ML20149H258
| ML20149H258 | |
| Person / Time | |
|---|---|
| Issue date: | 03/13/1985 |
| From: | Conti E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Rehm T NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML20149B626 | List: |
| References | |
| FOIA-87-728, FOIA-87-853 NUDOCS 8802190190 | |
| Download: ML20149H258 (10) | |
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,e *, ' NUCLEAR REGULATORY COMMisslON WASHINGTON. D. C. 20556 F.AR 13 1985 MEMORANDUM FOR: Thomas Rehm, Assistant for Operations, Office of the Executive Director for Operations FROM: Enrico F. Conti, Acting Director, Policy, Planning and Control Staff, Office of Nuclear Regulatory Research
SUBJECT:
RESPONSE TO HENRY FLYER QUESTION ON BUDGET Attached in question and answer format is the information verbally requested by Henry Myers, 7
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Enrico F. Conti, Acting Director Policy, Planning and Control Staff Office of Nuclear Regulatory Research
Enclosures:
As stated t
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QUESTION, WHAT WORK HILL BE DONE UNDER THE RES PROGRAM AND HOW WILL THAT WORK BE USED7 ANSWER, THE ATTACHED TABLE IDENTIFIES BOTH THE NATURE OF THE WORK AND THE DOLLAR ESTIMATES FOR THE RESEARCH PROGRAMS. ALSO ATTACHED IS A
SUMMARY
OF AREAS WHERE RESEARCH HAS BEEN A MAJOR CONTRIBUTOR TO REGULATORY ISSUE RESOLUTION AND AREAS WHERE RESEARCH IS CONTINUING TO SUPPORT OR REVISE REGULATORY POSITIONS. ALSO ATTACHED ARE TWO ITEMS OF COMPLETED RESEARCH EFFORTS WHICH ARE CURRENTLY UNDER REGULATORY EVALUATION.
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MYERS/RES 3/13/85 I
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FY 1986-1987 BUDGET 3/7/85 (DOLLARSINMILLIONS)
FY 1985 FY 1986 FY 1987*
REACTOR ENGINEERING $40.1 $40.3 $45.0 Seismic design margins of structures and piping systems, including fragility tests 7.3 6.0 Mechanical equipment qualification 1.8 2.8 Containment and penetration integrity 3.0 3.7 Reactor vessels and boundary components, including heavy section steel technology 9.4 8.5 Piping systems and steam generators 5.5 6.4 Nondestructive examination methodology including fracture mechanics 3.2 2.3 Plant instrumentation and control system -
safety implications 2.0 0 Electrical equipment aging & Shippingport components investigation 5.4 9.0 Hydrogen & fission product control systems 2.5 1.6 THERMAL HYDRAULIC TRANSIENTS 22.0 21.7 22.0 Semiscale 2.8 2.5 MIST 5.2 3.4 2D/3D 4.7 FIST 3.2 0.7 0.9 ROSA-IV 1.1 2.0 Hydraulic experiments & models on stean generator, two phase flow and counter current flow 2.6 3.2 TRAC and RELAP applications, assessment and maintenance 4.9 6.5 l
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- FY 1987 estimates by project detail have yet to be deterinined.
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. FY 1986-1987 BUDGET 3/7/85 (DOLLARSINMILLIONS)
FY 1985 FY 1986 FY 1987 ACCIDENT EVALUATION 40.0 30.7 31.0 Severe Accident Sequence Analysis 5.3 3.5 Power Burst Facility 12.0 4.0 Annular Core Research Reactor 1.4 3.0 NRU Reactor - Canada 1.9 3.2 Fission product release investigation and code development 8.3 10.4 Containment Loading including core melt phenomenology, aerosol transport and hydrogen behavior 8.8 6.6 LMFBR 0.7 0 HTGR 1.6 0 REACTOR OPERATIONS AND RISK 17.6 16.5 18.0 Risk Methodology 3.9 3.5 Data & data uncertainties 2.1 2.6 -
Regulatory & inspection applications 4.5 5.2 Severe accidents risk 5.8 5.2 Human factors 1.3 0 WASTE MANAGEMENT, EARTH SCIENCES & HEALTH 13.5 11.8 14.0 Seismic networks 3.5 3.5 Seismo. logy studies 1.0 1.0 Health effects 1.8 1.7 Occupational protection 0.6 0.8 High Level Waste site geochemistry, packaging materials and groundwater transport 5.0 3.0 Low Level Waste radwaste & containers, field applications of hydro / transport nadels 1.6 1.8 TOTAL $133.2 $121.0 $130.0
- FY 1987 estimates by project detail have yet to be determined.
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AREAS WHERE RESEAF ' HAS BEEN A MAJOR CONTRIB'JTOR ' ISSUE RESOLUTION PRESSURE VESSEL (PRESSURIZED THERMAL SHOCK) RESEARCH NAS PROVIDED TNE BASIS FOR COPNISSION DECISIONS ON PRESSURIZED THERML SHOCK (PTS) AND ALLOWING CONTINUED OPERATIONS OF ELEVEN PWR'S THAT HAD BEEN THOUGHT 70 BE AT RISK '
o MATERIALS RESEARCH PROVIDED METHODS FOR PREDICTING CRACK INITIATION, GROWTH, AND ARREST IN REACTOR YESSEL WALLS o CALCULATION OF h!XING 0F COLD EMERGENCY COOLANT WITH HOT REACTOR COOLANT l USING NEWLY DEVELOPED NRC THERMAL-HYDRAULIC CODES GAVE EXPECTED METAL WALL 1 TEMPERATURES.
o DEVELOPMENT OF PROBABILISTIC RISK ANALYSIS METHODS TO INTEGRATE THE RESULTS ;
0F MATERIALS RESEARCH, THERMAL-HYDRAULIC RESEARCH AND REAC10R DESIGN AND <
OPERATING DATA TO EVALUATE WHAT IS IMPORTANT TO PTS F.ISK.
o END RESULTS REVEALED THAT RUPTURE WOULD NOT BE POSSIBLE EXCEPT IN EXTREMELY UNLIKELY CASES. ,
PIPING RESEARCH YALIDATED THE COPHISSION'S LICENSING DECISIONS ON BWR PIPE CRACKS ,
o CRACKING IN LARGE DIAMETER STAINLESS STEEL PIPING WAS FIRST OBSERVED IN l 1982 (AT HINE MILE POINT).
i o VALIDATED ACCEPTABILITY OF INDUSTRY FIXES FOR CRACKED BWR PIPING SUCH AS WELDOVERLAYANDINDUCTION-HEATEDSTRESSIMPROVEMENT(IHSI)BASEDON RESEARCH STARTED IN 1980.
o PROVIDED TECHNICAL DATA BASE TO ESTABLISH "LEAK-BEFORE-BREAK".
o NON-DESTRUCTIVE EXAMINATION (NDE) RESEARCH IS PROVIDING PROCEDURES FOR I QUALIFYING NDE PERSONNEL, PROCEDURES AND EQUIPMENT FOR RELIABLY DETECTING l
AND ACCURATELY SIZING FLAWS IN PIPING. DEMONSTRATED ABILITY OF NEW SAFT-UT !
HETH00 TO DETECT AND SIZE FLAWS IN VERMONT YANKEE AND DRESDEN-3; RESOLVED i CONFLICT BETWEEN RESULTS OF KWU (GERMANY) AND LAMERT/McGILL/CRAWFORD.
THE COMMISSION PROMULGATED A FINAL RULE REQUIRING HYDROGEN CONTROL SYSTEMS FOR MARK III BWR'S AND ICE CONDENSER PWR'S. THIS RULEMAKING WAS SUPPORTED BY:
o NRC/EPRI-SPONSORED HYDROGEN BEHAVIOR AND COMBUSTION CONTROL RESEARCH
'?ROVIDED LARGE-SCALE EXPERIMENTAL DATA ON HYDROGEN DEFLAGRATION PROPAGA AS FUNCTIONS OF CONTAINHENT CONDITIONS.
o DEVELOPMENT AND USE OF RES-DEVELOPED THERMAL-HYDRAULIC CODES TO PREDICT PRESSURE BUILDUP IN CONTAINMENT AS A RESULT OF HYDROGEN DEFLAGRATION.
o REDUCED SCALE TEST! ON EQUIPMENT SURVIVABILITY DURING HYDROGEN BURN; PROMPTLY RESOLVED QUESTIONS RAISED BY EPRI LARGE SCALE TESTS 1
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AREAS WHERE RESEAP"' HAS BEEN A MAJOR CONTRIEJTOR
- ISSUE RESOLUTION 1
PROVIDE TECHNICAL BASES TO ELIMINATE REQUIREMEN' FOR PIPE WIP RESTRAINTS.
o PIPE WIP RESTRAINTS IMPEDE INSERVICE IN3PECT10N AP RESULT IN INCREASED -
OCCUPATIONAL EXPOSURE. CLOSELY SPACED RESTRAINTS CAN CAUSE HIGH THERMAL STRESSES DURING NORMAL OPERATION.
o RESEARCH STARTING IN 1979 DEVELOPED PROBAB*LISTIC METHODS FOR PREDICTING
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BEHAVIOR OF CRACKS IN PIPES . I o RESEARCH SHOWED THAT PROBABILITY OF PIPE BEING COMPl.ETELY SEVERED IS l EXTREMELY SMALL (I.E., DOUBLE ENDED GUILLO*INE BREAX IS UNLIKELY) o ALSO RESOLVED USI A-2 (ASYMETRIC BLOWDOWN LOADS) BY SHOWING THAT :
EARTHQUAKES DO NOT CAUSE PIPES TO BE SEVERED, THERE3Y ALLOWING CONTINUED l' OPERATION OF SEVERAL WESTINGHOUSE PWR'S THAT WOULD OTHERWISE HAVE TO BE SHUT DOWN, E.G., NORTH ANNA.
THE COMMISSION PROMULGATED A RULE WHICH IMPOSED REQUIREMENTS TO DEAL WITH i ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS). THIS RULEMAXING WAS SUPPORTED BY THE FOLLOWING RESEARCH: ,
o INFORMATION OBTAINED FROM PRA'S CARRIED Oltr UNDER THE NRC RESEARCH PROGRAM i IDENTIFIED THE DOMINANT ACCIDENT SEQUENCES.
J o DETAILED ANALYSIS OF ATWS FOR BWR'S UNDER THE SASA PROGRAM PROVIDED BASIC KNOWLEDGE OF THE INTERACTIONS OF MAIN STEAM ISOLATION VALVE AUTOMATIC AND MANUAL CLOSURE, POWER OSCILLATIONS WITH RECJCED WATER LEVEL, MANUAL SCRAM !
EFFECTS AND OPERATOR ACTIONS. TVA COOPERATION AND PARTICIPATION IN SIMULATOR RUNS PROVIDED A SOUND BASIS FOR FOTENTIAL OPERATOR ACTIONS.
o ATWSEXPERIMENTSCARRIEDOUTATLOFT[0RPW'SINVESTIGATED; l
LOSS-OF-FEEDWATER INDUCED LOCA'S FROM STUCK OPEN POWER OPERATED RELIEF VALVES WITH AND WITHOUT MITIGATING OPERATOF. ACTIONS; AND ATWS BY LOSS-OF-OFFSITE POWER.
SEISMIC RESEARCH PROVIDED THE BASIS FOR COMISSION DECISIONS ON ADEQUACY OF SAFETY MARGINS FOR EARTHQUAKES EXCEEDING THE DESIGN EARTHQUAKE FOR NUCLEAR POWER PLANTS.
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USE OF PROBABILISTIC METHODS DEVELOPED UNDER THE SEISMIC SAFETY MARGINS t
'itESEARCH PROGRAM (IN 1979 TO 1984) WERE USED TO ASSESS THE DESIGN ADEQUAC
) 0F THE MODIFICATIONS TO THE AUXILIARY ',EEDWATER SYSTEM AT SAN ONOFRE.
l o SITE SPECIFIC SPECTRA TO BETTER CH RACTERIZE THE DESIGN BASIS EARTHQUAKE WERE USED IN LICENSING DECISIONS.
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i;t ARE" WHERF CONTINUED RESEARCH IS NE' D I
A XEY PRODUCT FROM THE SEVERE ACCIDENT RESEARCH PROGRM WILL BE THE STAFF l REASSESSMENT OF METHODS FOR ESTIMATING ACCIDENT SOURCE TERMS IN SPECIFIC i ACCIDENT SEQUENCES. I o
THERE HAS BEEN A GENERAL BELIEF IN THE NUCLEAR COP 90NITY THAT WASH-1400 ESTIMATES OF RADIONUCLIDE RELEASES FOLLOWING ACCIDENTS WERE UNREALIST HIGH. THESE ESTIMATES IMPACT CURRENT REGULATIONS SUCH AS EMERGENCY PLANNING AND SITING MD FUTURE APPLICATIONS OF THE SEVERE ACCIDENT POLICY -
STATEMENT.
o STARTING IN 1982-83, AN NRC CONTRACTOR (BCL) UTILIZING CODES AND A DATA BASE DEVELOPED LARGELY FROM NRC RESEARCH PROGRM, WITH INPUT FROM COORDINATED GERMAN RESEARCH, DEVELOPED SOURCE TERM ESTIMATES FOR SIX REFERENCEU.S. PLANTS (BMI-2104). :
o SUPPORTING EFFORTS FOCUSSED ON UNCERTAINTY ANALYSIS (SMDIA), STATUS OF CODE VALIDATION (ORNL) MD CONTAINMENT BEHAVIOR.
o CONTRACTOR WORK SUBJECTED TO EXTENSIVE PEER REVIEW INCLUDING AMERICAN PHYSICALSOCIETY(DUEEARLY1985)ANDINTERACTIONSWITHINDUSTRY(IDCO o DRAFT STAFF REPORT (NUREG-0956) IN SPRING 1985 WILL StM4ARIZE METHODS,
. UNCERTAINTY, MAJOR ISSUES, VALIDITY MD APPLICABILITY.
o COPHISSION WILL USE THIS DOCUMENT AS ONE OF THE BASES FOR IMPLEMENTAT; REGULA10RY ACTIONS SUCH AS: y i ISSUE A REALISTIC REACTOR-SITING RULE.
DEVELOPEMERGENCYPLANNINGGUIDELINESTHATWILLPROVlDEThEBASISFOR A GRADED RESPONSE INCLUDINC MINIMUM REQUIRED EVACUATIONS.
MAKE FINAL DETERMINATION OF THE NEED FOR MITIGATING ENGINEERING FEATURES IN NEW PLANTS.
IN THk' LATE 1970'S, AFTER INCIDENTS AT RANCHO SECO AND THI, INDUSTRY AND NR RECOGNI2ED THAT REACTORS CAN BE REPRESSURIZED FOLLOWING A SEVERE THERMAl A PROPOSED PTS RULE WAS PROMULGATED IN 1984. FINAL RULEMAKING IS PLMNED IN 1985 MD WILL BE SUPPORTED BY RESEARCH EFFORTS ON THE FOLLOWING ISSUES. ;
o VALIDATE FRACTURE MECHANICS ANALYSIS METH005 FOR PROPAGATION AND ARR CRACKS BY TESTING A NON-UNIFORMLY (MBRITTLED VESSEL WITH HIGH COPPER TYPICAL OF OLDER REACTOR PRESSURE YESSELS (PRESSURIZED THERMAL SHOCK 4 EXPERIMENT-2). '
o MALYZE FULL-SCALE MIXING EDERJMENT TO BE RUN IN GERMANY IN 1986 AS PAR
, OF 2D/3D PROGRM.
o DURING 1985, HELP NRR DRAFT GUIDANCE FOR LICENSEE'S PLMT-SPECIFIC ANALYS!$ AND ACCEPTANCE ' CRITERIA FOR LICENSEE-PROPOSED CORRECT!YEi l
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.NAS WHERE CONTINUED RESEARCH IS NEEBED ISSUES AND CONCERNS REGARDING ADEQUATE DECAY HEAT REMOVAL CAPAl RESTART (APPLICABLETOALLB&WDESIGNEDPLANTS): 1 o
TEST TO ASSESS THE EFFECTIVENESS OF THE BOILER CONDENSER PROCESS '
HEAT FROM THE REACTOR COOLANT AND MAINTAIN NATURAL CIRCULATION.
o USE OF HIGH-P0lNT VENTS TO ASSURE NATURAL CIRCULATION AND LON CORE COOLING.
SEISMIC RESEARCH NEEDED TO PROVIDE THE BASIS FOR COPellSSION SAFETY DE!
CONTINUED OPERASILITY OF PLANTS THAT MAY BE SUBJECTED TO EARTNQUAKE !
BEYOND THOSE FOR WHICH THEY WERE DESIGNED. t o
EVALUATE THE IMPMT ON DESIGN AND OPERABILITY OF MORE SEVERE' EA LOADS THAT MAY BE APPLIED TO EASTERN PLANTS DESIGNED TO MINIMAL STA
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o PROVIDING METHODOLOGY TO DETERMINE MORE PRECISELY THE CAPACITY ' ,
SURVIVE EARTHQUAKES BEYOND THEIR INITIAL DESIGN i NUCLEAR PLANT AGING RESEARCH WILL PROVIDE A BAS!$ FOR RESOLU 1 RELATED TO THE IMPACT ON PLANT SAFETY OF SERVICE WEAR AND AGED SYSTEM l EQUIPMENT--!NCLUDING RELICENSING OF OLDER PLANTS. s - O '
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' EVALUATE MECHANISMS AND SIGNIFICANCE OF REACTOR AGING (C0RR0$10 EFFECTS, FATIGUE, WEAR) INCLUDING PRESENTLY llNRECOGNIZED NECHANISMS. SiljDY l
0F PEDIGREED COMPONENTS AND MATERIALS FROM TrlE SHIPPINGPORT REAl -
' CONTRIBUTE Ti UNDERSTANDING AGING. -
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o IDENTIFY AND RECOMMEND METHODS OF INSPECTION AND 5URVE!LLANCE NON! -
WHICH WILL BE EFFECTIVE IN IDENTIFYING AGING PROBLEMS BEFORE PLANT SAFETY. -
THEY A o
IDENTIFY AliD RECOMMEND ACCESTABLE MAINTENANCE PRACTICES THAT CAN EFFECTS OF AGING.
AGENCY CHARGED BY CONGRESS TO REVliW HIGH LEVEL WASTE SITE CHARACTE PLANS FOR THE PRESIDENT'S SELECTED THREE SITES. (EXISTING TECHNICAL BASE
- ISLIMITED.)
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! NEED TO ESTABLISH ACCEPTABLE DATA COLLECTION AND ANALYSIS METHODS SUITABILITY EVALUATION TO AVOID DELAYS IN DOE SCHEDULE
" OF SITE CHARACTERIZATION (FYB6). +
o NEED TO DETERMINE EXPECTED PERFORMkNCE ENVELOPES FOR CAN PACKAGE DESIGN TO AVOID EXCESSIVELY CONSERVATIVE REQUIREMENYS. '-
o NEED TO ESTABLISH EXPECTED RANGE OF THERMALLY-INDUCED PERTUR8dTION REPOSITORY ENVIRONMENT TO HAVE CONFIDENCE IN EXPECTED REPOSITORY PERFORMANCE.
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- 2 ARye WHERE CONTINUED RESEARCH IS NEF"O !
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R DIATION PROTECTION AND HEALTH EFFECTS RESEARCH WILI. PROVIDE NECESSARY TECHNICALSUPPORTFORREVISIONOFRADIATIONPROTECTIONSTANDARDS(10CFR20)AND APPLICATION OF NRC SAFETY GOALS FOR HEALTH PROTECTION.
o CONFIRM SUITABILITY OF INTERNATIONAL COMISSION ON RADIOLOGICAL i
(ICRP) DOSE MODELS FOR M001FYING PART 20 LIMITS. OBJECTIVE IS TO ESTABLISH TECHNICALLY SOUND HEALTH RISK BASIS FOR NRC RADIATION PROTECTION STANDARD AND OTHER REGULATORY ACTIONS.
o PROVIDE FOR VARIABLE NEUTRON QUALITY FACTOR DEPENDENT ON WORKER'S EXPOSURE, RATE. NEW DATA FROM CONTINUING JAPANESE STUDIES MAY REQUIRE CHANGES.
o IMPACT ON SAFETY GOAL OF ICRP COMBINED EXTERNAL AND INTERNAL DOSE REC 0tHENDATIONS NEEDS TO BE DETERMINED.
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.ESEARCH EFF0dTS UNDER REGULAT0s. EVALUATION i NRC RULE 10 CFR 50, APPENDIX "K" DESIGN BASIS ACCIDENT EVALUATION CRITERIA INDUSTRY NOW COULD SAVE MONEY THR0l!GH UTILIZING THE "K" APPENDIX REVISION BY INCREASING POWER IN SOME PLANTS (55) AND THROUG OPERATION AND FUEL MANAGEMENT.
o TWO-THIRDS OF NUCLEAR PLANTS ARE NOW LIMITED BY APPEND RESEARCH PERFORMED SINCE THE 1974 RULEMAKING HAS SHOWN T TEMPERATURES DURING A LOSS-OF-COOLANT ACCIDENT WOULD BE THAN CALCULATED USING EXISTING APPENDIX "K" MODELS.
o BASED ON COMPLETED RESEARCH, NRC HAS ALREADY APPROVED RELAXATION OF RESTRICTION 3 WITHIN EXISTING APPENDIX FOR "K" BOILING WATER REACTORS.
o STAFF WILL SUBMIT A REVISED RULE FOR COMISSION ACTION IN E UNRESOLVED SAFETY ISSUE A-44 STATION BLACK 0UT DEVELOPED APPROACH USING RELIABILITY ANALYS!$ AND RISK ASS DOMINANT FACTORS AFFECTING RISK AND COST-EFFICTIVE IPFRO PROVIDED TECHNICAL BASIS FOR MAINTAINING HININJM AC POW CAPABILITY FOR COPING WITH STATION BLACKOUT OF A SPECIF PLANTDESIGNANDLOCATION(NUREG-1032)(FY83-84).
STAFF WILL SUBMIT A RULE FOR C0m!SSION ACTION IN EARLY 1985.
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