NUREG-1365, Summary of 920625 Meeting of ACRS Subcommittee on Severe Accidents in Bethesda,Md to Continue Discussions on Revised Severe Accident Research Program Plan,NUREG-1365,Rev 1,Apr 1992
| ML20035E064 | |
| Person / Time | |
|---|---|
| Issue date: | 08/21/1992 |
| From: | Kerr W Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| RTR-NUREG-1365 ACRS-2831, NUDOCS 9304140232 | |
| Download: ML20035E064 (43) | |
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CER N H #H CERTIFIED BY W. KERR 8/21/92 SUINARY/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON SEVERE ACCIDENTS JUNE 25, 1992 BETHESDA, MARYLAND PURPOSE The ACRS Subcommittee on Severe Accidents held a meeting on June 25, 1992.
The purpose of this meeting was to continue the discussion of the revised Severe Accident Research Program (SARP)
Plan, NUREG-1365, Revision 1, April 1992. The discussion of this matter had been initiated at the Subcommittee Meeting on May 27, 1992. A copy of the meeting agenda and selected slides from the presentations are attached. The meeting began at 8:30 am and adjourned at 5:40 pm and was held entirely in open session. No written comments or requests for time to make oral statements were received from members of the public. The principal attendees were as follows:
ATTENDEES ACRS NRC STAFF W.
Kerr, Chairman B. Sheron, RES I. Catton, Member F. Eltawila, RES T.
Kress, Member C. Tinkler, RES D. Ward, Member A. Rubin, RES P. Davis, Consultant A. Notafrancesco, RES J. Lee, Consultant R.
Lee, RES D. Houston, Cognizant Staff R. Wright, RES 1
Engineer G. Holahan, NRR NRC STAFF CONSULTANTS S. Hodge, ORNL R. Summers, SNL R. Gauntt, SNL S. Thompson, SNL I
K. Washington, SNL C. Allison, INEL W.
Sha, ANL In all, there were 18 attendees other than IEC staff and included I
representatives of B&W, Bechtel and the DOE Defense Nuclear Facilities Safety Board.
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i Sev Acc Minutes, June 25, 1992 i
I DISCUSSION I
Chairman's Openino Remarks f
f In his opening remarks, Dr.
Kerr indicated that this was a continuation of the discussion of the SARP Plan which was initiated i
at the May 27th Subcommittee Meeting. He asked the other Members and Consultants to provide him with comments on the Plan for consideration by the Full Committee during their meeting in July.
PRESENTATIONS ON SARP PLAN Jntroduction Dr. Eltawila (RES) provided an update on those issues that had been discussed during the May 27th Subcommittee Meeting. He discussed changes in regard to Direct Containment Heating (DCH), Source Term and Severe Accident Codes. The DCH Program has been recharacterized as " Nearing Completion" and the revised schedule indicated that the last planned testing in this area would be in December 1992. One of l'
the new DCH tests involved testing in the one-sixth scale concrete containment facility at SNL.
For Source Term activities, he indicated that additional testing would be performed at ORNL to identify fission product releases during shutdown conditions. In the area of Severe Accident Codes, he noted that new sections would be added to the VICTORIA, IFCI and COMMIX codes and that additional i
diagrams would be incorporated to show the correlation between the i
various experiments and the different code models.
Core Melt Prooression
.j Mr. Wright (RES) discussed the aspects of in-vessel core melt progression from core uncovery to reactor vessel meltthrough. He
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indicated that the early metallic melt phase is reasonably well understood but that late ceramic melt phase is only generally understood. He stated that the focus of current research is in two key areas:
(1) core blockage or melt drainage in BWR dry core accidents and (2) ceramic pool meltthrough of the crust from a blocked coro (threshold and failure location). He noted that fuel coolant interactions and lower head failure were not addressed as part of this program. He described tests that are planned to be performed in the ACRR (SNL), NRU (Canada) and CORA (Germany).
As part of this discussion, Dr. Gauntt (SNL) provided a description of the ACRR facility and test program.
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A 1
Sev Acc Minutes. June 25, 1992 Molten Fuel-Coolant Interactions e
Dr. Eltawila (RES) discussed fuel-coolant interactions (FCIs) that can affect lower head failure and containment loading.
The technical issues of FCIs include hydrogen generation, steam explosions (in-vessel and ex-vessel), and thermal loads on the lower head and containment. He indicated that the data base was sparse for FCI melt quenching, particularly at high pressure. He described the experiments to be performed in the FARO facility (Italy) as prototypical of high pressure scenarios. The objectives of the FARO tests were related to the following phenomena: jet breakup and quench, penetration failure, structural interactions, premixing models, triggering at high pressure and propagation in realistic premixtures. He also noted additional FCI programs at the University of California-Santa Barbara and the University of Wisconsin. These studies were concerned with the formation of explosive
- mixtures, explosion propagarian ad the resulting energetics.
Debris coolability Mr.
Tinkler (RES) discussed debris coolability in regard to regulatory needs and the supporting research program. Regulatory needs included technical background for ALWR review, criteria for individual plant examination (IPE) review and criteria for l
containment performance assessment. He described the following tests that had been performed: SWISS, FRAG and WETCORE tests at SNL, MACE tests at ANL, and ACM tests at JAERI. He indi;ated that the SASM (scaling) methodology should be applied for future MACE (ANL) tests.
He also indicated that additional integral and separate ef fects tests are needed to resolve outstanding concerns.
MELCOR Code Development and Review Mr. Tinkler (RES) discussed the findings of the MELCOR Peer Review Group. These findings had been presented previously to the Severe Accident Subcommittee during the meeting in October, 1991 by Dr.
1 Boyack (LASL), Chairman of the Peer Review, and are documented in LA-12240, "MELCOR Peer Review, " March 1P92. The major findings were associated with
- numerics, missing
- models, deficient
- models, expanded assessment, and documentation. In each of these areas, Mr.
Tinkler described the basis for the findings, the recommendations and the NRC response or action. In some cases, the NRC action was already underway or planned before the Peer Review findings were developed. The primary missing models involved the following:
o primary system component natural circulation i
o high pressure melt ejection and DCH o ice condenser configuration o non-explosive interactions between debris and water o fission product vapor scrubbing o additional fission product deposition and surface reactions.
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I Sev Acc Minutes. June 25, 1992 Dr. Summers (SNL) indicated that MELCOR currently models the ice condensers.
In regard to expanded assessments, Mr.
Tinkler discussed sensitivity studies, benchmarking and code to code comparisons. He noted a long list of experiments both in this country and foreign that were to be assessed with the code. He also noted the assessment that would result from the Cooperative Assessment Procram which had participants from many foreign countries and several domestic users.
SCDAP/RELAPS Rgtivities Dr. Rubin (RES) discussed the development programs for SCDAP/
RELAPS. He noted that the code combination was applied to describe t
the response of the primary coolant system up to vessel or system failure and to model in-vessel core melt progression for various accident sequences in PWRs and BWRs. The current focus in this area is to develop a reliable and usable code for full-size plants.
He indicated that there is an ongoing peer review similar to that described above for MELCOR.
Further assessment and model development is on hold pending the completion of the peer review.
4 He also noted that many users of this code have experienced some difficulties in running it.
I CONTAIN Code frogram i
Mr. Tinkler (RES) described the CONTAIN code as NRC's best estimate j
mechanietic code for integrated analysis of severe accident containment phenomena for both PWRs and BWRs with the capability to predict containment loading and source terms. He discussed various areas of development and improvement that are currently being
-l pursued to revise the code. He indicated that a Peer Review Group would be established in the near future to review the code. This review would be organized and executed by LASL and the effort is scheduled to begin in the 2nd quarter of FY 93. He discussed in detail some nine areas of DCH related models that are now in CONTAIN.
ALWR Research Activities Dr. Rubin and Mr. Tinkler (RES) discussed the user needs by NRR for the ALWR review, specifically for the AP600 and the SBWR. The major objectives of this portion of the program were to provide:
i' dependent audit capability, assessment of. vendor test programs, n
identification of issues related to unique design aspects and performance under severe accident conditions. In performing this task, the RES staff plans to upgrade CONTAIN, COMMIX, MELCOR and SCDAP/RELAP5. Various contractors at SNL, ANL, ORNL, INEL and BNL were identified as participants in the overall program. Two other ALWR related programs were discussed:
(1) scaling analysis by Energy Research, Inc. of vendor containment test programs and'(2) assessment of hydrogen migration / combustion by SNL in the AP600.
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i Sev Acc Minutes June 25, 1992 User's Comments by NRR Dr. Holahan (NRR) discussed the general application of the research programs to needs within NRR for understanding of current plant performance and for review of ALWR plant applications.
He specifically indicated that NRR was greatly interested in the programs in regard to core melt behavior, fuel coolant interaction, i
debris coolability and the whole collection of computer codes.
He stated that there was an intangible benefit to having the severe accident
- program, that is, it produces staff expertise, information, experience, and involvement in international codes and l
research that wouldn't exist otherwise. He noted some specific NRR activities where the research results are being or will be applied, e.g., IPE reviews, accident management, containment performance, and operation with degraded steam generator tubes.
l Closino Remarks In closing, Dr. Kerr thanked the staff and their contractors for their participation in the meeting. He noted that this matter would be discussed during the Committee Meeting on July 9, 1992 and he asked the Subcommittee Members and Consultants to provide him with any comments for consideration during that meeting.
For the Committee Meeting, he asked RES to provide a presentation on accident progression and direct containment heating. He further asked the staff to describe how the results are applicable to existing and advanced plant designs and what the fallback position may be for each of the elements.
SUBCOMMITTEE CONCERNS I
During the meeting, the Subcommittee Members and Consultants expressed a number of comments and concerns related to the severe accident research program.
These comments and concerns are presented as follows in a random manner:
(1) In regard to core melt progression, Drs. Kerr and Lee asked i
whether melt progression is important if one calculated that the 4
vessel would not fail. Dr. Holahan indicated that it would be for j
assessing bypass sequences.
(2) In regard to questions about the capability of SCDAP/RELAP5, Dr. Allison (INEL) indicated that the code performed reasonably well in calculating the TMI-2 accident up through the heat-up l
phase. He also indicated that to build an input deck for the code to use on the ALWRs would be a 4-5 man-month effort and to build a deck from scratch for a specific plant would be a 1-1M man-year-effort.
(3) In regard to MELCOR, Mr. Davis asked about.the difference in
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the decontamination f actor between MELCOR and the Source Term Code Package (STCP). He noted that BNL had reported a' difference of a
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Sev Acc Minutes. June 25, 1992 I
factor of 400,000. Dr. Summers (SNL) indicated that they had been i
made aware of the dif ferences but could not explain them since both the BNL and the MELCOR calculations use the same STCP scrubbing model.
j i
(4)
In response to a question by Dr. Catton concerning steam generator tube rupture, Dr. Allison (INEL) indicated that they were performing analysis of this event for Dr. Shotkin (RES). to be l
applied in the assessment of accident management procedures. It was noted that this effort is in another Division of Research, thus would not appear in the SARP. Dr. Allison further indicated that SCDAP/RELAPS was being used to assess system failures by natural circulation, the aspects of intentional depresssurization, and evaluation of DCH under the various modes. He noted that a NUREG/CR l
would be issued this summer on this matter.
(5) In response to concerns expressed by Dr. Catton about the I
COMMIX
- code, Dr.
Sha (ANL) discussed the assessments being performed against the HDR tests and the HMS code. He indicated that ANL reports are available on this assessment and agreed to send i
copies of these reports to Dr. Catton.
(6) In response to a question by Dr. Kerr about the MACCS code, Dr.
I Eltawila indicated that this code was outside the scope of the SARP i
since it was applied and reviewed by another Division of Research.
FUTURE ACRS ACTION l
The SARP Plan will be discussed at the Full Committee during the July Meeting and an ACRS report on this matter will be developed at that time.
I ACTIONS, AGREEMENTS AW COMMITMENTS The following actions, agreements and commitments resulted from this meeting:
I (1) Dr. Eltawila agreed to have the PRA Branch (RES) provide a report on the validation effort concerning the MACCS code.
(2) Dr. Sha agreed to send leports on the assessment of HDR tests f
and the HMS code to Dr. Catton.
i (3)
Dr.
Eltawila agreed to have Subcommittee Members and Consultants contact Dr.-Rempe (INEL) directly is they had any i
questions concerning the lower head failure study.
l (4) The Subcommittee and the RES staff agreed upon the content-of a presentation for the Full Committee at the July Meeting.
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I Sev Acc Minutes. June 25, 1992 DOCUMENTS The review document for the Subcommittee meeting was as follows:
i Memorandum for R.
Fraley (ACRS) from B.
Sheron (RES) dated April 22, 1992.
Subject:
" Severe Accident Research Program Plan Update,"
i NUREG-1365, Revision 1, April 1992 CDRAFT PREDECISIONAL) l NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, NW, Washington, DC 20006, (202) 634-3273 or can be purchased from Ann Riley and Associates, Ltd.,
1612 K Street, NW, Suite 300, Washington, DC 20006, (202) 293-3950.
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SEVERE ACCIDENTS SUBCOMMITTEE MEETING BETHESDA, MARYLAND 5
ROOM P-110 l
- ACTUAL AGENDA -
Thursday. June 25, 1992 TIME A. Subcommittee Chairman's Opening 8 : 30 am Remarks - W.
Kerr B.
Severe Accident Rer' arch Program (SARP) Plan o Opening Remark F. Eltawila (RES) 8:35 am o Core Melt Progression R. Wright (RES) 8:50 am
- BREAK 10:05 am o Core Melt Progression (Contined)
R. Wright (RES) 10:20 am o Fuel Coolant Interaction F. Eltawila (RES) 11:15 am o Debris Coolability C. Tinkler (RES) 11:50 am
- LUNCH *********
12:20 pm B.
Severe Accident Research Program (SARP) Plan (Continued) o MELCOR Review /NRC Activities C. Tinkler (RES) 1:25 pm
- BREAK ****-****
3 : 00 pm o SCDAP/RELAP 5 Activities A. Rubin (RES) 3 :15 pm o CONTAIN C. Tinkler (RES) 3:55 pm o ALWRs Code Readiness C. Tinkler (RES)/
4 :30 pm A. Rubin (RES) 4 : 45 pm C. NRR Comments in Support of SARP G. Holahan (NRR) 5 :10 pm D. Summary Discussion and Plans for 5:30 pm Full Committee Presentation E. Adjourn 5:40 pm l
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1 SEVERE ACCIDENT RESEARCH PROGRAM-UPDATE (REVISIONS MADE SINCE MAY, 27, 1992, MEETING)
PRESENT TO:
ACRS SEVERE ACCIDENT SUBCOMMITTEE FAROUK ELTAWILA, CHIEF ACCIDENT EVALUATION BRANCH DIVISION OF SYSTEMS RESEARCH OFFICE OF NUCLEAR REGULATORY-RESEARCH JUNE 25, 1992-
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DIRECT'CONTAIMMENT HEATING DESIGNATED AS " HEAR COMPLETION" ISSUE ZION COUNTER PART TESTING COMPLETED (JUNE 1992)
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. NOVEMBER 1992 2
SURRY COUNTER PART TEST BEGIN OCTOBER 1992 SuRRY DCH LOADS DECEMBER 1992
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SEVERE. ACCIDENT' CODES ADDED : SECTIONS ON THE IICTORIA,, IFCI AND COMIX CODES ADDED DIAGRAM TO SHOW-THE EXPERIMENTAL DATA TO VALIDATE DIFFERENT MODELS
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CORE MELT PROGRESSION T 4 R.W. WRIGHT ACCIDENT EVALUATION BRANCH MEETING OF THE ACRS SU8C0604ITTEE ON' SEVERE ACCIDENTS BETHESDA, MARYLAND JUNE 25, 1992 3 ... ~ -....
^' i IN-VESSEL CORE MELT PROGRESSION O ' DESCRIBES THE STATE OF THE CORE FROM CORE UNC0VERY TO REACTOR VESSEL t MELTTHROUGH. l O PROVIDES THE INITIAL CONDITIONS FOR' ASSESSING CONTAINMENT LOADS: MELT MASS MELT COMPOSITION (METAL CONTENT) MELT TEMPERATURE (SUPERHEAT) RATE OF MELT RELEASE TIME OF RELEASE IN A GIVEN SEQUENCE 4 O MELT PROGRESSION ALSO PROVIDES* t IN-VESSEL HYDROGEN GENERATION THE CORE CONDITIONS THAT DETERMINE IN-VESSEL FISSION-PRODUCT l RELEASE,~ TRANSPORT, DEPOSITION,.AND REVAPORIZATION THE CORE CONDITIONS FOR ASSESSING THE CONSEQUENCES OF RECOVERY ACTIONS IN ACCIDENT MANAGEMENT, CORE REFLOODING IN PARTICULAR 0 RESEARCH IS NEEDED TO PROVIDE A TECHNICAL BASIS FOR GENERAL ~ APPLICATION OF THE LOW CONSEQUENCE THI-2 RESULTS. I.0W MASS OF CERAMIC MELT RELEASED FROM CORE (20% AT THI-2) VERY LOW' METAL CONTENT IN THE MELT POOL IN THE CORE AND IN THE MELT RELEASED INTO THE LOWER PLENUM 0 DEBRIS l COOLANT-' INTERACTIONS AND LOWER HEAD FAILURE NOT ADDRESSED HERE. 1
- -.=-- CORE MELT PROGRESSION: STATUS OF CURRENT UNDERSTANDING 0 EARLY.NETALLIC NELT PHASE-REASONABLY WELL UNDERSTOOD. CLAD BALLOONING OXIDATION HEATING AND HYDROGEN GENERATION IN INTACT CORE GEOMETRY U0 LIQUEFACTION (DISSOLUTION) BY MOLTEN ZIRCALOY p EUTECTIC MATERIAL INTERACTIONS AND RATES AMONG 00, ZR0, ZRY, 2 2 AND CONTROL MATERIALS AND THEIR OXIDES EARLY OPENING UP OF THE COMPARTMENTALIZED BWR CORE BY THE EUTECTIC INTERACTION OF CONTROL-BLADE MATERIAL WITH THE 2RY CHANNEL BOX WALLS MOLTEN-ZRY RELOCATION IS A NONCOHERENT, NONCOPLANAR, RIVULET-FLOW PROCESS THAT DOES NOT BLOCK STEAM FLOW AND HYDROGEN GENERATION. IS NOT A FILM FLOW PROCESS O LATE CERAMIC MELT PHASE-ONLY GENERAL UNDERSTANDING INFORMATION PRIMARILY FROM THE THI-2 CORE EXAMINATION l THI-2 RESULTS ALSO GENERALLY' APPLICABLE TO PWR uMRECOVERED ACCIDENTS CERAMIC MELT POOL GROWTH AND MELTTHROUGH FROM A BLOCKED CORE REFL90 DING PROBABLY STOPPED DOWNWARD POOL AND CRUST RELOCATION AT THI-2 TO GIVE SIDE MELTTHROUGH LIMITED MELT' MASS RELEASED.FROM CORE (20% AT THI-2) LOW METAL CONTENT IN CERAMIC MELT POOL 0 HYDROGEN GENERATION AND1 STRONG HEATING OF UNCOVERED CORE FROM ZRY 0xIDATION BY REFLOOD STEAM (LOFT FP-2 AND CORA) 8
b MOLTEN FUEL-COOLANT INTERACTIONS 9 PRESENTED TO: THE ACRS SEVERE ACCIDENT SUBCOMMITTEE FARoux ELTAWILA, CHIEF ACCIDENT Evaluation BRANCH DIVISION OF SYSTEMS RESEARCH OFFICE OF NUCLEAR REGULATORY RESEARCH MAY 27, 1992
6 'd FUEL-COOLANT INTERACTIONS (FCI) WHAT IS THE REGULATORY NEED THAT INVOLVES FCI? IN GENERAL, FUEL-COOLANT INTERACTIONS MIGHT BE CONSIDERED IN A NUMBER OF CIRCUMSTANCES THAT CAN AFFECT LOWER HEAD FAILURE, AND CONTAINMENT PERF0kMANCE: WHAT ARE THE'IECHNICAL ISSUES THAT INVOLVE FCI? WHAT IS THE' HYDROGEN GENERATION RATE DURING COOLANT REFLOOD? DOES THE FCI EXPLOSIVE BEHAVIOR ALTER THE NET AMOUNT OF HYDROGEN PRODUCED 7 CAN MELT GUENCH IN THE LOWER HEAD OF RPV? CAN MELT OUENCH OCCUR IN EX-VESSEL WITHOUT LARGE MECHANICAL LOADS? ~ DOES FUEL-COOLANT MIXING LIMIT THE VAPOR EXPLOSION YIELD AS THE SERG ASSUMED. i i I 2
FUEL-COOLANT INTERACTIONS (CONTINUED) l WHAT ARE THE NRC RESEARCH OBJECTIVES? O focus ON CERTAIN KEY, GENERIC PHENOMENOLOGICAL ASPECTS OF FCIS TO i ADDRESS THE COMPLEX VARIATIONS FOUND IN ACCIDENT MANAGEMENT. O DEVELOP QUALITATIVE SCENARIOS OF SEVERE ACCIDENT PROGRESSION WITH PARTICULAR REFERENCE TO WATER AVAILABILITY (RISK ASSESSMENT). a O APPLYING KNOWLEDGE IN ASSESSING ACCIDENT MANAGEMENT STRATEGIES. O ASSESS ALWRS UNIQUE DESIGN FEATURES, E.G., FLOODED REACTOR CAVITY IN AP600. 3
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FUEL-COOLANT INTERACTION (CONTINUED) QUENCEING OBJECTIVES OF PROPOSED FCI EXPERIMENTS AT FAR0 o JET BREAKUP AND QUENCH DETERMINE PENETRATION BEHAVIOR (BREAKUP) 0F MOLTEN CORIUM JET INTO A POOL OF WATER I l o PENETRATION FAILURE l DETERMINE THE POTENTIAL FOR STRUCTURAL ATTACK AND SMALL PENETRATION FAILURE BY HELT IMPINGEMENT AS FUNCTION OF WATER SUBMERGENCE o STRUCTURAL INTERACTIONS DETERMINE JET QUENCHING BEHAVIOR, POTENTIAL FOR STRUCTURAL ATTACK, AND RESULTING DEBRIS IN PRESENCE OF MAJOR STRUCTURES AS IN THE LOWER PLENUM OF A BWR e TEST PREMIXING MODELS ATTEMPT TO SIMULATE A LARGE MELT POUR INTO A PWR LOWER PLENUM, AND QUANTIFY RESULTING PREMIXING TRANSIENTS O TRIGGERING AT HIGN PRESSURE DETERMINE TRIGGERABILITY OF CORIUM WATER PREMIXTURES AT HIGH PRESSURE o TEST PROPAGATION IN REALISTIC PREMIXTURES DETERMINE TRIGGERABILITY AND RESULTING CONVERSION OF REALISTIC PREMIXTURES I 6
FUEL COOLANT INTERACTIONS ADDITIONAL NRC FCI PROGRAMS AT UCSB & UW i o TNIS RESEARCN'NAS SPECIFIC FOCUS DIRECTED TO GAIN FUNDAMENTAL UNDERSTANDING ON IMPORTANT FCI ISSUES SUCN AS MIXING, TRIGGERING, EXPLOSION PROPAGATION AND YIELD. OBJECTIVE 1. PROVIDE EXPERIMENYAL DATA TO TEST MULTIFIELD MODEL PREDICTIONS THAT PRESENTLY ASSUM2 TNAT LARGE POURS PRODUCE MODEST IN SIZE PREMIXTURES (DUE TO WATER DEPLETION IN TNE PREMIXING ZONE) OBJECTIVE 2. PROVIDE EXPERIMENTAL DATA ON HELT DROP FRAGMENTATION RATES WITNIN THE DETONATION WAVE OF AN EXPLOSION. OBJECTIVE 3. BASED ON ATTAINMENT OF ABOVE OBJECTIVES REFINE DEFINITIONS CN EXPLOSIVE PREMIXTURES AND TNEIR QUANTIFICATION. OBJECTIVE 4. REFINE QUANTIFICATION OF EXPLOSION PROPAGATION, AND CESULTING ENERGETICS, IN REALISTIC PREMIXTURES. APPLY TO VARIETY OF f SITUATIONS. L b 7
s. 1 DEBRIS C00 LABILITY RESEARCH PRESENTATION TO THE ACRS MAY 1992 CHARLES G. TINxLER 0FFICE OF NUCLEAR REGULATORY RESEARCH ACCIDENT EVALUATION BRANCH
l DEBRIS C00 LABILITY RESEARCH l WHEN HAS AN ACCEPTARLE ANSWER BEEN OBTAINED? l 0 ACCUMULATION OF DISPOSITIVE EXPERIMENTAL DATA FOR RANGE OF ACCIDENT AND PLANT CONDITIONS 0 ADDRESSED NON-PROTOTYPIC FEATURES IN THE EXPERIMENTAL FACILITY 0 RECONCILED CONTRADICTORY DATA SPECIFICALLY O COMPLETE MACE TEST MATRIX PLUS ADDRESS EFFECTS OF CORIUM COMPOSITION (METAL FRACTION), WATER DELIVERY (EARLY/ LATE) CONCRETE FRACTION 0 EVALUATE EFFECTS OF DEH TECHNIQUE ON EXPERIMENTAL RESULTS (OETERMINE POROSITY / PERMEABILITY OF QUENCHED MATERIAL) 0 RESOLVE ISSUES SURROUNDING MELT SPREADING FOR SLOW POURS OUT OF VESSEL WHERE APPLICABLE OR UNDER LOW SUPERHEAT/ SOLID DEBRIS CONDITIONS. t 4 .m. . - ~. -.., _ ~.. -. - -, -.. - ,-,.-.m. ,,.,-..m._, .._....~,-.,.-,-~~- -, m....-.. . -.w 4.--.. ..+.5- ,,..____~_-____m
r. DEBRIS C00 LABILITY RESEARCH RESEARCH PERFORMFD TO DATE 0 SWISS TESTS: SWISS-1 AND SWISS-2 PERFORMED AT SANDIA IN 1987; 304 STEEL MELT IN 22 CM DIA CONCRETE CAVITY; DEBRIS NOT COOLABLE. O FRAG TESTS: FRAG-3 AND FRAG-4 PERFORMED AT SANDIA IN 1987; HOT STEEL SHOTS IN 22 CH DIA CONCRETE CAVITY; DEBRIS NOT C00LABLE. 0 WETCOR TESTS: WETCOR-1 PERFORMED AT SANDIA IN 1991; OXIDIC HELT IN 32 CM DIA CONCRETE CAVITY; DEBRIS NOT C00LABLE. O HACE TESTS: H1 PERFORMED AT ARGONNE IN 1991; OXIDIC HELT IN 50 CM x 50 CM CONCRETE CAVITY DEBRIS NOT C00LABLE. H1B PERFORMED AT ARGONNE IN 1992; OXIDIC HELT IN 50 CM x 50 CH CONCRETE CAVITY; LIMITED DEBRIS COOLABILITY. O ACM TESTS: ACM002 PERFORMED AT JAERI IN 1991; THERMITE IN 10 CH DIA HGO CAVITY; DEBRIS COOLABILITY UNDETERMINED. ACH003 PERFORMED AT JAERI IN 1991; THERMITE IN 10 CH DIA HGO CAVITY; EFFECTIVE SURFACE COOLING OBSERVED. 6
a a L._ a 4 E-G.. ep & 9 e DEBRIS COOLABILITY RESEARCH i RESEARCH FINDINSS 4 i 0 EXPERIMENTAL DATA BASE SPARSE AND COCLABILITY INCONCLUSIVE. O SCALE EFFECT IN EXPERIMENTS NEED TO BE CONSIDERED. r 0 CooLABILITY MAY REQUIRE REDEFINITIbH. l 0 ANALYTICAL MODELING OF~ FUNDAMENTAL PHYSICAL PHENOMENA LIMITED. k 7 Gew-+4=u enseawourm=,-m-'=e-e-wTet*%e+sw+e ese-ew-N**Dwe 4+*4er5He**""""f4-7'O"uT9 + awO--D C'Yv4T++7*$aref'We"* w= M SM d'T+W PG WW W b49 W
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t-DEBRIS COOLABILITY RESEARCH FUTURE REC 0094ENR4TK9NS O SASM TYPE METHODOLOGY (MACE / TAC) WITH PARTICULAR APPLICATION TO DEBRIS-COOLABILITY SHOULD BE PURSUED TO STUDY POSSIBLE SCALE EFFECTS. 0 PHENOMEN0 LOGICAL STUDY OF FUNDAMENTAL COOLABILITY PROCESS AND MECHANICS SHOULD BE PURSUED LEADING TO IMPROVED ANALYTICAL MODELING. O ADDITIONAL INTEGRAL. TESTS NEEDED TO RESOLVE OUTSTANDING CONCERNS. b 0 SEPARATE EFFECTS TESTS MAY BE NEEDED 8
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MELCOR Egga REVIEW'S MAJOn FINDINGS i ON DEGREE OF COMPLETION REC 0094ENDATIONS HELCOR NUMERICS MooELS " MISSING" FROM HELCOR 1.8.1 MooELS NEEoING IMPROVEMENT EXPANDED ASSESSMENT T e DOCUMENTATION 2
NELCOR PEER REVIEW RECOMMENDATIONS i HISSING MODELS BASIS I e MODELS NEEDED~TO REPRESENT KEY PHENOMENA MISSINs M00ELS PRIORITIZED CONSIDERING IMPACT ON 4 e CoNTAISteENT FAILURE' TIME e SOURCE TERM RECOooeENDATIoMS FoR HIGHEST PRIORITY ATTENTION PRIMARY SYSTEM COMPONENT NATURAL CIRCULATION e HIGH PRESSURE MELT EJECTION AND ' DIRECT CONTAINMENT HEATING e e ICE CONDENSER- [ NON-EXPLOSIVE INTERACTIONS BETWEEN DEBRIS AND WATER e FISSION PRODUCT VAPOR SCRUSBING e ADoITroNAL FISSION PRODUCT DEPOSITION AND SURFACE REACTIONS NRC PREVIOUS ACTIONS-FIRST THREE MODELS IDENTIFIED AS MISSING HAD ALREADY BEEN SCHEDULED FOR DEVELOPMENT AND INCORPORATION INTO MELCOR AT THE -[ TIME OF THE PEER REVIEW -- -----------5-----------
~ L 's HELCOR PEER REVIEW RECOMMENDATIONS EXISTING MODELS NEEDING IMPROVEMENT BASIS ~ INDIVIDUAL MODELS WERE EVALUATED FOR TECHNICAL ADEQUACY e HIGN PnIoaITY ATTENTION'NEEDED CONSIDERING EFFECTS ON e e TIME OF CONTAINMENT FAILURE e MAGNITUDE OF-SOURCE. TERM RECOM4ENDATIONS FOR HIGHEST PRIORITY ATTENTION: EVALUATE WATER CONDENSATION / EVAPORATION MODCL FOR ITS APPLICATION TO AEROSOL PARTICLE GROWTH AND DEPOSITION IMPROVE REPRESENTATIONS OF CORE, CONCRETE CHEMICAL REACTIONS AND e PHASE DIAGRAMS IMPROVE MODEL FOR CONDENSATION IN CONTAINNENT -e POOL' SCRUBBING MODEL DECONTAMINATION FACTORS SHOULD BE EVALUATED AND REVISED AS NECESSARY 4 NRC PREVIOUS ACTIONS WATER CONDENSATION / EVAPORATION.MODEL DEMONSTRATION WAS SCHEDULED e AS A PART OF FY91 EFFORT. CORE CONCRETE CNEMICAL AND PHASE DIAGRAM IMPROVEMENTS WILL BE e PROVIDED BY IMPLEMENTING CORCON-M003 INTO NELCOR, PLANNED BY NRC IN 1991. NRC CONCURRENT-ACTIONS IMPROVEMENT IN CONDENSATION MODEL' SCHEDULED FOR FY93 e POOL SCRussING MODEL EVALUATION PLANNED-e 7
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s 4 'i. SCDAP/RELAPS NRC ACTIVITIES PRESENTATION TO THE ACRS SUBCOMITTEE ON SEVERE ACCIDENTS BY ALAN M. RUBIN ACCIDENT EVALUATION BRANCH DIVISION OF SYSTEMS RESEARCH l MAY 27, 1992 -. - - - ~ ~ - - - - - - - - -, - -. - - - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - - - - ~ ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' ,...._.....-,.-..... -.-----...-.- ---..~.- ---..--
r i SCDAP/NELAES O CODE OBJECTIVES e' DESCRIBE RESPONSE OF PRIMARY RCS DURING A SEVERE ACCIDENT UP TO REACTOR VESSEL OR SYSTEM FAILURE c MODEL IN-VESSEL MELT PROGRESSION PHENOMENA FOR e VARIOuS ACCIDENT SEQUENCES FOR PWRS AND BWRS SYSTEM THERMAL HYDRAULICS (RELAP) CORE' DAMAGE PROGRESSION HYDROGEN GENERATION FISSION PRODUCTS BEHAVIOR MODEL FULL-SIZE PLANTS AND' EXPERIMENTAL FACILITIES L e i FOR CODE ASSESSMENT f l 2 .. ~.. - ~.........
t SCDAP/RELAPS OVERALL STATUS e ONGOING TESTING OF SCDAP/RELAPS/M003[7x] BY SEVERAL CSARP MEMBERS I e COMPLETED INITIAL EARLY PHASE MODEL ASSESSMENT e IDENTIFIED MODEL IMPROVEMENTS 4 FOCUS ON IMPROVEMENTS IN RUNNING CODE FOR FULL-SIZE PLANT CODE RELIABILITY t CODE USABILITY INTERFACE BETWEEN RELAPS AND SCDAP EARLY PHASE COMPONENT MODEL BEHAVIOR LATE PHASE MODELING e ONGOING PEER REVIEW - INDEPENDENT ASSESSMENT OF MODELING CAPABILITIES AND LIMITATIONS AND ADEQUACY OF DOCUMENTATION i e DEVELOPMENT ASSESSMENT /MODEL DEVELOPMENT ON HOLD PENDING COMPLETION OF ABOVE WORK 3
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s THE CONTAIN CODE i i 'O NRC'S BEST ESTIMATE MECHANISTIC CODE-FOR INTEGRATED ANALYSIS OF SEVERE ACCIDENT CONTAINMENT PHENOMENA FOR BOTH PWRS AND BWRS O INTENDED FOR IN-DEPTH $TUDY OF SEVERE ACCIDENT SAFETY ISSUES i 0 CAPARILITY.TO PREDICT CONTAINMENT LOADING AND RADIOLOGICAL SOURCE TERMS i -t 0 LATEST VERSION - CONTAIN 1.12 9 0 NRC CONTRACTOR: SANDIA NATIONAL LAs0RATORIES (SNL) 2 er e-, .w, w-w evw e w-r..wmv-asr%-e v,s --v,.-.--w,-ve..ir,w-wr---we-,w---su r wr-w e w =--- w +, w,www vwww-m w-wrwe w -r w www w-me w ve-e.-www.-r--- + w.ww .s. wrw--
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i .~ e RECENT CONTAIN CODE ACCOMPLISHNENTS 0 CODE VALIDATION CONTRIBUTED TO ISP-29 (HDR E11.2) ACTIVITY PERFORMED CONTAIN CALCULATION OF HDR E11.4 (NRC/ GERMANY AsREEMENT) 0 CODE DEVELOPMENT / IMPROVEMENTS COMPLETION'0F HECTR-HYDROGEN BURN HODELLING UPDATES INITIAL APPLICATION AND ASSESSMENT OF CONTAIN FOR ADVANCED PASSIVE (PWR DESIGNS) 4
o i OVERVI.EW OF CONTAIN ACTIVITIES 0 PRO' GRAM OBJECTIVES: SUPPORT ALWR EFFORTS AND RESOLVE DCH ISSUE FOR' PLANTS WITH PW TYPE CONTAINMENTS. t 0 ALWR CONTAIN PROGRAM (DETAILS DISCUSSED UNDER ALWR PROGRAM) MODIFY CODE TO CAPTURE ALWR UNIQUE FEATURES VALIDATE CODE AGAINST EXPERIMENTAL TEST DATA PERFORM ALhNt CONTAINMENT ANALYSIS O CONTAIN CODE DEVELOPMENT AND MAINTENANCE PROGRAM = DCH RELATED CODE MODELING IMPROVEMENTS DCH CONTAIN ANALYSIS i UPDATE SELECTED CODE MODELS . CODE DOCUMENTATION USER SUPPORT / CODE MAINTENANCE 6 0 CONTAIN: CODE PEER REVIEW WILL BE ORGANIZED AND EXECUTED BY LOS ALAMOS NATIONAL LABORATORY FORMAT SIMILAR TO MELCOR AND SCDAP/RELAPS REVIEWS PLANS TO BEGIN IM.2ND-OUARTER OF FY93 i 3 -l ...J
5, asu ~ l ALWR RESEARCH ACTIVITIES 4 PRESENTATION TO THE ACRS MAY 1992 CHARLES G. TINKLER ALAN H. RusIN OFFICE OF NUCLEAR REGULATORY RESEARCH i ACCIDENT EVALUATION BRANCH
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.? ALWR PROGRAM 'O USER NEEDS IDENTIFIED BY NRR (DECEM8ER 1990) FOR AP600 AND SBWR 0' MAJOR. FOCUS AREAS 0 ' INDEPENDENT AUDIT CAPABILITY 0 ADEGUACY OF VENDOR TEST PROGRAMS O IDENTIFICATION-OF. REGULATORY ISSUES RELATED TO UNIQUE DESIGN . ASPECTS O SEVERE ACCIDENT PERFORMANCE n O INDEPENDENT AUDIT CAPABILITY t ASSESSMENT OF CONTAINMENT PERFORMANCE FOR BOTH DBAS AND SEVERE -0 ACCIDENTS ASSESSMENT OF PLANT RESPONSE / PERFORMANCE FOR SEVERE ACCIDENTS O 0 PLAN TO ASSESS AND: UPGRADE AS.NECESSARY EXISTING NRC CODES l 0 CONTAIN, COMIX, MELCOR, SCDAP/RELAPS 1 F t 2 l
OVERVIEW OF ALWR ACTIVITIES ' O CONTAIN CODE IMPROVEMENTS / ASSESSMENTS (SANDIA NATIONAL LABORATORIES) O COMIX CODE IMPROVEMENTS / ASSESSMENTS (ARGONNE NATIONAL LABORATORY) O MELCOR CODE INPUT DECK / SIMULATION STUDY SB6R (OAK RIDGE NATIONAL LABORATORY) AP600 (BROOKHAVEN NATIONAL LABORATORY) 0 SCDAP/RELAPS CODE INPUT DECK / SIMULATION STUDY AP600/SB6R (IDan0 NATIONAL ENGINEERING LABORATORY) 0 ASSESSMENT OF ADEOUACY OF VENO 0R ALWR CONTAINMENT PERFORMANCE TEST PROGRAMS 0 ASSESSMENT OF HYDROGEN HIGRATION/ COMBUSTION FOR THE AP600 (SANDIA NATIONAL LAs0RATORIES) t 3
i .ay OTHER ALWR RELATED PROGRANS 0 ENERGY RESEARCH, INC. PERFORM INDEPENDENT SCALING ANALYSIS OF VENDOR CONTAINMENT EXPERIMENTAL TEST PROGRAMS FOR AP600 AND SEWR PLANTS 0 SANDIA NATIONAL LamonATORIES ASSESSMENT OF HYDROGEN MIGRATION /COM8USTION INSIDE THE AP600 0 IDENTIFY POTENTIAL IMPACT FROM H RELEASES AT VARIOUS i g LOCATIONS DEPENDENT UPON POSTULATED SEOUENCES i GUANTITATIVE ASSESSMENTS OF HYDROGEN COMBUSTION INSIDE o THE AP600 s I i 6 i I - - - - -. ~... - -... - - -. -.... -. -. -. -.. -. - - - - - -.. ~. -. -......... - -. - - -. - - .. - - - - ~.. --}}