ML20136F913

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Informs of Addl Idcor Severe Accident Technical Issues Which May Not Receive Adequate Consideration at Nrc/Idcor 850326 Meeting.Issues Should Be Raised at Meeting
ML20136F913
Person / Time
Issue date: 03/25/1985
From: Hulman L
Office of Nuclear Reactor Regulation
To: Rosztoczy Z
Office of Nuclear Reactor Regulation
Shared Package
ML20136F919 List:
References
FOIA-85-781 NUDOCS 8504110380
Download: ML20136F913 (29)
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{{#Wiki_filter:I ( f DISTRIBUTIO!1 LLNIRAL FILL AEB R/F (dAR 2 51985 11EliORANDUM FOR: Zoltan R. Rosztoczy, Chief Research & Standards Coordination Branch Division of Safety Technology FROM: L. G. Hulman, Chief Accident Evaluation Branch Division of Systems Integration

SUBJECT:

ADDITIONAL IDCOR/NRR SEVERE ACCIDENT TECHNICAL ISSUES As a result of my review of IDCOR, NRC contractor and NRC staff severe accident assessments, and my participation in IDCOR meetings and source tenn analysis and policy activities, I have concluded that two important issues may not receive adequate consideration at the NRC/IDCOR meeting scheduled for March 26, 1985. Specifically, I conclude the following:

a. n.either NRC or IDCOR have established scrutable quality assurance / quality control procedures for evaluations and related code assessments; and
b. IDCOR has credited operator intervention of severe accidents in a much more optimistic manner than has the staff or its contractors.

(While it may be creditable to conclude that operator intervention would unlikely be successful for rapidly evolving accidents, a similar conclusion for slowly developing events does not appear appropriate). I request that these issues specifically also be raised at the IDCOR meeting. In particular, any differences of opinion in both areas should be clearly understood. Original !;ned by] : L. G. Hulman, Chief Accident Evaluation Branch Division of Systems Integration cc: R. Bernero G. Marino T. Speis J. Rosenthal D. Muller J. Read M. Silberberg L. Soffer TC : DSI: EB

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s .. t f~ETING AGENDA

SUBJECT:

NRC/IDCOR MEETING ON ASSESSMENT OF SEVERE ACCIDENT. UN, CERTAINTIES 4 l DATE: THURSDAY, SEPTEMBER 25, 1985 TIME: 9:00 A.M. - 2:00 P.M. LOCATION: RAMADA INN, BETHESDA AGENDA INTRODUCTION - NRC - ZOLTAN ROSZTOCZY 9:00 - IDCOR - ANTHONY BUHL THE USE OF UNCERTAINTIES IN REFERENCE PLANT-9:20 EVALUATIONS - ZOLTAN ROSZTOCZY

SUMMARY

OF NRC PROGRAMS ON THE ASSESSMENT OF 9:40 SEVERE ACCIDENT UNCERTAINTIES - JOSEPH MURPHY, NRC DISCUSSION 10:10 BREAK '10:40

SUMMARY

OF IBCOR PROGRAMS ON THE ASSESSMENT OF 10:50 SEVERE ACCIDENT UNCERTAINTIES - ROBERT HENRY, FAI COMMENTS ON THE IDCOR UNCERTAINTY ANALYSIS - 11:20 4 RICHARD BARRETT, NRC DISCUSSION 11:40 CLOSING STATEMENTS - IDCOR - ANTHONY BUHL 12:10 - NRC - ZOLTAN ROSZTOCZY A4-A4-

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g j 9 :. .-^ ~ 'l It THE-USE OF' UNCERTAINTIES IN REFERENCE PLANT EVALUATIONS PRESENTED BY ZOLTAN R. ROSZTOCZY U.S. NUCLEAR REGULATORY COMMISSION NRC/IDCOR MEETING ON ASSESSMENT OF SEVERE ACCIDENT UNCERTAINTIES BETHESDA, SEPTEMBER 26, 1985 i As-AC l F-J _-_--___---___----__.-.,_.-_-_--_----_Lu-_- a

-..,. ~ .:m~_.,..-,_ m,, _. _.. s D t THE USE OF UNCERTAINTJES IN REFERENCE PLANT EVALUATIONS ' STANDARD ENGINEERING TECHNIQUES ACKNOWLEDGE OUR INABILITY TO BE EXACT, TO COVER EVERY POSSIBLE OCCURRENCE. THEY ACCOUNT FOR THESE BY REQUIRING A MARGIN, FREQUENTLY CALLED SAFETY MARGIN. IN STRUCTURAL ENGINEERING, FOR EXAMPLE, A MARGIN OF A FACTOR OF THREE IS CUSTOMARY. WHEN THE CALCULATIONAL UNCERTAINTIES (INCLUDING' MODEL ING UNCERTAINTIES) ARE SMALL RELATIVE TO THE REQUIRED MAMGIN, IT IS COMMON PRACTICE NOT TO INCLUDE AN UNCERTAINTY ALLOWANCE ON-THE TOP OF THE MARGIN. IN THE EVALUATION OF SEVERE ACCIDENTS THE UNCERTAINTIES ARE EXPECTED TO BE LARGE, THUS DUE CONSIDERATION OF UNCERTAINTIES IS A DIFFICULT TASK. THE PROPOSED SAFETY GOAL IS, UNFORTUNATELY, ONCLEAR ON THE TREATMENT OF MARGIN AND UNCERTAINTIES. THE LATEST VERSION OF THE SAFETY GOAL, WHICH WAS PRESENTED TO ACRS ON JULY 11, 1985, RECOGNIZES THE SIGNIFICANT UNCERTAINTIES INHERENT IN PRAS ANDPROPOSESPROVISIONALGUIDELINESFORTgIALUSE. ONE OF THESE GUIDELINES IS AN UPPER LIMIT OF 10-PER YEAR ON CORE MELT FREQUENCY. IF THE CORE MELT FREQUENCY EXCEEDS THIS NUMBER, SAFETY IMPROVEMENTS SHOULD BE CONSIDERED, o THE QUESTION IS, IN VIEW OF THE PRESENT STATE-OF-THE-ART, HOW ARE WE GOING TO USE THE ESTIMATED UNCERTAINTIES OF THE' REFERENCE PLANT EVALUATIONS? AS IT IS OUTLINED IN'THE POLICY STATEMENT, THE EVALUATIONS WILL BE DONE BY A DETERMINISTIC APPROACH SUPPLEMENTED BY PROBABILISTIC ANALYSIS. CONSEQUENTLY, THE UNCERTAINTIES WILL BE USED IN CONJUNCTION WITH BOTH OF THESE METHODS. 1 I

an -- . -.,~,. - - m - - - - --t9 1 TREATMENT OF UNCERTAINTIES IN THE DETERMINISTIC EVALUATION I6E DETERP!NISTIC EVALUATION WILL ADDRESS-SIGNIFICANT CONTRIBUT. TO CORE MELT, FISSION PRODUCT-RELEASE'AND TRANSPORT. EARLY AND LATE CONTAINMENT FAILURES, AND ~ PROTECTIOr, OF THE PUBLIC IN CASE OF CONTROLLED AND UNCONTROLLED RELEASES FROM CONTAINMENT. THE GOAL OF THE EVALUATION IS FIRST,'TO IDENTIFY MAJOR CONTRIBUTORS, THEN EXAMINE WHAT HAS BEEN DONE AND WHAT MORE CAN BE DONE TO REDUCE THE LIKELIHOOD AND CONSEQUENCES OF THE-ACCIDENT, AND FINALLY MAKE-RECOMMENDATIONS ON WHERE~ TO DRAW THE LINE, WHICH OF THE' POTENTIAL IMPROVEMENTS ~NEED. TO BE IMPLEMENTED, AND WHICH ARE NOT. THE OVERALL EVALUATION WILL BE COMPOSED OF-A' SERIES-OF INDIVIDUAL DECISIONS BASED ON~ ENGINEERING' ANALYSIS AND JUDGMENT. THE UNCERTAINTIES ASSOCIATED WITH.THE ASSESSMENT OF EACH ISSUE WILL BE TREATED INDIVIDUALLY AND WILL BE FACTORED INTO THAT DECISION. FOR EXAMPLE, LET'S ASSUME THAT FOR A'GIVEN PLANT THE EVALUATION INDICATES THREE MAJOR-POTENTIAL !MPROVEMENTS (AN IMPROVEMENT IN THE ECCS, A. CHANGE IN THE: EMERGENCY PROCEDURES, AND BETTER DISPLAY INSTRUMENTATION) WHICH COULD REDUCE THE LIKELIHOOD OF~ CORE' MELT.- FURTHERMORE, LATE CORTAINMENT FAILURE IS' JUDGED TO REPRESENT SUFFICIENTLY LOW RISK TO THE PUBLIC, AND THERE ARE TWO WAYS (UPGRADED CONTAINMENT SPRAY SYSTEM, !MPROVED HYDROGEN ~ CONTROL SYSTEM) TO AVOID EARLY CONTAINMENT' FAILURE..THEN EACH OF THE FIVE POTENTIAL IMPROVEMENTS WILLHBE-JUDGED ON'THEIR OWN MERITS,; INCLUDING UNCERTAINTIES ASSOCIATED ~WITH THEIR-ASSESSMENT, AND DECISIONS WILL'BE MADE WHETHER TO' REQUIRE ANY OF THE IMPROVEMENTS. L Ar r i [ +:

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yc 3-1 TREATMENT OF UNCERTAINTIES IN THE PROBABILISTIC EVALUATION THE PROBABILISTIC ANALYSIS'HAS THE ADDED BFNEFITS OFl ? CONSIDERING A BROADER f?ECTRUM OF EVENTS, QUANTIFYING:THE RESULTS, AND COMBININC THEM.INTO A' RISK Er71 MATE. -THE c QUANTIFICATION AND Cot:21 NATION PERM TS TRADE 0! S 'IN THE MATHEMATICAL SENSE BETWEEN PREVENTION AND MITIGATION-AND i 1 ALSO PERMITS PROPAGATION OF UNCERTAINTIES. i THE~THREE MAIN RESULTS OF THE ANALYSIS ARE: ' CORE MELT FREQUENCY, THE AMOUNT AND TIMING ~0F RADI0 ACTIVE MATERIAL RELEASES; AND RISK. THUS, THE UNCERTAINTIES OF-THE ANALYSIS SHOULD BE DISPLAYED-IN TERMS OF THESE RESULTS. 2 BOUNDS'GIVEN~0N UNCERTAINTIES SHOULD BE CLEARLY DEFINED AS 4 A GOAL, EVEN IF IT IS HARD OR IMPOSSIBLE TO ARRIVE'AT EXACT VALUES. i TO ASSURE COMPLETENESS, NO-UNCERTAINTIES SHOULD BE ELIMINATED FROM CONSIDERATION IN AN ARBITRARt FASHION,.FOR i EXAMPLE DUE TO THE LACK OF INFORMATION. ELIMINATION SHOULD BE BASED ON THE MAGNITUDE OFLTHE EXPECTED i CONTRIBUTION T0-THE RESULTS. t. L THE ESTIMATED UNCERTAINTIES'WILL BE USED'(1) FOR COMPARING THE RESULTS'AGAINST PROBABILISTIC CRITERIA LIKE-THE-SAFETY GOAL; (2) COMPARING CALCULATIONS ~AGAINST~0THER CALCULATIONS, i FOR EXAMPLE IDCOR CALCULATIONS:VERSUS NRC CALCULATIONS; i (3) SEARCHING FOR' VULNERABILITIES IN PLANT-DESIGN AND PLANT { OPERATION; AND (4) PRIORITIZATION~0F FUTURE EFFORTS; t c I: I i 4 i t i L - ~ =.

.F gRREGog S' O O o { s o i, E n.M tt) 'q ,[ N g <q o s 4**** REVIEW OF THE IDCOR UNCERTAINTY ANALYSIS R. J. BARRETT NRC/IDCOR TECHNICAL EXCHANGE MEETING ON SEVERE ACCIDENT UNCERTAINTIES l SEPTEMBER 26, 1985 1 1 i 1 M. i As

~ s c 4 9 'l ( Sc0PE OF-IDCOR ANALYSIS THE IDCOR UNCERTAINTY ANALYSIS DEALS ONLY WITH: - CONTAINMENT PERFORMANCE- - FISSION PRODUCT BEHAVIOR i 5 A COMPLETE SEVERE ACCIDENT UNCERTAINTY ANALYSIS i SHOULD ALSO ADDRESS UNCERTAINTIES IN: - ACCIDENT SEQUENCE FREQUENCY - 0FFSITE CONSEQUENCE CALCULATIONS, [ l c i t

~. :a. u. ~ ~ = - O i CONTAINMENT AND FISSION PRODUCTS 4 GENERAL FEATURES OF IDCOR RESULTS UNCERTAINTY ANALYSES GENERALLY DID NOT LEAD TO QUALITATIVELY DIFFERENT OUTCOMES - DIRECT HEATING FAILURE FOR ZION - EARLY HYDROGEN BURN FAILURE IN SEQUOYAH TMLB l THE RANGE OF RESULTS OFTEN EXCLUDED RESULTS CALCULATED BY OTHER CREDIBLE ANALYSTS - THE FULL RANGE OF AEROSOL RELEASE FRACTIONS FOR THE ZION V-SE0DENCE IS ORDERS OF MAGNITUDE LOWER THAN THE BMI-2104 l RESULTS IN MANY CASES, NUMERICAL VARIATIONS IN CALCULATED RESULTS WERE' NARROW - IN SEQUOYAH TMLB, IN-VESSEL HYDROGEN PRODUCTION' VARIES FROM 459 T0 492 tsM; CONTAINMENT FAILURE TIME VARIES FROM 18.85 TO 19.24 HOURS 4 .h -w----.e.- g r-y v +-e-

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.w,. - :, + a.. .o _: I 1 CONTAINMENT AND FISSION PRODUCTS NRC CONCLUDES THAT THESE LIMITATIONS RESULT PRIMARILY FROM THE FOLLOWING FACTORS: - IMPORTANT PHENOMEN0 LOGICAL UNCERTAINTIES HAyE~NOT BEEN. QUANTIFIED i - THE POTENTIAL NEGATIVE EFFECTSlDF ACCIDENT SEQUENCE VARIATIONS, INCLUDING OPERATOR ERRORS OF OMISSION AND COMMISSION, HAVE NOT BEEN QUANTIFIED l - SOME IMPORTANT SEQUENCES HAVE NOT BEEN INCLUDED IN THE ANALYSIS 4 - FISS10N PRODUCT BEHAVIOR HAS NOT BEEN EXAMINED IN SUFFICIENT DETAIL THE FOLLOWING VIEWGRAPHS PRESENT SOME EXAMPLES OF EACH-TYPE OF OMISSION i i + ( .h$ 1

,,y-r __m.m PHENOMEN0 LOGICAL UNCERTAINTIES. i EXAMPLES. EFFECT OF CORE MELT PHENOMENOLOGY UNCERTAINTIES ON. IN-VESSEL HYDROGEN-PRODUCTION CATASTROPHIC' CONTAINMENT FAILURE VS.' LEAK-BEFORE-BREAK-NON-VOLATILE FISSION PRODUCT RELEASES IN. CORE-CONCRETE < INTERACTIONS DIRECT HEATING ACCIDENT SEQUENCE UNCERTAINTIES EFFECT OF AUTOMATIC AND OPERATOR ACTION ON IN-VESSEL HYDROGEN PRODUCTION V-SEQUENCE WITH PIPE RUPTURE RATHER THAN PUMP SEAL' FAILURE. EFFECT OF IMPAIRED SECONDARY BUILDING ON V-SEQUENCE RISK BWR LOCA SEQUENCES WITHOUT CRD PUMPS AVAILABLE' l AG 5

SygERGYSTic EFFECTS SEQUOYAH IMLB - ALLOW IN-VESSEL HYDROGEN TO VARY BY RAISING TZ00FF AND CONSIDERING ALTERNATIVE MODELS FOR MELT PROGRESSION - ALLOW VARIATION OF CSI AND CS0H RETENTION IN THE RCS - MODEL EX-VESSEL TELLURIUM RELEASE - CONSIDER HYDROGEN BURN MODELS THAT ALLOW HIGHER CONCENTRATIONS OF HYDROGEN ZION TMLB - - CONSIDER UNCERTAINTY IN RCS RETENTION OF CSI AND CSOH - RELEASE TE FROM THE FUEL EX-VESSEL - RETAIN THE ENTIRE CORE IN THE REACTOR CAVITY

.- ~_. _ ~ -..,.e ~... .m i 4 ADDITIONAL SEQUENCES A COMPLETE ANALYSIS OF THE PEACH BOTTOM TC SEQUENCE THE EFFECT OF DISABLING TZOOFF ON THE SEQUOYAH TMLB SEQUENCE j I DETAIL ON FISSION PRODUCTS FOR PEACH BOTTOM, ONLY CSI AND CSOH RELEASES ARE QUOTED I N0 oVANTIFICATION OF LANTHANIDE GROUP RELEASES IS PRESENTED i MORE DETAILED DESCRIPTION OF FISSION PRODUCT DISTRIBUTION IN THE PLANT AND TO THE ENVIRONMENT 4 i t t l Ac, ...w- ,r v.

~ CONCLUSIONS IDCOR ANALYSISISHOULD BE EXPANDED TO INCLUDE UNCERTAINTIES IN: - ACCIDENT SEQUENCE FREQUENCY I - 0FFSITE CONSEQUENCE CALCULATIONS A COMPLETE UNCERTAINTY ANALYSIS FOR CONTAINMENT PERFORMAhCE AND FISSION PRODUCT BEHAVIOR WOULD REQUIRE MORE ATTENTION TO: - PHENOMEN0 LOGICAL UNCERTAINTIES - ACCIDENT SEQUENCE VARIATIONS - SYNERGYST!c EFFECTS - ANALYSIS OF ADDITIONAL SEQUENCES - MORE DETAILED EXAMINATION OF FISSION PRODUCTS l b

n .w NUREG 1150 UNCERTAINTY ANALYSIS J. A. MURPHY-e SEPTEMBER 26, 1985 J b1

i ?:: =. GENERAL PRINCIPLES o - ASSESS IMPORTANCE-OF SOURCES OF UNCERTAINTY TO DECISIONS o NEED REASONABLE,. CREDIBLE RANGE IN '.WHICH - ACTUAL VALUE WILL BE FOUND -(90 PERCENT DEGREE ~ 'OF BELIEF)" ~ o NEED NOT BE EXPRESSED IN FORMAL STATISTICAL BOUNDS o. PURPOSE-IS TOL FOCOS ATTENTION ON IMPORTANT ASSUMPTIONS- -w

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l 1 ACCIDENT SEQUENCES AND CORE MELT FREQUENCY DE ~ERMt NE UNCERTA ~ T7 9 A N G ~i BY PROF > S GATIO N OF DA-5 l ' ! NC ERTAI NTIES ~ l ~ r o CONDUCT SENSITIVITY STUDIES BY VARYING KEY ' ASSUMPTIONS AND PARAMETERS AND RECOMPUTE UNCERTAINTY DISTRIBUTION FOR. EACH VARIATION o CONSIDER MULTIPLE VARIATIONS WHEN CORRELATED OR WHEN , LITTLE-CONFIDENCE (ABOUT 50 PERCENT) IN BASE CASE o .UNLIKELY.TO CONVOLUTE MORE THAN THREE INDEPENDENT-PARAM ETERS 1 I W. l 'i 4 .t_ L...

y INTERIM PRODUCT ] o. VARIATION IN MEAN OF DISTRIBUTION 1 o 5 AND 95 PERCENT BOUNDS FOR EACH SENSITIVITY STUDY AND BASE CASE o-IDENTIFICATION-OF FACTORS DRIVING UNCERTAINTY ESTIMATES AND VARIATION OF MEANS o HIGHLY DESIRABLE TO ASSIGN " DEGREE OF BELIEF" TO BASE CASE AND EACH SENSITIVITY STUDY AND COMBINE IN BAYESIAN PROCESS FOR NUREG-1150 o-BAYESIAN ANALYSIS PLANNED FOR NUREG-1150

TU E B E R T PR R E RO H AF T S S S I E H R S E G OA U F B R O )C T RI ED T L E H S B T T T I I I N M SM E SI SI E S L O V S S S E PA E AP T N T P O O N L N E T A D NR F M E T E E T N DN I P L I NE E M A EC B E0 T P 5 N E I C F l O DI OA 1 N T 1 C S E E I I E M EBG EI R E T RP G LR L HO E U I T( DWN o o )

SOURCE TERMS o TYPICALLY APPROXIMATELY 15 RELEASE BINS -TYPICALLY APPROXIMATELY 6 STCP RUNS / PLANT o o STCP RUNS DIRECTED TOWARD CENTRAL ESTIMATE o EXTRAPOLATION REQUIRED FOR REMAINING CENTRAL ESTIMATE OF L SOURCE. TERMS (ISOTOPIC RELEASE FRACTIONS, RELEASE ENERGY AND TIMING CHARACTERISTICS) AND ALL OPTIMISTIC AND PESSIMISTIC ESTIMATES FOR EACH PATHWAY THROUGH CONTAINMENT EVENT TREES o EXTRAPOLATION UNDER REVIEW BY SOURCE TERM REVIEW GROUP TO ASCERTAIN IF.ESTIMA'.ES. REFLECT BEST ESTIMATES OF SOURCE TERM GIVEN THE ASSOCIATED PHYSICS AND CHEMISTRY OF THE TREE ANALYSIS 4

e CONSEQUENCE CALCULATIONS 5/95 BASED ON PLANT-SPECIFIC METEOROLOGICAL DATA AND -o EMERGENCY PREPAREDNESS ASSUMPTIONS, BEST ESTIMATE VALUES FOR OTHER - PARAMETERS o ESTIMATE OF VARIATION OF MEAN CONSEQUENCES FROM PARAMETERS OTHER THAN METEOROLOGY AND EMERGENCY PREPAREDNESS ASSUMPTIONS -J

l i.

SUMMARY

OF UNCERTAINTY ANALYSES 1-1... STOCHASTIC UNCERTAINTIES ASSOCIATED WITH ACCIDENT i SEQUENCES- (5/95) 2. . VARIATION OF MEANS ASSOCIATED WITH ACCIDENT SEQU_ENCES (APPROXIMATELY 90%. DEGREE OF BELIEF) '3. OPTIMISTIC, CENTRAL, AND PESSIMISTIC PASSES THROUGH CONTAINMENT EVENT TREE (NO DEGREE OF BELIEF ATTRIBUTED IN SARRP BUT AN ATTEMPT TO CONDUCT . A ' LIMITED BAYESIAN ANALYSIS WILL BE ATTEMPTED FOR. NUREG-1150) e

SU M MARY OF U NC ERTAINTY AN ALYSES (CONTIN UED) i 4. . POTENTIAL FOR OPTIMISTIC, CENTRAL, AND PESSIMISTIC. SOURCE TERMS ASSOCIATED WITH EACH PASS THROUGH THE CONTAINMENT EVENT TREE: IF POSSIBLE, A LIMITED' BAYESIAN ANALYSIS WILL BE INCLUDED IN-NUREG-1150 5. STOCHASTIC UNCERTAINTIES ASSOCIATED WITH M ETEOROLOGY (5/95. FOR COST-BENEFIT ANALYSES; FULL RANGE FOR CCDFs)

6..

VARIATION OF MEANS FOR ALL OTHER CONSEQUENCE . PARAMETERS, EXCEPT USE SITE-SPECIFIC, NRC-SUPPLIED EMERGENCY RESPONSE ASSUMPTIONS 5 ].

l ?,. RISK CALCU LATIONS 1. PRELIMINARY UNCERTAINTY DISPLAYS-AND COST-BENEFIT ANALYSES o COMBINE THE FIVE CATEGORIES OF UNCERTAINTY ANALYSES AS DOUBLES (TWO PARAMETERS. AT EXTREMES OTHERS AT CENTRAL ESTIMATE) o .FOR EXAMPLE, COMBINE VARIATION OF MEANS OF ACCIDENT. SEQUENCES WITH. OPTIMISTIC, CENTRAL, AND PESSIMISTIC CONTAINMENT EVENT TREE PASSES--USE MEANS OF STOCHASTIC ' UNCERTAINTIES, CENTRAL ESTIMATES 'OF SOURCE TERMS, AND BEST ESTIMATE OF OTHER CONSEQUENCE PARAMETERS 2. CCDFs--SAME AS ABOVE. (1.E.,. ONLY. COMBINE AS~. DOUBLES), EXCEPT: o USE FULL DISTRIBUTION FOR METEOROLOGY o METEOROLOGY IS ALWAYS ONE OF THE DOUBLES D

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March 7,1995 MEMORANDUM FOR: Themis P. Speis, Director. Division of Safety Technology FROM: Zoltan R. Rosztoczy, Chief Research and Standards Coordination Branch

SUBJECT:

NRC/IDCOR MEETING ON OUTSTANDING TECHNICAL ISSUES ASSOCIATED WITH SEVERE ACCIDENTS DATE & TIME: March' 26,1985, 8:30 A.M. - 4:30 P.M. LOCATION: Room 1046, 1717 H Street, N.W. Washington, D.C. . PURPOSE: To discuss specific technical issues which have been identified either as major contributors to the uncertainty in plant severe accident evaluation or represent significant differences in modeling assumptions between the NRC and IDCOR studies. IDCOR will also describe a proposed approach for the generalization of reference plant results to other operating plants. PARTICIPANTS: NRC USpeis, R. Bernero, Z. Rosztoczy, M. Silberberg, R. Curtis G. Marino, J. Rosenthal, R. Barrett, G. Bagchi, J. Read IDCOR R. Henry, M. Fontana,' M. Kenton, J. Gabor .d.. " d T7.. i ZoltanR.Rosztoczy, Chief Research and Standards Coordination Branch Division of Safety Technology i .a ,g s e s w erib 4 ~,. __ _-_.- +

3 v n n. T? {? Summary of Major NRC/IDC0lk Technical -Issues for Severe Acciants T l In the followin'g paragraphs a' number of specific technical iskues are identi-fied that are either, major contributors to the uncertair,y in plant risk or represent signi,ficant differerices in modeling assumptions between the NRC and IDCOR. It is important to identify another difference between NRC and IDCOR methodclogies that is more basic. In The Severe Accident Risk Rebaselining and Risk Reduction Program (SARRP) uncertainties in phenomenological behavior are treated explicitly. Not only is a best estimate analysis parformed, but pessimistic and optimistic analyses are also performed to provide an under-standing of the range of possible outcomes. Although some sensitivity studies have been performed,in the IDCOR program, their purpose appears to be to support the selection of assumptions leading to point estimate results. It is the opinion of the NRC and its supporting contractors that the uncertainties in severe accident processes are quite large and that a meaningful evaluation of plant. safety and the possible need for plant modifications must include the explicit consideration of uncertainties. One significant source of uncertainty is the definition of the accident sequences. In several cases where IDCOR and NRC have analyzed the same or similar sequences, large differences in the calculated consequences have. resulted because of different assumptions about the sequence of events in the ~ accident. These differences often reflect a real uncertainty in the accident sequence definition. The NRC believes that it is particularly important to reflect these uncertainties in severe accident analyses. I Core Heatup Stage Issue #1 - Fission Product Release Prior to Vessel Failure Nk Uncertainties in fission product release prior to vessel failure can affect both the timing of release and the total quantity of release,from fuel. Whether the fission products are released in-vessel or ex-vessel can be particularly important because of the potential for retention of fission products on reactor coolant system structures for the in-vessel release com-ponent. Several sub-issues contribute to th'is issue; namely l y ._m._,. _,. _ _ _. .,.-_,_y ...,,.._,..m,. ,, _ ~ ,_,-.--e,,-

...w_.. .~ - -~- 2 (a) Assured temperature for beginning of fuel relocation. 4 IDCOR models a~ssume and employ a relocation temperature of 2800C (3100K); i.e., the melting point of UO., This high assumed temperature in conjection with 2 their fission product releese model (see below) allows all fol,atiie fission products such as I, Cs, at. Te to be released prior to core slumping and quenching in tt}e low'er plenum. As a result 100% of these F.P.'s are available to be deposited in the upper-plenum and other structures of the primary system, thereby resulting in'possible non-conservatism in the source term to the con-tainment for,certain sequences. i r The BMI-2104 models assume that melting occurs at an average eutectic tempera-ture of 2550K. In conjunction with the empirical CORSOR release coefficients, this temperature is high enough to release essentially all of the volatiles (except tellurium) in-vessel, but may underpredict the release of involatile materials. A more mechanistic fuel melt progression model is being developed for the MELP.ROG code based on experimental evidence from KfK PBF, and ACRR. (b)'Modeling of in-vessel release o'f fission products The IDCOR fission product release from fuel model is based upon the oxidation i of UO by steam. The model assumes that sufficient steam and contact area 2 (with UO ) is present at all times during the heatup to oxidize the UO to a 2 2 higher state, 'thereby significantly enhancing the release of fission products. Each of the species considered is released from the fuel at the same rate. Partial pressures of the vapor species are used to' determine the amount of released material that can be transported as a vapor. The condensed component is apportioned between aerosol and fuel surfaces according to an' input factor. ~ The BMI-2104 model, CORSOR, is entirely ' empirical based on a variety of.in-pile and simulant experiments. The BMI-2104 model does not attempt to account for differences in mass transfer limiting processes that may exist between the experiments and the' real system. AL e r 7- ,-,,.m-, .e 7 y

~ ~ j 3 \\ Recent PBF experi.nts indicate that the C0fiSOR and IDCOR models both probably overestimate in-vessel release of fission products. The VICTOR M code which is being developed as an element of the MELPROG package will provide a more 1 mechanistic model for fission product release which can be compared with integral experiments in PBF and A'CRR, [ (c) Te retention in. vessel A major difference exists in'the treaUent of tellurium release between the IDCOR model and the BMI-2104 model. Experiments at ORNL and at the PBF strongly indicate that the tellurium release is reduced during this stage of the accident if approximately 25% or more of the Zircaloy remains unoxidized. Because of the rapid heat-up of a core during boiloff, most sequences result in unoxidized Zircaloy contents ijreater than 25%. Therefore, in most sequences, the CORSOR release model allows most of the Te to be retained in the core while it is in the reactor vessel. The significance of this effect is that if the Te is released early - as IDCOR advocates - it will be deposited in the upper ~ plenum and not be available in the source term estimate; however, if it is retained in the molten fuel at this stage - as experimental evidence indicates 3, - it will be released when the core exits the R.V. and interacts with the concrete. Thus, the Te would be released to the containment volume without the potential for being deposited in the reactor coolant system and, perhaps, at a { time closer ta the time of containment failure. Although the COR50R model accounts for the interaction of tellurium with the Zircaloy cladding, the model is crude and the supporting data are sparse. The overall significance of this issue is that, for certain sequences in both PWR's and BWR's, significantly higher amounts of volatile fission products (I, Cs, and Te) may be available for release to the containment during the core / concrete interaction stage than computed in the IDCOR models. Depending on the containment failure. time, this effect can and does lead to muc'h larger source j terms than computed by IDCOR. The PIE of the PBF tests should help clarify j. this issue in early FY 1986. AL

w ~ ;m. m w.. _ Issue #2 - Recirculation of Coolant in the Reactor Vessel %g w..a. Neither IDCOR nor current NRC (BMI-2104) models are able to calculate recir-culation patterns of steam in the R.V. after core uncovery. Ba' sed on simpli- .fied analyses at SAI (EPRI) and Purdue, it appears that recirculating flow could have a significant effect on the core heatu' behavior of PWR's at least fc-high prenure sequences. It is speculated that recirculating flow can affec'. the core' heatup rate, in-vessel release of fission products, quantity of I hydrogen produced in-vessel, structure temperatures, and deposition of fission products in the reactor coolant system. Analytical studies have been initiated with the COBRA, C0tWIX, and TRAC codes to resolve this issue. These studies will be completed by December, 1985. Experimental results are being obtained by Westinghouse which should offer an opportunity for model validation. Issue #3 - Release Model for Control Rod Materials 3 ; 6,,'a

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A major discrepancy between the IDCOR and BMI-2104 models is the mode'of be-havior of silver and cadmium during core meltdown. The IDCOR model allows the control rod material to rapidly melt and runoff to cooler regions of the core where it freezes and takes no further part as a potential. source of inert aero- ~ sol material. In contrast, the BMI-2104 models ignore the potential for runoff of liquified control rod material prior to regional slumping and, as a result, may tend to overpredict the release of control rod silver and cadmium. l Neither the IDCOR or BMI-2104 analyses consider the possibility of chemical reactions with B C control material that could result in increased hydrogen 4 production or changes in the chemical form of fission product species. The SFD1-4 experiment in PBF is designed to study questions related to the behavior of silver-indium-cadmium control rods. Investigations at ORNI. are addressing B C control material behavior. 4 6> AL

N%d .W. ,s# w 4k 5 q; Ns(w ha ,s,,I' o% w 5 Issue #4 - Fission Product an'd Aerosol Deposition ....., o. c. in the Primary System ( ~ The uncertainties in the existin.g predictive capal,ility (TRAP-MELT 2 is the most detailed of the methods available) for deposition in the ; reactor coolant system are very' large. The vapor and aerosol transport processes are complex and interacting. The upper plenum geometries and flow regimes' e' treated in a crude approxima' tion of the prototype. In initial discussions with IDCOR, the. focus of NRC concern was the use of a' log normal aerosol size distribution in the RETAIN code. This formulation has now been replaced with an empirical aerosol model, the validity of which is of equal concern in the RCS geometry. Results of verification experiments at ORNL and the integral Marviken tests will provide a means to evaluate the existing models. II Melt Progression and Fuel Relocation Stage Issue #5 - Modeling of In-Vessel H, Generation' x,. c Substantial agreement exists in.the modeling of the steam /Zircaloy reaction in the reactor core as long as the original geometry is maintained.- However, when cladding melts'and slumps and flow channels begin to block, issues arise as'to 4 when blockage will occur, how effective it will be, whether cladding material will run.out of the hot zone, how much oxidation will occur as it moves, and-whether the relocated cladding will subsequently be reheated and exposed to steam. The IDCOR models employ.two parameters: a metal-water reaction cut-off temperature and a flow blockage parameter that effectively limit the extent.of 4 hydrogen production that occurs in-vessel. Similarly'the BMI-2104 code, MARCH 2, has externally controlled parameters that can limit the extent of metal-water reaction. In practice, however, the "best-estimate" assumptions ~made by. the IDCOR and BMI-2104 analysts have lead to substantially dif'ferent results. o AL _.-,._,r. .,._,.,.,_,.m ,w _._...__,,_..,,,,,,.,,4, _,m-.~...,ym_.m n..

~. -., - u. 6 For a PWR.SBLOCA sequence without ECCS, IDCOR models calculate 200 Kg of H2 produced, whereas, the the BMI-2104 calculation yields 450 Kg of H 3

  • 2 significant difference of a factor of 2.25.. Experimental data.of H2 production after fuel relocation 1s cifficult to obtain, but careful analysis of the PBF Phase I program.and planned ex.eriments in the NRU facility wherein full-length e'cments will be tested to slun. ping and held for long times will help to resolve this issue.

The MELPROG code will employ more mechanistic fuel slumping models and a two-dimensional fluid flow model.which will also provide a better under-standing of the issue. Issue #6 - Core Slump, Core Collapse, and Reactor Vessel Failure Models ~'d, cev.- ~7[ The IDCOR and BMI-2104 models of core slumping are greatly simplified repre-sentations of very complex processes. Some key features of the models are: (a) IDCOR, assumes that molten core material becomes isolated from the rest of the system until a slumping criterion is satisfied and then it instan-taneously slumps to the lower core support plate (LCSP). The NRC model treats approximately the effect of in-core fuel relocation and growth of the molten region. N l (b) IDCOR models assume failure of the LCSP after a user-input fraction of the l core is molten. NRC models calculate failure of lower core support structures due to heating by slumped core material. (c) IDCOR assumes immediate vessel failure after failure of LCSP. ' The NRC model permits' the user to select early local head failure o'r later gross head ov'erheating. The condition of the core material (mass, composition. and temperature) at the time of vessel failure can have a major influence. on the subsequent loads on containment (steam spike, direct heating) and the extent of dispersal of the core debris. l G. Az -p r--,- -,----.m- -,n-e,.-,-.o-n---n.---. ,w- .w ,-.n.-.----,-.,

7 s. The best. hope in resolving this issue is by' detailed best-estimate calculations of melt-progression using the MELPROG code. Since integral experiments in this area are clearly impractical, the code models will have to be validated with data from the ACRR debris relocation experiments scheduled to be completed in FY 1985. l Issus #7 - Alpha Modt :ontainment Failure by In-Vessel g , Steam Explosions The major significance of this issue is whether sufficiently energetic molten fuel / coolant interactions can occur in the lower plenum to produce a missile of sufficient energy to breach the containment (comonly referred to as the Alpha-mode failure). IDCOR contends that this;is not possible. There have been many discussions by NRC staff, NRC consultants, and IDCOR staff on the validity of IDCOR's contention. To date, however, the question has not been resolved to everyone's satisfaction except for a general feeling amongst all parties that " explosions in excess of 2000 MJ would be required to fail containment. Explosions of such high energy are deemed unlikely, but have not been demon-strated to be impossible." i \\ Work is continuing to resolve this issue by experiments at SNL, in the UK, and bybetter,more-detailedcalculationsofthemassesinvolvedusingtheMELPROG code (see issue f6). A report on this subject will be issued in March, 1985 Sy. an NRC-sponsored expert review group (NUREG-1116). 1 III Ex-Vessel State Issue #8 - Direct-Heating of Containment by Ejected Core Material i ~"Go <..- This issue is critical to containment failur'e timing for high-pressure sequences, such as TMLB', in large-dry PWR containments. In~the IDCOR calculation of ?.he TMLB' sequence for the-Zion Plant, it is assumed that half i i e ,--x. .-m.

.u. ..u o .m c,. n. .m ..x.. 8 of the molten material ejected from the vessel under pressure is swept out of the reactor cavity onto the containment floor. The analysis does not account for potential rapid heating of the containment atmosp.here by the cr.m debris or the cxidation of the core debris during transit and further heating of the atmosphere. The potential for " direct heating" of the containment atmosphere .is being investigated by the NRC in an experimental program which is underway at SNL. Results' will be available in FY 85. ~ TheissueofdirectheatingisrelatedtoIssue#2. If recirculating flow patterns within the. reactor coolant system result in alternative failure locations such as the hot leg, then the RCS will depressurize prior to melt-through of the lower head and the core debris will not be dispersed from the reactor cavity. As stated before, COBRA, TRAC, and C0fti!X calculations are being perfomed to study this possibility under the supervision of the CLWG. i Issue #9 - Ex-Vessel Fission Product Relaase Qbs l A potentially significant modeling difference between the IDCOR model 'and the NRC models'NORCON and VANESA) is related to the release of refractory fission i i products during the core / concrete interaction process. The NRC models allow for the production of volatile oxide.s and hydroxides of t.hese fission products by reaction with steam and carbon dioxides sparging through the melt. Uncer-tainties in the prediction of the ex-vessel release of fission products involve j the composition (masses of materials and oxidation state) at the time of vessel l failure, initi.al temperature of the melt, extent of core dispersal, modeling of core-concrete attack as well as the complex chemical behavior within the melt. Experiments are planned at SNL in which simulant fission products will be included in core-concrete attack te.*ts. These data should provide a basis for ~ i-testing CORCON and VANESA modeling assumptions by late 1985. i. Issue #10 - Ex-vessel Heat Transfer Models from Molten Core to Concrete / Containment . b ^,',

  • l This issue is associated with the magnitudes and mechanisms of energy transfer from the molten core debris to the concrete, to the containment atmosphere, and 1

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g l 1 l to overlying water pools (if present). The issue :ar. per.entially impact the mode and. timing of containment failure and t'he crms of fission products. This is an area in which major dif t.;rtr ' enst in the modeling

mptions used in the IDCOR and BMI-2104 anals -

~ general, the BMI-2104 at analyses involve more heat going into concrete attr4, more rapid production of non-condensible gases, more rapid' pressu;ization of 9a containment, but lower atmosphere temperatures, particularly in the BWR analyses. These differences 1 are partly the result of assumptions made regarding debris dispersal or spreading and,' partly, differences in core-concrete attack models. When water is present in the cavity, it'is assumed in the IDCOR ansiyses that a coolable debris bed will form. In the BMI-2104 analyses this possibility was treated - rametrically. Core-Concrete tests in the BETA facil q.and at Sandia " (involving sustained urania/ concrete heating tests) in FY 85 will provide an expanded data base for improving and validating models. Issue #11 - Revaporization of Fission rmiocts in the Upper Plenum O b '" I C y th the IDCOR and NRC models for in-vessel transp# d'eposition of fission ~oducts predict that a large fraction of fissier p meti: released from the e can be deposited on structures in the reactor - system. Associated ith these fission products is a significant portion J *e decay heat with the ctential to result in direct heating of structures an.: le revaporization of 6 posited fission products. In the most recent IDLOF. eslyses revaporization taken into account resulting in some enhancement of ;he environmental sour'ce " m for some reactor types and accident sequences. A o'rber of major uncer-

Jnties remain, however, including: a very. sparse d6La base related to the "c nanisms of reaction between fission product species nd surfaces (and the

.. Mntial for revaporization), the modeling of natural .. vection driven flow p.'tterns in the reactor coolant system prior to, and suavguent to, lower head we and the behavior of RCS insulation ma'terial in u.. accident environ-Coupling of the TRAP / MELT and MERGE codes has been achieved to permit ment. - is of the effects of revaporization.during the peri-prior to vessel At -r.

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~ ~ w-7 10 I L, meltthrough and in a parametric manner after meltthrough. A more rigorot treatment.will be provided by the coupled RELAP5 and TRAP / MELT codes whit being undertaken for the SCDAP code package. 'Results of analysis should be available in FY 85. Issue #12 . Deposition Model for Fission Products in Conteinment g,.-/-l 3 Because of the, limitations in the applicability of the log-normal size distri-bution assumption for aerosols, which,was incorporated in the RETAIN code, IDCOR has developed an empirical correlation for aerosol settling which has-been included in the MAAP code. This new model has also been criticized by NRC t contractors for not including a dependence on particle size and the lack of wide-range comparisons to experimental data. This issue will probably not be settled until direct comparison calculations.are made for specific plant / sequence events using both the NRC and IDCOR models. Issue #13 - Amount and Timing of Suppression j Pool Bypass <'~ ~ ]._ @ W., Analyses of suppression pool scrubbing in the BMI-2104 study indicate that the l effectiveness of suppression pools is quite good even when the temperature of 1 the pool is at saturation. In general, the extent of pool bypass was found to be a more impo'rtant source of uncertainty than the details of suppression pool modeling. In the IDCOR analyses suppression pool decontamination has been treated simplistically using constant decontamination factors and ignoring bypass under the assumption that bypass leakage would become plugged by l aerosols. The potential for bypass is quite plant-design dependent. However, since pool bypass will govern the magnitude of environmental releases in Mark III BWR designs, it should be explicitly considered in the analyses for this- ~ type of cont'ainment. 4 Issue #14\\ModelingofEmergency sponse v won The predic d consequences of sev re accidents can be ver sensitive to the modeling of ergency response. Th is _ particularly true r early fatalities 1 ~.

..c .x ~ j 11 heause of their threshold nature and the high dependenc'e.of dose to proximity to\\releasepoint. In the ba'h case IDCOR analyses, the population in the s evacuat h zone is removed at a gi.vSn rate without recognition of a straggler populationNhat is slow to evacuate or refuses to evacuate. In combination with long ass' d warning times.and the assumed mode of containment ailure. this assumpti ults in the prediction of qo early fatalities, even for sequences in wh'ich on the order of ten percent'o,f the volatile fission products i are assumed to be rel'essed from the plant. The modeling of emergency response ' - is complicated'by a number of factors including: tNe,humanelement(reluctance to leave home, panic, etc h site depIndence, and dependence on the weather and time of day. Nxx Key,sub-issues: q Evacu'htion Model; Definitio'n of Warning Time; Containmeni' failure N Mode. 'N Issue #15 - Containment Perfonnance [' The potential magnitude of the source term to the environment is largely controlled by the mode and timing of containment failure. If the containment i~ remains intact for a number of hours following melting of the core and the release of fission products, the potential consequences will be substantially reduced either'through natural deposition processes or by the action of con-j tainment safety features such as sprays, coolers, suppression pools, and ice l condensors. Over the past few years, the analyses of containment performance and model experiments at Sandia National Laboratories have verified the expec-tation that substantial margins exist between the design conditions and condi-tions at which major leakage of containment structures can be expected to - Considerable uncertainty still exists, however, as to how rapidly occur. j leakage will grow as a function of pressure and temperature for different containment designs, the mode by which a containment will fail, and.the location at which failure will be initiated. In the IDCOR analyses, contain-ment failure is typically characterized by a leak -rate sufficient to prevent l l At m.-.-,- y4, ,,e.,r.e mr ,,m.. w. y

_ _ _. w ~ ~ m m...~ a, w. .... e .mc 12 further pressurization.. This assumption has the effect of increasing the residence, time of fission products in the containment and enhancing the effec-tiveness of evacuation. In the BMI-2104 analyses, failure is typically charac-terized as a large hole that leads to depressurization of the containment, but does not involve total destruction o' f the bu'iding. Sensitivity studies have also been performed for leak rates that vary as a function of internal 4 r essure, i i The importance'of better determining the conditions leading to containment failure is plant-and sequence-dependent. Predicted consequences for a large, dry containment (such as Zion) can be insensitive to the pressure at which failure occurs, if the potential for early failure can be precluded. In con-trast, the Mark I source term results are found to be very sensitive to the location of failure and the conditions leading to failure. Experimental programs at Sandia will provide important data on the behavior of gasketed penetrations, electrical penetration assemblies, and the structural response of concrete containments. These, combined with further plant'-specific analys.es,sEo~uldservetoresolvetheoutstandingissues. ~ Issue #16 - Secondary Containment Performance 7 The IDCOR analyses tend to give greater credit for secondary containment per-fonnance than the BMI-2104 analyses. The differences are particularly evident in the IDCOR V sequences in which considerable deposition is predicted to occur in the auxiliary building even in the absence of water pool scrubbing. It is i also evident frora ORNL SASA analyses that the reactor building surrounding the Mark I primary containment could significantly mitigate the consequences of l severe accidents under a given set of conditions or assumptions. The principal l uncertainties regarding the effectiveness of secondary containment buildings relate to: the mode of primary containment failure and its impact on the sur-vival of the secondary building; potential for hydrogen deflagrations to occur in the secondary building; and the modes of leakage and failure of the i I h. ,-w-. - =, e- -w .~=<-4,,-,e, -#,--,w-,-,,.,,, ..-~,,,,m- ..-,,,~ww,-,v,---m. ,,,,y,--,,.=

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r 13 seccndary building. Secondary building performance is of particular interest in' accident sequences involving early failure or bypass of the primary contain-ment envelope and, thus, has the potential to mitigate consequences. The t uncertainties in secondary containment pcrformance could be reduced by adci-i tional analyses since the processes of interest are essentially the same as those-in the primary containrant for which methods have t:sn d,eveloped. To the extent that the. mode of failure of th'e primary containment influences the per-formance of the secondary buildings, the uncertainty will always remain large. Issue #17'- Hydrogen Ignition and Burning . a , :.a There are substantial differences between the IDCOR and NRC treatment of igni-tion, burning, and flame propagation in air-hydrogen-steam mixtures. -IDCOR analyses base ignition on a calculated flame tiemperature criterion which is a function of the composition of the atmosphere within the compartment. NRC analyses, in addition to considering hydrogen and oxygen concentrations within a compartment, give explicit consideration to steam inerting and the availa-bility of ignition sources. The IDCOR models appear to predict continuous burning in essentially all cases, whereas the NRC treatment tends to predict a . number of discrete burns. The NRC's approach tends to allow the buildup of higher hydrogen concentrations and hence can lead to the prediction of higher y containment pressures. The differences between the IDCOR and NRC treatment of e hydrogen ignition and burning are particularly pronounced in multi-compartment systems, such as the ice condenser containment, and in the absence of deli-berate ignition. A number of related sub-issues are noted below. ~ ' ~ Effect of Natural Convection l IDCOR models include consideration of natural convection driven flow between compartments which appears'to lead to enhanced hydrogen burning at low concen-i trations. NRC models do not include consideration of natural convection flows.- Flame. Propagation \\ Due to a combination of differences noted elsewhere, NRC treatment of hydrogen combustion indicates greater likelihood of flame propagation into the upper h1- - ~ < -..., -,, ._,y _,,..,,..~g., c.w -,._,,,, _ _.. _ _., ~..., s. i

14 compartmant of the ice condenser conte nment where impact on containment pressurization is the greatest. Potentially Detonable Concentrations NRC's treatmen.t of hydrogen. combustion indicates the possibility of developing potentially detonable compositions in local areas, e.g., the upper plenur W the ice condenser. IDCOR's treatment appears to preclude such localizc:i hydrogen buildup ~s.' Effect on Chemical Fonn of Fission Products ~ -~ There is experimental evidence that hydrogen combustion can alter the chemical form of airborne fission products, e.g., release of molecular iodine from cesium iodide aerosols due to hydrogen flames. It is not clear whether such ~ changes in chemical form increase or decrease the consequences of severe acci-dents. Neither IDCOR nor the'NRC analyses at present consider such changes in i fission product chemistry. Resolution of the outstanding issues will come, in part, from continued com-parisons betWeen experiments and analyses. It must be recognized, however,' that experimental data may not be available to address issues related to 4 burning in complex geometries. Issue #18 - Essential Equipment Performance ap s t Neither the BMI-2104 nor IDCOR analyses have provided a detailed assessment of the ability of essential equipment to survive the conditions associated with a severe accident environment (high temperature, high humidity, pressure differentials, flames, high radiation, and high aerosol loadings). This equipment has been qualified to survive the LOCA environment which is in many ways similar. However, experiments at Sandia have indicated that the more severe environment of core meltdown accidents co.:Id lead to the degradation of equipment, particularly cables. Once substantial core degradation has occurred, protection of containment integrity becomes the key safety function. e. -~,m. ,--...-,-,_._-,,.,--.~--..-v 7-,-y.--_--..-. - - -?---., ..-.--er

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15 Whether or not containment safety equipment is v.ulnerable to the severe acci-dent environment depends on the design of the plant. Not only does the issut-involve questions of the pt formance of the equipment, but also of the ability of the operatar to monitor conditions in the containment and to control essential equipment. Thus,'the. degradation of i.11toring and control systems could also potentially degrade the effectiveness of the containment or accident management stra'tegies to protect the public. e t e e e s i S e S i e A2. e

...~ u n.a. ,.:..u.......-.. w -. ~ ~ .......a.. s., ....,w ~ 9 MEETING OBJECTIVES 1. Discuss and perhaps agree on. criteria for evaluating issues. 1 2. Discuss ratings for each issue to obtain IDCOR viewpoint informally. Ultimately the rating will be an NRC product which will be provided to IDCOR as an aid for use in drafting guidelines for plant-specific analyses. 4 3. Perhaps identify and assign tasks to be undertaken over the next half year. e < S e ..b e 4 .) - e O e i 4 4 4 i l i i F A2 l-

PROPOSED CRITERIA FOR EVALUATING ISSUE'S 1. Degree of difference between NRC (BMI-2104, NUREG-0956, and SARRP) and IDCOR models and assumptions Comment: Irrespective of risk significance Rating: H. M, L 2. Impact of uncertainties and modelling assumptions on estimated risk Comment: Either as modelling assumptions lead to differences in. NRC and IDCOR risk estimates or conceptually as the un - certainty associated with an issue affects risk ~ Rating: H. M, L 3. Potential to impact plant-specific regulatory decisions l Comment: Identify plants and sequences for which the issue is potentially important-t Rating: Plant / sequence 4. Issue identified as important by American Physical Society i Rating: Y, N t 5. Need for long term research prior to severe accident closure l Comment: Is the potential outcome of this issue sufficiently + l critical, that closure for some plant types might have l to be deferred until some major research task (perhaps experimental) is completed? Recognize that even though the uncertainty associated with an issue may be large, it may not be necessary to obtain technical r~esolution prior to severe accident closure. Rating: Y, N. If Y, describe need 6. Need for short term research and analysis prior to severe accident closure Comment: Short term implies a time-scale of approximately six months, consistent with the development of guidelines for plant-specific evaluations in CY 1985 Rating: Y, N. If Y, describe need and perhaps as. sign responsibility AL l ._.7 + y .yy

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7. Expected need to improve treatment of this issue tumnd" current IDCOR assumptions or models for the development or spplication lof guidelines for plant-specific analyses Coment: This rating could be affected by the r6sults of short term research or analyses Rating: Y, N 8. Reconciliation with DRA0 identification and rating of PRA back-end issues. Rating: Identify corresponding DRA0 issues. Ylf, H, M,.L 4 ) e e *b o e g ~ A.

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O' ISSUE #9 - EX-VESSEL HEAT TRANSFER MODELS - FROM MOLTEN CORE TO CONCRETE / CONTAINMENT DEFINITION OF THE ISSUE NRC ANALYSES PREDICT MORE HEAT G0ING INTO CONCRETE ATTACK BECAUSE ASSUMPTIONS REGARDING DEBRIS DISPERSAL,.0R SPREADING, OR DIFFERENT CORE-CONCRETE INTERACTION MODELS. LIMITATIONS THE ISSUE IS GENERALLY APPLICABLE TO ALL SEVERE ACCIDENT SEQUENCES THAT PROGRESS TO REACTOR VESSEL FAILURE BY MOLTEN CORIUM. . EFFECTS ON RESULTS NRC'S MODEL RESULTS IN MORE RAPID PRGDUCTION OF NON-CONDENSIBLE GASES, MORE RAPID PRESSURIZATION, BUT~ LOWER ATMOSPHERE TEMPERATURES., PARTICULARLY IN THE BWR ANALYSES. UNCERTAINTY- .IDCOR'S ASSUMPTION 'THAT A.C00LABLE BED WILL BE FORMED'WHEN WATER IS PRESENT IN THE CAVITY IS NOT SUPPORTED BY RECENT SNL TESTS. MORE TEST DATA ARE NEEDED. RESOLUTION ' CORE-CONCRETE INTERACTION TESTS IN THE BETA FACILITY AND AT SANDIA IN FY 85 WILL PROVIDE EXPANDED. DATA: BASE FOR IMPROVING AND VALIDATING THESE MODELS. Y

1 i; !!SSUE #10 - EX-VESSEL FISSION PRODUCT RELEASE + i DEFINITION OF THE ISSUE NRC'S MODELS FOR THE RELEASE OF REFRACTORY FISSION PRODUCTS DURING THE CORE-CONCRETE INTERACTIO PROCESS ALLOWS FOR THE PRODUCTION OF VOLATILE OXIDES AND HYDROXIDES OF TilESE FISSION PRODUCTS BY ! REACTION WITH STEAM AND CARBON DIOXIDE SPARGING THROUGH THE MELT, i LIMITATION THE ISSUE IS GENERIC FOR ACCIDENTS PROGRESSED TO THE FAILURE OF THE REACTOR VESSEL BY M0LTEN CORIUM. l EFFECT ON RESULTS j NRC'S MODEL WILL RESULT IN A HIGHER RADIONUCLIDE RELEASE TO THE ATMOSPHERE. ? UNCERTAINTY l UNCERTAINTIES IN THE PREDICTION OF THE EX-VESSEL RELEASE OF FISSION PRODUCTS INVOLVE THE i COMPOSITION AT THE TIME OF VESSEL FAILURE, INITIAL TEMPERATURE OF THE MELT, EXTENT OF CORE ~ DISPERSAL, MODELING 0F CORE-CONCRETE ATTACK.AS WELL AS COMPLEX CHEMICAL BEHAVIOR WITHIN THE. MELT. RESOLUTIdN ON-G0ING EXPERIMENTS AT SNL SHOULD PROVIDE A BASIS FOR TESTING NRC ASSUMPTIONS BY LATE 1985, i l .p 't

~l i ISSUE #12 - DEPOSITION MODEL FOR FISSION PRODUCTS IN CONTAINMENT l i i I i DEFINITION OF THE ISSUE - THE NRC HAS SIGNIFICANT RESERVATIONS ABOUT THE IDCOR'MODEL BECAUSE IT DOES NOT INCLUDE EFFECTS AND PARAMETERS KNOWN TO BE IMPORTANT IN OTHER STUDIES. LIMITATIONS - THE ISSUE IS OF MOST IMPORTANCE IN SEQUENCES WITH LATE CONTAINMENT FAILURE l WHERE AEROSOL PROCESSES HAVE TIME TO OPERATE. ~ f EFFECTS ON RESULTS - NRC METHODS PROVIDE A MORE REALISTIC ASSESSMENT OF THE EFFECTS OF MULTIPLE RELEASES AND THE EFFECTS OF STEAM ATMOSPHERE ON AIRBORNE SOURCE TERM. UNCERTAINTY - FAILURE TO INCLUDE PHENOMENA KNOWN TO BE SIGNIFICANT CAUSE SKEPTICISM,' USE 0F MORE RIGOROUS.MODELS WOULD ADD CONFIDENCE TO CONCLUSIONS. RESOLUTION - ADVANCED ANALYTICAL METHODS SUCH AS CONTAIN WHICH ALLOW AN ARBITRARY. SIZE i DISTRIBUTION AND. INTERACTIVE TREATMENT OF AEROSOLS AND THERMAL-HYDRAULICS WHEN ASSESSED AGAINST DATA FROM NSPP, DEMONA AND LACE EXPERIMENTS WILL PROVIDE RELIABLE ANALYSIS OF [ ~ THIS PHASE-0F.A SEVERE ACCIDENT. l^ I a

a [ ISSUE #17 - HYDR 0 GEN IGNITION AND BURNING 4 q t i i t DEFINITION - 3 f-

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CODES USE DIFFERENT ASSUWTIONS Am DIFFERENT MDDELS FOR IGNITION, BURNING AE FLAE PROPAGATION l 1 i ECTR - HYDR 0 GEN OXYEN AND STEAM CONCENTRATIONS, MAAP - CRITICAL FLAE TEPPERATURE f I LIMITATIONS - T 1 L NOT AN ISSUE FOR BWR MARK'I OR II. ~

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] 1 EFFECTS ON ESULTS - } v i P0DELING DIFFEENCES ESULT IN ECTR PREDICTIONS OF HIGHER TEPFERATURE AE PRESSUE FOR BWR' l MARK III,. ICE COMENSER, AND LARGE DRY PWR CONTAllfENT TYPES, COPPARTPENTALIZATIOS IN MAAP CODE j DOES NOT ADEQUATELY TEAT TRANSPORT AND SUBSEQUENT EFFECT OF MIXING ON DIFFUSION FLAPES. i UNCERTAINTY - 5 IDCOR - MODELS NOT Awsn WITH EXERIENTAL DATA l - COPPARTENTS NOT EALISTIC REPRESENTATION.0F REACTOR CONTAINNT - MODELS IN ECTR VERSION _1.0 AE ELL DOCUENIED Y o l

2 ES0LllTION - o ' HYDR 0 GEN RULE FOR BWR MARK 111 AND PWR ICE CONDENSER EFFECTIVE FEBRUARY 28, 1985. o NEED FOR llYDR0 GEN CONTROL IN LARGE DRY PWR TO BE DETERMINED IN CY 85, -0 ANALYSIS NEEDED TO DEK)NSTRATE COWLIANCE WITH RULES TO BE DETERMIED, G I P G .y f -' ~

_. ~ _. _ -.. _ _,. _ .,..,,,,,,, ~.,, _ ..,, ~. _ -m ~ . ISSUE #11 - REVAPORIZATION OF FISSION PRODUCTS IN THE UPPER PLENUM DEFINITION o THE IDCOR FISSION PRODUCT REVAPORIZATION MODEL DOES NOT CONSIDER CHEMICAL'. INTERACTIONS OF DEPOSITED FISSION PRODUCTS AND STRUCTURAL SURFACES IN THE REVAPORIZATION PROCESS. LIMITATIONS o THIS ISSUE APPLIES TO ALL REACTOR TYPES AND ACCIDENT SEQUENCES. EFFECT OF RESULTS o CONSIDERATION OF THE AB0VE EFFECT-CAN DELAY THE TIMING OF REVAPORIZATION. THIS MAY RESULT IN A LARGER SOURCE TERM FOR DELAYED CONTAINMENT FAILURE ACCIDENTS. UNCERTAINTY o LACK 0F DATA BASE INTRODUCES LARGE UNCERTAINTIES - EXCLUSION OF CHEMICAL EFFECT NOT REALISTIC. 4 R550LUTION o IDCOR SHOULD CONSIDER THE REVAPORIZATION OF IRREVERSIBLY REACTED FISSION PRODUCTS IN THEIR MODEL. o THE NRC MODEL WILL BE AVAILABLE IN EARLY SUMMER. o SOME EXPERIMENTAL DATA WILL BE AVAILABLE IN 1985 (NRC, EPRI). At d

~ ISSUEJ360 . DIRECT HEATING OF C0tlTAINf1ENT BY EJECTED CORE MATERIAL e CORIUM DISPERSAL AND THE POTENTIAL FOR DIRECT HEATING 0F.. THE CONTAINMENT ATMOSPHERE BY TRANSFER OF THE LATENT HEAT OF THI DEBRIS AND BY EXOTHERMIC CHEMICAL REACTION NEEDS 1 TO BE CONSIDERED. e ISSUE IS BELIEVED TO BE RESTRICTED TO PWRs BASED ON LIKE-LIH00D THAT BWR CORE MELT AND SUBSEQUENT VESSEL FAILURE e WILL OCCUR AT LOW PRIMARY SYSTEM PRESSURE. e DIRECT HEATING IS A MECHANISM THAT PROVIDES A MECHANISM FOR EARLY CONTAINMENT FAILURE. e BASED ON QUALITATIVE AND SEMI-QUANTITATIVE ARGUMENTS, SUFFICIENT ENERGY RELEASE TO FAIL CONTAINMENT IS UNLIKELY TO OCCUR. l e NEITHER NRC NOR IDCOR HAS PRESENTED CREDIBLE QUANTITATIVE MODELS. QUALITATIVE MODELS HAVE BEEN EXPLORED BY THE l PARTIES. NRC IS FUNDING DEVELOPMENT OF MODELS AND PER-FORMANCE OF EXPERIMENTS. ~, _ _.,,, 9.. w-.

... ~... ~.. .... - ~, -. ~ i ~ e RESOLUTION-e IDENTIFY CONTAINMENTS / CAVITY DESIGNS IN WHICH J)EBRIS DISPURSAL LIKELY. ~ e DEVELOP SCREENING CRITERIA AND QUANTITATIVE MODEL 0F PHENOMENA, e PERFORM RISK. ANALYSIS USING SELF CONSISTENT SOURCE TERM. 'l. 'l 4 i d i O 9 O a 1 AL w w - 4 we,- ~w ,,v- -e -me--, s-r ur -w r

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_. ~ _ ..m~.._ _.. ~.... _. -. ~.. ISSUE #13B - FISSION PRODUCT RETENTION IN ICE BEDS D_EFINITION NRC MODEL (ICDEF) FOR F1SSION PRODUCT-RETENTION IN, ICE BEDS, o IS INCORPORATED INTO THE NAUA CODE.. PHENOMENA INCLUDED: BROWNIAN MOTION, GRAVITY, TURBULEhCE, INERTIA STEFAN FLOW, THERM 0 PHORESIS. o IDCOR REPORT FOR SEQUOYAH DOES NOT ADDRESS ICE BED RETENTION. SINCE THE ICE BED IS MODELED AS A SEPARATE COMPARTMENT IN THE CONTAINMENT DESCRIPTION, IT CAN BE INFERRED THAT THZ STANDARD CONTAINMENT MODELS APPLY: AEROSOL RETENTION IS BY STEAM CONDENSATION AND SEDIMENTATION. LIMITATION o APPLIES ONLY TO ICE CONDENSER CONTAINMENTS (PWR). EFFECT ON RESULTS FOR SEQUENCES WITH ICE (TML) AND FOR SEQUENCES WITHOUT ICE o (S HF) THE PREDICTEL RETENTION IN THE ICE BED WITH THE NRC 2 MODEL IS N10%. } UNCERTAINTY o APS REVIEW HIGHLIGHTS 2 MAJOR UNCERTAINTIES: SURFACE AREA 0F ICE DURING AN ACCIDENT OWING TO GLACIATION AND CHANGE WITH TIME, AND POTENTI AL FOR BYPASS FLOW. i l o NO EXPERIMENTAL STUDY OF AEROSOL REMOVAL PROCESSES IN ICE BEDS, BUT BELIEVED TO BE BASED ON SOUND SCIENCE. l RESOLUTION j o REVIEW EARLY DATA ON STEAM CONDENSATION IN ICE BEDS TO ADDRESS APS CONCERNS. CONDUCT EXPERIMENTS AT PNL, COMPARE WITH MODELS 9/86. o o-IDCOR SHOULD SPECIFICALLY MODEL ICE BED FISSION PRODUCT RETENTION. A2- ---,s,-,- y v. ,e.-,, .,e. w ,..mu- ,}}

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