTATION OF THE TMI FIXES AND AGGRESSIVE IMPLEMENTATIN OF THE COMMISSION'S SEVERE ACCIDENT POLICY STATEMENT.
It-CORE MELT IS LIKELY HOW CAN PLANTS BE SAFE EN0 UGH?
IF THERE IS A DISTINCT POSSIBILITY OF A SEVERE CORE-DAMAGE ACCIDENT IN THE NEXT 20 YEARS, EVEN ONLY ONE CHANCE IN 8, ARE REACTORS SAFE EN00G'..?
OVERALL WE BELIEVE THE ANSWER IS "YES",
~
THAT REACTORS PROVIDE A HIGH LEVEL OF PROTECTION FOR THE PUBLIC FROM THE RADIOLOGICAL RELEASES THAT MIGHT RESULT FROM REACTOR ACCIDENTS.
NOT ALL SEVERE CORE DAMAGE ACCIDENTS ARE LIKELY TO PROGRESS TO SUBSTANTIAL CORE MELTING, AND EVEN WITH SUBSTANTIAL CORE MELTING THE CORE MATERIAL MAY NOT BE RELEASED FROM THE REACTOR VESSEL.
THREE MILE ISLAND INVOLVED A SEVERE CORE DAMAGE ACCIDENT IN WHICH THE REACTOR CORE BEGAN TO MELT, YET MOST OF THE RADIDACTIVE MATERI ALS RELEASED FROM THE FUEL IN THE AC'C' REMAINED WITHIN THE PLANT, THE HEAVIER CORE AND FISSION
6 PRODUCT MATERIALS STAYED IN THE REACTOR COOLANT SYSTEM AND THE MORE VOLATILE MATERI ALS STAYED INSIDE CONTAINMENT.
BASED ON THE PRESIDENT'S COMMISSION REPORT ON THE ACCIDENT AT THREE MILE ISLAND, EVEN IF THE TMI CORE HAD MELTED DOWN COMPLETELY, AND MELTED THROUGH THE REACTOR VESSEL, THE TMI CONTAINMENT,WOULD HAVE
- HELD, ONLY A SMALL FRACTION OF THE LARGE-SCALE CORE-MELT ACCIDENTS ARE EXPECTED TO HAVE MAJOR OFFSITE RELEASES AND THUS SIGNIFICANT OFFSITE CONSEQUENCES, SINCE ONLY THOSE ACCIDENTS THAT INVOLVE THE EARLY FAILURE OR BYPASS OF THE REACTOR CONTAINMENT STRUCTURE COULD RESULT IN SIGNIFICANT RELEASES OF RADIDACTIVITY.
RESEARCH RESULTS ON SEVERE ACCIDENT TECHNOLOGY ARE SHOWING THAT CONTAINMENT STRUCTURES HAVE MORE SAFETY MARGIN THAN PREVIOUSLY ESTIMATED AND THEREFORE LESS RADI0 ACTIVITY WILL LIKELY BE RELEASED FROM CONTAINMENT STRUCTURES IN THE EVENT OF A CORE-MELT ACCIDENT THAN HAD PREVIOUSLY BEEN POSTULATED.
1 THIS IS NOT TO SAY THAT THERE ARE NO LOW PROBABILITY ACCIDENT SEQUENCES THAT COULD BREACH CONTAINMENT.
THERE ARE SOME KINDS OF ACCIDENT SEQUENCES THAT COULD CAUSE FAILURE OF ANY CONTAINMENT DESIGN, ALTHOUGH A SUBSTANTIAL FRACTION OF THE RADI0 ACTIVE MATERIAL WOULD BE TRAPPED WITHIN THE PLANT, 9
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ENCLOSURE 1
' SUWARY OF MAJOR DIFFERENCES TE FOLLOWING Stit%RIZES THE MMOR IDENTIFIED REASONS FOR THE DIFFERENCES BETWEEN THE PREVIOUS PRA RESULTS AND THE REBASELINED RESULTS FOR SURRY, PEACH BOTTOM, SEQUOYAH, AND GRAND GULF. SUCH A COMPARISON CANNOT BE f%DE FOR ZION, SINCE THE PREVIOUS STAFF EVALUATION OF THE LICENSEE'S PRA WILL BE USED FOR THE "REBASELINED" RESULT IN NUREG-1150.
THE SURRY PLANT HAS BEEN MODIFIED SIGNIFICANTLY SINCE 1975, REFLECTING AN EFFORT BY THE UTILITY TO RESPOND TO THE LESSONS LEARNED FROM WASH-1400, AS WELL
.I RESPONDING TO POST-TMI FIXES AND OTHER OPERATIONAL EXPERIENCE. SPECIFICAL-LY, THERE ARE NOW CROSS-TIES BETWEEN THE HPI AND AFW SYSTEMS OF UNITS 1 AND 2, THE BORON IlUECTION TANK HAS BEEN REMOVED, THE SHUNT TRIP HAS BEEN ADDED., ATkS TRAINING HAS BEEN IMPROVED, AND THE LPI VALVE TESTING FREQUENCY HAS BEEN MODIFIED TO LOWER THE LIKELIHOOD OF " EVENT V."
ADVANCES IN UNDERSTANDING OF THERMAL-HYDRAULIC BEHAVIOR HAVE LED TO SEVERAL MODELING DIFFERENCES AS WELL. FOR EXAMPLE, IMPROVED ANALYSIS OF REACTOR CAVITY AND CONTAltNENT STEP LEVELS DURING A St%LL LOCA HAVE INDICAT'ED. SEQUENCE S C 2
WILL NOT RESULT IN CORE D4%GE, THE SUCCESS CRITER10N FOR CONTAINMENT SPRAY OPERATION DURING RECIRCULATION HAS BEEN MODIFIED TO REQUIRE 1/4 TRAINS RATHER THAN 2/4, CREDIT FOR FEED AND BLEED CORE COOLING HAS BEEN ADDED, AND SE0'JENCE S G HAS BEEN ELIMINATED BECAUSE 8 DAYS ARE AVAILABLE FOR CORRECTIVE ACTION 2
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,,.,,.., -.. _ _... _ _ - _ _ -, - _ - -.. _ - ~
2 BEFORE CONTAINMENT FAILURE. THESE CHANGES HAVE LED TO A REDUCTION IN THE PREDICTED FREQUENCY OF SEVERE CORE DAt%GE AND SOME CHANGES IN DOMINANT SEQUENCES.
Tm SEQUOYAH ANALYSIS HAS BEEN SIMILARLY AFFECTED BY PLANT CHANGES AND THE CAPABILITY FOR IMPROVED MODELING. SPECIFICALLY, THE RPS SHUNT TRIP HAS BEEN-ADDED AND CREDIT GIVEN FOR IMPROVED Ah5 TRAINING, SYMPTOM ORIENTED PROCEDURES HAVE BEEN IMPLEMENTED, LPI CHECK VALVES ARE SUBJECTED TO INCREASED TESTING, AND ADDITIONAL PROCEDURES ARE NOW AVAILABLE FOR ENSURING THE REMOVAL OF THE REFUELING DRAIN PLUG. MODELING CHANGES INCLUDE CREDIT FOR BLEED AND FEED CORE COOLING, THE SUCCESS STATE FOR HPI HAS BEEN MODIFIED TO 1/4 TRAINS RATHER THAN 2/4, IN CERTAIN INSTANCES ONLY ONE RHR OR CONTAltNENT SPRAY HEAT EXCHANGER IS REQUIRED, RATHER THAN TWO, RCP SEAL LOCAS HAVE BEEN AIDED, AND THE STATION BLACKOUT ANALYSIS IS MORE DETAILED. HOWEVER, THE NEW ANALYSES ALSO IDENTIFIED A NEW bOMINANT SEQUENCE-A Com0N-CAUSE FAILURE MODE IN THE SER SYSTEM.
AT PEACH BOTTOM THE PLANT HAS ALSO BEEN MODIFIED SINCE 1975. PbST SIGNIF1-CANTLY, CREDIT HAS BEEN GIVEN FOR IMPLEMENTATION OF THE ATWS RULE, CONTAINMENT VENTING PROCEDURES ARE AVAILABLE, AND ADS NOW ACTUATES 01 LOW WATER LEVEL ALONE, RATHER THAN LOW LEVEL Af0 HIGH DRY WELL PRESSURE. CREdlTHASALSOBEEN GIVENFORltPROVEDREL1ABILITYOFDIESELGENERATORS,BASEDdNPLANT-SPECIFIC i
DATA RESULTING FROM A HIGHLY EFFECTIVE RELIABILITY PROGRAM, AND AN EXTENSIVE HtFAN RELIABILITY ANALYSIS HAS BEEN PERFORMED ON OPERATOR PERFORt%f4CE DURI j
ATWS.
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___-..~ _ _-, _ _...--. -__ _ _--.--......
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WE HAVE NOT YET CRITICALLY REVIEWED THE GRAND GULF RESULTS. BASED ON A PRELIMINARY REVIEW, IT APPEARS CHANGES RESULT FROM PLANT NODIFICATIONS ASSOCIATED WITH THE ATWS RULE, MORE DETAILED MODELING OF STATION BLACK 0UT SEQUENCES INCLUDING THE POTENTIAL FOR LONG-TERM LOSS OF HPCS AND RCIC DU LOSS OF CONTROL POWER OR PLNP SEAL COOLING, AND MORE REALISTIC CONSIDERATION OF RECOVERY FROM ACCIDENT SEQUENCES ItNOLVING LONG-TERM LOSS OF THE RHR SYST O
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WASHINGTON, D. C. 20555 h.l-
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t EDO PRINCIPAL CORRESPONDENCE CONTROL
. - - = - - -
- - - - - - - - - - - =
DUE: 06/13/86 EDO CONTROL: 001790 FROM:
DOC DT: 05/01/86 FINAL REPLY:
REP. BARBARA BOXER TO:
PRESIDENT REAGAN GREEN SECY NO: 86-524 FOR SIGNATURE OF:
EXECUTIVE DIRECTOR ROUTING:
DESC:
CHERNOBYL ACCIDENT RAISES QUESTIONS ABOUT NUCLEAR POWER PLANTS IN U.S.
DATE: 05/29/86 ASSIGNED TO: NRR CONTACT: DENTON SPECIAL INSTRUCTIONS OR REMARKS:
REPLY TO REP. BOXER I
NRR RECEIVED: May 30,1986 ACTION:
DSRO,:SPEIS~l NRR ROUTING:
DENTON PPAS MOSSBURG I
r CORRESPONDENCE CONTROL TICKET SECY NUMBER:
86-0524 LOGGING OATE 5/28/a6 ACTION OFFICE:
ED0 AUTHOR:
- Rep Barbara Boxer AFFILIATION:
U.S. House of Rep LETTER DATE:
5/1/86 FILE CODE C&R-2 BP ADDRESSEE:
Pres Ronald Reagan
SUBJECT:
Express concern re the nuc reactor accident at Chernobyl--
comparison to nuc plants in the U.S.
ACTION:
Direct Reply... Suspense: June 9 DISTRIBUTION:
OCA to Ack SPECIAL HANDLING:
Referral dtd 5/13/86 fm White House FOR THE COP 911SSION rah SIGNATURE DATE:
4 LDO - - 991790 Rec'd Off. EDO i
Date 0 2L FC Time -
-__._3/ci,_
5
? f0) l BARBARA BOXER sis CANNON serlmNG WASHINGTON, D C 205I$
,6TN DISTRKT,CAllFORNIA e
(202)225 516i DISTRICT OFFICES J
COMMITTU oN THE BUDGET 450 GOLDEN GATE A%T_
i y
SAN FRANC 15CO.CA 94102 COMMTTTEE ON (415) 6264443 v
GOVERNMENT OPERAllONS 901 IRWIN STREET SAN RAFAEl CA 94901
'"SOTANo FSE"E*'
601131ess of fl}c[Itttichhittics 823 MARIN. ROOM 8
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%nsljington, pi. 20515 soNOuA (707) 76.% 033 May 1, 1986 President Ronald R. Reagan The White House 1600 Pennsylvania Avenue Washington, D.C.
20505
Dear Mr. President:
The continuing disaster at the Chernobyl nuclear power plant in the Soviet Union raises a number of serious questions about nuclear power plants in this country.
First, I understand that we have five nuclear facilities in the U.S. controlled by the Department of Energy (DOE) which do not have containment structures and which do not meet standards for licensing otherwise applicable if they were commercial reactors.
One is located in Hanford, Washington and the other four in Savannah River, South Carolina.
The Hanford-N reactor has many similarities to the Chernobyl plant.
It has a graphite moderator and no containment structure.
Many are already calling for a shutdown of this facility.
Though the plants in Savannah River are considered " light water", and therefore not as comparable to Soviet construction, their lack of containment structures is a serious problem.
These five reactors, should at a minimum, meet NRC standards as do commercial production plants.
I believe you should order these plants closed until they l
comply with National Regulatory Commission (NRC) standards.
With regard to the 100 commercially licensed plants, I understand that the NRC testified last year before a subcommittee of the House Energy and Commerce Committee that there is a 45 percent chance of a core meltdown in one or more of these plants within the next 20 years.
That level of risk is not acceptable.
The NRC has not, in recent years, made data available which indicates which of these plants have the greatest risk.
I ask you to obtain the i
pertinent safety information from the NRC on these 100 plants and make this available to Congress and the public.
l l
9
Mr. President, as we are learning from the disastrous event in the Soviet Union, this issue requires the strongest possible leadership in order to insure that we never experience a "Chernobyl" in America.
By copy of this letter, I am asking the NRC and DOE to respond to the above points.
n terely,
/
ps-
- A-B Member of Congress BB/ swr ec: Secretary John S. Herrington cc: Chairman Nun _zio J.
Palladino l
May 13, 1986
Dear Mrs. Boxer:
On the President's behalf, I would like to acknowledge your May I letter expressing your i
concern regarding the nuclear reactor accident at Chernobyl, and raising questions with
[
respect to nuclear power plants in this country.
We appreciate your interest in sharing with us your concerns in this important area.
While you mentioned that you have asked the Department i
of Energy and the Nuclear Regulatory Commission to respond to your inquiry, I have taken the liberty of providing a copy of your letter to the Administrator of the Environmental Protection Agency.
With best wishes, Sincerely, a
William L. Ball, III Assistant.to the President 1
The Honorable Barbara Boxer House of Representatives Washington, D.C.
20515 WLB:KRJ:blb w/ copy of inc to Craig DeRemer, Cong cc:
Affrs, EPA - FYI
)
w/ copy of inc to Ted Garrish, Cong Affrs, cc:
r'_
- Energy --- FYI -
w/ copy of inc to Nunzio Palladino, NRC
' 'm cc:
FYI j
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