View source for IA-85-465, Partial Response to FOIA Request for Five Specified Summaries of Idcor/Nrc Technical Exchange Meetings.Forwards App a Documents.Records Placed in Pdr.Search Continuing for Other Meeting Summaries

{{Adams
| number = ML20134F720
| issue date = 08/01/1985
| title = Partial Response to FOIA Request for Five Specified Summaries of Idcor/Nrc Technical Exchange Meetings.Forwards App a Documents.Records Placed in Pdr.Search Continuing for Other Meeting Summaries
| author name = Felton J
| author affiliation = NRC OFFICE OF ADMINISTRATION (ADM)
| addressee name = Pedro J
| addressee affiliation = NUS CORP.
| docket = 
| license number = 
| contact person = 
| case reference number = FOIA-85-465
| document report number = NUDOCS 8508210231
| package number = ML20134F729
| document type = CORRESPONDENCE-LETTERS, NRC TO ENGINEERING/CONSTRUCTION/CONSULTING FIRM, OUTGOING CORRESPONDENCE
| page count = 2
}}

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[o,,                        UNITED STATES                                 b(
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AUS 011995 Mr. Jim Pedro Document Support Services Licensing Information Service NUS Corporation 2536 Countryside Boulevard                                 IN RESPONSE REFER Clearwater, FL 33575-2094                                  TO F01A-85-465

==Dear Mr. Pedro:==

This is in partial response to your letter dated June 26, 1985, in which you requested, pursuant to the Freedom of Information Act (F0IA), five specified summaries of the IDCOR/NRC Technical Exchange Meetings.
The enclosed Appendix A is a list of the meeting summaries requested at (c),
(d), and (e) of your letter. These records are being placed at the NRC's Public Document Room,1717 H Street, NW, Washington, DC 20555, in file folder F01A-85-465 for your inspection and copying.
The staff is continuing to search for copies of the meeting sumaries i         identified at categories (a) and (b) of your letter. We will notify you as soon as the search is completed.
l                                                  Si     erelyy y       7 pA
                                                       . M. Felton, Director Division of Rules and Records
                                              /     Office of Administration l

==Enclosure:==
As stated i
8508210231 B50001 PDR     FOIA PEDR085-465        PDR I
L

Re: F01A-85-465 APPENDIX A
: 1. 6/13/84   Sumary of NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Containment Loads - May 15-17, 1984 (item c).
: 2. 10/12/84 Sumary of NRC/IDCOR Meeting on Integrated Analysis of Fission Product Behavior - August 28-29, 1984 (item d).
: 3. 12/14/83 Sumary of NRC/IDCOR Meeting on Accident Phenomenology and Containment Loading - November 29 - December 1, 1983 (item e).
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       \*...+j June 13,1984                          -
MEMORANDUM FOR: Distribution FROM:              Themis P. Speis, Director
'                              Division of Safety Technology, NRR
!                              Robert M. Bernero, Director Accident Source Term Program Office, RES

==SUBJECT:==

==SUMMARY==
OF NRC/IDCOR MEETING ON INTEGRATED ANALYSIS OF SEVERE ACCIDENT CONTAINMENT LOADS - MAY 15-17, 1984 The Industry Degraded Core Rulemaking Program (IDCOR) is an effort on the part of nuclear utilities to develop the technical basis for determining whether changes in regulatory requirements are needed to reflect severe accident con-siderations. The NRC has recognized the potential benefit of factoring the IDCOR methods and results into the agency's decision process on severe accidents.
!           A series of meetings has been arranged for NRC to examine and evaluate IDC0R's methods, assumptions and results. The purpose of this interaction is to take advantage of the technical programs and information developed by IDCOR, under-stand its bases, and identify what use we can make of it.
The first two meetings, held in Harpers Ferry, W. Va., and Hunt Valley, Md.,
concentrated on the fundamental physical and chemical processes governing accident progression, ccntainment loading, and fission product behavior. This memorandum is a summary of the third meeting, held in Rockville, Md., which dealt with integrated analyses of containment loads for a variety of plants and accident sequences. The principal technical findings are described in Enclosure 1. A meeting agenda and a list of attendees are included as Enclosures 2 and 3, respectively.
The technical presentations covered three main topics: (1)presentationof the accident sequences chosen for analysis; (2) a description of the IDCOR MAAP integral analysis code, and (3) comparison of IDCOR and NRC containment load results for several types of containments and sequences. At the end of the meeting, summary presentations were given by NRC and IDCOR contractor representatives, in which both parties outlined the principal areas of A
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agreement and disagreement for the PWR and BWR plants analyzed and discussed at the meeting. Enclosure 4 contains viewgraphs from the summary talks.
Enclosure 5 contains viewgraphs from all other technical presentations.
Although the meeting resulted in general agreement between NRC and IDCOR on the majority of technical issues, the meeting summary presentations high-lighted several areas of disagreement and several issues requiring further study. The main points of technical disagreement are briefly summarized in Enclosure 1.
The next technical excharge meeting, tentatively scheduled for late August 1984, will concentrate on integral analysis of fission product behavior. Subsequent meetings will deal with the quantification of risk from severe accidents and the assessment of possible improvements in plant design, operation, and emergency preparedness.
d       GW-                  N' Themis P. Speis, Director Divisien of Safety Technology, NRR W
f Robert M. Bernero, Director l
Accident Source Term Program Office, RES

==Enclosures:==
: 1. Technical Findings
: 2. Agenda
: 3. List of Attendees
: 4. Summary Viewgraphs
: 5. Technical Presentation Viewgraphs CONTACT:     R. Barrett x27591 C. Peabody x37691

    . en                                    ...                                ..    ..
ENCLOSURE 1

==SUMMARY==
OF THE PRINCIPAL TECHNICAL RESULTS OF THE MAY 15-17, 1984 IDCOR/NRC MEETING The meeting revealed some areas of disagreement between IDCOR and NRC and many topics that would require further analysis by both parties in order to reconcile differences of understanding. These insights were outlined in the summary statements presented on May 17 (Enclosure 4).
The discussion below briefly describes the most important areas requiring further work, namely:
Hydrogen Production Hydrogen Combustion High Pressure Melt Ejection Core-Concrete Interaction Containment Response In-Vessel Fuel Coolant Interactions In-Vessel Melt Progression Sensitivity and Uncertainty Analyses Hydrogen Production - IDCOR's MAAP model for hydrogen production includes two effects that tend to truncate all hydrogen production after the onset of core liquefact' ion; (1) the assumption that the zirconium-water interaction stops at a user-specified temperature (usually 2300*K) and (2) the assumption that melting fuel causes blockage of the coolant channels and cuts off the steam supply for further reactions. IDCOR has performed sensitivity calculations in which one or the other of these restrictions has been disabled, but no calculations have been done for the case where both are turned off.
NRC's MARCH code allows Zr-H,0 interactions to continue beyond 2300 F and incorporates the assumption that fuel melted early will enter the water in the lower plenum and boil off additional steam for hydrogen production.
The difference betweer. the two approaches is illustrated by the results of calculations with MAAP and MARCH of the TMLB' sequence for Sequoyah. In the time period prior to the onset of core slump, both codes calculated hydrogen production equivalent to oxidizing about one fourth of the zirconium. After the onset of core slump, however, calculations employing MAAP showed essentially zero hydrogen production, while calculations with MARCH showed continued                           i production at a high rate. Overall, MAAP calculated 23% zirconium oxidation j            in-vessel, while NRC consultants u:;ing MARCH calculated 49%.
Ex-vessel hydrogen production is also calculated differently by the two codes.

MAAP assumes instantaneous quench when water is present, and there is no time for hydrogen production. MARCH allows for extended quench during which signiff-cant quantities of hydrogen can be produced.
Hydrogen Combustion - The MAAP code model for combustion of hydrogen in contain-ment uses a calculated flame temperature criterion based on the concentration of steam and hydrogen. If the flame temperature exceeds the critical value of 1310 F, a global burn is assumed to take place. The model assumes an ignition source and does not account for the suppression of combustion above 60% steam concentration. By allowing combustion at low hydrogen concentration in most cases, this model precludes the higher pressures that would result from combustion initiated at higher hydrogen concentrations.
IDCOR has examined the combustion of hydrogen in the reactor cavity in some detail. The combination of high temperatures and high hydrogen concentrations caused by core-concrete interactions promotes the burning of hydrogen with oxygen drawn into the cavity by natural convection currents. In some of our work as well as work performed by NRC's consultants, steam inerting or oxygen starvation precluded combustion in the cavity. This is a phenomenon that NRC needs to evaluate more closely.
Finally, the point was made that both IDCOR and the Containment Loads Working Group (CLWG) have ignored the potential for detonable hydrogen mixtures near the outlet of ice condenser cabinets.
High Pressure Melt Ejection - In high-pressure sequences, it is likely that portions of the molten core will be expelled from the reactor cavity. In a worst case scenario, a large fraction of the molten core could be suspended as an aerosol in containment, directly exchanging heat witn the atmosphere and chemically reacting with oxygen. NRC staff and consultants are divided on the question of the likelihood and severity of such an event. IDCOR does not consider this worst-case scenario credible.
In the IDCOR calculation of TMLB for Zion, the high ejection pressure pushes half of the core out of the cavity where it spreads out evenly on the floor of the steam generator room without significant energy transfer to the containment atmosphere. The resulting thin layer never gets hot enough to attack the concrete floor of the SG room.
The NRC Containment Loads Working Group continues to work toward a technical resolution of the question of direct heating of the containment atmosphere includ-ing the performance of true prototypic experiments.
A related issue under review by NRC is whether the reactor coolant system will fail prior to vessel lower head failure. Neither NRC nor IDCOR has developed

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a position on the likelihood that this phenomenon will preclude high-pressure melt ejection, and the problem is under study by both groups.
Core-Concrete Interaction - For several eouences, there was disagreement between IDCOR and the NRC contractors on the partitioning of energy during core-concrete interactions. In general, the NRC calculation showed greater energy transfer to concrete structure. The latter results produced more noncondensible gases and higher containment pressures. These differences are due to modeling assumptions and initial conditions (such as the temperature and thickness of the melt).
Containment Response'- Although containment response was not a subject of this meeting, there was a great deal of discussion of differences between NRC and IDCOR on this question. The NRC consultants suggested that IDCOR consider the possibility of containment leakage before containment failure, (i.e., threshold type of failure) especially for MARK I containment. It was also pointed out that liner burn-through in the MARK I and pedestal failure followed by reactor vessel collapse in the MARK III were failure mechanisms that should be considered by IDCOR.
In-Vessel Fuel Coolant Interactions - Although not discussed in depth at this meeting, fuel coolant interactions (FCI) continue to be a source of uncertainty with respect to accident progression, steam and hydrogen production and fission product behavior. Molten fuel entering the lower plenum can interact with the remaining water in a number of ways ranging from stable boiling to steam explosions.
The IDCOR methodology essentially precludes energetic fuel coolant interaction in-vessel. This is a major point of technical disagreement. The NRC MARCH code allows for energetic fuel coolant interactions, but the phenomenon has not been included in the CLWG standard problems. An assessment of this phenomenon and its impact on accident progression is needed.
In-Vessel Melt Progression - It is becoming increasingly apparent that the mode and timing of containment failure is sensitive to the details of in-vessel melt progression. For instance, differences tietween IDCOR and NRC approaches to modeling fuel relocation lead to large differences in hydrogen production. More effort by both parties would be needed to better characterize in-vessel melt phenomena and to reflect those phenomena in the codes in order to resolve these differences.
Sensitivity and Uncertainty Analyses - Although PCOR has performed several studies of the sensitivity of containment loads to individual parameters, NRC believes it is necessary to evaluate the effect of simultaneously varying multiple parameters.
An example of this is the sensitivity of hydrogen production to the assumptions that (1) the Zr-water interaction is cut off at 2300 K and (2) fuel melt will block steam flow channels and inhibit Zr-water reactions. To realistically evaluate

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the effect of these assumptions, both restrictions should be simultaneously relaxed. Future sensitivity studies should cover such synergistic effects.
Furthermore, the effects of modeling assumptions as well as parameter choices should be explored more completely. It is our understanding that such studies are currently underway at IDCOR and will be presented in the future.
In addition to sensitivity analyses, there was agreement that uncertainty analyses should be performed.      In sensitivity analyses the independent variables can be varied over arbitrary ranges. In uncertainty analyses, an effort is first made to determine the ranges over which the independent variables can reasonably be expected to vary. The resulting variations in dependent parameters then represent an estimate of the real uncertainty.

DISTRIBUTION gh#
NRC/NRR                    NRC/RES            ACRS H. R. Denton               R. Minogue         R. Tripathi E. Case                    0. Bassett         S. Seth R. Mattson                 G. Arlotto         R. Cushman R. Houston                 R. Curtis J. Hulman                  T. Walker R. Vollmer               y G. Marino Z. Rosztoczy               G. Marino J. Rosenthal               W. Morrison J. Shapaker              y J. Han C. Tinkler                 R. Wright R. Palla                   M. Cunningham W. Lyon                    M. Silberberg      TEC A. El-Bassioni             T. Lee
:.R. L Ba rrettr'            C. Peabody         A. Buhl P. Niyogi          M. Fontana T. Eng             E. Fuller Other NRC                  B. Agrawal         J. Carter R. Meyer           H. Mitchell J. Conran, DEDROGR         B. Burson          S. Asselin M. Taylor, DEDR0GR          J. Larkins         K. Meyer J. Telford P. Baranowsky J. Glynn EPRI DOE FAI                M. Leverett F. Witmer                                     R. Sehgal H. Fauske          D. Squarer R. Henry J. Gabor Bechtel                     M. Kenton          NUS S. Blazo                                       P. O'Reilly P. Fulford Siegel, AIF                                              l f
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Sandia Nat'l Lab. EGG-Idaho D. Moore, EI                  D. Dahlgren       S. Behling J. Young, El                  M. Berman         R. Gottula S. Thompson M. Lloyd, Middle South        R. Cole Services           J. McGlaun        Battelle Columbus D. Paddleford, W              J. Hickman P. Nakayama, Jaycor           K. Bergeron       P. Cybulskis L. Azarello, Duke Power       D. Kunsman        R. Denning W. Mims, TVA                  D. Aldrich M. Cosella, Coned             J. Sprung         Oak Ridge Nat'l Lab J. Meincke, CPC0              J. Walker W. Iyer, NYPA                 J. Griesmeyer     S. Hodge A. Marie, PEC0                F. Harper G. Krueger, PECO              D. Powers R. Smith, Scandpower          V. Behr L. Engstrom, OKGAB Sweden     J. Linebarger C. Ader, Stone & Webster      S. Dingman L. Rib, LNR Associates        A. Camp J. Metcalf, Stone & Webster   A. Benjamin Brookhaven Nat'l Lab.
W. T. Pratt M. Khatib-Rahbar R. Newton T. Ginsberg G. Greene M. Corradini, Univ. of Wisc.
I. Spiewak, American Physical Society S. Niemczyk, UCS T. Theofanous, Purdue University

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ENCLOSURE 2 AGENDA NRC/IDCOR MEETING J
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INTEGRATED ANALYSIS OF SEVERE ACCIDENT CONTAINMENT LOADS 1
l                                            MAY 15-17, 1984 Tuesday, May 15 Introduction 1
Welcome, Purpose, Ground Rules, Schedule - T. Speis, R. Bernero        l i
NRC Introduction - M. Fontana, E. Fuller - IDCOR l             Accident Sequence Plant Categorization & Sequence Selection - J. Young, EI Break Plant Categorization & Sequence Selection - A. Eng, NRC F. Harper, SNL Lunch Overview of MAAP l             -
Introduction to MAAP - R. Henry, J. Fabor, M. Kenton, - FAI Lunch Verification of MAAP - P. Nakayama, Jaycor l             -
Break PWR Large Dry Containment IDCOR Methods & Results - M. Kenton, FAI Break Overview of CLWG & Standard Problems - M. Silberberg, NRC CLWG Methods & Results - T. Theofanous, Purdue Univ.

l Tuesday, May 15 (Cont'd) l 1
Direct Heating - T. Ginsberg, BNL Likelihood of Hig1 Pressure Scenarios - W. Lyon, NRC Adjourn Wednesday, May 16 PWR Ice Condenser IDCOR Methods & Results - M. Kenton CLWG Methods & Results - P. Cybulskis, BCL Break BWR Mark I & Mark II IDCOR Methods & Results for Mark I - J. Gabor Lunch
         -    CLWG Methods & Results for Mark I & Mark II - W. T. Pratt, BNL Break BWR Mark III
         -    CLWG Methods & Results - A. Camp, SNL 10COR Methods & Results - J. Gabor Adjourn Thursday, May 17 Summaries
          -    PWR Containment Loads - M. Berman, SNL & E. Fuller, TEC
          -    BWR Containment Loads - S. Hodge, ORNL & E. Fuller, TEC
          -    Status of Issue Papers Related to Accident Phenomenology and Containment Loads (Issues 1.1 and 1.2) - R. Bernero, NRC Parting Remarks - T. Speis, R. Bernero, M. Fontana Adjourn

          . ,L i
ENCLOSURE 3 ATTENDANCE I
NRC/NRR               NRC/RES            ACRS T. Speis              R. Bernero         R. Tripathi Z. Rosztoczy          T. Walker          S. Seth J. Rosenthal          G. Marino          R. Cushman J. Shapaker           W. Morrison
!              C. Tinkler            J. Han R. Palla              R. Wright W. Lyon               M. Cunningham      TEC         l A. El-Bassioni        M. Silberberg R. Barrett            T. Lee             A. Buhl C. Peabody         M. Fontana P. Niyogi          E. Fuller Other NRC             T. Eng             J. Carter B. Agrawal         H. Mitchell (
f             J. Conran, DEDROGR    R. Meyer           S. Asselin M. Taylor, DEDROGR    B. Burson          K. Meyer J. Larkins J. Telford P. Baranowsky J. Glynn           EPRI DOE
;                                    FAI                M. Leverett F. Witmer                                R. Sehgal H. Fauske          D. Squarer R. Henry J. Gabor Bechtel               M. Kenton          NUS S. Blazo                                 P. O'Reilly P. Fulford

7
* Industry                               Sandia Nat'l Lab.          EGG-Idaho D. Moore, El                           D. Dahlgren                S. Behling                 '
J. Young, EI                           M. Berman                  R. Gottula S. Thompson M. Lloyd, Middle South                 R. Cole Services                    J. McGlaun                 Battelle Columbus D. Paddleford, W                       J. Hickman P.Nakayama,Jaicor                      K. Bergeron                P. Cybulskis L. Azarello, Duke Power                D. Kunsman                 R. Denning W. Mims, TVA                           D. Aldrich M. Cosella, Coned                      J. Sprung                  Oak Ridge Nat'l Lab J. Meincke, CPC0                       J. Walker W. Iyer, NYPA                          J. Griesmeyer              S. Hodge A. Marie, PECO                         F. Harper G. Krueger, PECO                       D. Powers                  Brookhaven Nat'l Lab.
R. Smith, Scandpower                   V. Behr L. Engstrom, OKGAB Sweden              J. Linebarger              W. T. Pratt C. Ader, Stone & Webster               S. Dingman                 M. Khatib-Rahbar          j L. Rib, LNR Associates                A. Camp                    R. Newton J. Metcalf, Stone & Webster            A. Benjamin                T. Ginsberg G. Green M. Corradini, Univ. of Wisc.
I. Spiewak, American Physical Society S. Niemczyk, UCS                                                                            !
T. Theofanous, Purdue University

.    '3 -
ENCLOSURE 4

==SUMMARY==
PRESENTATIONS PWR Containment Loads - M. Berman, SNL PWR Containment Loads - E. Fuller, TEC BWR Containment Loads - S. Hodge, ORNL BWR Containment Loads - E. Fuller, TEC

7.
t REPORT OF THE PWR GROUP K. BERGERON M. CORRADINI P. CYBULSKIS M. BERMAN D. POWERS                             ,
E. HASKIN                             {
T. THE0 FAN 00S                       i J. TELFORD R. WRIGHT J. ROSENTHAL V. BEHR, T. GINSBERG, Et. al.
GROUNDRULES ADDRESS CONTAINMENT LOADING ONLY AND NO SOURCE TERM ASPECTS:
TIMING & FAILURE MODE WILL NOT DISCUSS W-FAILURE MODE (COVERED AT HARPERS FERRY)
GENERAL AGREE ON MANY THINGS ALMOST T00 NUMEROUS TO MENTION WE EMPHASIZE " UNCERTAINTY" VS " SENSITIVITY" ANALYSES (RANGES OF BEHAVIOR, MODEL UNCERTAINTY, SYNERGISM) 1 1

AREAS OF DISAGREEMENT AND FURTHER WORK I. IN-VESSEL MELT PROGRESSION A.                   MATERIAL MOTION (IN-CORE MELTING & RELOCATION PATTERN, IN-PLENUM FCIs)
B.                    FISSION PRODUCT RELOCATION (THERMAL EFFECTS)
II. HYDROGEN A.                    PRODUCTION (IN-CORE ZR OXIDATION, RADIATION DOWN, EX-VESSEL PROLONGED CONCRETE ATTACK, FCIs, CONCRETE DEGASING IN RC)
B.                    COMBUSTION (FLAME T CRITERIA, IGNITION, LOCAL INERTING AT IGNITERS, H, FLOW PATH TO AND FROM RC, MIXING WITH 0XIDANT AT tARGE SCALE, LOCAL DETONATION)
III. DIRECT HEATING A.                     CHEMICAL REACTION (AVAILABILITY OF REACTANTS)
: 8.                     SEPARATION PHENOMENA (WATER IN CAVITY, GE0 METRY EFFECTS, CONCRETE PARTICIPATION)
C.                      INTEGRITY OF RPV BOUNDARY PRIOR TO CORE MELT (RE-CIRCULATION PHENOMENA, CORE BLOCKAGE)

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PWR 

==SUMMARY==
RESULTS E. FULLER
)
TEC 1
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FOCUS IS ON ACCIDENT SEQUENCES SELECTED FOR REFERENCE PLANTS AND ON THE MODULAR ACCIDENT ANALYSIS PROGRAM (MAAP)
WILL DESCRIBE PROCESS OF SELECTING KEY AND DOMINANT SEQUENCES WILL DESCRIBE THE MAAP-PWR AND MAAP-BWR CODES WILL DISCUSS PROCESS USED FOR MAAP VERIFICATION AND VALIDATION WILL PRESENT SOME RESULTS OBTAINED WITH CODES SUPPORTING MAAP ON TOPICS IDENTIFIED AT HARPERS FERRY WILL ILLUSTRATE USE OF MAAP PRESENTING ANALYSIS OF ONE SEQUENCE FOR EACH REFERENCE PLANT

t IN-VESSEL HYDROGEN GENERATION IS TREATED IN DIFFERENT MANNERS BY IDCOR AND NRC CONTRACTORS IDCOR SHOWS THAT RATES AND AMOUNTS OF HYDROGEN GENERATED ARE SEQUENCE DEPENDENT WHEN CALCULATED WITH MAAP, AND ARE INFLUENCED BY CORE BLOCKAGES MARCH ASSUMES THAT HIGH-TEMPERATURE MATERIALS ARE STILL AVAILABLE FOR OXIDATION WHEN SLUMPING TAKES PLACE.                     SIGNIFICANT AMOUNTS OF HYDR 0 GEN ARE CALCULATED TO BE GENERATED DURING SLUMPING i
i

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NATURAL CIRCULATION IN PWR PRIMARY SYSTEMS IS MORE IMPORTANT THAN PREVIOUSLY THOUGHT BOTH IDCOR AND NRC ARE BEGINNING TO INVESTIGATE NATURAL CIRCULATION EFFECTS PATTERNS CAN DEVELOP PRIOR TO CORE DEGRADATION FLOW DIVERSIONS FROM GE0 METRIC DISTORTIONS PROBABLY ARE SIGNIFICANT WHATEVER CONCLUSIONS ARE DRAWN MUST BE CONSISTENT WITH OBSERVATIONS MADE FROM THE TMI ACCIDENT

k, IDCOR HAS CARRIED OUT SIGNIFICANT UNCERTAINTY AND SENSITIVITY ANALYSES l
* SMALL VARIATIONS IN KEQ QUANTITITES OF INTEREST TO CHANGES IN KEY PARAMETERS GENERALLY SEEN MORE THAN ONE KEY PARAMETER MAY NEED TO BE VARIED AT A TIME, BASED ON JUDGMENT NATURAL CIRCULATION IS IMPORTANT     l IN CONTAINMENT BUILDINGS         l BOTH MAAP/PWR AND MAAP/BWR HAVE NATURAL CIRCULATION MODELS BOTH WITHIN AND BETWEEN COMPARTMENTS l
l

DIRECT HEATING OF CONTAINMENT IS TREATED BY IDCOR BY USING MAAP IS NOT SIGNIFICANT IN THE PWR REFERENCE PLANTS
* INFLUENCES ARE FROM CAVITY CONFIGURATIONS AND FRDM AVAILABILITY OF WATER IS SIGNIFICANT IN PEACH BOTTOM BECAUSE UPWARD RADIATIVE HEAT TRANSFER CAN LEAD TO HIGH DRYWELL TEMPERATURES
* BNL ASSUMES UPWARD RADIATIVE HEAT TRANSFER TO BE SMALL, WHEREAS SANDIA CALCULATE" IT DIRECTLY
* b 3
CONSIDERATIONS REGARDING ANALYSES OF BWR SEVERE ACCIDENT CONTAINMENT LOADS R. BARRETT                                              E. HASKIN A. CAMP                                                 S. H0DGE M. CUNNINGHAM                                           T. PRATT P. CYBULSKIS                                            J. ROSENTHAL G. GREEN                                                J. TELFORD PRESENTATION AT NRC/IDCOR MEETING 0,.3 INTEGRATED ANALYSIS OF SEVERE ACCIDENT CONTAINMENT LOADS ROCKVILLE, MARYLAND MAY 17, 1984

I                                                  .
COMMENTS WILL BE PRESENTED IN FOUR CATEGORIES          -
GENERAL OBSERVATIONS SIGNIFICANT AREAS OF AGREEMENT WITH IDCOR APPROACH SIGNIFICANT AREAS OF DISAGREENENT WITH IDCOR APPROACH AREAS REC 0 MENDED FOR FUTURE DISCUSSION

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GENERAL OBSERVATIONS PROGRESSIVE RELOCATION OF THE CORE IS BELIEVED TO BE THE MORE REALISTIC APPROACH FOR BWR ANALYSES PARAMETRIC VARIATION OF COMPLETE FAILURE CORE PLATE LEAKAGE BEFORE FAILURE OVERPRESSURE FAILURE OF CONTAINMENT WILL NOT ALWAYS RESULT IN LOSS OF INJECTION.                                     POSSIBILITIES ARE NO EFFECT ON INJECTION LOSS OF INJECTION LOCA LOCA AND LOSS OF INJECTION PRESSURE SUPPRESSION POOL CONSIDERATIONS ARE IMPORTANT THERMAL STRATIFICATION CON 0ENSATION OSCILLATION AND CHUGGING SPEED OF PASSAGE OF CORIUM THROUGH LOWER PLENUM IS SUBJECT TO DEBATE

p- -
I                                                                 'd e SIGNIFICANT AREAS OF AGREEMENT WITH IDCOR APPROACH CONTROL R0D DRIVE HYDRAULIC SYSTEM (CROHS) CAN PREVENT CORE PLATE FAILURE IN SOME CASES EFFECT OF CRDHS FLOW AFTER REACTOR VESSEL FAILURE SHOULD BE CONSIDERED IN APPROPRIATE CASES PRESSURIZED BLOWO('WNS NEED TO BE CONSIDERED FOR BWRs MARK I DRYWELL INTEGRITY IS THREATENED BY HIGH TEMPERATURES AFTER REACTOR VESSEL FAILURE                ,

SIGNIFICANT AREAS OF DISAGREEMENT WITH 10COR APPROACH EXTENT OF H GENERATION IN CORE 2
EFFECT OF CHANNEL BLOCKAGE AND HIGH TEMPERATURE CUTOFF EFFECT OF CONTINUED STEAM SUPPLY FROM LOWER PLENUM CORIUM LEAKAGE THROUGH CORE PLATE RADIATION FROM BOTTOM OF CORE OXIDATION OF STAINLESS STEEL DESPOSITION OF CORIUM EXITING VESSEL DRYWELL SUMPS (MARK I)
DIRECT HEATING FOR CASES WITH PRESSURIZED EXIT l
FOR MARK III CONTAINMENT AT HIGH TEMPERATURES MUST CONSIDER SEAL FAILURE IN ORYWELL WALL DEGASSING FROM UNLINED CONCRETE

AREAS OF DISAGREEMENT (Cont'd)
CLWG SEES LARGE SENSITIVITY OF CALCULATED TENPERATURES AND PRESSURES IN MARK I CONTAINMENTS OUE TO MODELING ASSUMPTIONS INITIAL CONDITIONS PLANT SPECIFIC CONSIDERATIONS 10COR RESULTS IN GENERAL SHOW HIGHER TEMPERATURES LOWER PRESSURES GENERAL RESULT IS MORE HEATING OF ATN0 SPHERE LESS HEATING OF CONCRETE THE CORE-CONCRETE INTERACTION MODELS REMAIN OPEN TO QUESTION REGARDING:
GENERATION OF NONCONDENSIBLES i
RA0!ATION RECEPTORS FROM CORIUM EFFECT OF AEROSOLS l

e ,
AREAS RECOMMENDED FOR FUTURE DISCUSSION MARK I CONTAINMENT LEAKAGE BEFORE FAILURE (NUREG-1037)
LINER BURN-THROUGH AS MARK I DRYWELL FAILURE MODE DIRECT HEATING OF THE CONTAINMENT ATMOSPHERE POSSIBILITY OF PEDESTAL FAILURE COMBUSTION MODELS DIFFUSION FLAMES C0 CONSIDERATIONS FLAME TEMEPRATURE CRITERIA IGNITOR MODELING CAPACITY OF AEROSOLS TO PLUG LEAKS IN MARK III DRYWELL WALL SOURCE OF LEAKS SIZE OF LEAKS 1

puumi iI FINAL REMARKS ALL COMENTS CONCERNING THE MARK III CONSIDERATIONS ARE BASED ON REVIEW OF THE TQUV WITH ADS SCENARIO ONLY A DIRECT COMPARISON OF IDCOR AND MARCH-BASED RESULTS FOR AN IDENTICAL PROBLEM WOULD BE HIGHLY DESIRABLE.
i

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BWR 

==SUMMARY==
RESULTS E. FULLER TEC 1
l l
l

L RAMIFICATION OF EXTENSIVE CORE DEBRIS - CONCRETE ATTACK IN GRAND GULF PEDESTAL SHOULO BE CONSIDERED PEDESTAL FAILURE AND VESSEL COLLAPSE ALTERNATE FISSION PRODUCT TRANSPORT PATHS FISSION PRODUCT DISTRIBUTION AT THE TIME OF VESSEL MOVEMENT

i TWO ADDITIONAL ASPECTS IN PEACH BOTTOM ANALYSES THAT WERE IDENTIFIED WILL BE CONSIDERED BY IDCOR THE INFLUENCE OF THE SUMP BELOW THE PEDESTAL LINER FAILURE MAY BE A CONTAINMENT FAILURE MODE
           -     MAY NOT REALLY BE A FAILURE MODE, SINCE A GLASSY SEAL MAY FORM WILL BE INVESTIGATED FROM THE STANDPOINT OF FISSION PRODUCT RELEASE l
I l
f

NRC STATES THAT EFFECTS OF IGNITERS NOT WORKING IN T yQUV IN GRAND GULF SHOULD BE ADDRESSED MUST DETERMINE POTENTIAL AND CONSEQUENCES OF GLOBAL BURNS MUCH TIME IS AVAILABLE TO RECOVER POWER PRIOR TO REACHING 8% LIMIT NEW EPRI DATA ON PROBABILITY OF STATION BLACK 0UT MUCH LESS THAN CONSIDERED IN THE GRAND GULF RSSMAP

ENCLOSURE 5 PRESENTATIONS OF TECHNICAL RESULTS I.      Overview of Meeting Topics - E. Fuller, TEC II. Selection of Dominant Core Damage & Fission Product Release Sequences - J. Young, EI III. NRC Plant Categorization & Sequence Selection - T. Eng, NRC F. Harper, SNL IV. MAAP Introduction - R. Henry, J. Gabor, M. Kenton - FAI V.      MAAP Verification - P. Nakayama, Jaycor VI. Zion Station: Integrated Accident Analysis - M. Kenton, FAI VII. Overview of Containment Loads Working Group - M. Silberberg, NRC VIII. Short-Term Containment Loading in Zion-Type Large Drys -
T. Theofanous, Purdue Univ.
IX. High Pressure Melt Ejection Containment Loading - T. Ginsberg, BNL     ,,
X.      Likelihood of High Pressure Scenarios - W. Lyon, NRC XI. Sequoyah Nuclear Plant: Integrated Accident Analysis - M. Kenton, FAI XII. Ice Condenser PWR Standard Problem - P. Cybulskis, BCL XIII. Peach Bottom Atomic Power Station: Integrated Containment Analysis -
J. Gabor, FAI XIV. BWR Mark I and Mark II Standard Problems - W. T. Pratt, BNL XV. CLWG Standard Problems: BWR Mark III Diffusion Flames - A. Camp, SNL XVI. Grand Gulf Nuclear Station: Integrated Containment Analysis - J. Gabor, FAI

I   l OVERVIEW OF MEETING TOPICS EDW ARD L. FULLER                  l EPRI AT TEC IDCOR/NRC MEETING ON INTEGRATED ANALYSIS OF SEVERE 1
ACCIDENT CONTAINMENT LOADS               l ROCKVILLE, MARYLAND MAY 15-17,1984 31221A34 l
l

FOCUS IS ON ACCIDENT SEQUENCES SELECTED FOR REFERENCE PLANTS AND ON THE MODULAR ACCIDENT ANALYSIS PROGRAM (MAAP) e WILL DESCRIBE PROCESS OF SELECTING KEY AND DOMINANT SEQUENCES e WILL DESCRIBE THE MAAP-PWR AND M A AP-BWR CODES
* WILL DISCUSS PROCESS USED FOR MAAP VERIFICATION AND VALID ATION l
l e WILL PRESENT SOME RESULTS OBTAINED WITH CODES SUPPORTING MAAP ON TOPICS IDENTIFIED AT HARPERS FERRY
* WILL ILLUSTRATE USE OF MAAP PRESENTING AN.ALYSIS OF ONE SEQUENCE FOR EACH REFERENCE PLANT l
31221A33 l

SEQUENCE SELECTION PROCESS i
;               e ORIGINATED IN PROBABILISTIC RISK i                 ASSESSMENTS FOR THE FOUR i                  REFERENCE PLANTS e FUNCTIONAL CLASSIFICATION PROCESS TO GROUP SEQUENCES 4
e DOMINANT AND KEY SEQUENCES SELECTED 4
31221A35 i
f 1

THE MAAP CODE WAS DEVELOPED FOR SEVERE ACCIDENT ANALYSIS l
e KEY ACCIDENT SEQUENCES ANALYZED FOR REFERENCE PLANTS e SEPARATE SYSTEM MODELS FOR BWRs AND PWRs e PHENOMENOLOGY FROM VARIOUS IDCOR TASKS INCORPORATED e EVENT CODE LOGIC ENABLES MANY OPERATOR ACTIONS TO BE EVALUATED e UN. CERTAINTY ANALYSIS FEATURE ALLOWS FOR EVALUATION OF KEY PARAMETER VARIATIONS l
31221A38

MAAP VALIDATION AND VERIFICATION TASK PROVIDED AN INDEPENDENT CHECK OF THE MODELS AND CODING STRUCTURE e MODELS INDIVIDUALLY ASSESSED e INDEPENDENT FLOW DIAGRAMS PREPARED e LINE-BY-LINE VERIFICATION CARRIED OUT e APPARENT DIFFERENCES DOCUMENTED AND RECONCILED l
l 31221A37

SEVERAL iMPORTANT AREAS IDENTIFIED AT HARPERS FERRY WERE INVESTIGATED e STEAM AND HYDROGEN GENERATION FROM SRV CYCLING e WATER ABSORPTIVITY e DOWNWARD VIEW FACTORS i
* BUOYANCY-DRIVEN CIRCULATION 1
l e DEBRIS QUENCHING MODEL IN MAAP e UNCERTAINTIES IN KEY MAAP MODELS 31221A38 1

USE OF MAAP FOR ANALYSIS OF EACH REFERENCE PLANT WILL BE ILLUSTRATED eTMLB          I FOR ZION eS    2 HF      FOR SEQUOYAH eS    1 E FOR PEACH BOTTOM e T QUV 3
FOR GRAND GULF e EACH SEQUENCE WILL BE USED TO POINT OUT BEST-ESTIMATE BEHAVIOR AND UNCERTAINTIES                             i IN KEY MAAP MODELS 31221A39
* 6 5
SELECTION OF DOMINANT CORE DAMAGE AND FISSION PRODUCT RELEASE SEQUENCES JON YOUNG IDCOR/NRC MEETING ON INTEGRATED ANALYSIS OF SEVERE ACCIDENT CONTAINMENT LOADS ROCKVILLE, MARYLAND MAY 15-17, 19811 l

l ORIGINAL PRAS                                                           IDENTIFY DOMINANT FOR                                                            ;    SEQUENCES REFERENCE PLANTS                                                              -CORE DAMAGE l                                                                                 FREQUENCY l
V l
SELECT INITIAL KEY SEQUENCES m       -CORE DAMAGE l                                                                                  FREQUENCY l                                                                        3                                  RISK PROFILE
                                                                                 -FUNCTIONAL FAILURES
                                                                                 -RELEASE CATEGORY SPECTRUM DESCRIPTIONS OF FUNCTIONAL FAILURE PATTERNS FROM                                                                           V 14 PRAS ACCIDENT PROCESS AND FISSION PRODUCT TRANSPORT TASKS A
V UPDATE KEY SEQUENCES SELECTION OF DOMINANT AND KEY SEQUENCES               .

e       ,
I PROBABILISTIC RISK ASSESSMEtlT STUDIES SURVEYED FOR IDCOR                         i 1
PLANT                               TYPE OF STUDY        PLANT TYPE PWR ARKANSAS NUCLEAR ONE-1           IREP                 B8W CALVERT CLIFFS                  RSSMAP               CE CRYSTAL RIVER-3                 IREP                 B&W INDIAN POINT-2                  UTILITY-SPONSORED
* W INDIAN POINT-3                  UTILITY-SPONSORED
* W          ,
OCONEE                          RSSMAP               B&W        l SEQUOYAH                        RSSMAP               W SURRY                           REACTOR SAFETY STUDY W ZION                            UTILITY-SPONSORED
* W BWR BROWNS FERRY-1                  IREP                 GE GRAND GULF                      RSSMAP               GE LIMERICK                        UTILITY-SPONSORED    GE MILLSTONE POINT-1               IREP                 GE PEACH BOTTOM                    REACTOR SAFETY STUDY GE I
l
* DETAILED EXTERNAL EVENTS ANALYSIS INCLUDED.
l l
I

FUNCTIONAL FAILURE CLASSIFICATION FUNCTIONAL FAILURE                                         ABBREVIATION REACTOR INTEGRITY                                       RI CORE INVENTORY MAKEUP                                 CMO CORE HEAT REMOVAL                                     CHO CONTAINMENT PRESSURE SUPPRESSION                      CNP CONTAINMENT HEAT REMOVAL                              CHN CONTAINMENT INTEGRITY                                 CNI CONTAINMENT RADIOACTIVITY REMOVAL                     CNR NOTE:   A SUBSCRIPT "L" INDICATES A LATE FUNCTIONAL FAILURE.
l l
l l.._                                                              . _ _ _ _ _       .

O    8 PWR Key Plant Features By Functional Group Reactor Integrity (RI)
: 1. Reoctor Pressure Vessel
: 2. Reactor coolant pressure boundary
: 3. Safety / relief volves Core Makeup (CoM)
CoM-Early                                              CoM-Late
: 1. Accumulators and upper head                      I. High pressure recirculation system injection system
: 2. High pressure injection system                   2. Charging system
: 3. Chorging system                                   3. Low pressure recirculation system
: 4. Low pressure injection system S. Internal vessel vent volves Core Heat Removal (CoH)
:      CoH-Early                                           CoH-Late
: 1. Reactivity control systems                      1. Residual heat removal system
: o. Control rods                                o. High pressure recirculation system
: b. Boron injection system                          b. Low pressure recirculation system
: c. RHR heat exchangers
: d. Cooling water systems for RHR heat exchangers
: 2. PORV, code safety volves and                   2. PORV, code safety volves and high point high point vents                                  vents
: 3. Steam system                                   3. Steam system                               I
: o. Steam generators                                o. Steam generators
: b. Feedwdter systems                               b. Feedwater systems
: c. Power conversion system                         c. Power conversion system
: d. Secondary system steam relief                   d. Secondary system steam relief
: 4. Charging / HPI and letdown                     4. Charging / HPI and letdown system system (feed and bleed                            (feed and bleed capability) capability)

DOMINANT SEQUENCE DESCRIPTION SHEET Plant Name                                                 Type PRA Study Sequence Name                                          Frequency Functional Failures H        I E   L  EOL      EOL     Ent    EnL     EnL      Enk Initiating Event Description Initial System Failures (By System Name)
Dependent System Failures (By System Name)
DOMINANT SEQUENCE DESCRIPTION SHEET

Sequence:
Plant:
Accident Processes Core Behavior Containment Response Additional Phenomenological Considerations i
'l R_elease Categories 1
     -n -m,   e  ,             -- --,- - - -   - - - - - . - , .. , , - ,    ,,~ . - - - - - - - -    ---n-      --- + .- . .- - - - - - --- w, - - -- , - --- - ---

i DOMINANT FUNCTIONAL FAILURE CATEGORIES IN CORE DAMAGE SEQUENCES PWR FUNCTIONAL                                                                           AVERAGED CONTRIBUTION FAILURE                                                                              TO TOTAL CORE CATEGORY                                 REPRESENTATIVE SEQUENCE                     DAMAGE FREQUENCY RI,C OML                                 LOCA WITH FAILURE OF RECIRCULATION (S2H)            32%
COM,COH,CnP,CnH,CNR                      STATION BLACKOUT (TMLB')                            20%
CH O                                       LOSS OF FEEDWATER (TML)                             13%
RI,C OM                                                                                      12%
LOCA WITH FAILURE OF INJECTION (S2D)
COM,COH                                  LOSS OF FEEDWATER WITH FAILURE OF FEED                9%
AND BLEEDJ AND ATWS ALL OTHERS                               (11 CATEGORIES)                                     lll%
e
                                                                                                                                =
O

DOMINANT FUNCTIONAL FAILURE CATEGORIES                                    '
IN CORE DAMAGE SEQUENCES BWR FUNCTIONAL                                                                                                        AVERAGED (ONTRIBUTION FAILURE                                                                                                           TO TOTAL LORE CATEGORY                                                       REPRESENTATIVE SEQUENCE                            DAMAGE rREQUENCY CnHe                                                           TRANSIENT WITH FAILURE OF LONG-TERM                          35%
HEAT REMOVAL (TW)
COM,CoH                                                        ATWS (TC)                                                    2 11 %
i
! CoM,CoH,CnP,CN H,CNR                                           STATION BLACKOUT                                             11%
; CMO                                                            TRANSIENTS WITH FAILURE OF MAKEUP (TQUV)                     10%
RI,C HNL                                                       LOCA WITH FAILURE OF LONG-TERM HEAT                            9%
!                                                                REMOVAL (SI) l
) RI,C OM                                                        LOCA WITH FAILURE OF INJECTION (SE)                            6%
l I ALL OTHERS                                                     (6 CATEGORIES)                                                 5%
1 l

OCCURRENCE OF FUNCTIONAL FAILURES IN CORE DAMAGE SEQUENCES AVERAGED OCCURRENCE i
FREQUENCY IN FUNCTION FAILURE                                                     CORE DAMAGE SEQUENCES PWR REACTOR INTEGRITY, EARLY (RI)                                             57%
CORE INVENTORY MAKEUP, EARLY (COM)                                        50%
CORE HEAT REMOVAL, EARLY (CO H)                                           44%
CORE INVENTORY MAKEUP, LATE (COLM)                                        30%
BWR CORE INVENTORY MAKEUP, EARLY (COM)                                        56%
l                                                                                                   CONTAINMENT HEAT REMOVAL, LATE (C           H)
NL                             43%
CORE HEAT REMOVAL, EARLY (COH)                                            39%
b e
e

l IDCOR REFERENCE PLANTS j             PLANT (PRA)                   TYPE                         CONTAINMENT TYPE SEQUOYAH (RSSMAP)             PWR - WESTINGHOUSE           ICE CONDENSER ZION (PSS)                    PWR - WESTINGHOUSE           LARGE DRY l
PEACH BOTTOM (RSS)            BWR - GENERAL ELECTRIC       MARK I GRAND 6ULF (RSSMAP)           BWR - GENERAL ELECTRIC       MARK III l
l
)l I

DOMINANT CORE DAMA6E SEQUENCES - SEQUOYAH (RSSMAP)
CORE DAMAGE SEQUENCE           FREQUENCY              FUNCTIONAL FAILURE S2H                2E-05                  RI,CoMt S1H                IE-05                  RI,CoMt S2D                6E-06                  RI,CoM S2HF               SE-06                  RI,CoMt ,CnPt ,CnHt ,CnRt
;  V                  SE-06                  RI,CoM,Cul
;  SID                fiE-06                 RI,CoM S1HF               3E-06                  RI,CoMt ,CNt P ,CnHt,CnRt T23M_              3E-06                  CoH AH                 SE-07                  RI,CoMt l  TIB 3M_B'13        3E-07                  CoM,CoH,CnP,CnH,CnR AD                 2E-07                  RI,CoM i  AHF                9E-08                  RI,CoML,CnPt,CnHt,CnRL TOTAL FREQUENCY    5.7E-05 k
4

DONINANT CORE DAMAGE SEQUENCES - ZION (PSS, REY. 0)
CORE DAMAGE SEQUENCE           FREQUENCY            FUNCTIONAL FAILURE 1                  2E-05                RI,CoMt 2 (TMLB')          6E-06                 CoM,CoH,CnP,CnH,CnR 3                  5E-06                 RI,CoMt 4                   SE-06                RI,CoMt 5                   4E-06                CoH 6                   3E-06                CoH 7                   2E-06                RI,CoHL 8                   1E-06                RI,CoMt,CnHe 9                   1E-06                RI,CoM 10                  4E-07                RI,CoM 11                  3E-07                CoM,CoH 12                 2E-07                CoM,CoH 13                 2E-07                CoM,CoH 14 (TMLB')         2E-07                CoM,CoH,CnP,CnH,CnR 15                 2E-07                CoM,CoH 16                 1E-07                RI,CoM,CnI TOTAL FREQUENCY  4.8E-05

DOMINANT CORE DAMAGE SEQUENCES - PEACH BOTTOM (RSS) i CORE DAMAGE l SEQUENCE              FREQUENCY                      FUNCTIONAL FAILURE TC                    1E-05                          CoM,CoH TW                    IE-05                          CnHe                           ,
TQUV                  5E-07                          CoM l SEI                   2E-07                          RI,CoM
! AE                    1E-07                          RI,CoM S2HI                  1E-07                          RI,CoMt j S2J                   1E-07                          RI,CnHL R                     1E-07                          RI,CoM l TQVW                  1E-07                          CnHe j SE2                   SE-08                          RI,CoM S1HI                  flE-08                         RI,CoMt
! SJ1                   3E-08                          RI,C NHL l AHI                   1E-08                          RI,CoMt                        :
AJ                    1E-08                          RI,CnHe i AEG                   ----
RI,CoM,CuI i
TOTAL FREQUENCY     2.1E-05                                                   .
4 O

O DOMINANT CORE DAMAGE SEQUENCES - GRAND GULF (RSSMAP)
CORE DAMAGE SEQUENCE              FREQUENCY                FUNCTIONAL FAILURE T23 0W                  IE-05                  C,H e T 0W                    SE-06                  C,H e 1
SI                      SE-06                  RI, C,He l TC 23                     SE-06                  CoM, CoH T23 PQI                 I4E-06                 RI, C,He T1PQI                   2E-06                  RI, C,He T10VV                   2E-06                  CM o
T23 PQE                 SE-07                  RI, C oM T1PQE                   2E-07                  RI, C oM TOTAL FREQUENCY       3.5E-05

SEQUOYAH (RSSMAP) KEY ACCIDENT SEQUENCE SELECTION INITIAL KEY                                                                                               ADDITIONAL SEQUENCES
* SEQUENCES
* SH 2
(RI, Co,M,)
SD 2
(RI, C oM)
S2HF                                                     (RI, CoeM , C,Pt , C,HL , C,Rt )
V                                                        (RI, C oM, C,1)
T23 ML                                                   (CoH)
T13 B M_B'13                                               (CoM, Co H, C,P, C,H, C,R)
SH y      (RI, C ot M)
SD 1
(RI, C oM)
* FUNCTIONAL FAILURES ARE SHOWN IN PARENTHESES.
4,'
e D

4 SEQUOYAH (RSSMAP)
COVERAGE OF ORIGINAL PRA RESULTS BY KEY AND DOMINANT SEQUENCES INITIAL      WITH ADDITIONAL DOMINANT KEY SEQUENCES     KEY SEQUENCES  AND KEY i
CORE DAMAGE            68%               93%       $100%
FREQUENCY RELEASE CATEGORY 1                  2%              80%        n100%
     ~
2              n100%             +100%        n100%
3                 88%              88%        n100%
4                 36%              97%        n100%
5                  0%              95%        n100%
l l
t l
l

ZION (PSS) KEY ACCIDENT SEQUENCE SELECTION i
INITIAL KEY                                        ADDITIONAL SEQUENCES
* SEQUENCES
* 1      (RI, C ot M) 2      (Co M, Co H, C,P, C,H, C,R) 5      (CoH) 6      (CoH) 7      (RI, C og H) l 9      (RI, C oM) 10     (RI, C oM) 1 14   (Co M, Co H, C,P, C,H, C,R) 16          (RI, CoM, C,I)
SyCDY       (RI, C oM)
THF         (RI, CoeH , C,Rt)
* FUNCTIONAL FAILURES ARE SHOWN IN PARENTHESES.
k 9

ZION (PSS, REv. 0) i COVERAGE OF ORIGINAL PRA RESULTS BY KEY AND DOMINANT SEQUENCES
         \
              '                        INITIAL      WITH ADDITIONAL                       I)OMINANT KEY SEQUENCES     KEY SEQUENCES                         AND KEY CORE DAMAGE              61%                                          61%    85%
FREQUENCY RELEASE CATEGORY Z-1               52%                                          79%    79%
2                 3%                                          98%    98%
2R                99%                                          99%    99%
Z-3               73%                                          73%    86%
SR                70%                                          70%    93%
Z-5               71%                                          71%    78%
6                98%                                          98%    98%
7                61%                                          61%    89%
8A                 6%                                          34%    34%
8B                57%                                          57%    83%
l l
l

l                                                                        ,
PEACH BOTTOM (RSS)
COVERAGE OF ORIGINAL PRA RESULTS BY KEY AND DOMINANT SEQUENCES INITIAL       WITH ADDITIONAL DOMINANT KEY SEQUENCES      KEY SEQUENCES  AND KEY CORE DAMAGE                           99%                99%       *100%
FREQUENCY RELEASE CATEGORY 1               99%                99%      ~100%
2               98%                98%      $100%
3               98%                98%      *100%
4                0%                20%         20%

O r
PEACH BOTTOM (RSS) KEY ACCIDENT SEQUENCE SELECTION INITIAL KEY           ADDITIONAL SEQUENCES
* SEQUENCES
* TW        (C,Hg )
TC        (CoM, CoH)
TQUV      (CoM)
:                 SE        (RI, C oM) 1 TQW       (C,Ht )
AEG      (RI, CoM, Cg I) l
* FUNCTIONAL FAILURES ARE SHOWN IN PARENTHESES .
1

GRAND GULF (RSSMAP) KEY ACCIDENT SEQUENCE SELECTION INITIAL KEY              ADDITIONAL SEQUENCES
* SEQUENCES
* l                                                                                          T23 0W        (C,Hg )
l l                                                                                          T10W         (C,Hg )
TC 23 (CoM, C H) o SI          (RI, C,Hg )
T100V    (CoM) l                                                                                          T23PQI      (RI, C,Hg )
T1PQI      (RI, C,Hg )
* FUNCTIONAL FAILURES ARE SHOWN IN PARENTHESIS.
i i

a GRAND GULF (RSSMAP)
COVERAGE OF ORIGINAL PRA RESULTS BY KEY AND DOMINANT SEQUENCES INITIAL       WITH ADDITIONAL DOMINANT KEY SEQUENCES      KEY SEQUENCES  AND KEY CORE DAMAGE            97%                97%      *100%
FREQUENCY RELEASE CATEGORY 1             *100%              *100%        *100%
2                94%                94%          94%
3                83%                83%       *100%
4                85%                85%       *100%

RESULTS KEY AND DOMINANT SEQUENCES IDENTIFIED BASED ON:
         -CORE DAMAGE FREQUENCY
         -FUNCTIONAL FAILURE PATTERNS
         -RELEASE CATEGORY SPECTRUM l
l l
I

l l                                             RESULTS i                       KEY AND DOMINANT SEQUENCES PROVIDE HIGH COVERAGE OF ORIGINAL PRA RESULTS
                             -CORE DAMAGE FREQUENCY
                             -RELEASE CATEGORY SPECTRUM i
l                        KEY SEQUENCES INCLUDE MAJOR FUNCTIONAL FAILURES OF l                        CURRENT PRAS i
i 1
l

SEQUOYAH (RSSMAP) DOMINANT AND KEY SEQUENCES DOMINANT / KEY               FUNCTIONAL                              ACCIDENT FISSION SEQUENCES                    FAILURES                               PROCESS  PRODUCT _ CONSEQUENCES SH                           RI, C M                                    X 2                                            ot SH                           RI, C M 1                                            oe i
SD 2
RI, C oM                                   X S2HF                         RI, Cot           M , C,Pt, C,He, C,R L    X        X       X
,                                           V                            RI, CoM, C,I                               X        X       X SD1 RI, C oM l
;                                           S1HF                         RI, Coe           M , C,Po, C,Ht, C,R t
                                            'T 23 ML(Z)                    CH o                                         ~X AH                           RI, C M ot T13B M.B'13-                 Co M, Cn H, C,P, C,H, C,R                 X,I      X,I     X, I AD                           RI, C oM                                   X AHF                          RI, Coe           M , C,Pt, C,He, C,R t
!                                           "I" INDICATES SEQUENCE WITH CONTAINMENT ISOLATION FAILURE.
l
* INDICATES SEQUENCE CHANGE TO INCLUDE FEED AND BLEED FAILURE.                                         s

                                                                                              ~
ZION (PSS, REV. 1) DOMINANT AND KEY SEQUENCES              .
DonINANT/ KEY     FUNCTIONAL                 ACCIDENT      FISSION SEQUENCES        FAILURES                  PROCESS       PRonUCT CONSEQUENCE 1             RI, C ot M                      X 2              Co M, Co H, C,P, C,H, C,R    X,I           X, I    X,I 3              RI, C ot M                      X
 !            4              RI, C ot M
             *5              C M, C H o     o
             '6              CoM, CoH 7              RI, C og H
8              RI, C og M , C,H e 9             RI, C oM 10              RI, C oM 11              C M, C H o     o 12              C M, C H o     o 13              C M, C H o     o 14              Co M, Co H, C,P, C,H, C,R     X             X    X
.            15              CoM, CoH 16              RI, CoM, C,I                  X             X    X SICDY             RI, C oM THF               RI, CoeH , C,R t "I" INDICATES SEQUENCE WITH CONTAINMENT ISOLATION FAILURE.
* FIRE SEQUENCES SUBSTITUTED FOR A M SEQUENCES IN PSS, REV. 1.
I

PEACH BOTTOM (RSS) DOMINANT AND KEY SEQUENCES DOMINANT / KEY                       FUNCTIONAL                          ACCIDENT                        FISSION SEQUENCES                 FAILURES                           PROCESS                         PRODUCT    CONSEQUENCE TC                                   C M, CoH                               X                               X          X o
TW                                  C,H e                                  X                               X          X TQUV                                CM o
SE                                 RI, C oM                               X                               X          X 1
1 AE                                 RI, C oM l
S2HI                              RI, C ot                          M l
SJ          2                     RI, C,H e TQW                               C,H e                                  X                               X          X j
j                      SE                               RI, C oM 2
AEG                              RI, CoM, C,I                                                                                   ,
* SEQUENCES REMOVED FROM ORIGINAL PRA LIST:                                           R, S1HI, S1J,   AHI,  AJ.
l                                                                                                                                                    .

                                                                                              ~
GRAND GULF (RSSMAP) DOMINANT AND KEY SEQUENCES 4
DOMINANT / KEY    FUNCTIONAL          ACCIDENT               FISSION SEQUENCES
* FAILURES            PROCESS                PRODUCT CONSEQUENCE TgQW              C,H e                  X                      X       X 1
T10W              C,H e SI                RI, C,He TC g                Co M, Ca ll            X                      X       X T3 PQI            RI, C,He TyPQI             RI, C,He T100V             CM o                     X                      X       X Tg PQE            RI, C oM j         T1PQE             RI, C oM
!         *AE               RI, C oM               X                      X       X l         'AI               RI, C,He
          *ADDED TO RISK PROFILE.

d
.i j                                                                  er 1T-i PRESENTATION ON THE NRC PLANT CATER 4RIZATION AND SEQUENCE SELECTION FOR THE NRC/4DCOR MEETING MAY 15-17, 19818 i
l                                          BY TONY ENG, NRC FRED HARPER, SNL 3
i

PLANT CATEGORIZATION OBJECTIVE FORMULATE A LIMITED AND REPRESENTATIVE NUMBER OF PLANT CLASSES FOR THE CHARACTERIZATION OF CURRENT LWR RISK TO BE USED BY THE NRC SOURCE TERM AND SEVERE ACCIDENT RISK REDUCTION WORK.            FOR EACH PLANT CLASS, IDENTIFY.AND DESCRIBE THE DOMINANT ACCIDENT SEQUENCES AND THE REASONS WHY THEY ARE DOMINANT, k
e

SUPPORTING RESEARCH AND APPROACH PERFORMED BY THE ACCIDENT SEQUENCE EVALUATION PROGRAM (ASEP)
PROVIDE UPDATING 0F SEQUENCE LIKELI-HOOD ESTIMATES FOR REFERENCE PLANTS USING INSIGHTS FROM PRAS, SPECIAL STUDIES, OPERATING EXPERIENCE, AND CONSIDERATION OF TMI FIXES EXTEND THE REFERENCE PLANT CONCEPT (CONTAINMENT TYPES) CONSIDERING ACCIDENT SEQUENCE LIKELIHOOD CHARACTERISTICS IN THE DEVELOPMENT OF GENERIC PLANT GROUPS

ASEP  APPROACH
;                                                                                                                                                  I l       EvEn TREE DEVELOPMENT m m gy PRA DOMINANT                                            1 ACCIDENT l
SEQUENCE                       l mm*NF            INSIGHTS TO           PN                          PWT SYSTEMS                   C CATECORIZATION FR       P    l II ACCIDEBTI                      l              l SEQUENCE                       i             T LIKELIHOOD           ,         I REASSESSMENT SYSTEM                           q y l
i   '   MODEL I b                         "
DEVELOPMENT                 ACCIDENT I                                   SEQUENCE DESCRIPTION
;                                                                                                                                                                 l BY PLANT GROUP SPECIAL STUDIES,                             T OPERATING EXPERIENCE, l       ACCIDENT TMI FIXES                              SEQUENCE ANALY3IS l                                         GENERIC 1r nane st,tas or               SOURCE TERM e-          -
SEQUENCE                  -  AND LIKELIHOODS    PLANT-        SARRP POR REFERENCE                ASSESSMENTS py ,           SPECIFIC 9

l REFERENCE PLANTS BASED ON ASSUMPTION THAT PLANTS CAN BE
^
CATEGORIZED BY CONTAINMENT TYPE FOR FISSION PRODUCT RELEASE SELECTED PLANTS HAVE A TYPICAL CONTAINMENT DESIGN AND AVAILABILITY l
0F DESIGN DETAILS j                                       PEACH BOTTOM       (BWR MARK I)
)                                       LIMERICK       (BWR MARK II) l                                       GRAND GULF       (BWR MARK III)
)                                       ZION       (PWR LARGE DRY)                                  ,
;                                       SEQUOYAH       (PWR ICE CONDENSER)
SURRY       (PWR SUBATMOSPHERIC) l.
l i

PLANT CATEGORIZATION GROUP PLANTS FROM TWO PERSPECTIVES
!                           TOP DOWN:       -
LWR TYPE PLANT FUNCTION RESPONSE TO
:                                                  INITIATING EVENTS PLANT SYSTEM, RESPONSE TO l                                          -
l                                                  INITIATING EVENTS
;                        BOTTOM UP:        -
DIVERSITY OF DESIGN OF KEY l                                                  SYSTEMS NECESSARY TO PREVENT DOMINANT ACCIDENT SEQUENCES RISK CHARACTERISTICS OF DOMINANT ACCIDENT SEQUENCES l
1 1                                                                               .
e 9

TOP DOWN APPROACH DIVIDE LWRS INTO PWR AND BWR GROUPS DEVELOP FUNCTIONAL EVENT TREES IN REEP0NSE TO GENERAL CATEGORIES OF INITIATING EVENTS (LOCA AND TRANSIENT) 1               -
GROUP PLANTS BASED ON DIFFERENCE IN           ,
:                  PREVENTIVE OR MITIGATIVE FUNCTIONAL
;                  CAPABILITIES 1
i                                                                t l
1 1
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3 TOP DOWN APPROACH (CON'T)                 ,
GROUP PLANTS BASED ON THEIR ACCIDENT PREVENTION AND MITIGATION STRATEGY IN RESPONSE TO THE INITIATING EVENTS (STRATEGY MEANS THE SYSTEMS TO PERFORM THE FUNCTIONS)
CLASSIFY PLANTS BASED ON POSSIBLE SYSTEM COMBINATIONS REQUIRED TO PERFORM THE FUNCTION FOR THE SPECIFIC INITIATING EVENT DEVELOP SYSTEMIC EVENT TREES REFINE PLANT GROUPS CONSIDERING VARIATION OF SYSTEM DEPENDENCIES AMONG PLANTS

BOTTOM UP APPROACH GROUP PLANTS BASED ON DESIGN SIMILARITY OF THEIR SYSTEMS NECESSARY TO PREVENT DOMINANT ACCIDENT SEQUENCES IDENTIFY DOMINANT ACCIDENT SEQUENCE FROM TWELVE EXISTING PRAS IDENTIFY PLANT SYSTEMS AND SUPPORT SYSTEMS IMPORTANT TO DOMINANT SEQUENCES COLLECT SYSTEM DRAWINGS FOR ALL PLANTS SIMPLIFY SYSTEM DRAWINGS BASED ON PAST PRA INSIGHTS (MAJOR FLOW / ENERGY PATHS, flAJOR ACTIVE COMPONENTS, IMPORTANT PASSIVE COMPONENTS, ETC.)
FORMULATE " GENERIC" SYSTEM CONFIGURATIONS BY COMPARING ALL SIMPLIFIED SYSTEM DRAWINGS (E.G., ALL PLANT-SPECIFIC AFW SYSTEMS)
CONSIDERING DIFFERENCES IN REDUNDANCY, DIVERSITY, AND SUPPORT SYSTEM DEPENDENCIES             :
j
                                                                                                                                    ~
FOR EACH SYSTEM                                        ,
                                                                                                                                     =
1

BOTTOM UP APPROACH (CON'T)           .
COALESCE PLANT GROUPS (BASED ON SYSTEM DESIGN DIFFERENCES) BY THEIR RISK CHARACTERISTICS DEVELOP FAULT TREE MODELS FOR EACH GENERIC SYSTEM CONFIGURATIONS FOR DOMINANT ACCIDENT SEQUENCES l
DEVELOP RISK CHARACTERISTIC INSIGHTS THROUGH I                                     ESTIMATED ACCIDENT LIKELIHOOD, SENSITIVITY, i                                     UNCERTAINTY AND RECOVERY ANALYSES SEQUENCE FREQUENCY RANGES 4
l DOMINANT FACTORS DRIVING SEQUENCE I                                          FREQUENCIES COLLAPSE PLANT GROUPS BY SIMILAR RISK 1 '
CHARACTERISTICS

PLANT CLASS HIERARCHY MERGE PLANT GROUPS FROM TOP DOWN AND BOTTOM UP APPROACHES INTO A C0HERENT HIERARCHY REFINE INTERACTIVELY ALL LEVELS OF PLANT GROUPS ACCOUNTING FOR LOWER LEVEL CHARACTERISTICS IN HIGHER LEVEL PLANT GROUPS EXPAND PLANT CLASS HIERARCHY TO INCORPORATE CONTAINMENT TYPES FOR EACH PLANT GROUP AT VARIOUS LEVELS t
O

.        b
    =
PLANT CLASS HIERARCHT/.
LWR                                                   LIGHT WATER REACTOR BWR                     PR                                     TWO TYPES OF LWR i
l BWRL1,1                         BWRLI.2                                        PLANT CLASSES BASED ON SIMILAR FUNCTIONAL RESPONSE TO l                                                                      INITIATING EVENTS i
BWRL2.5               BWRL2.6                                PLANT CLASSES BASEE ON SIMILAR SYSTEM AESPONSE TO SPECIFIC INITIATING EVENTS BWRL3.3                         BWRL3.4                                         PLANT CLASSES BASED ON SIMILAR RISK CHARACIERISTICS OF DOMINANT ACCIDENT SEQUENCES BWRL4.4                 BWRL4.5                                     PLANT CLASSES BASED ON SIMILAR DESIGNS OF EIT DEFENSE-IN-DEFTH I                                                               SYSTEMS NECESSARY TO PREVENT          1 l                                                               THE DOMINANT ABCIDENT SEQUENCES BWR 3             BWR 20    BWR 15             12         BWR 1                       SPECIFIC PLANTS AT EACH LEVEL, PLANT GROUPS CAN BE FUR'DIER CATEGORIZED BY CONTAINMENT TYPES

CURRENT STATUS PLANTS CAN NOT GROUP INTO FEW CATEGORIES BY SIMILARITY OF SYSTEM DESIGN (LACK 0F STANDARDIZED DESIGNS IN THE U.S.)
29 PWR PLANT GROUPS (FROM 72 PWRS), 15 BWR PLANT GROUPS (FROM 31 BWRS) (WITHOUT ACCOUNTING FOR SERVICE WATER SUPPORT SYSTEM VARIATIONS)
COMPLETE TOP DOWN APPROACH BY 7/84 COMPLETE BOTTOM UP APPROACH BY EARLY 1985 COMPLETE REBASELINE OF SEQUENCE LIKELIH00DS FOR REFERENCE PLANTS BY 7/84 9

l l
There are two sets of sequences which j       will be analyzed l
: 1) 10 sequences in the Interim ASEP analysis
: 2) 19 sequences in the source term reassessment studies I
I l
1 i
1 i
i

1 4
i i
i i        i i
i          The sequences analyzed by ASEP are dominant to core melt frequency
;           1) SARRP
: 2) SARP l           3) Other programs i
l l
l i
e
* L P

l                                                                                              .
!                                                                                                 s I                                                                                              -
i I
;                The first cut at updating sequence info j                            was the reassessment report i
j               1) Used past PRA information l               2) Applied new insights from other studies l               3) Back of the envelope type calculations 4
j               4) Only did plants with PRAs i
4 l
r I
i I
1

I 24 sequences were defined to represent the spectrum of sequences from past PRAs 1
: 1) 8 BWR sequences
: 2) 16 PWR sequences J
O

Table 2.1    ASEP PWR Sequence Classes Sequence Class                     Sequence Class Description 1       All types of transients with no core cool'ing but with containment systems available 2       TLOOP and TAC /DC bus with no core cooling and no con-tainment cooling 3       All types of transients with no core cooling and with sprays operable / fans failed 4       TLOOP and TAC /DC bus with no core cooling and with fans operable / sprays failed 5       All types of transients with failure to scram (ATWS) 6       Many types of transients with SRV stuck open and no       )
core cooling during injection but with containment        J systems available 7       Man types of transients with SRV stuck open and no core coo ing during recirculation but with containment systems available 8       Many types of transients with SRV stuck open and nd core cooling during recirculation and with fans operable but sprays failed during recirculation 9       Spurious safety injection transient with no core cooling during recirculation and with sprays operable / fans failed 10       "V" sequence (interfacing systems LOCA) 11       All size LOCAs with no core cooling during injection but with containment systems available 12       Small LOCA with no core cooling during injection but with fans operable / sprays failed 13       All size LOCAs with no core cooling during recirculatiop but with containment systems operable                   i 14       Intermediate /small LOCAS with no core cooling during re-circulation but with fans operable / sprays failed 15       Intermediate /small LOCAS with no core cooling during recirculation and with no containment cooling during recirculation 16       Small LOCA with eventual core cooling loss and with no containment cooling caused by the initial loss of containment spray 9

Table 2.2    ASEP BWR Sequence Classes sequence Class                     Sequence Class Description 1       All types of transients with no core cooling, with VSS available, and with/without suppression pool cooling 2       All types of transients with SRV stuck open and with no core cooling, but with VSS available, and with/without suppression pool cooling 3       All types of transients with core cooling, VSS available, but with no suppression pool cooling 4       All types of transiertts with SRV stuck open, with core cooling and VSS available, but with no suppression pool cooling
:   5       All types of transients with no suppression pool cooling leading to loss of core cooling before containment failure and with VSS available 6       All types of transients with failure to scram (ATWS) and without adequate cooling, but with VSS available, and with/without suppression pool cooling 7       A NO PCS transient with failure to scram (ATWS) and with adequate core cooling and VSS available but with inadequate suppression pool cooling 8       Small LOCA with core cooling and VSS available but no suppression cooling 10 l
l l
l

Several issues were incorporated into sequence values in the rebaselining report i                                                                           1) Battery depletion blackout sequences
: 2) Offsite and onsite AC reliabilities
: 3) Pump seal LOCAs
: 4) Auxiliary feedwater AC dependencies
: 5) Initiating events
: 8) Stuck open safety / relief valves
: 7) Recovery timing on long term sequences

Other issues were mentioned, but not quantified
: 1) Feed and bleed
: 2) Cold shutdown before recirculation
: 3) ATWS implementations
: 4) Fast cooldown of primary e
e 6

0 1
d t en wa ot hr s op tm n
ei me                        _
sb s
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s                         _
s s        s  s a e        e  e e
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s e
s Aes        R W
R W
e          P  B h          5  5 T          )
1
                        )
2
        ;    !   i;!       ! l : ! !

The 5 BWR sequences included both
,                                               early melt.s and late melts
: 1) Transient with loss of injection (TQUV)
: 2) Transient with loss of decay heat removal (TW)
: 3) ATWS (TC)
: 4) Transient with stuck open SRV and loss of injection (TPQE)
: 5) Transient with stuck open SRV and loss of decay heat removal (TPW)
O e
e

The 5 PWR sequences were early melt, high RCS pressure sequences
: 1) Imsa of all feedwater (TML)
: 2) Station blackout. (TMLB')
3                3) ATWS (TMKU)
: 4) Transient induced LOCA with loss of injection (TMQ-D)
: 5) Small LOCA with loss of injection (S,D)

1 1
The dominant sequences for ASEP and the precursor study are similar
: 1) Some minor differences a  BWR small LOCAs b PWR ATWS e P W R S ,H
                                                 . f kNe(/f/}}
s a
                                           'y kl O
e A

O ASEP dominant sequences compare well with Q
IDCOR functional groups DWR i
ASEP            IDCOR TW TW T    TWS)       TCgTWS)
QI            T UV Blackout.)
TPQE              QI ( ,,S,)
b        \#
                                                                      ~           .
w

t i ASEP dominant sequences compare well with IDCOR functional groups PWR IDCOR                      #
sad
          $ (TvQ-D)
Tuw                        $h TML v
Tuc -o TDU (ATWS)             ];
Tuta-
                                                         /a
                                                          .O-j{ 7
                                                          -C O
e e

i The ten sequences are analyzed
,         quantitatively and qualitatively i
i          1) Interim ASEP effort
)
: 2) 100 plants l
: 3) Pipe segment level of detail
: 4) Simplifying assumptions--electrical, procedural, actuahon, control L   $   jLH

l TMI fixes are incorporated
: 1) Difficult to find out the status of changes
: 2) Auxiliary feedwater AC dependencies
: 3) Room cooling, pump seal integrity--2 hrs without AC
: 4) Anticipatory reactor trip on loss of feedwater or turbine vip
: 5) Automatic RCIC switchover from CST
(
a
                                                                                                                             +
                                                                                                                                         = ,r -
                                                                                                                            ~)h,;/. .
3'_1 1             -
E e
O

i Many of the other sequences will be analyzed during the Final ASEP effort
                         .[
IQ   .
                                              ;l
                       . g.
(n . s-r ,   y g- --

l I
The source term sequences are not necessarily dominant
: 1) Many are risk dominant
: 2) Some were chosen to cover the range of phenomenological situations l
l A
e

o l             There are ten BWR sequences considered in the source term reassessment studies Plant. Sequence   Frequency % Core melt Comments Grand Gulf    TC         5E-6       23       RSSMAP TQUV       4E-6       17       Rebaselined, long term blackout SzE         E-8        1       Rebaselined, pump seal leak TPQI       2E-7        1       Rebaselined, timing, IE Limerick      TQUV       IE-5       86       Limerick PRA TPE        2E-7        i Limerick PRA TC         IE-6        9       Limerick PRA Peach Bottom  TW         8Ev3       30       Rebaselined, timing IE i
TC         7E-6       30       WASH 1400-except IE AE         2E-7        1       WASH 1400

There are nine PWR sequences considered in
:          the source term reassessment studies l
l Plant. Sequence Frequency % Core melt Comments Sequoyah    SzHF     5E-6            5   RSSMAP, old TMLB'    IE-6            2   Rebaselined, procedureslong term blackout TML       3E-6            3   RSSMAP, no feed and bleed Surry       SsD      9E-5           58   Rebaselined, pump seal leak AB        <E-8           <1   WASH 1400 V        <E-6            <1   Rebaselined, new procedures TMLB'    2E-5           13   Rebaselined, long term blackout Zion        S2D      DE-4           50   ZPSS review, loss of CCW TMLB'    6E-6             1  ZPSS, seismic l
l
                                                                                    ^
l                                                                                 -

S
(                                       h MODULAR ACCIDENT ANALYSIS PROGRAM INTRODUCTION R. E. Henry IDCOR/NRC Meeting on Integrated Analysis of Severe Accident Containment Loads Rockville, Maryland May 15-17,1984 l
Y                                         Y

PURPOSE OF THE CODE e       To calculate the primary system and contain-ment response for postulated severe core damage initiators using realistic models of the key physical processes.
l l
e       Phenomenological models are based upon first
!             principle physics with sufficient detail to char-acterize the major influences on the overall response.
* Concentrate on best estimate approach, but-when uncertainty and/or sensitivity analyses show a phenomenon to be of little importance, the mathematical representation can be rele-gated to a more simple formulation or assump-tion. This focuses the attention on the key physical processes through the use of parallel uncertainty and sensitivity analyses.
l                                                                Y l
1

1 i
DESIGN PHILOSOPHY OF MAAP e Develop simple physical models for the indi-1 vidual phenomena.
e incorporate the fullest possible representation of the feedbacks between models.
e  Establish the adequacy of the physical models with both individual and integral uncertainty and sensitivity analyses.
4
    ~~''   ' -''

MAAP FEATURES e    Fully parallel - phenomenology not constrained to occur sequentially, e    Modular - relatively easy to add new compart-ments or improve phenomenological models.
* Models wide variety of phenomena, engi-neered safeguards, and other plant systems.
e    Models wide variety of designs.
e    Phenomenological models simple but backed up by work of IDCOR subtasks.
* Extensive capability for operator interventions.
e    Fast-running.
l

MAAP CODE STRUCTURE Hig5 Level Subroutines Main Program input / Output Integration System and Region Subroutines Event Codes Primary System and Containment Nodes Core Heatup Phenomenology Subroutines Hydrogen Combustion Concrete Decomposition Vessel Failure Property-Library Subroutines Steam Tables Y    .
1 I

i Main Program i
input                                                                Output Subroutines                                       b ente             Subroutines Int e gr a t or JJ i
Region 1                       Region 3                     R e glom N Subroutine                     S u b t outine               Subroutino l
r---        ,            r---           ,
_. Ph. nom....
S u b r o u tin. i         l4g, , , , _ _ ,_ 3 l      l4                   n l   g ,,,,, _ ,_ _ 3 I                          I I                          e l                Property Subroutines
          ~
Phenomenon Subroutine 2 l
l
(      I J

MAAP ANALYSES Major Intervals of Accident Evaluation e initial transient - depletion of primary system coolant.
O Core uncovery and heatup.
O In-vessel material migration.
O Containment analyses (debris in containment).
i l

MAAP ANALYSES Initial Transient and Potential for Cooling With No Core Damage e Considerable detail on system geometry.
e includes loss of water inventory and accumulation in the containment.
e includes limited water sources.
1 i

l     .    .
(                                                                         D MAAP ANALYSES Core Uncovery and Heatup e Calculates depletion of water inventory.
O Considers specific core geometry (segmented BWR core versus open lattice PWR design).
e includes the influence of limited water sources.
O Calculates hydrogen generation due to cladding oxidation - 12.1.
e Models the behavior of core spray and upper head injection - 15.1.
O Evaluates the influence of in-core blockages, due to degradation, on the remainder of the accident progression - 15.1.
k

MAAP ANALYSES In-Vessel Material Migration e Less detail in the phenomenological description, will rely on well characterized physical principles.
: 1. Material migration into the lower plenum - 15.1.
: 2. Fragmentation - 15.1.
: 3. Steam generation - 14.1, 15.1.
: 4. H 2 formation - 12.1.
: 5. In-vessel coolability - 15.2.
: 6. Vessel failure - 15.2.

MAAP ANALYSES Containment Analyses e Physical description based upon well characterized physical principles.
I
: 1. Ex--vessel steam generation - 14.1.
: 2. Additional H 2 formation - 12.1.
: 3. H 2 combustion - 12.2 and 12.3.
: 4. Core material distribution - 15.2.
: 5. Core debris coolability - 15.2.
: 6. Concrete attack - 15.3.
: 7. Establishment of heat transport path.
k

1Y' THE MAAP-BWR SEVERE ACCIDENT ANALYSIS CODE l
J. R. Gabor IDCOR/NRC Meeting on Integrated Analysis of Severe Accident Containment Loads Rockville, Maryland May 15-17,1984 t

                                                    ~
* l l
l MAAP/BWR APPLICATIONS TO DATE e Reference plant analysis (IDCOR)
Peach Bottom Grand Gulf e BWR core heatup code benchmarking e Others

ACCIDENT SEQUENCES ANALYZED TO DATE e Large break loss of coolant accidents (AE)
Liquid line break Steam line break e  Small break loss of coolant accidents (S E)
Liquid line break Steam line break e Station blackout (TQVW) e Anticipated transient without scram (TC) e Transients with loss of injection (T QUV) 1 e  Transients with loss of containment heat                                     j removal (TW) i

[                                                  D l
l 1
PRIMARY SYSTEM SAFETY FUNCTION IN MAAP HPCI high pressure coolant injection RCIC reactor core isolation cooling l
HPCS high pressure core spray Suppression pool alignment to HPCI, RCIC, HPCS RHR +1 and #2 heat exchangers RHR *1 and #2. pumps RHR #1 and #2 alignment to LPCI RHR +3 pumps (always aligned to LPCI)
Y

PRIMARY SYSTEM SAFETY FUNCTION IN MAAP LPCS low pressure core spray CRD control rod drive flow High pressure service water injection SLC standby liquid control Safety relief valves (5 sets)         i ADS automatic depressurization system Reactor protection system i
MSIV main steam isolation valves      l l
l l
1 l

CONTAINMENT SAFETY FUNCTIONS IN MAAP Drywell vacuum breakers l
Drywell purge system (Mark lil) l l          Containment purge system (Mark lil)
Containment venting system Hydrogen igniters i

  ~   '
CONTAINMENT SAFETY FUNCTIONS IN MAAP RHR +1 and #2 alignment to SP RHR +1 and +2 alignment to drywell sprays (Mark I and ll)
RHR +1 and #2 alignment to wetwell sprays (Mark I and 11)
RHR +1 and #2 alignment to containment sprays (Mark Ill)
Y

1 SRV o
a    MAIN STEAM UPPER PLENUM                     ]      z
* STEAM E C C S--*a                                              TO ECCS SPRAYS                                  '(               9TURSINES u
ECCSINJECTION b=
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r LPCl(BWR/6)               ;
ff . -
                                                                - JET PUMP l               CORE                               :
C A S LC -*=
FLOW LOWER PLENUM l    .:: l 0.'.*. " .*.". L ...
     ~b....b--.E..Y..E*ECRD FLOW BWR Primary System Nodalization

          -            +         . - -          .a -                  -   - -                     g g
: r.            -     .
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r            -
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:                      PRIMARY SYSTEM DRYWELL
                                       ''g;-
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T O R U S_.
                                                          " " '                        l oh SHELL
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NATURAL CIRCULATION CONSIDERED BWR MARK 1 CONTAINMENT

1
                                                                 ,._r d        O
[/           ____.
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PRIMARY SYSTEM k6                       f' r
ORYWELL E                 ?           '/ .
7
                                                                      ;             _   ?             /:        WETWELL PEDESTAL CAVITY
                              'jff_is_.' /;  //// /
hf                                 :
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                              !                                                  //// '                              h ,!
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:.'.''7                                                                  .,..,,.m
                        . . . . . . .               .. ,.. ...                        s
                                                         .....                        i PRWARY                                                        ,,, ,, ,,,,,,,,
SYSTEM                  ,                DRYWELL                    ,,,,,,         WETWELL l    . . . . . . .,,...
                                 ''"" !r,.*,.,.                                                                                  1 I                                 jyy*jyAL           ...
                                                            ..         _                           ...        .        '#::'a**-
1 BWR Mark l Containment Nodalization 1
1
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                                                               , , [7%HL//H/f h
ll                                                           l COMPARTMENTS                 l/                  '
flg,
                                          ,.4
                                                                                                   .Vi
                                          ' 't                                                      i''i y         .                                           .,  ,
                                                    /                                                  /
lE                      f
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                                                                            ~
z   -j      2       PRIMARY SYSTEM
                                          /'t'                      I               l f         ({ ,
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                                         / ;)            /                          l   ;
I               ORYWELL j
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WETWELL-> * / ,I                              '
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                                                            ,             _                         < /. PEDESTAL CAVITY r
2        91                                          'f V////////////////////////////4
                 ,,,.                        **"'''""'",',,[****              UPPER CONTAINMENT COMPARTMENT S                       -
MOOLE CONTAINMENT COMPARTMENT A                           j
* T.',',,""     i l=
             ....,u.
                            ,,,y WETWELL l
SYSTEM                         ,,,,""              DRYWELL
             .n...              . . ~
T
                          "'*l.".""'.                                                                                   .....
PEoEsTAL                   ""*"'"'"*-
CAVffY
                                                       .....,...                                         .....                                                                      1
                               ..                                                     .....                                                                                         l a mamma                                                e.eus                                                                                         j f                                                             i l
BWR Mark lil Containment Nodalization

    .   .                                                          tl M
[
THE MAAP-PWR SEVERE ACCIDENT ANALYSIS CODE i
M. A. Kenton IDCOR/NRC Meeting on integrated Analysis of Severe Accident Containment Loads Rockville, Maryland May 15-17,1984

l MAAP/PWR APPLICATIONS TO DATE e                         Reference plant analyses (IDCOR).
Zion Sequoyah e                         TMI-2 benchmarking (EPRl/IDCOR).
* EPRI heatup code benchmarking.
o                         Others.
1 1
l l
      -___m_,_-------.m-w_,,-m  --            - , ++v. -,- - , - - . .,,,. w-- ,, , , ---. __-..---w ym     - - - - - -       - - - - - - - ,-

KEY PWR SYSTEMS MODELED IN MAAP l
PRIMARY SYSTEM e   injection (LPI, HPI, Charging) e   injection- and Recirculation Phase Pump Lineups e  Accumulators (Cold Leg, UHI) e  Pressurizer and Steam Generator Relief Valves, Steam Dumps e  Main and Auxiliary Feedwater l
l
                      -,---,-.--,--w-,,.,--,,-,--,-w,             -,-- - - - -, ,, ., -- . - - , - - - , --- - - ,----- - - , . - . - - - - - - - - - , - -- - , --- - - - . --

l KEY PWR SYSTEMS MODELED IN MAAP CONTAINMENT e                   Sprays e                    Fan Coolers, Fans e                   igniters, incomplete and Complete Burning Models e                      Ice Condenser k

EXAMPLE SEQUENCES RUN WITH MAAP/PWR Loss-of-Coolant Accidents:
e   Large break.                               )
1 e   Small break.
Transients:
e  Blackouts (TMLB').
* Loss of feedwater with and without injection, (e.g., T23ML).
Others:
e  Loss of shutdown cooling.

EXAMPLE LOCA SEQUENCES RUN WITH MAAP/PWR e Large break without injection, (e.g., AD, AH).
* Small break without injection, (e.g., S2D, S2H, S2HF).
* Stuck-open pressurizer relief valve.
* V-sequences.
i i
e Steam generator tube ruptures.

o                                                     o M            N AfoRS                              E M G NERATOR
                               /                                                   /
                           \TI                                                   (mi                       )
                                                                      'R 3 UNORoKgN Hof LE                     yppgg , g y g, Ho N()e LOWER PL NUM                                                          i
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                                     /                                              ,
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[!EE5'fre 4et NoR E!!NE bus MAAP-PWR Primary System Nodalization (Westinghouse Plants)

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COMPARTMENT
                                               ,                                           f
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s IN BROKEN AND UNSROKEN LEGS v
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                                                            ]    /             V'          /          COMPARTMENT COMPARTMENT             ,
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PRIMARY SYSTEM sg l                 _
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                                                             /// /
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l     (                   PWR Large, Dry Containment Nodalization                                                                           j

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                                                   ....~.                                                                     . .-
                                                                                          . . _                                                          .r.AM O.N. 4,o. ..
PWR Ice Condenser Containment Nodalization

f l
l
[                                                                                )
i ON-GOING WORK WITH MAAP/PWR e                 Extensions of TMI benchmarking work.
e                 incorporation of more sophisticated primary sys-tem thermal-hydraulic models to produce inputs 4
i for fission product transport calculations.
l e                 incorporation of more sophisticated fission prod-uct release and transport models.
                                                                                                                          )
l l
_ _ _ - - - - - - - - , - - - - - , ,    .---,-.___,_rr_,          . . _ ,_ _ _ _ _ - - ._,m___. ._. _ , . ___ ___ _

1 l
MAAP SUPPORT P. I. NAKAYAMA IDCOR SUBTASK 16.1 IDCOR/NRC MEETING ON INTEGRATED ANALYSIS OF SEVERE ACCIDENT CONTAINMENT LOADS ROCKVILLE, MARYLAND MAY 15 - 17, 1984 11011 Toneyona Road - Post Omco Box 85154                                   .         San Diego, Californio 92138                                                                                 .
(619) 453 6580 1

E                                                                   !
VERIFICATION OF MAAP OBJECTIVE: e TO INSURE THAT CODE DID WHAT MANUAL SAYS IT DID.
4 i    APPROACH:
* PRODUCED CODE DOCUMENT INDEPENDENT OF CODE DEVELOPERS.
* CODE FLOW DIAGRAMMED.
e MODEL EQUATIONS INDEPENDENTLY DERIVED.
I e LINE-BY-LINE CODE CHECK.
l RESULT:    e RECONCILED DIFFERENCES WITH CODE DEVELOPERS.
t

e               e t
EXAMPLE OF LINE-BY-LINE VERIFICATION l
i e coelva ernacostu tescreescosts tur amtcat=6 s t noc e s t vet .
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l 1

ANALYSIS TO SUPPORT MAAP PHENOMENOLOGICAL MODELS SRV CYCLING:                   ADDITIONAL STEAM AND HYDROGEN PRODUCTION WATER ABSORPTIVITY:            HEAT TRANSFER TO WATEP. COULD AFFECT PARTICLE SIZE DOWNWARD VIEW FACTOR:          ADDITIONAL STEAM AND HYDROGEN PRODUCTION BUOYANCY DRIVEN FLOW:          FLUID, HEAT AND FISSION PRODUCT TRANSPORT DEBRIS QUENCH:                 MULTIPLIER OF TWO USED IN MAAP TO ACCOUNT FOR SIDE QUENCHING l
l l
l

STEAM SPIKES DURING VALVE LIFT 45 -
Steam Generation Rate 40 -
35 -
30 -
u 25 -
                            .?
20 -
15 -
W to -      k s
5-          N 400          800       1200  1600 2000       2400 2800        3200 Time (sec)

NO DIFFERENCE IN MAXIMUM FUEL TEMPERATURE AND HYDROGEN PRODUCTION 3000                  -
i 200 l
l l                           2500                  -
Maxima Fuel Temperature
                                                                                                                                                         - 175 i 150 2000                 -
Total Hydrogen Generated -                          125 K
kg
                                                                                                                                                         - 100 1500                 -
75 l
t 50 l
1000                 -
                                                                                                                                                         - 25 l
                                                                     }
                                                                      .                       i   i          e i         e                           i 0
400    800      1200        1600   2000 2400 2800                   3200 Time (see) i l
  , _ _ . - . - _ - - - -     _ _ _ _ , _ _ . _ . . _ - _ - _ _ . -              .___,___,__m         __,m

SOURCE:                              HANDBOOK OF OPTICS 100.0 s
3 10.0              -
o                  -
lc e                                                          t ~100%
                                                   ~
l 8
E e
c 8
                                                   ;                                                               t~75%
O j             1.0  -
                                                                                                                      ~50%
                                                                      .                                                             1 I
                                                                      .                                                                f 0.1                1                                         I           il                                                                                I 0.8           1.0                                1.2                1.4       1.6                              1.8                              2.0                             2.2 Microns                                                                                                            l 1

10.0 e
i       1.0  -
              .d u                                                            ;500*K
                                                                        /
eQ w .c a       .
Iw w
E "e       .
              =8   a 0.E oa
              >y           -                                     3000*K we c-e be
              *3 0.1  -
                ?.         .
m.
                           .                                                                           i 0.01                        I           i   e       i            i 0.8 1.0                           1.2 1.4     1.6      1.8 2.0 2.2 Microns

4 WATER ABSORPTIVITY e l'00% ABSORPTION IN 1 CM OF WATER OF A GREATER THAN
: 1. 3 7 y M e 63% RADIATED AT A GREATER THAN 1.37 uM FOR 2500*K I                          o 15% RADIATED BETWEEN 1.1 AND 1. 37 UM WHERE 50 TO 75%
ABSORBED e OXIDE SURFACE SHIFTS EMISSIVITY TOWARD LONGER WAVE LENGTH                                                l e 70 PLUS % OF ENERGY WOULD BE ABSORBED IN CM e MAAP TREATMENT IS SUFFICIENT

DOWNWARD VIEW FACTORS
        ..f I
R=           L=
1      1 1             A= L      +R  - 1 I      I I      I B=L     2
                                                       -R2+1
                       -                                      ~
F 12 = 1-    COS      i i-h      ( A + 2)    - ( 2R) COS   l
                              +B SIN     I j  -
THE SECOND TERM IS THE " ESCAPE" FRACTION WHICH IS DIVIDED EQUALLY UPWARD AND DOWNWARD CL ADD ING O.D. = 1. 09 CM ROD PITCH         = 1.44 CM 1                 = 366 CM l                        ESCAPE FRACTION ~ 10-3 l

BUOYANT FLOW OBJECTIVE:  e TO ASSESS EFFECT OF BUOYANCY DRIVEN TWO-DIMENSIONAL FLOW e LET FLOW DEVELOP, NOT SPECIFIED e OBTAIN MEASURE OF FLOW FROM UPPER PLENUM TO
      .               CORE REGION APPROACH:
* USE AVAILABLE TWO-DIMENSIONAL FLUID-DYNAMIC CODE, MIRTH e USE ENERGY INPUT FORM CORE HEATUP CODE LIMITATION: e INCOMPRESSIBLE. WITH BOUSSINESQ AND PDV WORK e SINGLE FLUID e NO GEOMETRY CHANGE e NO TEMPERATURE FEEDBACK ON 0x!DATION

MASS, MOMENTUM AND ENERGY EQUATIONS
       ~                      ~
3,   g   g 3. g R. U    38 ,V Z
5T * . R    3R
                      +
8Z
                                    =0                           CO ai;.Ra.za=.3a.(2.g-,.                                            <,>
3i;.R.a.za=-3.....                      1
                                      ... g T-1>            <,>
hoi.i1h.,,Rui.h.z,vi}=-;/'b"".h.z]
i           v
                                              + H h (T M - T)   (4)

       - -        +                           2-e e CORRECTION TO CORE METAL TEMPERATURE T    =T     + T'           (1)
DT M CP C    C DT        = HA( T g   -T)+
C       (2)
DT M CP C    C DT
                                       *           ()
M CP C     C         = HA ( T g   -T)   C
()
i i
T N#
T'   =fT      g C
p C
(T g - T C) DT  (5) g

MIRTH NODALIZATION OF SEQUOYAH REACTOR PRESSURE VESSEL 4
sd i e--1.74 m
* a                 a d I Upper Plenum Inner Region 3.084 m Reactor Pressure u,,g(t)  Vessel Outlet Upper Tie-Plate Control Rod-Spide,r Upper Assembly                            0.61 m Upper End Fitting g                             g   Fuel Rod Fission Gas Plenum Active Fuel                               3.65 m Assembly 4__               m            :   r vio(r,t)

POROSITIES AND FRICTION FACTORS FOR SEQUOYAH REACTOR PRESSURE VESSEL POROSITIES         FRICTION FACTORSX 8      Sg     SZ      FUCS-1)    F yCS-1)
ACTIVE FUEL ASSEMBLY AND        0.5     0.5    0.5        7          2.0 FUEL ROD FISSION GAS PLENUM UPPER END FITTING AND           0.5     0.5    0.5        5          3.3 CONTROL ROD SPIDER UPPER TIE PLATE                 0.5     0.5    0.32       5          5.5 UPPER PLENUM                    0.752   0.752  0.752      3          0.15 MFRICTION FACTORS USED IN MIRTH ARE DIMENSIONAL AND HAVE UNITS OF SEC~I. (SEE EQS. 2 AND 3) i

VELOCITY VECTOR, GAS AND METAL TEMPERATURE AT 145 MINUTES l
o b
66 0     5 ea    ~ a*..
a'..                                                   /
1.30+003 64..-n'.,
s,s.'ns 4ssi-a es esei-n ai 64.**a es 6  ...a 4                              1.20+003                 -        1.20+003 s...,! o.
I
:L; /.>.}
             ,,,,h
: j.                       1.20+003                    'i 1.30+003              0 6++ + 4 : ,, ,. ,,. .; m o
                        '"                            1,40 003                          1.30+003
              !Ii;.* *. l . l. l ,. :
* 1.50+003                  \      1.40+003
                  *:::::l 1:18:883 7n+
                                                      ,,gg;003gg -d
                                                                    / p/
1.40+003 1.20+003
                          .                                            w                1.00+003 gDO.g 800.0 2 188:8 800.0 800.0 l
(
                                               )) ,         700.0
* g.g.,      JN            300.0
                             ;          l W 'n [ l L'
: 1. 0+003 fil
                                            ].                 1.30+003 1.50*003 ll$0'003             l            1.60+003 1.gD+00 Jai -  800.0
                                                               !s!!!!
800.0 500.0
                                                      . 588 4   vQ

l e   .
nOE Id2Cx yA52N3 0
_              0 0,
0                                 5$ gC. &3 4                                 i
                                                 ?
9% fCH                       l$2EN n3 !
O O.
               )
X 93 F*     I'l
(
5 3
0 0
0 0
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         )
K
(    0 e0 r                                             O   W u 5 0                                         D t
a2                                        w r
e p
o    %
m                         O e0 T 0 s   0, u0 m2 i
x a
M 0
0 0,
5 1
0 0
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1 0
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5                                       L$*a% E2$ E FS F%E n {A l
7E ';>=gI   :
: 3 0                                       3Eo= %k:               -
0            -         -       ,-          -      _            .
5     0PO       %P        u O'       EO     5o     so.O   -
a*g {L                            -

e   e l
FLUX AT TIE PLATE AT 145 MINUTES O.                                                              '
l co d
9 O
U M_
LJ O (A
N M
* N X  g-E X
D J 9.
L_ o
      ~
l     d-
,     9 l
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i       6   6       6       4     4       4     6   e     e 0.0 0.2 0.4     0.6     0.8   1.0     1.2   1.4 1.6   1.8 RADIUS (M)

BUOYANT FLOW RESULTS:   e  HOMOGENEOUS GAS AND METAL TEMPERATURES IN CORE AND PLENUM REGION.
l i
e  TEMPERATURES IN CORE REGION COOLER THAN HEATUP.
J METAL ~ 300K, GAS ~ 500K AT 145 MINUTES e_ REVERSE FLOW FROM PLENUM TO CORE ~ 10 KG/SEC e  AT  145  MINUTES   UPPER  ASSEMBLE  TEMPERATURES REACH STEEL MELT TEMPERATURE e  CALCULATIONS BEYOND THIS POINT SHOULD TAKE INTO ACCOUNT TEMPERATURE FEEDBACK TO OXIDATION, IN-   !
l l                 CREASE STEAM AVAILABILITY,'AND GEOMETRY CHANGE   )
DUE TO CLADDING BALLOONING.
1 t
i

DEBRIS BED GEOMETRIES
(                      ATMOSPHERIC PRESSURE                                                               ATHOSPHERIC PRESSURE o
a   i,il d i                   .ss4 Il                                                                                  Il WATER AT 100'C WATER AT 100*C m     .
2.15 m                                                            .           ...
2.15 m o                                                                                   n
          ,,,,      o                                                         i         ,
                                                                                           ,....         n DEBRIS BED AT 500*C                                  l 1  .....
L    ,         65 kg STAINLESS STEEL                                 i         ,    :'.':.
          ,,,    0.75 m                           DEBRISBEDAT500*Cyl 0.75 m d = 3.1 m c = 0.4                 65kgSTAINLESSSTEELl                    l l
d = 3.1 mm                   l         l l""
o i  ;...          o 4,,,,,                                         c . 0,4              4                   1....-
-l l :                                                              0.075m 0.12m ll 0.075 m ONE-DIMENSIONAL DESRIS                    TWO-DIMENSIONAL DEBRIS BED GEOMETRY                            BED GEOMETRY

o   .
MOMENTUM VECTORS OF ONE-DIMENSIONAL DEBRIS QUENCH 44                                   4 a.
                  ..J                                       k
                                                           }\
                  ..J                                       ,
                   .J
                   .J l
                 ..J l
                ..J                                     l jl i 1
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l
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                .. i
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: d. :.
4                                    G GAS MOMENTUM VECTORS              LIQUID MOMENTUM VECTORS AT 160 SEC                           AT 160 SEC VGMAX        '+ 7 M/S                VMAX = 2.26 M/S

                                                                   .O CONTOUR PLOTS OF ONE-DIMENSIONAL DEBRIS QUENCH o
O f                                   44 100.0   _   10 0.0
                    .400
      .1000    ,;
375.0
        .200    l                           125.0 l
: 4.                                   4.
VOID FRACTION CONTOURE                    BED TEMPERATURE AT 160 SEC                           AT 160 SEC l

D ,   ,
MOMENTUM VECTORS OF TWO-DIMENSIONAL DEBRIS QUENCH qu                                   .
: h.                                     k ll i
: k.                                      .
h1
: k.                                         [,,
Ik
: h. .                                i      f 6.. .                          $\/      4
    ,                          b..                                    t:
: h. .                              b,'
: b. .                            M. t, h                               b...,          j q                                      ...
ll                          ==t...         l
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f...
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                            ,    ilt.                            ..+...
ll[                               .i..-
                            ,    11g
                                                                     **.'     \
                              'l 4... l l4.                             .
                                                                     +...
1 14                                   s..
4L                                 42              1 GAS MOMENTUM VECTORS                         LIQUID MOMENTUM VECTORS  i AT 95.0 SEC                         AT 95.0 SEC       i VGMAX = 6.82 M/S                      Vgx = 3.5 M/S       j 1

CONTOUR PLOTS OF TWO-DIMENSIONAL DEBRIS QUENCH O
o t9 4a                               .+ w
           .300 i
l 400                           10 0.0
           .500                           12 5.0    7 a                                I l
J                              x J .
4 O
o o
v VOID FRACTION CONTOURS                  BED TEMPERATURE AT 95.0 SEC                         AT 95.0 SEC i
1

MOMENTUM VECTORS OF TWO-DIMENSIONAL DEBRIS QUENCM l'
4u                                     }al!,.
I ..                                       l d
I)k 5~                                               tv
                                                         )   ,
lh A
                                                         )          l f
: k.                                             f (                   c A          l(
k..                                        .
                                                                           \
: k.                                               ,,
: k. .                                           t L                                   A.'4          ,6
: h.                                   b...,
                       *..                                 m ...
                   ,    IL.
e...
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t...
p..                                    ..t...
f...
t...
t
: t. . .                                        f...
                                                               ..t...
                                                               .,t ee.
4M                                      4M GAS MOMENTUM VECTORS              LIQUID MOMENTUM VECTORS AT 300 SEC                           AT 300 SEC VGMAX = 6.77 M/S                     VMAX     3*                  M/S

CONTOUR PLOTS OF TWO-DIMENSIONAL DEBRIS QUENCH O
O M
4-                                ,-
e i
      .400 10 0.0
      .500 {  i 12 5.0  7 I
l
    .1000 tR:8 4-                                4-O o
VOID FRACTION CONTOURS                 BED TEMPERATURE AT 300 SEC                         AT 300 SEC

?
DEBRIS QUENCM ONE-DIMENSIONAL EXPERIMENT                BED TEMPERATURE             = 500*C RUN NO. 4                                 WATER TEMPERATURE           = 100*C CROSS-SECTIONAL AREA = 0. 01767 M 2 OVERALL RESULTS                                  MIRTM            EXPERIMENT QUENCM FRONT SPEED, CM/SEC                        0.175                 0.23 2
DOWNWARD QUENCM RATE KW/M                         773.0               836.0 i
COMPARISON OF QUENCM RATES 2
KW/M TIME (SEC)                       MIRTM 1-D                MIRTM 2-D 100                             890                        3736 i                  200                             779                        2419 300                             773                        1945                       l MAAP USE OF FACTOR OF ~2+ JUSTIFIED.
l I
                                                           --      ,- -                 ~-  - - - - - - -

l                                           E 1
e i
.1 i
ZION STATION IDCOR TASK 23.1            ,
INTEGRATED ACCIDENT ANALYSIS M. A. Kenton IDCOR/NRC Meeting on         l Integrated Analysis of Severe Accident Containment Loads i
Rockville, Maryland May 15-17,1984             l 1

                                                    .o ZION TMLB' ACCIDENT DEFINITION e Loss of all AC power and auxiliary feedwater at time O.
1 o Loss of component cooling water leads to f ailure of main coolant pump seals at .75 hours (about 150 gpm total).

                                                                                              /t
                                                                                              /
s UPPER
                                                                                              /
s COMPARTMENT
                                        /                                                    /
STEAM GENERATORS IN BROKEN AND                  -          O' UN8ROKEN LEGS                  <                                  e:                '
                                                                                             /
PRESSURIZER In c7d
[ /
                                                                      /'/
                                                                        "       bc   9       /
                                                                                             /         LOWER ANNULAR                         -                         /            :
                                                                                     ,'      /         COMPARTMENT COMPARTMENT                     -                                         -
PRIMARY SYSTEM
                                        /     _          r      ;
OUENCH TANK
                                                \
V      // /////////]
CAVITY Y''/' .N///i-l]            :       /
{
l UPPER COMPARTMENT l                                                                                            ...
                                                      ...ut .
                                                            ,we=r eone                                               .............,
6o.a.               l Coup..fus=7             ,
                       ;  ~".~."..~"                                                    .........
                       '   ...                                                                            :    aun.c ,...
c"'"                                         ...'..                                                  .....'......
f                ,
I                                         l
                                                                                           .-                   ..e s a w.  =
                                                         ..,u..,
      .                                                   SYSTEu
                                 '''                                                                       '    3.ca t. 6g 3 6
j Sft.u of.g..r3
                            ..........          L-           .....
J   .........m.,
                               . . . .                                                      .-               ,,uo.....,e..,
                                                              .                  i
(              PWR Large, Dry Containment Nodalization s

l                                                ,
                                                     ,o l
l l
ZION TMLB' EVENT 

==SUMMARY==

TIME                      EVENT O sec     Loss of all AC and auxiliary feedwater 45 min    Pump seals f all 58 min    S/G inventory 10 percent 1.5 hr    Rupture disk falls 2.0 hr    Core uncovered 3.7 hr    RV falls 8.9 hr    Cavity dry 10.8 hr   Lower compartment floor dry 22 hr     Containment falls l
l

i i
ZION TM_B-
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BW M
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l l                                                                                         t i

l e      e i
ZION TILB-m -.
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6 5
ZION Tit _B-n   g    .IN-CORE               . IN-CONTMT O     :
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* 6 ZION TM_B-a     e_
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{                                       TIE HR N                                                                             )    .

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710N TM_B-
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i ZION TM_B-
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(
UNCERTAINTY / SENSITIVITY ANALYSIS FOR TMLB' - ZION Selected Results i
l Base      20%     80%   Cut-Off Case    Molten   Molten 2500 K Time of Containment Failure      22.2      24.7    21.7    22.1 (hr)
Fraction Clad Reacted            .075      .068    .075    .083 Time Between Vessel and          18.6      21.6    17.2    18.4 Containment Failure (hr)
G P

OVERVIEW 0F CONTAINMENT LOADS WORKING GROUP M. Silberberg, NRC

l . ~.
10~
CONTAINMENT LOADS WORKING GROUP (CLWG)
OBJECTIVE l       o    IDENTIFY RECOAO,4 ENDED MODELS FOR CONTAINMENT LOADING ANALYS1S o   DEVELOP ESTlWATES OF CONTAINMENT LOADING FOR SPECIFIC PLANTS o   INTERACT WITH CONT. PERFORMANCE WORK GROUP SCOPE o   SIX REFERENCE PLANTS o    IMPORTANT ACCIDENT SEQUENCES EARLY CHALLENGES BY THRESHOLD FAILURE OR BY-PASS EVENTS PHENOMENOLOGICAL UNCERTAINTIES APPROACH                                               i o  ANALYSIS OF STANDARD PROBLEMS BY NRC CONTRACTOR SPECl ALiSTS + OT'HERS o  REV1EW AND CONSENSUS BY WG

CLWG STANDARD PROBLEM PROCESS FLOW CHART COMENTS            NRC STAFF NRC STAFF PREPARES                                                       CLWG REVIEWS                                                                  CLWG
                                                                    ~                                                    :
DISTRIBUTES PREllHINARY                                                       PRELIMINARY                                                           ANALYZES SP SP                             SP                           FINAL SP ANALYZE
:         SPECIAL ISSUE f M.!A INPUT FROM LABS                                                                                                                                                     , ,
m IDENTIFY CLWG DRAFT                    _
CLWG DEVELOPS       ,        CLWG REVIEWS SP -    SPECIAL
                                                                                                    ~
REPORT                               SP CONSENSUS                   RESULTS             ISSUES i f REFINED ANALYSES SP L
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CONCLUSIONS i
FOR LARGE DRY CONTAINMENTS UNCERTAINTIES CAN BE NARROWLY BOUNDED AND " STEAM-SPIKE" TYPE FAILURES DO NOT APPEAR PHYSICALLY REASONABLE.
IN THE PRESENCE OF WATER C00 ABILITY OF CORIUM DEBRIS BEDS VERY LIKELY.
SUPPLY OF MODEST WATER QUANTITIES, OVER THE LONG TERM, CAN PREVENT CONTAINMENT FAILURE.
i r
Y I
e

g

==SUMMARY==
OF EFFORTS OF CONTAINMENT LOADS WORKING GROUP SUBCOMMITTEE ON DIRECT HEATING:
HIGH-PRESSURE MELT EJECTION CONTAINMENT LOADING PRESENTED BY T. GINSBERG BROOKHAVEN NATIONAL LABORATORY DEPARTMENT OF NUCLEAR ENERGY UPTON, NY   11973 NRC/IDCOR MEETING-0N INTEGRATED ANALYSIS OF CONTAINMENT LOADS ROCKVILLE, MD MAY 15, 1984 BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(IIll

SUBCOMMITTEE ON DIRECT HEATING:    MEMBERS K. BERGERON, SNL                    B. R. SEGHAL, EPRI W. B0HL, LASL                       J. SIENICKI, ANL D. CHO, ANL                         B. SPENCER, ANL M. CORRADINI, UNIV. OF WISCONSIN    D. SQUARER, EPRI T. GINSBERG, BNL                    T. G. THE0FANUS, PURDUE UNIV.
M. PILCH, SNL                       D. WILLIAMS, SNL D. POWERS, SNL                      R. WRIGHT, NRC (CHAIRMAN)
BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(ILll

TODAY'S PRESENTATION e BACKGROUND e SUBCOMMITTEE OBJECTIVES e THE DIRECT-HEATING PROBLEM e SCOPE OF EFFORT e MAJOR PHENOMEN0 LOGICAL UNCERTAINTIES e CONTAINMENT LOAD PARAMETER ESTIMATION PROCESS e GENERAL STATEMENT OF SUBCOMMITTEE VIEWPOINTS e QUANTITATIVE RECOMMENDATIONS:     TWO VIEWS e 

==SUMMARY==

i BROOKHAVEN Nail 0NAL LABORATORY l} gj l A5500ATED UNIVERSITIES, INC.(Illl

BACKGROUND e INITIAL CONSENSUS ON LARGE DRY PWR STEAM SPIKE ANALYSIS (THE0FANUS) e SNL HIGH-PRESSURE EXPERIMENTS e DIRECT HEATING OF CONTAINMENT ATMOSPHERE e CHEMICAL REACTION ENERGY e PARAMETRIC CONTAINMENT LOADS STUDIES e FORMATION OF SUBCOMMITTEE ON DIRECT HEATING l
BROOKHAVEN NATIONAL lA8 ORATORY l} g)l A5500ATED UNIVERSITIES, INC.(llli l
l

SUBCOMMITTEE OBJECTIVES e FOCUS ON HIGH-PRESSURE MELTDOWN SEQUENCES IN LARGE DRY PWR'S:
STANDARD PROBLEMS 1 & 2 (ZION, SURRY) e ASSESS APPROACHES TO PREDICTION OF LILELIH00D OF PRIMARY SYSTEM DEPRESSURIZATION PRIOR TO                        I MELT EJECTION FROM VESSEL                                          )
e ASSESS APPROACHES TO PREDICTION OF CONTAINMENT REPONSE TO HIGH-PRESSURE MELT EJECTION SCENARIO e RECOMMEND TO NRC A PROCEDURE FOR CALCULATION OF CONTAINMENT RESPONSE DURING HIGH-PRESSURE DIRECT                   i HEATING SEQUENCE OF EVENTS l
i i
BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(Illl

THE DIRECT HEATING PROBLEM e  PREDICT CONTAINMENT LOAD:
g:-'                                                                              d)_;.
HIGH PRESSURE BLOWDOWN OF CORE MELT INTO CONTAINMENT ATMOSPHERE l                                                                                                                     AND - AS A CONSEQllENCE 4
THERMAL AND CHEMICAL ENERGY TRANSFER TO C                         C                                                    -
WATER (IN P0OLS AND ATMOSPHERE) l                                                   -
r                                                                                AND TO l                 [ _1                                                             1
: m.                           cc' w-CONTAINMENT ATMOSPHERE (DIRECT HEAT)
                                                                                     ~
e SP-1 AND SP-2 BASE CASE INITIAL CONDITIONS i                                                        )].
Ghu   ;--
               -.                   v         ;,;_.-l'.Q ~~.. "'- 1.;:.yk5)
TUNNEL M
: b. :c;,7.,f;.g.-lggi;
                                . f                 % REACTOR BROOKHAVEN NATIONAL LABORATORY l} g} l CAVITY                                                                                          -
A5500ATED UNIVERSITIES, INC.(Itll a

i                                                                                              .
;                        Energy Balance for Dispersed Debris Scenarios                       .'
i I
!                                                                "0UENCH" IN i                                                                ATMOSPHERE j
Direct Heat Chemical Energy         Unreacted j
V' Radiate to Reacted                                        Heat Sinks l
                                                  -       i j                                                   y Quench in Pool Sticks to Structures Thermal Energy                                To Concrete
;                 To Cavity l
i Quench in Pool l                                                                                  -

l DIRECT HEATING WHY LOOK AT PROBLEM AT ALL7 e CALCULATIONS SHOW THAT DIRECT HEATING ALONE PRODUCES 2 5-3 TIMES THE AP AS DOES STEAM PRESSURIZATION e J.E ALL AVAILABLE METAL WERE OXIDIZED, CHEMICAL ENERGY SOURCE WOULD BE FROM 0 5 To 1 7 TIMES THE MELT THERMAL ENERGY (SP-1) e  INITIAL PARAMETRIC CONTAINMENT LOAD CALCULATIONS SHOWED THAT THE ABOVE CAN BE A PROBLEM UNDER SOME CONDITIONS.
i l
BROOKHAVEN Nail 0NAL LABORATORY l} g)l
;                                                  A5500ATED UNIVERSITIES, INC.(llll I
l l
i _ _

SCOPE OF EFFORT-1 e                EXPERIMENTS
                                                    - EXAMINED ANL/SNL TEST RESULTS e               PARAMETRIC SCOPING CALCULATIONS e               MELT AND WATER DROPLET TRANSPORT
                                                   - FRAGMENTATION PROCESSES
                                                   - EXAMINED PLANT-SPECIFIC PATHWAYS FOR MELT TRANSPORT AB0VE CAVITY
                                                   - EXAMINED PROPOSED MECHANISMS FOR MELT DROPLET DEPOSITION ON STRUCTURES
                                                   - MELT VS. PATER TRANSPORT BROOKHAVEN Nail 0NAL IABORATORYl} g)l A5500ATED UNIVERSITIES, INC.(Illl
   , - - - - , , - - -     , -       , , . , , , . , - - - - . - - - - ~ _ . . . . - - - - . - . - - _ . _ _ - . - . . _ . _ _

SCOPE OF EFFORT-2 e                  CHEMISTRY
                                  - CONSIDERED RANGE OF POTENTIAL CHEMICAL ENERGY SOURCE MAGNITUDES
                                  - CALCULATIONS OF 02-LIMITED CHEMICAL REACTION IN WELL-MIXED ATMOSPHERE e                  ENERGY PARTITION IN ATMOSPHERE
                                  - INITIAL EFFORT TO EXAMINE MELT-WATER MIXING AND HEAT TRANSFER IN ATMOSPHERE l
l l
BROOKHAVEN Nail 0NAL LABORATORY l)l~gl A5500ATED UNIVER5lilES, INC(llll

MAJOR PHENOMEN0 LOGICAL UNCERTAINTIES e MELT TRANSPORT
        - EFFICIENCY OF MELT EJECTION FROM CAVITY IN SP-2 (SURRY)
        - DISTRIBUTION, DEPOSITION AND FALLOUT AB0VE CAVITY e CHEMISTRY
        - P0TENTIAL FOR H2 RECOMBINATION WITH 02 IN AIR-STEAM ENVIRONMENT
        - AVAILABILITY OF 02 FOR METAL REACTION e ENERGY PARTITION IN CONTAINMENT ATMOSPHERE
        - SUSPENDED WATER / MELT HEAT TRANSFER COUPLING IN ATMOSPHERE
        - LOSSES TO STRUCTURES, ETC.
BROOKHAVEN NATIONAL LABORATORY l} lj l ASS 00ATED UNIVERSITIES, INC.(Illl l

CONTAINMENT LOAD PARAMETER ESTIMATION PROCESS e  GLOBAL CONTAINMENT ENERGY BALANCE CALCULATION e  SPECIFY ON BEST JUDGMENT BASIS:
     - FRACTION OF CORE MELT STORED THERMAL ENERGY TRANSFERRED TO WATER
     - EXTENT OF METALLIC CHEMICAL ENERGY RELEASE OF MELT QUENCHED IN WATER
     - FRACTION OF CORE MELT STORED THERMAL ENERGY TRANSFERRED DIRECTLY TO CONTAINMENT ATMOSPHERE
     - EXTENT OF METALLIC CHEMICAL ENERGY RELEASE OF MELT IN ATMOSPHERE (STEAM OR OXYGEN) e  ESTIMATES:    " LOW," "BEST ESTIMATE," "HIGH" i
I BROOKHAVEN Nail 0NAL LABORATORY l} g3l A5500ATED UNIVERSITIES, INC.(Illl l
I

l l
Table 1 Sample Parameter Table Best                                                                   l Low                  Estimate                                          High                   !
Thermal   Chemical     Thermal                            Chemical      Thermal     Chemical          l l
Direct Heat                                                     x%                                y% 02 uater Quench                                                    w%                                z% STM BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC. (E lll u       . _ _ _ _ _ _ _ _ _ _ _ _ _ _

I GENERAL STATEMENTS OF SUBCOMMITTEE VIEWP0INTS GROUP A CONCLUSION:
l CURRENTLY AVAILABLE DATA SUPPORT' DIRECT HEATING EFFECTS INVOLVING NO MORE THAN 2% OF THE EJECTED MATERIAL STORED THERMAL ENERGY.
GROUP B CONCLUSION:
IT IS NOT POSSIBLE AT THIS TIME TO RULE OUT OCCURRENCE OF SUFFICIENT DIRECT HEATING TO PRESENT A SEVERE CHALLENGE TO CONTAINMENT.
BROOKHAVEN NATIONAL iABORATORYl} g)l A5500ATED UNIVERSITIES, INC.(llll l

QUANTITATIVE RECOMMENDATIONS e  SUBJECTIVITY e GROUP B:
            - RANGE OF ESTIMATES
            - LARGE UNCERTAINTY
            - SHOW REPRESENTATIVE ESTIMATES BROOKHAVEN NATIONAL LABORATORY ASSOCIATED UNIVERSITIES, INC(1lll I
l

Table 4 Group A:   Parameters'for SP-1 and SP-2 Best low                  Estimate                 High Thermal   Chemical     Thermal     Chemical     Thennal    Chemical Direct
        <2%
                   ---          <2%         ---         <2%           ---
Heat Water s80%        ---         480%        ---         s80%           ---
Quench BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNNERSITIES, INC.(llll

Table 5-1 1
Representative Parameters for SP-1 Group B:
Best Low                  Estimate                    High Thermal      Chemical     Thermal    Chemical       Thermal      Chemical 0           0         27.5%      50% STM        50%         50% 0 Heat                                                                        2 100%          0         72.5%      30% STM        50%         30% STM Quench Table 5-2 Group B: Representative Parameters for SP-2 Best Low                   Estimate                 High Thermal      Chemical     Thermal    Chemical       Thennal     Chemical 5%         50% STM      25%        50% STM        50%         50% 0 Heat                                                                          E Wa er 85%                       75%        30% STM        50%         30% STM Quench BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES,INC.(IIll

==SUMMARY==

e SUBCOMMITTEE REVIO'5D AVAILABLE EXPERIMENTAL EVIDENCE AND ANALYTICAL ARGUMENTS e ARRIVED AT DIVIDED OPINION e DRAFT REPORTS DISTRIBUTED FOR COMMENT e NEED EXPERIMENTS WITH APPROPRIATE SIMULANT AB0VE-CAVITY STRUCTURES AND WITH WATER TO HELP RESOLVE DIFFERENCES l
l l
BROOKHAVEN NATIONAL LABORATORY l}l)l A550 GATED UNIVERSITIES, INC(l(Il l

k LIXELIHOOD OF HIGH PRESSURE SCENARIOS W. LYON, NRC i
i 1
t l
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4

HIGH PRESSURE BLOWDOWN OF MELT THROUGH THE LOWER VESSEL HEAD IS CONSIDERED LESS LIKELY THAN A LOW PRESSURE MELT-THROUGH. THIS IS DUE TO A COMBINATION OF MULTI-D NATURAL CONVECTION AND FISSION PRODUCT MIGRATION.
THESE RAISE THE TEMPERATURE OF OTHER RCS PRESSURE BOUNDARY COMPONENTS AND INCREASE THE LIKELIHOOD THEY WILL FAIL.
CALCULATIONS MUST INCLUDE THESE PHENOMENA TO CORRECTLY COMPUTE TIME AND POSITION OF FAILURE.
l i.
I

CONSIDERED:  VESSEL ONE-DIMENSIONAL (10), MULTI-DIMENSIONAL (MD);
EX-VESSEL 10, MD; F.P. MIGRATION; INSULATION; HYDROGEN; OTHER CHEMICAL BEHAVIOR; CORE MELT; COMPONENTS; STRUCTURAL; EXPERIMENTAL.
PARTICIPANTS:
ANL EPRI INEL NYPA PNL PURDUE RMA SAI SANDIA

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0.6 -
* RECIRCULATING FLOW : CORE /EX-CORE                                                                          -
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(131ND dW31 SYS*9'S

RCS PRESSURE B0UNDARY WILL FAIL AT LOW TEMPERATURE PUMP SEAL ~500F OTHER ~1300F NATURAL CIRC. WITH NO RCS LOOP SEALS:    MAX RCS AT = 250-500 F (W TYPE)
WITH LOOP SEALS, U-TUBE SG. (W TYPE): HOT LEG TO CORE AT IS
  ~500 TO 1200 F DUE TO MULTI-DIMENSIONAL FLOW.
I EXPECT B&W TO BE SOMEWHAT HIGHER.
PORV WILL TEND TO REDUCE AT.
BUT:
TMI AN ANOMOLY-NOT FULLY UNDERSTOOD (ALSO NOT " PURE" ACCIDENT) l l
l l

I RESULTS PRELIMINARY, ALTHOUGH THINKING IS SOUND.
THEREFORE, ALTHOUGH WE BELIEVE HIGH PRESSURE MELT THROUGH OF THE L.P.
IS LESS LIKELY THAN PREVIOUSLY THOUGHT...I DO NOT BELIEVE IT CAN BE ELIMINATED FROM OUR CONSIDERATION TODAY.
CODE DEVELOPERS HAVE A LOT OF WORK TO DO TO CORRECTLY CALCULATE:
TIMING FAILURE LOCATION PERTURBATIONS l
l

g 9 AND WORK IS NEEDED TO REOUCE UNCERTAINTY IN REGARD TO OUR MAJOR CONCLUSION.
I I
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                                            )
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4 SEQUOYAH NUCLEAR PLANT
,                   IDCOR TASK 23.1 INTEGRATED ACCIDENT ANALYSIS M. A. Kenton IDCOR/NRC Meeting on integrated Analysis'of Severe Accident Containment Loads Rockville, Maryland May 15-17,1984 i

[
l l
SEQUOYAH S2HF ACCIDENT DEFINITION l
e Small (2 inch diameter) break at time O.
e Attempt to switch to recirculation line-up falls when RWST level reaches 4.5 feet due to either:
Case 1: Valve failure in system; spray water in recirc sump.
Case 2:   Failure to remove drain plugs in re-fueling pool, spray water held in refueling pool.
Y

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UPPER COMPARTMENT seen
                 -e                                                             - - ;=:-                                            .?.'*"e'',
a-aveneens ICS CoasettetSA                                                        A8HeULAN                                               s.eemans UPPER PLeseges                                                    C048PASThitmf                                              e-easessee,se,4,e,s e *e                                                                   e*.*e
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LOWSA ConsPARTesSNT I
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PReetARY                                                               P#E00U#3IEE ST8 Teel
                                         ..                       e  _
GROEEttLEG
                                                                                                                         **'"                 ~
STE Ats OEseER ATOR
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u .ROae. Lame,
                                                                                  .                                                               .T.A .....A,0...         1 1
;           .                                                                                                                                                              1
)                                                                                                                                                                          l l                PWR Ice Condenser Containment Nodalization l
l l

SEQUOYAH S2HF (DRAINS OPEN)
EVENT 

==SUMMARY==

TIME                       EVENT 0 sec       Break occurs O sec       Reactor scram 1 min       Main coolant pump trip 23 min      Switch to recirc falls; injection lost 1.3 hr      Core uncovered 2.1 hr      First burn (upper pienum) l 4.4 hr      RV fails 4.7 hr      Ice depleted 8.9 hr      Containment falls

i                 SEQUOYAH S2HF (DRAINS PLUGGED)
EVENT 

==SUMMARY==

TIME                                                        EVENT 0 secs             Break occurs O secs             Reactor scram 1 min              Main coolant pump trip 23 min             Switch to reciro f ails, injection lost 1.3 hr              Core uncovered 2.1 hr              First burn (upper plenum) 4.4 hr              RV falls 4.5 hr              Ice depleted 5.6 hr             Cavity dry 18 hr              Containment f ails

t     f SEQUOYAH S2HF n   II.I -
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f.........l........l.........l.........l....m .I O. I                                                                          2         3  4     5 TIE HR i
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W W
e e
W f    ... .....l.. ... ..l.........l.........l....                           % ...I O.          1        2                                               3 4       5 TITE HR I
                                                                                              ]

4      9 S         S SEQUOYAH S2HF a
e_
O
      =
                             ~
      -*                     b Z                      -
b
      &n                     E.
6 N
W                      !
O                       -
b U                       :                 [
k" b
f                     .. .... .I.... ...l   .......iI. .......l...     .....I O.          1          2          3         4                  5 TITE HR
                                                                                           - Y

o      .
SEQUOYAH S2HF/ DRAINS OPEN o-
     .O
      =
* N     [
L
     @ CD J
           ~
T<                                                               -
            ~
W -
4     -
2 e : -
Yn . " .
f    . . . . . ili s e s in i s ili s i n i i n iili n i n e i n ili s i . . ..iilis i ni s iiil . ii . . li n i ii, m iliiii . . . . li e ,iii, i I O.              1           2             3             4           5               6              7               8              9               10 ,
l TITE HR

SEQUOYAH S2HF/ DRAINS OPEN
* n O     -
 ==    .
X    .
O W
m 19 19 W
le IS WU
*'* m M
M Im W
Im 19 m
f 1........l.........l.. ......l. .... ..l. . . . l.........l.n.. ...l...... .l.. ... . l...... ..I
: c.       I          2        3         4        5        6         7        8         9       10 TIPE HR

i
  )
J 4
SEQUOYAH S2HF/ DRAINS OPEN
         "   e- -
          .o     -
x     :
w     -
          ^     _
Ee    __
i 1
,I              _                                                                                              l
 !           e  _
v     .
N m     -
U  n W     :
i f5,,,,...l.........l.........l........l.........i.........i........l........l........,C.........I
: o.        I        2         3        4         5         6        7         8       9       10 TIPE HR                                             I 1
l 1
l l
l 1

  -w                                                                w SEQUOYAH S2HF/ DRAINS OPEN 8_ W W
M up m
W i    ==     -
W m
m W
M M
m M
M BW Q  .
m se W
W      -
se O
J     L_ -
W I
            =
1 l          =
[      niin a i r e lin e sini si!n ein s i nin ln ni s tie nili nii n einilt ri e n tiin!i a t t e in n e lsin a n n t s els in ni n e r e lmie n sis e s!
0               1               2             3               4               S               6             7              8              9          10 TIPE HR i

0 o      .
SEQUOYAH S2HF/ DRAINS OPEN re  g    e,I fE51t CSFT                     .LPPER CMPT C.
m ki    1 f1 L m
       >=    2 4     :
I h
F      :
th    ~
_     )
       <m (3
hh hl hlkkkllll llllllllj gllgllggg jggjggggg gg g g g O.           I             2           3           4           5            6 7   8     g to TIPE HR i
l I

SEQUOYAH S2HF/ DRAINS OPEN a   8?  _
oa m
N     =
N   -
b T
Z"      :                                                                                   i 7                                                                                   i E       E w8, ZO
    ==      .
I f     n.n...linn....l.n.....li......l  . nili.ni nliii..in.In.nniil nn .nl......nl i          o.        1      2       3        4     s       s      7       8      s        to l
TIE HR l
l

SEQUDYAH S2if/ DRAINS Ct.OSED oN X
               ~
N  -
b In :
J  _-
m       :
N 5
2-      :
l-      :
u-@   .
o :.
2,,      .I             l.......     .I         ,,,i f      .       .......                      ,,..
O.        S            10             15            20 TIPE HR l

SEQUOYAH S2W/ DRAINS CLOSED                  .
    .o.      :
N       I i
I a.=  m   '
    >=      -
a u
r:                        sf
                              /
         .     ,, , ,. ,, , ,   l,,,,,   .  .I,,...,,,,!,,,,,,,,,l C.                  S            10              15      20 TIME HR i
(

SEQUOYAH S2W/ DRAINS CLOSED
* e_*
O.
N    -
W p
l e                                                     1 m
O m
CD J     -
W W
u
     ==     -
O m
W 4
W m
            ~
I,,,,,,,. l>          ,I,,,,,,,,,I f    ,,,,,,,,                  >,,,,,
: c.          5            10            15          20 TIPE HR

                                                                                                      .o
(
l SEQUOYAH S2iF/ DRAINS CLOSED l "    o-O*
   ==
m  L I                                                                                   '
le     i e,n W  L 5      i v      :
u      :
b
          ~
             *I  I I   I I I    I I     I I I   I I t 3 I g     3 g 3 g g g g    g g  l g g ,       g O.                          5                      10                      15           20 TIPE HR l
l l

I l
SEQUOYAH S2W/0 RAINS CLOSED a      o .g LENER COWT       .UPPUt CtMPT O*
         =
t X      "
m        -
l,1. O "-
                                                                           )
b 5
W  L
                                                        -T-5 E
m        :               b_   -
                              ..J A
_V l
U*                S                 10       15        20 TIPE HR
(
                                                                         )

SEOUOYAH S7tf/ DRAINS CLOSED a    e ,-
o" :.
X     :
5      i b
F-     :
I      :
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b W
5 N      -
I*     7 A       :
b       .
Ui N   .
2      ;
f     .......    .I.........l...      .....l.........I
: o.             5           10                  15                                   20 TIME HR

                                                                     ~
UNCERTAINTY / SENSITIVITY ANALYSIS l                  FOR S2HF (DRAINS OPEN) - SEQUOYAH            >
Selected Results i
Base    20%    Cut-Off   Exit Temp. Drag
!                       Case  Molten   2500 K      150*F    Coeff. 1 Time of Contain- 9.1       9.7       8.7       10.5      9.2 l
l     ment Failure (hr)
I    Fraction Clad       .21     .21      .22        .19      .20 Reacted i
i    Time of Ice De-     4.4     4.8      5.0        6.3      4.3 plation (hr) l    Time Between        5.0     6.6      4.0        6.2      5.3 Vessel and Con-l     tainment Failure (hr) i i

[
f ICE CONDENSER PWR STANDARD PROBLEM APPROACH e TMLB' SEQUENCE SELECTED AS BASIS FOR ANALYSIS.
1 e MARCH 2 USED TO DETERMINE IN-VESSEL HDYROGEN GENERA. TION AND RELEASE TO CONTAINMENT, e  INDEPENDENT ANALYSES CONDUCTED BY SANDIA NATIONAL LABORATORIES AND BATTELLE'S COLUMBUS LABORATORIES TO DESCRIBE EFFECTS OF HYDROGEN BURNS.
e ANALYSES INCLUDED CONSIDERATION OF:
VARIOUS IGNITION CRITERIA EFFECTS OF VARYING IN-VESSEL HYDROGEN PRODUCTION RATE OF HYDROGEN RELEASE FROM PRIMARY SYSTEM
      - MAGN I r eJOE o t= s rf.4M        S PING i
C4BaHelle               )
(                                       -     -             -
c., _ ,.,<. .,....., y

ICE CONDENSER PWR STANDARD PROBLEM BCL APPROACH e ICE CONDENSER CONTAINMENT MODELED AS FOUR COMPARTMENT SYSTEM.
e BURN ROUTINE IN MARCH 2 USED TO EVALUATE BURNING OF HYDROGEN AS WELL AS CARBON MONOXIDE.
e MACE ROUTINE IN MARCH 2 USED TO DESCRIBE CONTAINMENT PRESSURE AND TEMPERATURE RESPONSE.
                                                                                 ~
OBaHelle cei e., <. ,.. ,,,, y 

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ICE CONDENSER PWR STANDARD PROBLEM SNL APPROACH e HECTR CODE USED TO DETERM.INE CONTAINMENT RESPONSE, INCLUDING HYDROGEN BURNING, 4 do dvRN/N6 8 ICE' CONDENSER CONTAINMENT MODELED AS NINE INTERCONNECTED VOLUMES.
e MutTICOMPARTMENT TREATMENT OF ICE CONDENSER, INCLUDING MECHANISTIC STEAM CONDENSATION.
                                                                              ~
OBattelle            -
(                                                            <ei_,.,use,..,.,y.

l l
I
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[                                                                                       l at 1 - comparta.nt sumder 1 - Flow Junction Number
                                                        - - One-way Flow Junction
                                                       ~-       Two-way Flow Junction HECTR Ico-condens er conta in men t model.                            l

h h MARCH 2.0 CHRON0 LOGY FOR CLWG ICE CONDENSER CONTAINMENT STANDARD PROBLEM EVENT                                 TIME (S)
: 1. STM GEN DRY                            3885
: 2. CORE UNC0VER                           5550
: 3. START MELT                             7350
: 4. CORE SLUMP                             8640
: 5. START HEAD HEATUP                      8745
: 6. BOTTOM HEAD FAIL                       9465
                             ~

m                 ..- , .        a     u. ., _.
9          .
MARCH 2.0 CHRONOLOGY FOR CLWG ICE CONDENSER CONTAINMENT STANDARD PROBLEM EVENT                                                                                       TIME (S)
: l. STM GEN DRY                                                                                  3885
: 2. CORE UNC0VER                                                                                 5550
: 3. START MELT                                                                                   7350
: 4. CORE SLUMP                                                                                   8640
: 5. START HEAD HEATUP                                                                            8745
: 6. BOTTOM HEAD FAIL                                                                             9465 l
l l

cLwo so=6 case o.06 MFlRCH Input.
7000              .
e   .
                 .p 6000          -
O           ..
O           -
(m    5000 -
3             .    .
3O    4000 g            ..
G         ..
[
3000 --
O          --
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(f)         .    .
    @    2000       -
y           ..
m                -
1000 -      -                                                                 -
4     .
l           l           l         l 0,           ,
e r
t 4
8 9460           9470        9480        9490      9500        9510       9520 TLme      (seconds)

cLwg son 6 case o.06                               ,
MARCH Input.
20             .    ,   ,
17.5 -- ..
g          ..
g                                                                                                  .
g         ..
15 -      -                                                                                -
o s.3        ..
3                                                                                                   .
m               .
c) 12.5 -      -
J          -    -
O         -    -
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e    10 -     -
o             .
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7         .    .
2.5 -..
o
                   .     ., .               ,   s,          ,      ,     ,     .     .   .
9460          9470     9480                      9490         9500        9510         9520 TLme        (seconds)

cLwg son 6 case c.06 Compartment 1 350                    .   .
                            .    .                                                                                    m
                            .    .                                                                                    g
                            .    .                                                                                    m 300 -                -                                                                                -
                           .     .                                                                                    g
                           .     .                                                                                    su
      -     250 -               -                                                                                 -
O                  -    -                                                                                    -
Q_                  .     .                                                                                    .
      .X                  .     .
[
p 200 --
O                 _    .                                                                                     .
U) g                 .     .                                                                                     .
(_                .     .                                                                                    .
Q-     150 -              -
                                                                           -- J                                  -
as    .                                                                                    g 100 -             -                                                                                  -
                        . p                                                                                     .
                                     .         h                                   h 0                3000        6000                    9000              12000       15000 TLme             (seconds)
I l
t

cLwg so#6 case c.06 Cornpartment 4 600      . .
i  .    .    ,  .  .
i   .   .
c      --                                                            -
. .)
-.3 M
m O      ..                                                            .
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3 0
t      ..                                                             .
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M 200      : :   l  :    :    l   :  :    l   :  :         :  :
0        3000         6000       9000        12000        15000 TLme   (seconds)

cLwo son 6 case a.06 v            .                           .
Ice RemOLnLng 1 __                  .           .
0.9 ..-                                                                                        .
v>            .    .
: 0. 8 --
U)                                                                                                -
O             --
r               -:                                                                                      :
0      0. 7 --                                                                                    -
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0       0.6 - -
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: 0. 4 --
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0.1 --                                                                                    -
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: 0.         :   :
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:      l  :   :
l
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0                3000              6000              9000            12000         115000   l l
i Ilme (seconds)

CLWG ICE CCNLEhSER STANDARD FROBLEM II ERTING BASED ON SNL CASE 0.06 - NO BURNING COMPARTMENT                                                           TIME AFTER NO                DESCRIPTION                                   VESSEL BREACH 1                  DOME
                        < 4.1% HYDROGEN             UNTIL                    3  SEC 2                  IC UPPER PLENUM s     5% OXYGEN               SY                     17  SEC 1 55% STEAM                   BY                     18  SEC 9                 TOP ICE REGION 1    5% OXYGEN                BY                      6  SEC 1 55% STEAM                   BY                     10  SEC 8                 NEXT-TO-TOP ICE REGION s       5% 0XYGEN              BY                      l  SEC 1     55% STEAM                BY                      3  SEC 7                 NEXT-TO-BOTTOM ICE REGION 6                 BOTTOM ICE REGION 3                  IC LOWER PLENUM 4                 LOWER COMPARTMENT 1       5% 0XYGEN 1     55% STEAM S                 DEAD-ENDED REGION 1       5.3% OXYGEN

Ta:-le 2. Case itsers;ticr,s ar.: ?.nar, cf "iss. s Feat Leadirts                             Fretsrc Ccee Case In '.'essel Steas Ignition Typee                       Nuater of Bcns by Ces;artment           Pressare.             Teaserature.              Fressee.
No.       Zr Cnid. S;ite acd Limit (!H21                  1 2 3 4 5 6 7 8 9 LFa (Ccs;t.4! F(Ccapt.ll 15                                               LFa     IH2
    .........- -           ................................._                      .......... ..=---             ..- ---            = =......................
0.00         49.41 H ;t. VB+5s                  4.1       2 3 0 0 1 0 0 1                        4 60; (1-3.6-9) 14C0 (2)                 546         5:s     4.1 2.000 4 9. 4*.      Hi;h W. sect. 10.0                    1 8 0 0 1 0 0 2 : 667 fall)                                     1667 (2)        88:         :06     6.8 0.01         4?.41 Hic O!e.:!C                  8.0       1 7 0 0 1 0 1 2 6 729 (1.2)                                     16:E (2)        789         ::8     8.0 0.02         49.4% High 01 all                  8.0       1 10 0 0 1 0 1 9 17 574 (all)                                   !!77 (2)        79;         30      5.1 0.020        49.4*  High Nanaech, 12.0                    1 8 0 0 1 0 0 2 4 652 (411)                                     1762 (2)       1837         MO      6.6 C.0:c 49.4% High Noratch. 30.0                            0 1 0 0 0 0 0 0 1 ;5: talls                                     2045 (2)        42          ---     ---
0.06         49.a; High None                100.0         0 0 0 0 0 0 0 0 0 33: tall)                                      4:7 (4)        418         ---     ---
C.07         49.41 High V8                      8.0       2 3 0 0 0 0 0 1                        1 618 (all)              2132 i:1        476         toi     4.1 0.08         49.4*  High VD5s                   8.0       1 2 0 0 0 0 0 1 3 535 tall)                                     1:05 (:)        475         2:8     6.1 C.09         49.41 High '<B+10s                 8.0       1 2 0 0 0 0 0 2 2 659 (all)                                     1460 (91        512         :%      8.0 0.10         49.4! High V8+20s                  8.0       1 2 0 0 0 0 0 t                        i   E50 (1.2)            1505 (B)        500         286    11.0 R.00         49.41 Los VB+5s                    8.0       1 2 0 0 0 0 0 1 3 455 (all)                                     1305 (2)        474         228     6.1 S.01         9".E:  Los Vl+5s                   8.0       1 3 0 0 1 0 0 2 2 704 (1.2)                                     20r. (2)       1077         244     8.7
    !.0          ??.S*  Lou Vb!hr                   8.0       1    1 0 0 1 0 0 1                     1 1459 (1.2)             2h9 t!)        1810         246    22.7 1.01         39.4*  High Vi+5s                  8.0       2 2 0 0 0 0 0 0 1 60: (1,2)                                     1440 (2:        470         455     4'.1 U.03         ;9.41 Lce VB+5s                    8.0       1 2 0 0 0 0 0 0 3 407 (all)                                     1440 (2)        450         *:a     4.6 V.00         af.:, Miar, Vir5s                  8.0       0 0 0 0 0 0 0 0 1 1051 :1.2)                                    1677 (!)        506         24E    14.5 V.01         48. : High C! all                  8.0       1 8 0 0 0 0 0 5 8 661 (1-3.6-9) 1804 (2)                                        521         :*:     8.0 e D!estC--Geliberate Ignition except is Ice Cse;artsents ar.d Icner pierun D! all- Deliterate Igr.ition in all ccepartnerts VE --Vessel Breach 1
l l

ICE CONDENSER PWR STANDARD PROBLEM TMLB' SEQUENCE WITH DEBRIS QUENCH EVENT                         IIME, MIN.
STEAM 6ENERATOR DRY               61.0 CORE UNCOVER                      93.2 START MELT                       115.8 START SLUMP                      138.2 CORE COLLAPSE                    139.8 HEAD FAIL                        149.8 i
CONCRETE ATTACK                  339,5 i                                                                                                                                       ICE MELTED                       404.2 l
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(                ICE CONDENSER PWR STANDARD PROBLEM TMLB' SEQUENCE WITH DEBRIS QUENCH HYDROGEN BURNS FOR H20N = 0.08 TIME, MIN.   ~ COMPARTMENT    H2 BURN, LB. PRESSURE, PSIA 139.5               3             26.4       22.1      24.1 140.1               3             16.2       24.4      25.5 140.6               3             12.1       25.8      26.6 142.8               3             41.0       26.6      29.8 144.3               3             20.5       29.6      30.3 148.8               3             35.0       28.1      31.1 148.9               3             16.4       31.7      32.8 169.1             3&4            639.3       34.2      76.1 174.9               3             37.7       51.8      54.0 179.5               3             32.2       46.2      48.0 184.0               3             30.6       43.7      45.3 190.2               3             29.9       41.4      43.1 198.3               3             27.0       39.8      41.2 211.1               3             28.0       38.4      40.1
                                                                                    ~
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ICE CONDENSER PWR STANDARD PROBLEM l                       TMLB' SEQUENCE WITH EARLY CONCRETE ATTACK EVENT                               TIME, MIN.
!                                 STEAM GENERATOR DRY                     61.0
:                                CORE UNCOVER                            93.2 START MELT                            115.8 START SLUMP                           138,2
 .                                CORE COLLAPSE                         139.8 l
l HEAD FAIL                             ll49.8
 !                                ICE MELTED                            529,2 l
fleBaHelle
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ICE CONDENSER PWR STANDARD PROBLEM TMLB' SEQUENCE WITH EARLY CONCRETE ATTACK HYDROGEN BURNS FOR H20N = 0.08.
TIME, MIN. COMPARTMENT     H2 BURN, LB. C0 BURN, LB. PRESSURE, PSIA 139.5               3              26.4                      22.1      24.1 140.1               3              16.2                      24.4      25.5 140.6               3              12.2                                25.6 142.8               3              41.0                                29,8 ,
144.3               3              20.5                                30.3 148.8               3              51,4                                32.8             -
151.7             3&4             313.3              35.0    35.0      53.0
 !      165.8               3              38,0                      35.2      35.9 179.1               3              16.4                                32.7 189.1               3              21.1                                33.2 201.5               3              24.7                                35.3 286.0             3&4             472.9           4,191. 40.9      92.7 299.9               3              19.0             375. 52.5      55.9 403.1               4             258.0           5,919. 51.3     105.0
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ICE CONDENSER PWR STANDARD PROBLEM

==SUMMARY==
OF ICE CONDENSER PWR STANDARD PROBLEM RESULTS IGNI TION AK PRESSURE, PSIA
;   IN-VESSEL        STEAM        THRESHOLD        IGNITION Za 0xIDE, %      SPIKE          V/O HZ        Time / TYPE    MARCH        HECTR 49            HIGH           NONE            NONE          30            48 i       49            HIGH              8             DI           76           104 49'           HIGH*             8*            DI*          68*           95*
49            HIGH              8             VB           76            88 49'           HIGH*             8*            VB'          70           150*
49            HIGH              8         VB+5 SEC         77            77 49            HIGH              8         VB+20 SEc        82           122
;       49            HIGH            10          NON-MECH.        80            95 49             Low              8             DI           53 49             LOW              8         VB+5 SEc                       70 49             low            10          NON-MECH.        48 l    100            HIGH              8             VB          135 100             low              8         VB+5 SEC                      101 39            HIGH              8         V8r5.5ed,                      86 39             LOW              8         VB+5 SEc                       58
 ;      49            HIGH            4.1         VB+5 SEC                       86 EARLY RELEASE OF H2 FROM PRIMARY.
                                                                                             ~
                                                                                               ~
!                                                                      OBaHelle m .       ., _ , j ..

ICE CONDENSER PWR STANDARD PROBLEM DIFFERENCES BETWEEN MARCH AND HECTR MODELING OF ICE CONDENSER CONTAINMENTS e MARCH IS LIMITED TO SERIES FLOW PATHS -- ALL FLOW FROM LOWER COMPARTMENT TO THE DOME PASSES THROUGH ICE CONDENSER.
e HECTR MODELS SERIES AS WELL AS PARALLEL FLOW PATHS -- FOR ICE CONDENSER THIS INCLUDES SOME FLOW FROM LOWER COMPARTMENT TO DOME THAT~ BYPASSES ICE BED.
* MARCH TREATS ICE CONDENSER IN THE JUNCTION BETWEEN TWO COMPARTMENTS, CONDENSING ESSENTIALLY ALL STEAM PASSING THROUGH IT.
e HECTR EXPLICITLY MODELS THE ICE CONDENSER COMPARTMENT, INCLUDING MECHANISTIC MODEL FOR STEAM CONDENSATION.
e MARCH INTERCOMPARTMENT FLOWS MODELED ON BASIS OF PRESSURE EQUILIBRATION.
e HECTR MODELS PRESSURE DRIVEN AND BUOYANT FLOWS, ACCOUNTING FOR FLOW RESISTANCES AND ONE-WAY ICE CONDENSER DOORS.
OBattelle c.,_s,<.-,...,..,y

SENSITIVITY TO EXTENT OF IN-VESSEL OXIDATION FOR CLWG ICE CONDENSER CONTAINMENT STANDARD PROBLEM CASE            IN-VESSEL             PEAK        PEAK NO            ZR OXID.            PRESSURE     TEMPERATURE 1               (KPA)          (K) 39.4 U.00                                  407          1440 i
R.00              49.4                488          1305 l
S.01              99.8                704          2022 l
l 4
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o             .
I 5                                              SENSITIVITY TO STEAM SPIKE MAGNITUDE FOR CLWG ICE CONDENSER CONTAINMENT STANDARD PROBLEM PEAK                                           PEAK CASE                      IN-VESSEL                                                       PRESSURE                                       TEMPERATURE No                     7R 0xfD.                                  I'40T                   (KPA)                                            (K)
U.00                               39.4%                                           2        407                                             1440 T.01                               39.4%                                100                 602                                             1440 R.00                               49.4%                                           2        488                                             1305 0.08                              49.4%                                100                  535                                             1305 I
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PEACH BOTTOM ATOMIC POWER STATION
!                       IDCOR TASK 23.1               ;
.             INTEGRATED CONTAINMENT ANALYSIS 1                          J. R. Gabor i
l IDCOR/NRC Meeting on Integrated Analysis of Severe Accident Containment Loads Rockville, Maryland May 15-17,1984 k

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a           MAIN STEAM UPPER PLENUM              ,   ]               =
* STEAM ECCS -                                                  TO ECCS SPRAYS                                   /(                 UTURBlNES u
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DOWNCOMER                                                                  FW,HPSW (OUTER SHROUD)-                  :
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BWR Mark l Containment Nodalization

_                      =
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'                                                                           1 DRYWELL                              I 2
WETWELL PEDESTAL CAVITY
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1                    PEDESTAL W ALL 2                    DRYWELL W ALL 3 7////,             DRYWELL FLOOR 4(i                  TORUS ROOM W ALL Mark l Containment Heat Sinks                                                        !
k                                                                                                      l

4 BWR OPERATOR INTERVENTION CONDITIONS 1         Primary system pressure 2         Pedestal pressure 3         Drywell pressure 4         Wetwell pressure 5         Pedestal temperature 6         Drywell temperature 7         Wetwell temperature 8         Vessel water level 9         H   mole fraction in wetwell 10         Suppression pool temperature 11         Problem time 12         Event code status change 13         Suppression pool water level l
Y

i PEACH BOTTOM - S E Sequence Definition i                                        2 l
e Steam line break (0.1 ft ) at time = 0.
e Failure of all core make-up.
e CRD is available.
e Suppression pool cooling available at 10 minutes.
l e Ultimate containment pressure limit assumed to be 132 pala.
* Ultimate containment temperature limit assumed to be 1200'F.
l l
l

PEACH BOTTOM - S,E EVENT 

==SUMMARY==

TIME                   EVENT 0   Break in steam line (0.1 ft )
6.8 sec Reactor scrammed 67 see  MSIVs closed, feedwater tripped 10 min  Suppression pool cooling on 1.2 hr Automatic depressurization on (ADS) 1.3-7 hr Core uncovered and reflooded with CRD flow CRD flow ceases 15 hr 18 hr  Top of core uncovered 23 hr  Start of core melt 24 hr  Vessel f ailure 31 hr  Containment f ailure (overtemperature)
I l

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           <       :                 Containment Failure                                                                                                              :
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 !         $5       I Vessel Failure                                                                                              } _
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SMALL LOCA -- PEACH BOTTON 8
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1 1

MAAP-BWR UNCERTAINTY / SENSITIVITY                                              -
l ANALYSIS PARAMETER SET Paraenter                                                                Befault   Minieue   Minieus
;                   ahnber Nees          Peraester Bescription                       Ibtits     value     value     Uslue 1   FRCDEF  Frictiuse coefficient for corius in UFAIL                   5.000E-03 1.000E-03 1.000E-et 2   FMMCP   Fraction   of total core mass which must sett to fail the core piste                                      0.200 i
0.100     0 440
!                      3   MTRLAS  Fuel sAanmel to centrol blade heat transfer coefficient                                       W/e882-K   50.0      5.00      500.
4   NTFB    File boilias heat transfer coefficient            W/e482- K  300.       100. 400.
5 FELOCK    Fuel channel bles:kans model switch 0=use blockaan modelf1= turn blockade model off             0.000E400 0.000E400   1.00 l                     4   TZXF F  Suidation cut-off temperature                     Kelvin    2.300D 03 2.0950 03 2.500Ef03 7 FacPF     Fraction of area of core plate failed                                                         i 0 300     1.000E-02   1.00 l                     8   Caert   Flame buemanew dras coefficient in the redestal              5.00     0.500      10.0 9 C900M     Flame buswanew dras coefficient in the drwuer!!              5.00     0.500      10.0 10   CSSIAf  Flame buswanew dras coefficient in the uutuell               5.00     0.500      10 0 31   CSOCA   Flame buemanew dras coefficient in comparteent A             5.00     0.500       10.0        '
12   CSOCS   Flame buswanew dras coefficient in cuerartment 8             5.00     0.500      10.0 13   XCM     Cerium reference thereal boundarv laver thickness seter     0.100     0.100       1.00 1                     14   NTOER   Corium-crust heat transfer coefficient in BEC0fr  W/e882- K 1.000D03   500. 5.000D03 15   XCast   Nanteue certue thickness on drwwell floor and
'l                                 pedestal floor (Nork II ontv)                     meter     5.000E-02 5.000E-03 0 100 16   X3CNEP  Paricle size (diameter) for corium as at falls i
into suppressian pool (MarkII ontv)               meter     1.000E-02 1 000E-02 0.100 17   TCFLAft critical flame teeeerature                        Kelvin     983.      900. 3.200003 I                     18   FCNils  Churn-turbulent cratical flow parameter                      1.53       1.00     3 00 19   FBROP   Broplet critical flou parameter                              3.70      3.00      5.00 20   FFLOSS  Floodina flow parameter                                      3.00      2.00      4.00
!                     21   FSPat   Parameter for bottua-sparsed stese void fraction             1.00      3.00      4.00 1
22 FUOL      Parameter for volume source void fraction oudel              2.00      1 00      4.00 l                    23   TTENTR  Entratneent effect*ve emetwsnt time               secuswi   0.500     0.100      10.0 24   EU      Emissivitw of water                                         0.900     0.000      1.00 l
25   EtR. Emissivity of wall                                          0.950     0.700      1.00 26 ECN       Eelssivitw of cortue                                        0.350     0.700      1.00 27 EG        Emissivitw of sas                                           0.600     0.500      1.00 1
28 EED       Emissivitw of eeuspoent-                                    0.850     0.700      1.00 29   FOUER   Frutaan of core serav flow a!!oved to tNPass cure           0.500     0.000E+00  1 00 i                     30 aff       Number of Penetration failed in lower head                   1.00      1.00      10.0 31   FCBC3W   Dounconer perimeter per meter from pedestal door Olart Il ente)                                               2.00      1.00      5.00
,                     32  FCHF     Coeffacient for CHF correlation in PLSTN                    0.140     0.120     0.300
.                     33 FC99RK    Diuharse coeff acaent for Pire break                        0.750     0.100      1.00 1                    34   FENTR    nultaplaer for kutateladze criterion for cavity blouout (GT 1.0 = dafficult3LT 10 = easser)                 0.330     0.200      100.
4 35  SCALU    Scalans factor for all burnins velocities                    1.00      1.00      103.
I                     34   SCALH   Scalans factor for heat transfer coefficients to passive heat sanks                                           1 00     0.500      10.0
!l t
37  FuMIN    Manteue burn velocitu                             e/s        1.00     0.500      10.0 1
1 i

UNCERTAINTY / SENSITIVITY ANALYSIS FOR TQVW - PEACH BOTTOM
!                                 Sequences Analyzed Base Case           e   Station blackout.
I                               e   20% core melt required to fall core plate (FMAXCP = .20).
e   With core blockage (FBlock = 0).
e   Zircaloy oxidation cut-off tem-perature, TZOOFF = 2300 K.
Case #1            e   Same as base case, except FMAXCP = .40.
Case #2
* Same as base case, except FBlock = 1.
Case #3
* Same as base case, except TZOOFF = 2500 K.
l l
l

I UNCERTAINTY / SENSITIVITY ANALYSIS FOR TQVW - PEACH BOTTOM Results Base       Case                                                              Case Case Case         #1                                                               +2    +3 Time of Contain-                                          16.45      16.62                                                           16.70  16.44 mont Failure (hr)                                                                                                                            !
Fraction of Zr                                              .06        .06                                                               .07  .065 Reacted in-Vessel Time of Core Un-                                            8.4        8.4                                                               8.4   8.4 covery (hr)
Hydrogen Gener-                                              690.       750.                                                             780. 750.
ated at Vessel Failure (Ibs)
Time of Vessel                                               12.36      13.33                                                          12.43   12.36 Failure (hr)

BLACKUUT - -- PEACH BOFTOM                                    CFMAXCP:.1) n               .s
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1 l
I ey -- ---,---.--,--,y.-                                            . -

          -e          O BLACKOUT -- PEACH B0fION                                (FMAXCP= 4)
               "        m          .................m                    .,.
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0   0 70 0.40 0.60 0 80          1    1.2        1.4             1.G      18        2.0 IINE HR                                              xlO '
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BLACKOUT -- PEACH BOTTON                                                                    (FBLOCK:1)
        "       a                   ..... ... p.:                .....p . ...                    ,. . . ..... ,......                       ... .             .,.     ...p.......p......_
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    .__                    ~.                 .                      ._.                    __                                              -_.              _
BLACKOUT -- PEACH BOTTOM                                                            CFBLOCK:1)
                       .n g            .
                                                    ... p ...... ......          ,,. .......,, .......,........ g                     .;.....p.;..                 ...p...... . p ......
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O.                O.20 0 40 0.60 0.80                                                       1     1.2                1.4                1.6                18       2.0 TIME HR                                                                        xlO '

BLACKOUT -- t'EACH HOTTOM                                                                       i TZOOFF=4040) n     1 o_       - .......i.........,........., . . ....i... .. .i... .... 3......... p........                                               .........i........ .
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: o. 0.20 0.40 v.60 0.80                                                       1                          1.2       1.4       1.6             18          2.0 TIME IR                                                                                  x10' l

HLACKOUT --- PEACH HOTTOM                                    (TZOOFF=1010) e          n l
                                      ..... J.iii iiiig ........gi. . .. igii .. ...j,. . ..       ,g.   ...... g........,g........                       g.,,,,,,,,,
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BLACKOUT -- PEACH BUTTOM                     (TZOOFF 4040)
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BLACK 00I -- PEACH BOTTOM                                  (NPF=10)
           ,       .8
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                                                  ,       f                 i,    .!    - f.,.  ,,l.     !.                   '          3 0      0.70 0.40 0.60 0.80                                    1   1.2       1.4   1.6                 18         2.0 TIME HR                                           xlO '

BL.ACKOUT --- PE ACH BOTTOM                 (NPF=lO)
* 8
                  .  .o ,,,,,,u og. . . .  , .,;. ...  .g.... .;. o.m.         ...i,u... u3m ,,,,,,,,
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1 I

e       O o       se 1
:                                                                                      BL ACKui> T -- PEACH BOTTOM                                                (NPF=lO)
)             9 i            , ., - .  ,-                      ... p...u. o p o o                                                        u.;.. .. . . p ..... s.,.  .p .    .o u g u..o u p u.o o p o .....
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0                     O.20 0.10 0 60 0.80                                                                                     1 1.2          1.4                      1.G                           1.3                         2.0 TIME HR                                                                                    x10 '
l l
l i

o
                                                                                                 ])(l Y AN OVERVIEW 0F THE ACTIVITIES OF THE CONTAINMENT LOADS WORKING GROUP (CLWG)
RELATED TO THE BWR MKI AND II STANDARD PROBLEMS PRESENTED BY W. T. PRATT BROOKHAVEN NATIONAL LABORATORY llPTON, NEW YORK          11973 PRESENTED AT THE NRC/IDCOR MEETING ON INTEGRATED ANALYSIS OF SEVERE ACCI. DENT CONTAINMENT LOADS ROCKVILLE, MARYLAND MAY 15-17, 1984 BROOKHAVEN NATIONAL LABORATORY l)
A5500ATED' UNIVERSITIES, INC(Illl WTP:2.1

OllTLINE e OBJECTIVES e APPROACH e BACKGROUND e DEFINITION OF STANDARD PROBLEM e CALCULATIONAL METHODS e BWR MARK I STANDARD PROBLEM e RWR MARK II STANDARD PROBLEM e SPECIAL ISSUES:
e   INTER VS. CORCON l              e   MARK I CONTAINMENT FAILURE VIA l                  CORIUM ATTACK e 

==SUMMARY==

BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(ILll WTP:2.2

o     .
OBJECTIVES OF THE CWG e TO MECHANISTICALLY MODEL CONTAINMENT BEHAVIOR llNDER SEVERE ACCIDENT CONDITIONS e TO SYSTEMATICALLY ADDRESS A NilMRER OF STANDARD PROB-LENS APPLICABLE TO REPRESENTATIVE PLANTS FOR THE SIX CONTAINMENT TYPES UNDER CONSIDERATION e FOR EACH STANDARD PROBLEM THE GROUP WILL:
e  ESTABLISH STANDARD METHODOLOGY WHERE POSSIBLE e  PROVIDE A BROAD CONSENSUS VIEW 0F AREAS WHERE CALCULATIONS CAN BE PERFORMED WITH CONFIDENCE e  IDENTIFY WHERE UNCERTAINTIES EXIST AND PERFORM SENSITIVITY STUDIES BROOKHAVEN NAll0NAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(I til WTP:2.3

APPROACH e STANDARD PROBLEMS SELECTED TO ADDRESS ACCIDENT PHENOMENOLOGY WITH POTENTIALLY SEVERE CONTAINMENT LOADING:
e   SELECTION OF PROBLEMS BASED ON INSIGHTS t
GAINED FROM EXTENSIVE ANALYSES BY NRC CONTRACTORS AND INDUSTRY e EACH SAMPLE PROBLEM ANALYZED BY SEVERAL DIFFERENT ORGANIZATIONS e RESilLTS ARE COMPARED IN OPEN FORilM WITH EXTENSIVE t
PEER REVIEW i
e CONTAINMENT LOADS THEN PROVIDED TO CONTAINMENT PERFORMANCE WORKING GROUP (CPWG) l l
l                                          BROOKHAVEN NATIONAL LABORATORY l ASSOCIATED UNIVERSITIES, INC.(IIll WTP:2,4

e   '.
(
BWR WITH A MARK I CONTAINMENT e  AVAILABLE DOCUMENTATION:
o  PEACH BOTTOM REACTOR SAFETY STilDY (WASH-1400)
ASTP0 STUDIES (BMI-2104) e  BROWNS FERRY SASA WORK AT ORNL STATION BLACK 0llT (NilREG/CR-2182)
LOSS OF DHR (NUREG/CR-2973)
LOSS OF INJECTION (NUREG/CR-3179)
ATWS (NUREG/CR-3470) 1 BROOKHAVEN Nail 0NAL LABORATORY l}lgl A5500ATED UNIVERSITIES, INC.(1til WTP:2.5

BWR WITH A MARK Il CONTAINMENT l
e AVAILABLE DOCllMENTATION:
e   LIMERICK GENERATING STATION - PROBABILISTIC RISK ASSESSMENT (LGS-PRA)
REVIEWED BY BNL (NllREG/CR-3028) e   LGS - SEVERE ACCIDENT RISK ASSESSMENT (INCLllDES EXTERNAL EVENTS)
REVIEWED BY BNL (NUREG/CR-3493)
BROOKHAVEN Nail 0NAL LABORATORY l} gy l A5500ATED UNIVER5 tiles, INC.(1til WTP:2.6

POTENTIALLY DOMINANT ACCIDENT SEQUENCES                            1 i
e  ATWS WITH CONTINUED COOLANT INJECTION (IDCOR SEQUENCE TC) e  TRANSIENTS WITH LOSS-OF-0ECAY-HEAT REMOVAL (IDCOR SEQUENCE TW) e  TRANSIENTS AND LOCA's WITH LOSS-OF-COOLANT MAKE-IIP (IDCOR SEQUENCES TQVW AND SIE)
BROOKHAVEN NATIONAL LABORATORY l} ggl A5500ATED UNIVERSITIES, INC(1til WTP:2.7

ATWS WITH CONTINilED COOLANT INJECTION e   REACTOR DOES NOT SCRAM e   COOLANT INJECTION CONTINilES e   S!!PPRESSION POOL STARTS R0ll.ING e   CONTAINMENT FAILilRE OCCllRS RAPIDLY DilE TO STEAM PARTIAL PRESSilRE e   COOLANT INJECTION FAILS e   CORE MELTS INTO A FAILED CONTAINMENT e   UNCERTAINTY RELATED TO PRIMARY SYSTEM T/H AND NEUTRONICS e   CONTAINMENT RESP 0NSE CALCULATIONS STRAIGHT-FORWARD BROOKHAVEN Nail 0NAL LABORATORY l} g)l A5500ATED UNNERSITIES, INC.(1lll WTP:2.8

TRANSIENTS WITH LOSS-OF-DECAY-HEAT REMOVAL e  REACTOR SCRAM SUCCESSFUL e  COOLANT INJECTION CONTINUES e  SUPPRESSION P0OL STARTS BOILING e CONTAINMENT FAILURE OCCIIRS LATE (DAYS) DIlE TO STEAM PARTIAL PRESStlRE j
e COOLANT INJECTION FAILS e CORE MELTS INTO A FAILED CONTAINMENT e CONTAINMENT RESPONSE CALCIILATIONS STRAIGHT-FORWARD BROOKHAVEN Nail 0NAL LABORATORY l}lyl A5500ATED UNIVERSITIES, INC(ILll WTP:2.9 l

TRANSIENTS WITH LOSS-0F-COOLANT MAKE-IlP e    REACTOR SCRAM SUCCESSFill e   COOLANT MAVE-IIP FAILS e   CORE MELTS INTO INTACT CONTAINMENT e   SilPPRESSION POOL SilRC00 LED e   CONTAINMENT FAILS DUE TO HIGH PRESSURES i                   DR TEMPERATURES DURING CORillM/ CONCRETE INTERACTIONS l             e     llNCERTAINTY ASSOCIATED WITH CORIUM/ CONCRETE i                    INTERACTIONS e     HENCE, SELECTED AS BASIS FOR STANDARD PROBLEM BROOKHAVEN Nail 0NAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(1Ul
! WTP:2.10 l
l  _       __ _        _ - _ _ _ _

POTENTI AL CONTAINMENT FAILIIRE MODES e                    FAllllRE MODES CONSIDERED:
e   FAllllRES INI)UCED BY HIGH PRESSURES OR TEMPERATURES e   FAILURES INilllCED BY BASEMAT PENETRATION e                   FAllllRE MODES NOT CONSIDEREll:
e   STEAM-EXPLO".10N INDilCED FAIL llRES e  HYDROGEN BilRN INiillCED FAILilRES i
BROOKHAVEN Nail 0NAL LABORATORY l} gyl      l A5500ATED UNIVERSITIES, INC.(1til WTP:2.11 i

DEFINITION OF STANDARD PR0RLEM e FOCllS OF MARK I AND MARK 11 STANDARD PROBLEM:
e   PRESSilRE/TEMPERATilRE RESPONSE 1)llRING CORillM/ CONCRETE INTERACTIONS e ISSUE TO BE ADDRESSED (BY CPWG):
e   MODE (OVERPRESSURE VS. TEMPERATURE) AND TIMING OF CONTAINMENT FAILURE e SENSITIVITY STUDIES:
e   INITIAL CORIUM TEMPERATURE o   ZlRCONillM, STEEL, AND l10           MASS IN CORIUM e   METAL OXIDATION IN-VESSEL e   EX-VESSEL CORIUM DISPERSAL e   CONCRETE TYPE BR00KHAVEN Nail 0NAL LABORATORY l)lyl ASSOCIATED UNIVERSITIES, INC. (1 til WTP:2.12

CALCllt.ATIONAL METHODS e ORNL:
e   MARCH 1.lB (INTER llSED TO M01)EL CORillM/
CONCRETE INTERACTIONS) e BCL:
e   MARCH 2 (WITH MODIFIED INTER) e BNL/PilRilllE:
e   MARCH 1.1B (STAND ALONE CORCON N0D 1) e MARCHl.1 (STAND ALONE CORCON MOD 1) e   MARCH 2 (STAND ALONE CORCON M0I) 1) l                                                         e SANDIA:
e  MARCON (MARCH 2 LINKED WITH CORCON MOD 2 Pl.llS OTHER Mal)lFICATIONS)
BROOKHAVEN Nail 0NAL LABORATORY lg gyl A5500ATED UNIVER5till5, INC.(llli l   WTP:2.13

1 I
RWR MARK I STANDARD PRORLEM e PLANT GE0 METRY l
e PHENOMENGLOGY AND VISIIALIZATION OF PROCESS e COMPARISON OF CALCllLATIONAL METHODS F
!                                e SENSITIVITY STilDIES:
a   CORillM SPREADING e   CONCRETE COMPOSITION e   CORillM TEMPERATilRE i
I BROOKHAV[N Nail 0NAL LABORATORY l3lgl A5500Ai[D UNIVERSITIES, INC.(llli i
WTP:2.14 l

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a j                                                                                              Mark I (TQUV) 1               la            lb            2                  3    3a                                4 l        Corium Spread (m)                                                5                                           3                  5                                      3
:         Debris Temp (,F)                                              4130                                       2700         4130                                 2700 i         Concrete Type                                                    L                                           L                  B                                      B i         Free H,0 (%)                                                     3              6                            3                4      8                                4 i         Steel in Corium (ib)                                          140K                         85K            140K        140K                     _
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I f   .           .
l OBSERVATIONS REGARDING MARK I STANDARD PROBLEM l
l e  WITH CORIUM EX-VESSEL AND WITH0llT OPERATOR ACTION, CONTAINMENT INTEGRITY WILL EVENTilALLY BE LOST e  MODE AND TIMING 0F CONTAINMENT FAllllRE SENSITIVE T0:
e    INITIAL CONDITIONS e   PLANT SPECIFIC FEATllRES e MODELING ASSUMPTIONS BROOKHAV[N Nail 0NAL LABORATORY l}lyl ASSOCIAi[D UNIV [R5lil[5, INC.(1lll WTP:2.17

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;                                            Failure                                                                                                          VOLUME NO.1 i
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l _-_       _ _ _ _ _                                                                                                                           _         _                   _.

SENSITIVITY OF CONTATHMENT RFSPONSE MODELING ASSilMPTIONS o CORCON VS. INTER i
e  llPWARD HEAT TRANSFER FROM TOP 0F CORIUM e CONVECTIVE MIXING e
e  DEGASSING OF CONCRETE BROOKHAVEN Nail 0NAL LABORATORY lgl}l ASSOCIAi[D UNIVERSITIES, INC.(1Ill WTP:2,24

nWR MARK 11 CONTAINMENT BROOKHAVEN
                             ~~              Nail 0NAL LABORATORY l}
ass 00ATED UNIVERSITIES, INCU       ;
WTP:2,25

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WTP:2,26

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c RWR WITH A MARK II CONTAINMENT e  WETWELL (AND SUPPRESSION P0OL) DIRECTLY llNDER-NEATH DRYWELL (AND REACTOR VESSEL) e  DIAPHRAGM FLOOR SEPARATE WETWELL AND DRYWELL e  WATER CANNOT ACCUMULATE ON DIAPHRAGM FLOOR e  HENCE, INITIAL CORE / WATER INTERACTIONS WILL BE LIMITED e SilBSEQUENT ACCIDENT PROGRESSION DEPENDS ON HOW CORE HATERIALS PASS THR0llGH DIAPHRAGM BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED lJNIVERSITIES, INC. (I El I WTP:2.28

BWR WITH A MARK II CONTAINMENT (CONT.)
e  IF MAJORITY OF CORE MATERIALS REMAIN ON DIAPHRAGM FLOOR, EXTENSIVE CORE / CONCRETE INTERACTIONS OCCIIR e CONTAINMENT INTEGRITY WILL BE CHALLENGED BY LONG-TERM PRESSilRE AND TEMPERATURE BilILD-llP e IF CORE MATERIALS RAPIDLY PASS THR0llGH DIA-PHRAGM FLOOR INSIDE PEDESTAL WALL, RAPID STEAM GENERATION WILL OCCllR e STEAM GENERATION RATES MAY CHALLENGE CONTAIN-MENT INTEGRITY e COMBUSTION OF COMBUSTIBLE GASES PREVENTED BY INERTING l
l BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC(1lll WTP:2.29

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            .h Mark II (TQUV) 5 Sa        Sb    Sc Sd    6    7      7a       8 Corium Spread (m)                                                5                          3    5                3 t                                                              Debris Temp ( F)                                              4130                       2700 4130           2700 L    B               B l                                                              Concrete Type                                                    L 3  6                       3    4       8       4 Free H90 (%)                                                                                                 140K Steel in Corium (lb)                                          140K          85K          140K 140K 0                 25 50    0    0               0 Pool Losses (%)
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l OBSERVATIONS REGARDING MARK II STANDARD PROBLEM e   DUE TO LARGER VOLUME THE MODE AND TIMING OF CONTAINMENT FAllllRE IS LESS SENSITIVE T0:
e INITIAL CONDITIONS e MODELING ASSilMPTIONS e   HOWEVER, MODE AND TIMING OF CONTAINMENT FAILilRE IS SENSITIVE T0-l e  PLANT SPECIFIC FEATIJRES l
l                                              BROOKHAVEN NATIONAL LABORATORY l)l)l l                                                  A5500ATED UNIVERSITIES, INC.(ILll WTP:2.35 1

SPECIAL ISSilES BROOKHAVEN NAll0NAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(llll WTP:2.36

SPECIAL ISSilES:   INTER VS. CORCON e INTER:   INTEGRATED INTO MARCH 2 CODE e CORCON:   STAND ALONE STATE-0F-ART CORIUM/ CONCRETE INTERACTIONS MODEL e DIFFERENCES:
e   DECAY HEAT DISTRIBUTION e   CHEMICAL REACTIONS, RATES, AND HEAT RELEASES e  FREE WATER TREATMENT e  CORIUM SURFACE RADIATION HEAT TRANSFER l
e DIFFERENCES WILL BE RESOLVED IN TIME FOR CONSENSUS REPORT                                                                  1 i
BROOKHAVEN NATIONAL LABORATORY l)l)l A5500ATED UNIVERSITIES, INC.(1L11 WTP:2.37

SPECIAL ISSllES:
MARK I CONTAINMENT LIHER FAllllRF VI A CORIllM ATTACK e  EXTENT OF CORIUM SPREAD DEPENOS ON INITIAL CONDITIONS e  CORIUM ER0 DES INTO CONCRETE e  BASALT ERODES OlllCKLY e  LIMESTONE ERODES SLOWLY e  CORIllM ATTACK IS FIRST MOLTEN, THEN SOLID e  BASALT CONCRETE CASE:    FAST EROSION VERTICALLY SLAG SOLIDIFIES ON LINER LINER FAILURE DELAYED s  LINESTONE CONCRETE CASE:    SLOW EROSION VERTICALLY l
SLAG DOESN'T INSilLATE LINER LINER FAILS EARLY BROOKHAVEN NATIONAL LABORATORY l l                                                    A5500ATED UNIVERSITIES, INC(Illl l
l WTP:2.38 1

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548.8 ft -  k.* ' / . 4 ,, t h.hi 4
drywell liner - concrete shield wall gap WTP:2.39
                                                                ,y

l

==SUMMARY==

e BASED ON STANDARD PROBLEM CALCULATIONS:
e  BWR MARK I, WITH CORIllM EX-VESSEL AND WITH-DUT OPERATOR ACTION, CONTAINMENT INTEGRITY WILL EVENTUALLY BE LOST e  BWR MARK II, EX-VESSEL INTERACTIONS AND CHAL-LENGES TO CONTAINMENT INTEGRITY DEPEND ON CONTAINMENT CONFIGilRATION e OUTSTANDING ISSilES WILL BE RESOLVED AND INCORPORATED INTO CONSENSUS REPORT BROOKHAVEN NATIONAL LABORATORY l} g}l A5500ATED UNIVERSITIES, INC.(Itil WTP:2 A0

CLWG STANDARJ PRO.BLE:V: 6
!. BWR MK ::I:: DI:TKSIOX :?LAYES 1
FMRC     -
R. Zalosh
!                        E. A. Ural LANL    -
J. R. Travis
;            SNL   -
A. L. Camp
!                        S. E. Dingman J. E. Shepherd I                        J. C. Cummings C. A. Hickox   g i
d K

i PCRPOS3                              -
l Calculate pressure and temperature loads and heat fluxes f or a range of postulated hydrogen injection rates.
e

j I
f l
i l
i l                                               GRAND GULF NUCLEAR STATION i                                                          (BWR MARK 111)
E L. 2 90'- 5 "
DOME i
CO NT AIN M E NT4m E L.2 6 7 '-9" l
S.L. E L.2 3 7'-9" UPPER POOL EL. 2 0 8'- 10 "
E L.18 4 '- 6 "
VESSEL S HIELD -              -                 _
E L.161 '- 10' WETWELL -        I:      E__
E L.14 2 '-4 "
D R.Y W E L L -              _
(j            '
PEDESTAL-                                               !       .I .              =                      - m E L.1 17 '-4 "
SUPPRESSION          g/yp#
2                   .
t               '
POOL              @    .
                                                                         . lb                  ..      '.5' .'. .                ' '/ /           E L.9 3 '-O" A _ - - - --- - _ _ _ _ _   -                            ___-------.------,,m--~_                                                 - ,.               ---_-..--------.-,---I

Table 2 Initial Conditions
:                                                                                                  (at start of H2 injection)
Parameter                                                                                                                                                     Value l                           Gas Temperature                                                                                                                                               120 F
;                           Pressure                                                                                                                                                      17 psia l                            Steam Mole Fraction                                                                                                                                          0.19 I                           H2 Injection Temp.                                                                                                                                            155 F                                            ,
Supp. Pool Temp.                                                                                                                                             155 F Other Surfaces                                                                                                                                                120 F j                            _____________________________________________
i i
l I
I l
t 1
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r

Table 3
.                                 Hydrogen Source Terms Case                                        Location                                      Release Rate                      Duration Ibm / min                         min A                           8 non-adjacent SRVs                                                   40                           75 plus 1 adjacent SRV B                           8 non-adjacent SRVs                                                  '75                           40 plus 1 adjacent SRV C                           8 non-adjacent SRVs                                                 100                            30 plus 1 adjacent SRV 1                           8 non-adjacent SRVs                                                   50                           30 plus 1 adjacent SRV 2                           8 non_ adjacent SRVs                                                100                            10 plus 1 adjacent SRV 3                           1 SRV                                                               100                            30
:     4                           1 SRV                                                                 50                           10 1     _________________________________________________________
i I
2

1 ASSUMPTIONS
: 1. Diff usion flames                                                    will exist f or postulated      ,
hydrogen            injection rates.
i
! 2. Hydrogen is released in a 10 f t diameter circle above each sparger.
; 3. Sprays are inoperable.
1
: 4. Drywell purge system is inoperable.
2 I

l 1
i I
l                                                       METHODS i
l
!                               FMRC and SNL1 - Experimental correlations for diffusion
)                                                        flame behavior l                               SNL2 - Lumped-parameter code (HECTR) l l
i LANL - Finite-difference code (HMS)
)
I I

l 4
COXVERS::0X FACTORS l
1 BTU /HR/FT**2       =  3.1546 W/M**2 101.325 kPa =   1     ATY       14.7 psi l
l 4
4

I l
i Case A Results                                                 )
l Parameter                                            FMRC      LANL                SNL1                      SNL2  ;
Peak Pres. (psia)
Maximum                                           29           --            --
25 Minimum                                           21           --            --
20 Best Estimate                                     --           --            --
22   l Peak Gas Temp. (F)
Maximum                                        1832            --            --
2420 Minimum                                       ---              --            --
458 Best Estimate                                 ---              --           1832                     1826 Peak Heat Fluxes (1000 BTU /hr/ft**2)
Drywell Wall - 10' Maximum                                     8.0              --           5.7                       7.9   i Minimum                                     2.9              --           2.9                       1.8 Best Estimate                               ---              --            --                        --
Drywell Wall - 20' Maximum                                     6.2              --           2.9                       6.3 Minimum                                     2.9              --           1.7                       1.6 Best Estimate                               ---              --            --                        --
Drywell Wall - 30' Maximum                                     5.6              --           2.3                       5.7 Minimum                                     2.9              --           1.5                       1.4 Best Estimate                               ---              --            --                        --
Wetwell Wall - 10' Maximum                                     1.0              --           4.8                       4.4 Minimum                                     0.7              --           1.6                       0.6 Best Estimate                               ---              --            --                        --
Netwell Wall - 20'
   ,           Maximum                                     0.73             --           3.0                       4.1 Minimum                                     0.5              --           1.3                       0.5 Best Estimate                               ---              --            --                        --
Wetwell Wall - 30' Maximum                                     0.52             --           2.3                       3.8 Minimum                                     0.3              --           1.1                       Q.5 Best Estimate                               ---              --            --                        --

e   .
                                                                                                              )
                                                                                                              \
Case B Results Parameter                                     FMRC           LANL  SNL1          SNL2 Peak Pres. (psia)
Maximum                                       30           --    --            28 Minimum                                       --           --    --            21 Best Estimate                                 --           27    --            24 Peak Gas Temp. (F)
Maximum                                    1832            --    --          2420 Minimum                                    ---             --    --           624 Best Estimate                              ---            2372  1832         1826 Peak Heat Fluxes (1000 BTU /hr/ft**2)
Drywell Wall - 10' Maximum                                 12.             --  10.8          12.0 Minimum                                 3.8             --   5.1           3.2 Best Estimate                           ---             12.   --            --
Drywell Wall - 20' Maximum                                 8.8             --   3.8           8.9 Minimum                                 3.8             --   2.1           2.9 Best Estimate                          ---             6.    --            --
Drywell Wall - 30' Maximum                                 7.7             --   2.8           7.6 Minimum                                 3.7             --   1.7           2.5 Best Estimate                          ---            3.5    --            --
Netwell Wall - 10' Maximum                                 1.9             --   6.7           7.0 Minimum                                 1.3             --   2.0           0.9 Best Estimate                          ---            1.5    --            --
Wetwell Wall - 20' Maximum                                 1.5             --   3.8           6.3 Minimum                                 1.0             --   1.5           0.8 Best Estimate                          ---            1.6    --            --
Netwell Wall - 30' Maximum                                 1.1             --   2.9           5.7 Minimum                                 0.7             --   1.2           0.7 Best Estimate                          ---            1.7    --            --

Case C Results Parameter                                                    FMRC          LANL                                   SNL1  SNL2 Peak Pres. (psia)
Maximum                                                      32              --                                 --
29 Minimum                                                      24              --                                 --
21 Best Estimate                                                --              25                                 --
25 Peak Gas Temp. (F)
Maximum                                                  1832                --                                 --
2420 Minimum                                                  ---                 --                                 --
730 Best Estimate                                            ---           2552                                   1832 1826 Peak Heat Fluxes (1000 BTU /hr/ft**2)
Drywell Wall - 10' Maximum                                              14.                 --
16.1 13.9 Minimum                                              4.3                 --
7.2   4.4 Best Estimate                                        ---
13                                 --               --
Drywell Wall - 20' Maximum                                              10.                 --                               4.7  10.1 Minimum                                              4.3                 --                               2.4   3.8 Best Estimate                                        ---           6.5                                     --              --
Drywell Wall - 30' Maximum                                              9.0                --
3.2   8.9 Minimum                                              4.3                --
1.8   3.5 Best Estimate                                        ---           3.5                                     --              --
Wetwell Wall - 10' Maximum                                              2.4                --
7.9   8.9 Minimum                                              1.6                --
2.3   1.1 Best Estimate                                      ---             1.6                                     --              --
Netwell Wall - 20' Maximum                                            2.1                  --                                4.6   7.9 Minimum                                              1.4                --                                1.7   1.0 Best Estimate                                        ---           1.7                                     --             --
Notwell Wall - 30' Maximum                                              1.5                --                                3.3   7.0 Minimum                                             1.0                 --                                1.3   0.9 Best Estimate                                      ---            1.8                                      --            --

Case 1 Results Parameter                                           FMRC                             LANL   SNL1                SNL2 Peak Pres. (psia)
Maximum                                             27                            --      --                  25 Minimum                                             --                            --      --                  20 Best Estimate                                       --                            --      --
22 Peak Gas Temp. (F)
Maximum                                          1832                             --      --              2420 Minimum                                         ---                               --      --
496 Best Estimate                                    ---                              --      --
1826 Peak Heat Fluxes (1000 BTU /hr/ft**2)
Drywell Wall - 10' Maximum                                      S.2                               --      --
9.2 Minimum                                       3.2                              --      --
2.3 Best Estimate                               ---                               --      --                  --
Drywell Wall - 20' Maximum                                      7.0                               --      --
7.0 Minimum                                      3.2                               --      --
2.0 Best Estimate                               ---                               --      --                  --
Drywell Wall - 30' Maximum                                      6.2                               --      --
6.3 Minimum                                      3.2                               --      --                1.7 Best Estimate                                ---                               --      --                  --
Wetwell Wall - 10' Maximum                                      1.3                               --      --                4.8 i
Minimum                                      0.9                               --      --                0.6 Best Estimate                                ---                               --      --                  --
Wetwell Wall - 20' Maximum                                      0.93                              --      --                4.4 Minimum                                      0.6                               --      --
0.5 Best Estimate                                ---                               --      --                  --
Metwell Wall - 30' Maximum                                      0.67                              --      --
3.8 Minimum                                      0.4                               --      --                0.5 Best Estimate                                ---                               --      --                  --
i l
i

Case 2 Results Parameter                                                     FMRC   LANL   SNL1      SNL2 Peak Pres. (psia)
Maximum                                                    28      --    --
28 Minimum                                                    --      --    --
21   ,
Best Estimate                                              --      --    --
24   l Peak Gas Temp. (F)
Maximum                                                 1832      --     --
2420 Minimum                                                ---        --     --
701     1 Best Estimate                                          ---        --     --
1826 Peak Heat Fluxes (1000 BTU /hr/ft**2)
Drywell Wall - 10' Maximum                                             14.        --    --
13.9 Minimum                                             4.3        --    --
4.4 Best Estimate                                       ---        --    --     --
Drywell Wall - 20' Maximum                                             10.       --     --
10.1 Minimum                                            4.3        --     --
3.8 Best Estimate                                       ---       --     --     --
Drywell Wall - 30' Maximum                                            9.0        --     --
8.9 Minimum                                            4.3        --    --
3.5 Best Estimate                                      ---        --    --      --
Wetwell Wall - 10' Maximum                                             2.4       --     --    8.2 Minimum                                             1.6       --     --
1.0 Best Estimate                                       ---       --     --     --
Wetwell Wall - 20' Maximum                                            2.1        --     --    7.0 Minimum                                             1.4       --     --
0.9 Best Estimate                                      ---        --     --     __
Wetwell Wall - 30' Maximum                                            1.5        --    --     6.0 Minimum                                            1.0        --    --     0.7 Best Estimate                                      ---        --    --      --
1

                                                                     /                  t
                                                                                 /
Case 3 Results Parameter                                       FMRC         LANL    SNL1        SNL2 Peak Pres. (psia) z   Maximum                                          32         --       --
                                                                                 / 32 Minimum                                          24         --       --
22 Best Estimate                                    --
25       -_         27 Peak Gas Temp. (F)
Maximum                                      1832           --          --
2420 Minimum                                      ---            --       --
539 Best Estimate                                ---
2732     1832      1826 Peak Heat Fluxes                                                                        '
(1000 BTU /nr/ft**2)
Drywell Wall - 10' Maximum                                 28.             --
31.4       31.4 Minimum                                  13.            --       --
3.8             ,
Best Estimate                           ---             30       -_        __
Drywell Wall _ 20' Maximum                                 29.            --
                                                                       '33.0       33.0 Minimum                                  13.           --        --
4.8 Best Estimate                           ---
21        -_        __
Drywell Wall - 30' Maximum                                 30.            --
31.1       31.1 Minimum                                 13.            --       .--       5.7 Best Estimate                           ---
12.5      --         __
Wetwell Wall - 10'
!         Maximum                                 4.0            --
11.1        5.7 Minimum                                 2.7            --       --        0.7 Best Estimate                           ---
4.0       __         __
Wetwell Wall - 20' Maximum                                 4.4            --
12.4        6.7 Minimum                                 2.9            --       --
0.8 Best Estimate                           ---           4.75      --         __
Wetwell Wall - 30' Maximum                                  4.5            --
10.8        7.6 Minimum                                  3.0            --       --
1.0 Best Estimate                           ---
5.5       --         _-
I l

Case 4 Results Parameter                         FMRC       LANL         SNL1      SNL2   l Peak Pres. (psia)                                                          l Maximum                           24        --            --
25 Minimum                           22        --            --
20 Best Estimate                    --        --            --
22 Peak Gas Temp. (F)
Maximum                       1832          --            --
2420 Minimum                       ---           --            --
399 Best Estimate                 ---           --
1832   1826 Peak Heat Fluxes (1000 BTU /hr/ft**2)
Drywell Wall - 10' Maximum                   21.           --
23.8    23.8 Minimum                   9.2           --            --
2.4 Best Estimate             ---           --           --           ,
Drywell Wall - 20' Maximum                   22.           --
24.1    24.1 Minimum                   9.2           --           --
3.2 Best Estimate             ---           --           --        --
Drywell Wall - 30' Maximum                   20.          --
19.7    19.7 Minimum                   9 .* 2        --           --
3.8 Best Estimate            ---           --           --        --
Netwell Wall - 10' Maximum                   2.7           --
10.8     2.5 Minimum                   1.8           --           --
0.3 Best Estimate             ---           --           --        --
t Wetwell Wall - 20' Maximum                   3.0          --         10.5      3.2 Minimum                  2.0           --           --
0.4 Best Estimate            ---           --           --        --
Wetwell Wall - 30' Maximum                  2.9           --
7.9       3.5 Minimum                  1.9           --           --
0.4 Best Estimate            ---           --           --        --
l I
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Y MAX AND MIN TEMPERATURE IN WET WELL 1800                               ,        ,          ,           ,     ,            ,          ,     ,
'1 1600 -                                                                                                            -
C 1400 -                                                                                                                      -
1200 -                                                                                                             -
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                                                                     ^
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o 3 1400       -
N 1200   -
t 4 1000 1                                                                        \
                 $ 800       -
E                                                   I 600- -                                                                             -
x y      400- -
f                                       -
E'
                             ~
0 0
hfW 50          100                 150     200    250
: y's 300     350 400
                                                                                                          ~
AZIMUTHAL ANGLE (DEGREES)
CASE    C       TIME =                       60 1 SECONOS 1 = 10 FEET                        3: 30 FEET         * = SPARGER LOCATIONS

HEAT FLUX TO THE INNER WALL 1000                               ,                ,     ,            ,                           ,                              ,                             ,                     ,
o
            >       800  -                                                                                                                                                                                   -
eJ
            ?
u 600 -                                                                                                                                                                                    -
1 E       400                                                                                                                                                                                      -
A
            $       200-  -
                                                                    '         2                                                                       N-                                                     -
w 5
_J O!  -                                                                                                                                                                                  ,
k e -200- -
            $ -400                                    *                *      *           *                            *                              *                             *
* O                    200                  400  600           800 1000 1200 1400 1600 1800 l                                                                                        TIME (S)
CASE                             C                AZIMUTHAL ANGLE =                                                                             157 50 DEGREES 1 = 10 FEET                                       3 = 30 FEET l
l l
l l

TEMPERATURE OF THE INNER WALL 500         ,    ,     ,           ,     ,            ,   ,     ,
450< -                                                                            -
     ^      400  -                                                                            -
U 350- -                                                                            -
     $      300- -
h 250- -                                                                            -
s tr     200  -                                                                            -
W
: a. 150  -                                                                            -
I W      100- -                                                                            -
H 50-                                                                              -
0 0 200    400  600          800 1000 1200 1400 1600 1800 TIME (S)
CASE      C     AZIMUTHAL ANGLE =                   157 50 DEGREES
,               1 = 10 FEET     3 = 30 FEET l

CONCLUSIONS l
l    1. Significant uncertainty remains in large-scale diffusion l          flame behavior. FMRC 1/4 scale tests should provide a i         better understanding of the phenomena.
!    2. Diffusion flames should not directly threaten containment j         due to overpressure.
!    3. Peak gas temperatures could be as high as 2000 f, but will j
be well below maximum theoretical values due to high gas       '
entrainment rates.
: 4. Peak heat fluxes could range from less than 1000 BTU /HR/FT**2 j         to more than 30000 BTU /HR/FT**2; however, the peak values will only be maintained for short periods of time.             l l
I
: 5. Care should be exercised when extrapolating these results to predict the loads from particular accident scenarios and l
the response of specific equipment and penetrations.
                                                                       -l 1
l i

TVT GRAND GULF NUCLEAR STATION lDCOR TASK 23.1 INTEGRATED CONTAINMENT ANALYSIS J. R. Gabor IDCOR/NRC Meeting on Integrated Analysis of Severe Accident Containment Loads l
l Rockville, Maryland            l May 15-17,1984                1

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BWR/6 Mark 111 Safety and Other Systems

                                                                                                       .xv.i, .
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    . e
(
i GRAND GULF - T QUV Sequence Definition e Accident initiated by loss of off-site AC.
e Failure of power conversion system.
e Faliure of all core make-up.
e Failure of containment heat removal.
e Containment hydrogen igniters available.
e Ultimate containment pressure assumed to be 72 psia.
l i
l f
l

GRAND GULF NUCLEAR STATION T QUV - BASE CASE ACCIDENT CHRONOLOGY TIME                                                                                    EVENT
.                                  0.0 sec                                    initiating Event: Loss of off-site power; Loss of main feedwater; Turbine stop valve and turbine bypass valve closures                                                                                                ,
l 8.7 sec                                     Reactor scram completed i
13.3 sec                                    MSIVe closed 21.8 sec                                    RPV Level 2 LOCA setpoint reached 28.7 min                                    RPV Level 1 LOCA setpoint reached; Vessel depres-
!                                                                             surization manually initiated 29.2 min                                    DW purge system actuat'es l'                                30.4 min                                    Core begins to uncover 59.3 min                                    Suppression pool make-up actuates 2.2 hr                                      Fual melting begins 2.8 hr                                      Core plate failure followed by vessel failure                                                            ,
41.8 hr                                     Contahment fature
   --e-, --rw-- ---- - . - ---       -..,,--n-_--.-.,_--,,-.,,_,n,_,                        , _ , . _ , _ - , , -   , - - - - - , - - - - - - . - ,

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            ~
UNCERTAINTY / SENSITIVITY ANALYSIS FOR T QUV - GRAND GULF Sequences Analyzed                 '
Base Case  e  Same as Task 23.1 base case, except (1) no automation depressuriza-tion actuation.
(2) drywell - Compartment A leakage path.
(3) MSlVs close ~ 30 min.
* Burn drag coefficient, (CDs =
1.0).
e  20% core melt required to fall core plate, (FMAXCP = .20).
e  With core blockage (FBlock = 0).
e  Zircaloy oxidation cut-off tem-perature (TZOOFF = 2300 K).
l               e  Number of lower head penetra-tions f ailed (NPF = 1).

J :
UNCERTAINTY / SENSITIVITY ANALYSIS FOR T QUV - GRAND GULF Sequences Analyzed Case #1
* Same as base case, except NPF =
10.
Case #2
* Same as base case, except FMAXCP
                 =.40.
Case +3  e Same as base case, except FBlock
                = 1.
Case +4  e Same as base case, except TZOOFF
                =   2500 K.
Case +5  e Same as base case, except CDs
                = .5.
k

,                    UNCERTAINTY / SENSITIVITY ANALYSIS                              ) '
FOR T QUV - GRAND GULF Results a
Base               Case     Case  Case   Case   Casa Case                +1       +2     +3    +4     #5 Time of Contain-       44.7               44.6    62.4   40.8   44.8   44.5 ment Failure (hr)
Fraction of Zr Re-     .047               .047    .047    .16   .051   .047 acted in-Vessel Time of Core Un-        .66               .66      .66    .66    .66   .66 covery (hr)
H     *"  *     *           *                 *        '      *
* 2 Vessel Failure (Ibm)
Time of Vessel          3.74               3.74    4.27   3.62   3.74   3.74
( Failure (hr)                                                                     j

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         #              o                           UNITED STATES E'           /o                 NUCLEAR REGULATORY COMMISSION
       -h.            -
WASHINGTON, D. C. 20555
        \..../
4 OCT 12 1984 MEMORANDUM FOR:    DISTRIBUTION FROM:              Themis P. Speis, Director Division of Safety Technology, NRR Robert M. Bernero, Director Division of Systems Integration, NRR

==SUBJECT:==

==SUMMARY==
OF NRC/IDCOR MEETING ON INTEGRATED ANALYSIS OF FISSION PRODUCT BEHAVIOR - AUGUST 28-29, 1984 This memorandum is a summary of the fourth meeting between NRC and the In-dustry Degraded Core Rulemaking Program (IDCOR) held in Rockville, Md. which dealt with integrated analysis of fission product behavior. A summary of the principal technical results of this meeting are described in Enclosure 1. A meeting agenda and a list of attendees are included as Enclosures 2 and 3, respectively.
The technical presentations covered (a) some containment load issues from the May 1984 meeting that needed additional attention from both the staff and IDCOR, (b) fission product reemission and resuspension, (c) fission product release and transport as modeled by IDCOR, and (d) releases to the environment from selected sequences for four plants representing four containment types.
The plants were: Zion (large, dry); Peach Bottom (Mark I); Sequoyah (ice con-denser); and Grand Gulf (Mark III).
At the end of the meeting, summary presentations were given by NRC and IDCOR contractor representatives, in which both parties outlined the principal areas of agreement and disagreement for the PWR and BWR plants and sequences analyzed and discussed at the meeting. Enclosure 4 contains viewgraphs from the summary talks. Enclosure 5 contains viewgraphs from all other technical presentations.
Several days after the conclusion of the meeting, IDCOR requested that its comments on NRC contractor summary points be included with this meeting summary; the comments are included as Enclosure 6.
Although the meeting resulted in general agreement between NRC and IDCOR that the release of fission products to the environment is significantly less than calculations in the Reactor Safety Study, the meeting summary presentations highlighted several areas of disagreement and several issues requiring further study. The main points of technical disagreement are briefly summarized in Enclosure 1.
I

==Contact:==
J. Mitchell, 49-29402 C. Peabody, 42-74632
  ,. ,                                                     DN t              . -
s 4
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                                                                  ,v , t ,
Wi!
The Industry Degraded Core Rulemaking Program (IDCOR) is an effort on the part of nuclear utilities to develop the technical basis for determining whether changes in regulatory requirements are needed to reflect severe accident considerations. The NRC has recognized the potential benefit of factoring the IDCOR methods and results into the agency's decision process on severe accidents. A series of meetings has been arranged for NRC to examine and evaluate IDCOR's methods, assumptions and results.                       The purpose of this interaction is to take advantage of the technical programs and infor-mation developed by IDCOR, understand the bases, and identify what use we can make of the results.
The first three meetings, held in Harpers Ferry, W. Va., Hunt Valley, Md.,
and Rockville, Md. concentrated on the fundamental physical and chemical processes governing accident progression, containment loading, fission product behavior, and integrated analyses of containment loads for a variety of plants and accident sequences.
The next technical exchange meeting, tentatively scheduled for November 1984, will concentrate on (1) integrated risk analyses (probability times consequences), (2) operator procedures for accident mitigation, and (3) MARCH /MAAP comparisons.
                                   ---        /
: c. OLL&l                3ltc 1~
Themis P. Speis, Director Division of Safety Technology Office of Nuclear Reactor Regulation lc                              &           M        v Robert M. Bernero, Director Division of Systems Integration Office of Nuclear Reactor Regulation

==Enclosures:==
: 1. Summary of Technical Results
: 2. Meeting Agenda
: 3. List of Attendees
: 4. Summary Viewgraphs
: 5. Technical Presentation Viewgraphs
: 6. Additional IDCOR Comments e                                                   _ _ _ -
                                                                -           _ _ - _ _     _. -           /

t h                                               #A N'
DISTRIBUTION NRC/NRR            NRC/RES         ACRS H. R. Denton       R. Minogue      R. Tripathi E. Case            0. Bassett      S. Seth R. Bernero         G. Arlotto      R. Cushman R. W. Houston      R. Curtis       G. Quittschreiber L. G. Hulman       T. Walker R. Vollmer         G. Marino Z. Rosztoczy       J. Glynn        TEC B. Sheron          W. Morrison J. Rosenthal       J. Han          A. Buhl C. Tinkler         R. Wright       M. Fontana R. Palla           M. Cunningham   E. Fuller W. Lyon            M. Silberberg   J. Carter, III A. El-Bassioni     T. Lee          H. Mitchell J Y Barrett         C. Peabody      S. Asselin T. Speis           P. Niyogi       K. Meyer J. Mitchell        T. Eng C. Allen           B. Aggarwal     EPRI P. Easley          R. Meyer S. Acharya         B. Burson       M. Everett J. Read            J. Larkins      B. R. Sehgal F. Akstulewicz     J. Telford      D. Squarer W. Pasedag         P. Baranowsky   R. Vogel F. Gillespie    C. Fuller Other NRC          R. VanHouten J. Martin     ~
Battelle Columbus J. Conran, DEDR0GR L. Chan M. Taylor, DEDROGR L. Soffer       P. Cybulskis J. Austin, OCM     C. Ryder        R. Denning J. Gieseke DOE                FAI             P. Owczarski K. Wiegardner F. Witmer          H. Fauske R. Henry        EG&G Idaho AIF                J. Gabor M. Kenton       S. Behling J. Siegel                          R. Gottula J. Broughton NRC PDR k

b Sandia National Lab      Other Affiliations D. Dahlgren              R. Breeding, EI M. Berman                D. Moore, El S. Thompson              J. Young R. Cole                  M. Lloyd, Middle South Services J. McGlaun J. Hickman               D.Paddleford,[W L. A. Wooten, K. Bergeron              P. Nakayama, Jaycor D. Kunsman               L. Azarello, Duke Power D. Aldrich               W. Mims, TVA J. Sprung                M. Cosella, Coned J. Walker                J. Meincke, CPC0 J. Griesmeyer            W. Iyer, NYPA F. Harper                J. Davis, NYPA D. Powers                A. Marie, PEC0 V. Behr                  G. Krueger, PEC0 J. Linebarger            H. R. Diederich, PECO S. Dingman               R. Smith, Scandpower A. Camp                  J. Engstrom, OKG AB/ Sweden A. Benjamin              J. Liljenzin, CTH/ Sweden S. Webb                  L. Rib, LNR Associates C. Leigh                 J. Metcalf, Stone & Webster A. Peterson              C. Ader, Stone & Webster D. Williams              M. Corradini, University of Wisconsin P. Mast                  I. Spiewak, American Physical Society E. Haskin                S. Niemczyk, UCS R. Habert, UCS Oak Ridge National Lab   T. Theofanous, Purdue University J. Kelly, University of Virginia S. Hodge                 K. Araj, Harvard University I. Catton, UCLA Brookhaven National Lab  R. Seale, University of Arizona S. Beal, SC&A W. T. Pratt              R. Paccione, Long Island Lighting Co.
M. Khatib-Rahbar         K. Holtzclaw, GE R. Newton                R. Smith, NuCon Corporation T. Ginsberg              A. Pressesky, AWS G. Greene                M. Ryan, Inside NRC R. Jaung                  P. O'Reilly, NUS Wen-Shi Yu                P. Fulford, NUS H. Ludewig               G. Kaiser, NUS S. Blazo, Bechtel

- u.
I                                       ,
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Enclosure 1

==SUMMARY==
OF THE PRINCIPAL TECHNICAL RESULTS IDCOR/NRC MEETING AUGUST 28-29, 1984 The August 28-29, 1984 meeting between IDCOR and NRC focused on the integra-ted analyses of severe accident fission product behavior.
Several of the containment performance issues identified as areas of disagree-ment at the NRC/IDCOR technical exchange meeting of May 15-17, 1984 were also discussed. The principal issues discussed were:    hydrogen production and combustion, core / concrete interactions, and sensitivity analysis. Based on these discussians, there is a better understanding of the technical differences and their bases between IDCOR and the NRC contractors and consultants.
10COR analyses of specific sequences for four representative plants
* were per-formed using the MAAP 2.0 code. This code includes natural circulation in the primary system, incorporates codes previously used as stand-alone codes (FFRAT and RETAIN), and incorporates an aerosol deposition correlation, which is discussed below. The NRC contractor results for the same four plants are
* Zion with a large dry containment, Peach Bottom with a Mark I containment, Sequoyah with an ice condenser containment, and Grand Gulf with a Mark III containment.
Enclosure 1 w

extracted from calculations presented in the BMI-2104 document on source term evaluation, which was published in draft form in July 1984.
The updated version of MAAP contains a new empirical aerosol deposition correlation to account for agglomeration and settling of aerosols.       A sedi-mentation rate is applied based on a single parameter, aerosol cuncentration.4 Two fitted constants have been determined from experiments and compared with other experiments. Water absorption by hygroscopic materials is included.
This model was criticized by the NRC contractors during the summary presenta-tions for not including a dependence on other known important factors (e.g., particle size) and was further criticized for lack of a wide range of comparisons to experiments.
10COR has concluded that resuspension of settled fission products will not be important. Reevolution is calculated in MAAP based on partial pressure differ-ences and convective flows. Similarly, the TRAPMELI code, used by the NRC con-tractors, has been changed to include reevolution using Raoult's Law (basically the same mechanism) and the deposited fission product beta and gamma energy.
For both 10COR and BMI-2104 calculations, sequence-dependent distributions (by nuclide in some cases) of fission products within the plant are included in the viewgraphs of the technical presentations.      In several cases, different assump-tions were made concerning the specifics of the sequence definition.        This is important to recognize in any comparison of the ultimate distributions, as stressed by both 10COR and NRC contractors in their presentations and summaries.
For details, see the viewgraphs of sequence definitions and ultimate distribu-

k                                                                                      l W l]
tions in Enclosure 5.
Zion For Zion, 10COR evaluated the station blackout event with and without a pump seal LOCA, with a purge line open, and the V sequence. Containment failure was calculated at about 30 hours for the TMLB and was assumed from the beginning for the open purge valve and V sequence cases. IDCOR found a sensitivity to assumod vapor pressure for Cs0H, the correlation parameters in the aerosol model, and the hole size at containment failure. Except for the case of a large hole size at failure or the case of failure to isolate (open purge line), releases to the environment were calculated to be of order 10-3 or less (noble gas release fractions are always close to unity and are not further discussed in this summary). For Csl and Cs0H, for a 0.5 square foot containment failure or failure to isolate containment, releases were of order 2x10-2; releases of other particulate nuclides were <10 -3  . A sensitivity study to thermal hydraulic parameters was performed with trivial sensitivity shown except to changes that altered the sequence or where a parameter change revealed a model weakness by giving erroneous results (e.g., core slump model).
During the summary discussions, the NRC contractors stated the belief that the uncertainties were "much larger than the 10COR sensitivity studies are indi-cating." The BMI-2104 results for Zion were for a small break LOCA and the station blackout. The calculations were not carried out until overpressure failure was confirmed or denied. However, they were continued long enough to see that, at the rate of change of pressure with time, if the failure occurred, it would be at extremely long times after the start of the sequence. The base-mat melt-through failure was evaluated instead with releases to the environment b

B*
lL.
v of CsI, Cs0H, and Te of order 10 -4 or less.                                          For isolation failure (comparable
                                                                                                                                                                ~2 to an open purge vahe) the releases of Cs and I were of order 10                                                                   .      Because of the assumption of Te release from the core / concrete interaction, its release fraction to the environment was calculated to be 0.2.
Peach Bottom For Peach Bottom, IDCOR evaluated the ATWS, small break LOCA with injection failure, and the transient with no makeup and failure to remove decay heat sequences. The containment failure times varied from 12 to 23 hours (sequence dependent) from overtemperature and were attributed to failure of components that provide the containment boundary.                                            Release fractions for Cs and I ranged from 0.01 to 0.07, Te from 0.02 to 0.09, with other nuclides of order 10 -4 and less. For the transient sequence with failure to remove decay heat, but with early injection, IDCOR assumed that injection would fail at containment failure only for containment failure sizes large enough to fail the pumps via dynamic response.                        This assumption is based on tests performed by General Electric that show operability of pumps in the environment following a small size failure of the containment.                                          In contrast, previous calculations have assumed that any failure of the containment for this sequence would cause injection failure.                              Cs and I release fractions were calculated to be of the same order as given above.
The station blackout event was also analyzed for different assumptions of the area over which core / concrete reactions could take place.                                            Although minor details varled, containment failure time (about 20 hours) and Cs and I release fractions (0.07) were unchanged.                                          A sensitivity analysis for thermal hydraulic
                                                                                                                                                                                                         /

l l
l                                      i
; parameters similar to that for Zion was performed with similar results.      It was noted following this presentation and reiterated in the summary presen-tationbytheNRCcontractorsthatextraordinaryoperatbractionswere assumed for the analyses of this plant. The IDCOR response was that the operator actions only shift the time scale of the accident, but do not change the release fractions.
l The evaluations in BMI-2104 were for the ATWS, the large-break LOCA with loss of injection, and the transient with failure of decay heat removal sequences.
l For the large break LOCA, containment failure was calculated to be atiabout 30 minutes, with release to the environment of about 30% of the Cs and I and
; about 60% of the Te,    The releases for an ATWS sequence with failure into the reactor building were about 10% of the Cs and I and about 25% of the Te.      As expected, releases directly to the environment are larger:      about 20% of the Cs and I and about 40% of the Te. For the transient with failure of heat removal and release directly to the environment, the release fraction for Cs and I is l about 5% and for le is about 20%.
! Sequoyah For Sequoyah, the differences between IDCOR and BMI-2104 results are due in large measure to hydrogen treatment. As discussed above, the positions on hydrogen generation and burning are unchanged from the previous meeting in Rockville (May 1984). IDCOR analyzed nine sequences:   four representing con-tainment failure by overpressure and four sequences for which containment failure is not predicted. The last sequence is an open purge line sequence and represents the largest release fractions:     about 2% of the Cs and I and l
L

l 1 >
4       r*
about 0.5% of the Te. For other sequences, the release fractions are 10" or less.
The results presented from BMI-2104 were for three sequences:      station blackout, small break LOCA with recirculation failure, and transient with failure of secondary side heat removal. For the station blackout, if the containment fails by ignition of hydrogen as the molten core exits the vessel, 4.id for the small break LOCA case, release fractions are about 2%. However, if ignition does not occur in the station blackout sequence and containment failure is by (velayed) overpressure, release fractions are about 10-3 or less.      The transient with failure of secondary side heat removal is calculated to have release fractions less than 1%, decades less in the case of delayed overpressure failure.
Grand Gulf for Grand Gulf, the differences are again due to hydrogen treatment.        IDCOR analyzed four sequences and three were reported from BMI-2104.      Both analyses included significant credit for suppression pool fission product scrubbing.
The techniques for taking scrubbing into account were different.       10COR used one of two constant values depending on pool depth (release through x quenchers or vents), while a more elaborate model was used in the BMI-2104 calculations, depending on several parameters that vary over the course of the sequence.
IDCOR assumed a nominal pool bypass until vessel failure (affecting only the large break LOCA sequences). Subsequent to vessel failure, the bypass path has been previously evaluated to be blocked by aerosol particles.      For all sequences,

v l W     e.
releases were found to be of order 10 -3 and less.
The results from BMI-2104 include evaluation of the small break LOCA with failure of injection, with also a nominal bypass of the pool, and with an open bypass path (vacuum breaker assumed to be open). For the through pool cases, environmental releases were of order 10 -2 to 10 -3  . For the nominal bypass case, releases were slightly higher. For the large bypass case, Cs and I releases
                                   -3 were of the order of 10      but Te releases were of about 10%.
In addition to the areas of disagreement already discussed above, there_. remains a disagreement on-the Te-Zr reaction.      IDCOR' assumes that the.Te is released in-vessel with the volatile nuclides, while the results in BMI-2104 are based on the model that relates Te retention in the core to the amount of unoxidized Zr. Retention with the core material allows for Te release at the time of Zr oxidation during concrete attack.
The summary viewgraphs from both IDCOR and NRC contractor presentations have been included as Enclosure 4.

65   .                                                                         ENCLOSURE 2 AGENDA NRC/IDCOR MEETING 0.lN INTEGRATED ANALYSIS OF SEVERE ACCIDENT FISSION PRODUCT BEHAVIOR AUGUST 28-29, 1984 Tuesday, Auoust 28 Introduction 8:15-8 45 A.M.         -     Welcome, Purpose, Ground Rules, Schedule - T. Speis, R. Bernero (NRC)
                              -     Introduction - A. Buhl, M. Fontana (IOCOR) 8:45-9:30 A.M.        -     Containment Loads Issues Remaining from the May 15-16 Meeting         (IDCOR)
In-Vessel Hydrogen Production Hydrogen Combustion Behavior Core Concrete Interaction In Mark I BWR's In-Vessel Fuel Coolant Interactions Sensitivity and Uncertainty Analysis 9:30-10:15 A.M.         -
Containment Loads Issues     (NRC) 10:15-10: 30 5.M.       -
Break Fission Product Methodology 10: 30-11: 00 A.M.      -    Correlation for Fission Product Deposition (IOCOR) 11: 00-11: 30 A.M.      -
Reevolution of Fission Products     (IDCOR) 11:30-12:00 A.M.        -
Reevolution of Fissicn Products     (NRC Contractor) 12: 00-1: 00 P. M.       -
Lunch 1:00-2:00 P.M            -    MAAP Modeling of Fission Product Release & Transport In the Primary System and Containment Enclosure 2 L          _

Enclosure 2 Tuesday, August 28 (Cont'd)
PWR Large Dry 2:00-3:00 P.M.         -
10COR Results 3:00-3:15 P.M.         -
NRC Contractor Results                              -
l                                            3:15-3:30 P.M.         -
Break l
l                                                   BWR Mark I 3:30-4:30 P.M.         -
10COR Results 4:30-5:00 P.M.        -
NRC Conractor Results
              ,      - 5:00 P.M.                                   -
Adjourn f
7:00-                 -
NRC Working Group Meetings Wednesday, August 29 PWR Ice Condenser
                                          -8:00-9:00 A.M.          -
10COR Results 9:00-9: 30 A.M.       -
NRC Contractor Results 9:30-9:45 A.M.        -
Break BWR Mark III 9:45-11:15 A.M.        -
10COR Results 11:15-11:45 A.M.       -
NRC Contractor Results 11:45-12:30 P.M.        -
Lunch 12: 30-1:45 P.M.       -
NRC Working Group Meetings l.
Enclosure 2

v hT'   o Enclosure 2 Wednesday, August 29 (Cont'd)
Meeting Preliminary Summaries 1:45-2:00 P.M.       -
Containment Loads Issues     (NRC Contractor) 2:00-2:15 P.M.       -
Fission Product Modelling in MAAP    (NRC Contractor) 2:15-2:30 P.M.       -
PWR Fission Product Results     (NRC Contractor) 2:30-2:45 P.M.       -
BWR Fission Product Results     (NRC Contractor)
          . 2:45-3: 30 P.M.      -
IDCOR Summary
          '3:30-4:00 P.M.        -
Parting Remarks - R. Bernero, T. Speis, A. Buhl Enclosure 2 il'

v w-  -e ATTENDEES / AUGUST IDCOR MEETING               Enclosure 3 NRC                              BNL S. Acharya                      G. Green F. Akstulewicz                  R. Jaung C. Allen                        H. Ludewig J. Austin                       R. Newton R. Bernero                      T. Pratt B. Burson                       Wen-Shi Yu L. Chan J. Conran                       EPRI M. Cunningham R. Curtis                       M. Everett P. Easley                       C. Fuller
          .F. Gillespie                    B. R. Sehgal J. Han                          R. Vogel L. Hulman T. Lee                          Sandia J. Martin R. Meyer                        V. Behr J. Mitchell                     A. Benjamin W. Pasedag                      K. Bergeron C. Peabody                      D. Dahlgren G. Quittschreiber               J. Griesmeyer J. Read                         E. Haskin J. Rosenthal                    D. Kunsman C. Ryder                        C. Leigh M. Silberberg                    P. Mast L. Soffer                       J. McGlaun T. Speis                         A. Peterson M. Taylor                        D. Powers J. Telford                       J. Sprung K. Tripathi                      S. Thompson R. VanHouten                    J. Walker T. Walker                        S. Webb D. Williams Battelle/ Columbus ORNL P. Cybulskis R. Denning                      S. Hodge J. Gieseke Battelle/PNL P. Owczarski K. Winegardner Enclosure 3 i
i t

Y' o    re Enclosure 3 2-Other Affiliations K. Araj, Harvard University S. Asselin, lEC L. Azzarello, Duke Power Company S. Beal, SC&A -
S. Blazo, Bechtel Power Corporation R. Breeding, Energy Inc.
J. Broughton, EG&G Idaho A. Buhl, TEC J. Carter, III, TEC I. Catton, UCLA M. Corradini, University of Wisconsin J. Davis, N.Y. Power Authority H. R. Diederich, Philadelphia Electric J. Engstrom, OKG AB/ Sweden M. Fontana, TEC-J. Gabor, Fauske & Assoc.
R. Habert, UCS R. Henry, FAI K. Holtzclaw, GE G. Kaiser, NUS J. Kelly, University of Virginia M. Kenton, Fauske & Assoc.
G. Krueger, Philadelphia Electric J. Liljenzin, CTH/ Sweden M. Lloyd, Middle South Services, Inc.
A. Marie, Philadelphia Electric J. E. Metcalf, Stone & Webster H..Mitchell, TEC
: 5. Niemczyk, UCS P. O'Reilly, NUS R. Paccione,~Long Island Lighting Company A. Pressesky, AWS M. Ryan, Inside NRC R. Seale, University of Arizona R. Smith, NuCon Corporation T. Theofanous, Purdue University L. A. Wooten, Westinghouse Enclosure 3 n -
                                          -p ,   - . , - - .

aw       e Enclosure 4 l
l NRC CONTRACTOR AND IDCOR 

==SUMMARY==
VIEWGRAPHS*
           *VIEWGRAPHS AND PORTIONS OF VIEWGRAPHS IN THIS TYPEFACE WERE HANDWRITTEN AT THE MEETING AND RETYPED FOR CLARITY.
Enclosure 4
                             - + _ _ --_ -

                                                                                                        . .w   :
IDCOR AND BMI-2104 RESULTS                                                        ,
NOT GREATLY DIFFERENT                                                           ,
e PROBLEM IS VERY COMPLEX e SOME DETAILS DIFFER, BUT OVERALL THE SIMILARITIES PREDOMINATE                                       i e EXTENT OF AGREEMENT GRATIFYING
 ~
t
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J l
i IDCOR 3 :. .

o BOTH IDCOR AND BMI-2104 ANALYSES ARE ADVANCES FROM WASH-1400 e MECHANISTIC TREATMENT OF ACCIDENT PROGRESSION BASED ON INTEGRATING PHENOMEN0 LOGICAL MODELS, PLANT MODELS, AND SAFETY SYSTEM MODELS e BETTER CHARACTERIZATION OF CHEMICAL FORMS OF FISSION PRODUCTS, PARTICULARY FOR IODINE AND CESIUM e SIGNIFICANT DEPOSITION OF FISSION PRODUCTS IN PRIMARY SYSTEMS CALCULATED WITH MODELS ACCOUNTING FOR AEROSOL FORMATION AND SETTLING t
IDCOR i

J     e4 4
HYDROGEN PRODUCTION e IDCOR UNCERTAINTY AND SENSITIVITY ANALYSES SHOW NO EFFECT ON CONTAINMENT FAILURE AND FISSION PRODUCT RELEASE e IDCOR BENCHMARKING 0F TMI SHOWS MAGNITUDES OF HgGENERATION THAT AGREE WELL 5
i IDCOR

W zw c TE RELEASE e BMI-2104 TREATS TE RELEASE BOTH IN-VESSEL AND IN CORE DEBRIS-CONCRETE ATTACK e IDCOR CALCULATES COMPLETE TE RELEASE EARLY e EXPERIMENTAL EVIDENCE SEEMS TO SUPPORT SOME TE RELEASE DURING CORE-CONCRETE ATTACK e IDCOR WILL CARRY OUT A SENSITIVITY ANALYSIS TO DETERMINE EFFECTS 0F TE RELEASE OURING CONCRETE ATTACK i
e SHOULD ONLY AFFECT CASES WHERE CONTAINMENT FAILS EARLY l
i l
l I
i IDCOR i:
I

m
: o.   ,e REVAPORIZATION e IDCOR CONSIDERS UNCERTAINTIES IN VAPOR PRESSURE o ' VAPOR PRESSURE MAY BE LOWERED ONCE EXPERIMENTAL DATA ARE OBTAINED AND EVALUATED
* 1 IDCOR l
  ~ -:-,.w,-   -       _ _ _ , . , , _ _ , . . _ , . , _ . , _ , , , , _ , , , , , _ , _ , _ , . , _ , _ , , _ , _ _ _ ._ __    _ _ _

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        \
EARLY DRYWELL FAILURE FROM DRYWELL LINER ABLATION e NRC POINTED OUT POSSIBILITY BUT THEY ASSUMED HIGH DEBRIS       ,.
INVENTORY e IDCOR CALCULATES THAT DEBRIS INITIALLY QUENCHED BEFORE DRYOUT AND REHEATING e IDCOR SENSITIVITY CASE SHOWED LITTLE EFFECT IDCOR l

o             a SGTS IN PEACH BOTTOM e PEACH BOTTOM TEAM HAS EVALUATED THE SYSTEM e 'IDCOR WILL LOOK AT THE DESIGN AND OPERATION AGAIN t
I IDCOR l
~
                                                          . , - , , , - . , - . . - - - - . . - - . - - - , - + , . . - - - - - - - . . . , , . , ,
                                                                                                               ~

he g CONTAINMENT LOADS 

==SUMMARY==

IN-VESSEL HYDR 0 GEN PRODUCTION HYDROGEN COMBUSTION BEHAVIOR MODE OF MELT EJECTION CORE / CONCRETE INTERACTION STEAM EXPLOSIONS DIRECT HEATING OF CONTAINMENT ATMOSPHERE SEQUENCE DEFINITION SENSITIVITY UNCERTAINTY ANALYSIS l
i s
NRC CONTRACTORS

IN-VESSEL HYDROGEN PRODUCTION DIFFERENCE HYDR 0 GEN GENERATION DIFFERS BY FACTORS OF 1.4 TO 10.0 BECAUSE OF BLOCKAGE ASSUMPTIONS, LACK OF DOWNWARD RADIATION HEAT TRANSFER AND THE SHORT TIME BETWEEN SLUMP AND VESSEL FAILURE SIGNIFICANCE AFFECTS PROBABILITY AND TIMING OF CONTAINMENT FAILURE NRC CONTRACTORS

h e, HYDROGEN COMBUSTION BEHAVIOR DIFFERENCE 4
FLAME TEMPERATURE CRITERIA USED PREVENTS HIGH HYDR 0 GEN CONCENTRATION, NATURAL CIRCULATION OF GASES FROM CAVITY TO CONTAINMENT CAUSES MORE COMPLETE BURNING.
SIGNIFICANCE MAJOR DIFFERENCE IN PREDICTION OF CONTAINMENT PRESSURE (EARLY VS, LATE CONTAINMENT FAILURE).
NRC CONTRACTORS

MODE OF MELT EJECTION DIFFERENCES CLWG USES A C0HERENT RELEASE OF ALL CORE MELT AVAILABLE AT VESSEL FAILURE.
IDCOR RELOCATES CORE SEQUENTIALLY AS REGIONS SLUMP.
SIGNIFICANCE THIS RESULTS IN DIFFERENT INITIAL CONDITIONS FOR CORE-CONCRETE INTERACTIONS, PREVENTING ABLATION ATTACK ON LINER AND LIMITING MELT AVAILABLE FOR DIRECT CONTAINMENT HEATING CONSIDERATION.
i l
l NRC CONTRACTORS i

o CORE / CONCRETE INTERACTIONS DIFFERENCES IDCOR MODEL RESULTS IN THICK INSULATING CRUST BETWEEN MELT AND CONCRETE.
CORCON MODEL RESULTS IN THINNER, LESS STABLE CRUSTS, SIGNIFICANCE SOLID, NOT M0LTEN, DEBRIS ATTACK WHICH AFFECTS UPWARD HEAT FLUX INTO CONTAINMENT GAS GENERATION FROM CONCRETE FISSION PRODUCT RELEASE FROM MELT DEGRADATION OF OVERHEAD STRUCTURES INCLUDING DE-GASSING 0F UNLINED CONCRETE INTERACTION WITH OVERLYING WATER P00L NO FURTHER OXIDATION OF METAL COMPONENT OF MELT NRC CONTRM: TORS

                                                          ,      s STEAM EXPLOSIONS POSITIONS UNCHARGED SINCE HARPER'S FERRY MEETING NRC WILL ORGANIZE AN EXPERTS GROUP TO REVIEW THE ISSUE,  MEETING IS PLANNED FOR OCTOBER 1984.
NRC CONTRACTORS

T h b
                                                                             \
DIRECT HEATING 0F CONTAINMENT DIFFERENCE NRC CONSIDERS HEATING AS A RESULT OF PRESSURE EJECTION OF CORE DEBRIS FROM VESSEL AND DISPERSAL OUT OF CAVITY, INTEREST BASED UPON RESULTS OF SNL TESTS, IDCOR NEGLECTS BASED UPON ANL TESTS, SIGNIFICANCE POSSIBILITY OF EARLY CONTAINMENT FAILURE EITHER BY DIRECT HEATING ALONE OR IN CONCERT WITH HYDROGEN.
NRC CONTRACTORS 1
i~

o d SEQUENCES DEFINITION DIFFERENCE THE IMPORTANCE OF SEQUENCE DEFINITION IS RECOGNIZED BY BOTH NRC AND IDCOR.
SIGNIFICANCE THIS IS AN EXTREMEL  IMPORTANT ISSUE AND CAN INFLUENCE PROGRESSION OF CORE MELT AND MODE AND TIMING OF CON-TAINMENT FAILURE.
NRC CONTRACTORS l

n s
                                                                             \
SENSITIVITY - UNCERTAINTY ANALYSIS DIFFERENCES IDCOR STUDY BASED UPON INCOMPLETE SET OF PARAMETERS DOES NOT INCLUDE VARIATION OF PARAMETERS BASED UPON ENGINEERING JUDGMENT EFFECTS OF VADVING MORE THAN ONE PARAMETER PER CASE NOT EXAMINED.
SIGNIFICANCE CONCLUSION OF THE STUDY ARE DIFFICULT TO ACCEPT WITHOUT A MORE COMPLETE EXAMINATION OF INPUT PARAMETER AND THEIR RANGE.
NRC CONTRACTORS i

o   <~.
SEVERE ACCIDENT ANALYSIS MODELS
: 1. AEROSOL MODEL NRC CONTRACTORS AND STAFF HAVE SIGNIFICANT RESERVATIONS ABOUT THE FAI AEROSOL MODEL, PARTICULARLY FOR USE WITHIN THE RCS:
THE CORRELATION DOES NOT INCLUDE EF.ECTS AND PARAMETERS KNOWN TO BE IMPORTANT FROM OTHER STUDIES (E.G., AN EXPLICIT TREATMENT OF SIZE DISTRIBUTION)
THE AGREEMENT SHOWN WITH EXPERIMENTAL DATA IS NOT COMPELLING BECAUSE OF THE SIMILARITY OF THESE EXPERIMENTS TO THE CORRELATION DATA BASE AND, FOR THE WET EXPERIMENTS, THE NEED TO SET TUNABLE PARAMETERS COMPARIS0N WITH EXPERIMENTS AT DIFFERENT SCALES OR WITH MORE RIG 0ROUS MODELS WOULD HELP TO DISPEL SKEPTICISM NRC CONTW\CTORS

CO   6
: 2. REVAPORIZATION MODELING l
I THE NRC CONTRACTORS AND STAFF AGREE THAT REVAPORI-ZATION IS IMPORTANT, THE IDCOR MODEL IS VERY SIMPLISTIC AND IT IS UNLIKELY THAT IT PRODUCES REALISTIC PREDICTIONS OF THE TIMING AND EXTENT OF RELEASE FROM SURFACES. ALTHOUGH THE ENVIRONMENTAL RELEASE TERMS OBTAINED BY THE IDCOR ANALYSES WERE FOUND TO BE SMALL FOR THE ACCIDENT SEQUENCES ANALYZED, THE SIMPLIFIED TREATMENT OF RCS BEHAVIOR MAY BE UNACCEPTABLE WHEN MORE REALISTIC ASSESSMENT IS MADE OF RETENTION EXTERNAL TO THE RCS OR TO RCS THERMAL-HYDRAULIC BEHAVIOR.
: 3. RESUSPENSION MODELING ALTHOUGH WE FEEL THAT THERE IS SOME POTENTIAL FOR RESUSPENSION WITHIN THE RCS THAT SHOULD BE RESOLVED, WE AGREE WITH THE IDCOR APPROACH T0 IGNORE RESUSPENSION FROM THE RCS AND CONTAINMENT WITH THE CURRENT LEVEL OF UNDERSTANDING, NRC con 2n c79gg

                                                                      , os
: 4. TELLURIUM RELEASE THE IDCOR MODELING OF TELLURIUM BEHAVIOR DOES NOT RECOGNIZE EXPERIMENTAL DATA INDICATING THAT A LARGE COMPONENT OF THE TELLURIUM INVENTORY WOULD REMAIN WITH THE CORE MATERIAL AND BE RELEASED DURING CONCRETE ATTACK.
: 5. CHEMICAL REACTIONS OF FISSION PRODUCTS THE IDCOR ANALYSES IGNORE THE POTENTIAL FOR CHEMICAL REACTIONS THAT COULD CHANGE THE CHEMICAL FORMS OF FISSION PRODUCTS AND AFFECT THEIR SUBSEQUENT TRANSPORT, SUCH AS:
REACTIONS WITH CONTROL MATERIALS FORMATION OF METHYL IODIDE RADIOLYSIS RADIATION EFFECTS REACTIONS WITH SURFACES OXIDATION AT HIGH TEMPERATURE ALTHOUGH SOME OF THESE EFFECTS MAY NOT BE VERY LARGE, THEY MAY DOMINATE SEQUENCES WITH SMALL RELEASE FRACTIONS AS ANALYZED BY IDCOR.
NRC CONTIMCTORS

n a 6,  CORE-CONCRETE RELEASE FROM OUR ANALYSES WE BELIEVE THE RELEASE OF FISSION PRODUCTS AND AEROSOLS DURING CORE-CONCRETE ATTACK IS IMPORTANT,          IDC0R'S APPROACH NEEDS CLARIFICATION, 7,  P0OL SCRUBBING WE DISAGREE WITH THE APPROACH,           THE DF'S SELECTED ARE CONJECTURAL,          THE STATE-0F-THE-ART OF DATA AND MODELS PERMITS A BETTER TREATMENT,
: 8. ICE BED DECONTAMINATION WE BASICALLY AGREE THAT THE TWO MECHANISMS BEING INVESTIGATED BY IDCOR ARE THE MOST IMPORTANT BUT ARE CONCERNED ABOUT THE ABILITY TO MODEL SEDIMENTATION WITHOUT AN EXPLICIT TREATMENT OF PARTICLE SIZE.
: 9. FISSION PRODUCT DECAY CHAINS THE LACK OF EXPLICIT TREATMENT OF FISSION PRODUCT DECAY CHAINS (E.G.,Te132 ,, 1 132) COULD AFFECT THE TRANSPORT OF FISSION PRODUCTS AND THEIR PARTICULAR RELEASE TO THE ENVIRONMENT, PARTICULARY OVER LONG TIME PERIODS OR FOR SMALL SOURCE TERMS.
I l
l l
NRC CONTRACTORS w

a     .'.
: 10. AEROSOL GENERATION MECHANISMS                                                          ,
THE POTENTIAL FOR ADDITIONAL AEROSOL GENERATION MECHANISMS (STEAM
   ,    EXPLOSIONS, HIGH PRESSURE BLOWDOWN) SHOULD BE RECOGNIZED AS CONTRIBU, TING TO THE UNCERTAINTY IN ENVIRONMENTAL RELEASES.
1 1,
j.
(
l NRC CONTRACTORS 1
t
: i. .
l UNCERTAINTY / SENSITIVITY ANALYSES BASED ON OTHER STUDIES, WE BELIEVE THAT THE UNCERTAINTIES ARE MUCH LARGER THAN THE IDCOR SENSITIVITY STUDIES ARE INDICATING. THE IMPLICATION MAY BE THAT THE IDCOR MODELS DO NOT INCLUDE THE MOST SENSITIVE PARAMETERS OR DON'T ALLOW THEM TO BE VARIED. WE EXPECT UNCERTAINTY ASSOCIATED WITH UNMODELED PHENOMENA TO DOMINATE.
h NRC con trucIORS

4   s$
PLANT ANALYSIS THERE ARE MANY DIFFERENCES IN RESULTS,          IT IS NOT CLEAR YET TO WHAT EXTENT THESE ARISE FROM DIFFERENCES IN MODELING VERSUS ACCIDENT SEQUENCE DEFINITION,            IN THOSE Cl4SES WHERE THERE IS APPARENT AGREEMENT, IT MAY BE FORTUIT0US.
BASIC' ASSUMPTIONS ON ACCIDENT SEQUENCE DEFINITION CAN MAKE A BIG DIFFERENCE.        THERE IS AN APPARENT LACK OF CONSISTENCY IN THE IDCOR TREATMENT OF HUMAN BEHAVIOR.
THIS SHOULD BE,A SUBJECT OF DISCUSSION AT THE NEXT MEETING, FOLLOWINGC0bPLETIONOFMARCH-MAAPCOMPARIS0NCALCULATIONS AT BCL, INTERACTIONS WITH IDCOR STAFF WOULD HELP TO UNDER-STAND DIFFERENCES.
THE IDCOR MARK I RESULTS DEPEND HEAVILY ON THE RETENTION CAPABILITY OF THE REACTOR BUILDING.         WE ARE CONCERNED ABOUT OUR UNDERSTANDING 0F THE FAILURE MODES OF THE PRIMARY CONTAINMENT AND THE REACTOR BUILDING,          SIMILARLY, THE PWR V SEnUENCE RESULTS RELY HEAVILY ON RETENTION IN THE AUXILIARY BUILDING.
WE BELIEVE THE TREATMENT OF DRYWELL LEAKAGE AND P00L BYPASS IN THE MARK III CASES IS INADEQUATE.
NRC CONTRACTORS l

to c2 Enclosure 5 s
1 VIEWGRAPHS FROM TECHNICAL PRESENTATIONS i
I 2
Enclosure 5

fo    o H
OVERVIEW 0F MEETING TOPICS EDWARD L. FULLER EPRI/IDCOR IDCOR/NRC MEETING ON INTEGRATED ANALYSES OF SEVERE ACCIDENT FISSION PRODUCT BEHAVIOR c
4 ROCKVILLE, MD AUGUST 28-29, 1984 4
i

                                                             . s.
THREE PREVIOUS MEETINGS HAVE LED TO THIS ONE 0   THERMAL-HYDRAULIC PHENOMEN0 LOGY DISCUSSED AT HARPERS FERRY, WEST VIRGINIA IN DECEMBER 1983 0   FISSION PRODUCT RELEASE AND TRANSPORT PHENOMEN0 LOGY DISCUSSED AT HUNT VALLEY, MD IN FEBRUARY 1984 0   INTEGRATED ANALYSES OF THERMAL-HYDRAULIC BEHAVIOR DISCUSSED AT ROCKVILLE, MD IN MAY 1984 ELF /ts P
t
: r. .
THE MAAP 2.0 CODE IS NOW USED TO PERFORM THE COMPLETE ANALYSIS 0 PROPER MODELING OF EFFECTS OF REVAPORIZATION REQUIRES TREATING NATURAL CIRCULATION IN THE PRIMARY SYSTEM.                            THE CIRC CODES WERE DEVELOPED FOR THIS 0 USE OF MAAP 1.2, FPRAT, CIRC, AND RETAIN IN TANDEM SHOWED THAT MANY IMPORTANT FEEDBACKS NEEDED TO BE BETTER ACCOUNTED FOR 0 KEY FEATURES OF FPRAT, CIRC, AND RETAIN WERE INCORPORATED INTO MAAP.
AN AEROSOL REPOSITION CORRELATION WAS ALSO DEVELOPED.                                       THE RESULT lS MAAP 2.0 l
l l
1

                                                                                                                                                       . e FOCUS IS ON FISSION PRODUCT BEHAVIOR DURING POSTULATED SEVERE ACCIDENTS IN THE IDCOR REFERENCE PLANTS 0 WILL REVISIT SEVERAL ISSUES REMAINING FROM PREVIOUS MEETINGS 0 WILL DESCRIBE THE MAAP MODELS THAT TREAT FISSION PRODUCT BEHAVIOR 0 WILL PRESENT RESULTS FOR KEY ACCIDENT SEQUENCES IN THE FOUR REFERENCE PLANTS
-                          0 WILL PRESENT LIMITED UNCERTAINTY AND SENSITIVITY ANALYSIS RESULTS i
ELF /Is I
I
                                                                                                                  - ~ . _ _ , _ - _ _ _ _ _

n       .
SEVERAL ISSUES DISCUSSED AT PREVIOUS MEETINGS WILL BE BRIEFLY REVISITED 0    IN-VESSEL HYDROGEN PROUCTION 0    HYDROGEN COMBUSTION BEHAVIOR 0    CORE DEBRIS-CONCRETE INTERACTIONS 0    IN-VESSEL FUEL / COOLANT INTERACTIONS 0    SENSITIVITY AND UNCERTAINTY' ANALYSIS 1
i i
ELF /Is l
l
    - - .   .   - - -   ,.     -.-      .       _ _ ~ . . - . . - - . - . . . . . . - _   .   . _ , _ - . . . -.

                                                                     . 4 THE KEY MODELS IN MAAP FOR FISSION PRODUCT RELEASE, TRANSPORT, DEPOSITION, AND REVAPORIZATION WILL BE DISCUSSED 0    RELEASE FROM FUEL BASED ON MODELS IN THE FPRAT CODE 0    FISSION PRODUCT TRANSPORT BASED ON TIGHT COUPLING WITH MAAP THERMAL-HYDRAULIC MODELS 0    FISSION PRODUCT DEPOSITION BASED ON CORRELATIONS DEVELOPED BASED ON OBSERVED EXPERIMENTAL BEHAVIOR 0   FISSION PRODUCT REVAPORIZATION BASED ON THERMODYNAMIC, HEAT TRANSFER, AND THERMAL-HYDAULIC CONDITIONS J
W d
ELF /Is

i,. .
RESULTS OF INTEGRATED ANALYSES AND EFFECTS ON FISSION PRODUCT RELEASE TO ENVIRONMENT WILL BE PRESENTED FOR THE FOUR IDCOR REFERENCE PLANTS 0   ZION: Tf1LB'         AND V SEQUENCES 0   PEACH BOTTOM:          TW, TC,.TQVW, AND SIE SEQUENCES 0   SEQUOYAH:    S2HF, TMLB', AND V SEQUENCES 0   GRAND GULF:          T230W, T23C, AE, AND T100V SEQUENCES ELF /Is

CONTAINMENT LOADS ISSUES FROM THE MAY 15th - 16th IDCOR/NRC INTERACTION MEETING Robert E. Henry Fauske & Associates, Inc.
16WO70 West 83rd Street Burr Ridge, Illinois 60521 (312) 323-8750 NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Fission Product Behavior Rockville, Maryland August 28 - 29, 1984 i

                                              /
IN-VESSEL HYDROGEN PRODUCTION e Consider bounds of outer cladding surf ace and outer plus inner surf ace available for oxidation.
Upper bound results in about a 50% increase in H generated before core slump.
2 e Upper bound (twice cladding surface area) results in about twice the H2 generated following core slump.
e Water swell covering damaged regions can also increase the H2 generated before vessel failure.
e The IDCOR results are insensitive to these bounds for in-vessel hydrogen generation.

HYDROGEN COMBUSTION BEHAVIOR
          ~                                                      -
Considered issues related to H2 reaction at high temperatures and high H 2O partial pressures, flame
_ temperature and H 2 fl w path to and from the RC. _
e   While other reactions may occur at high temperatures, there is no apparent cutoff due to high steam partial .
pressures.
e   incorporated the Westinghouse flame temperature correlation based upon measurements with high steam partial pressures. Containment responses are insensitive to large variations in the flame temperature.
e   incorporated natural circulation between the primary system and containment after vessel failure including H-2 e Containment analyses do not demonstrate H 2 concentrations sufficient for detonation in any compartments.
i 1

IN-VESSEL FUEL-COOLANT INTERACTIONS (HYDROGEN GENERATION)                               ,
l e Calculate the steam generation by debris slumping into the lower plenum.
e Calculate the level swell of lower plenum water due to steam generation.
e Calculate the overheated regions which can be covered by the level swell.
e Calculate additional oxidation due to the level swell.

[
CORE-CONCRETE INTERACTION IN MARK 1 BWRs e Considered Several Different Debris Configurations
                 - uniformly distributed within the pedestal and drywell
                - distributed over the pedestal floor and l                          one-fourth of the drywell floor
                - restricted to the pedestal l
l 9 No Significant Changes for Time to Containment Failure or Calculated Releases to the Environment k                                                                                                   ~
l l
l l      . .. .      - - - -      ---

SENSITIVITY AND UNCERTAINTY ANALYSES
    # Uncertainty analyses were performed with MAAP 1.2 on individual parameters (models) for thermal-hydraulics. Results presented at Meeting 3A.
    # No large influences found with respect to whether containment f ailure occurs and if so the time of occurrence.
O Additional uncertainty analyses have been performed with MAAP 2.0 - integrated fission product release and deposition.
e Uncertainties considered
       - Sequence definition
       - Vapor pressures of volatile fission products
       - Aerosol sedimentation                                    ,

f I
>                                                                                                                                                                               .+
3 IN VESSEL HYDROGEN PRODUCTION
                                                                                   - A COMPARISON OF NRC CONTRACTOR AND IDCOR ESTIMATIONS i
JAMES T. HAN, NRC-RES NRC/IDCOR ltETING ON INTEGRATED ANALYSIS OF FISSION PRODUCT BEHAVIOR i
I
+
ROCKVILLE, MARYLAND i                                                                                           AUGUST 28-29,1984                                                                     '
l i
r-f i
e I
4

  .- 3 THE PRESENTATION IS BASED ON THE FOLLOWING REFERENCES o NRC CONTRACTOR CALCULATIONS USING THE PARCH CODE:
: 1. GIESEKE, J. A., ET AL., "RADIONUCLIDE RELEASE UNDER SPECIFIC LWR ACCIDENT CONDITIONS," BMI - 2104, VOLS, I - VI (DRAFT PUBLISHED IN JULY 1984).
: 2. HANDOUTS FROM THE FIRST NRC/IDCOR MEETING (IN HARPER'S FERRY, WEST VIRGINIA, NOVEMBER 29 - DECEMBER 1, 1983) AND FROM THE THIRD NRC/IDCOR MEETING (IN ROCKVILLE, MARYLAND, MAY 15 - 17, 1984).
e IDCOR CALCULATIONS USING THE MAAP CODE *:
: 1. IDCOR TECHNICAL REPORTS 23.1 FOR ZION, SEQUOYAH, PEACH BOTTOM, AND GRAND GULF (DRAFT PUBLISHED IN JULY 1984),
: 2. SAME AS NO, 2 AB0VE,
: 3. IDCOR TECHNICAL REPORT 12,1, " HYDROGEN GENERATION DURING SEVERE CORE DAMAGE SEQUENCES," (JULY 1983).
4,   IDCOR TECHNICAL REPORT 15,1A, "IN-VESSEL CORE ITLT PROGRESSION PHENOMENA," (JULY 1983).
* IDCOR TECHNICAL REPORTS 23,1 SUPERSEDE OTHER IDCOR RESULTS SHOULD ANY DIFFERENCES ExlST.

QUANTITATIVE COMPARISON OF IN-VESSEL HYDR 0 GEN PRODUCTIONS CALCULATED FOR VARIOUS ACCIDENT SEQUENCES e NlEBERS GIVEN ARE THE IN-VESSEL HYDROGEN PRODUCTIONS EQUIVALENT TO THE FRACTION OF ALL ZIRCALOY IN THE CORE OXIDIZED.
e HYDROGEN PRODUCTION FOR ALL ZIRCALOY IN THE CORE OXIDIZED:
ZION - 1950 LBs          SEQUOYAH - 2230 LBs GRAND GULF - 76G0 LBs               PEACH BOTTOM - 6330 LBs e PWR - TMLB' PLANT          ZION                               SEQUOYAH BCL            0.51 (0.28 BEFORE CORE SLUMP)      0.49 (0.25 BEFORE CORE SLUMP)
IDCOR         0.15                               0.34 i                  RATIO
* 3.4                                1.4
* RATIO = BCL RESULT /IDCOR RESULT e PWR, ice CONDENSER CONTAINMENT - S HF 2
0 66      57 BEFORE CORE SLUMP)
IDCOR         0.39 - 0.40 RATIO          1.7 e PWR - S 2D PLANT          ZION OR SEQUOYAH BCL            0.85 (0.73 BEFORE CORE SLUMP) FOR ZION IDCOR        0.30 FOR SEQUOYAH
                                                                                  ,rm   --w-=     - + -   w--,

        -QUANTITATIVE C0FPARISON OF IN-VESSEL HYDROGEN PRODUCTIONS CALCULATED FOR VARIOUS ACCIDENT SEQUENCES (Corn'D) e BWR/6, VARK III CONTAINMEfR - TQUV PLANT          GRAND GULF BCL            0.40 (0.18 BEFORE CORE SLLNP)
IDCOR          0.056 (0.0013 FOR A CASE WITH DEPRESSURIZATION)
RATIO          7.1 e BWR/4, PARK I CONTAINMENT - W PLANT          PEACH BOTTOM BCL            0.61 (0.60 BEFORE CORE SLlFP)
IDCOR          0.068 RATIO          9.0 e BWR/4, MARK I CONTAINMENT - TC PLANT          PEACH BOTTOM BCL            0.47 (0.26 BEFORE CORE SLUMP)
IDCOR          0.047 RATIO          10.
i j
L

e  IDCOR ESULTS AE OBTAINED BY ASSUMING (1) AS A CUT-OFF TBfERATURE OF 2300 K (OR USER INPUT) IS PEACHED, BLOCKAGE FORMS IN THE CHANNEL AND SHUTS OFF STEAM SUPPLY AND STOPS HYDROGEN PRODUCTION, AND (2) HYDROGEN PRODUCTION DUE TO CORIUM - WATER INTEPACTION IN THE VESSEL LOER PLENUM IS NEGLIGIBLE. THEREFORE, LOWER HYDROGEN PRODUCTIONS AE CALCULATED, e  ON THE CONTRARY, BCL RESULTS ARE OBTAINED BY ASSUMING (1)
NO BLOCKAGE F0WS AND IELTING NODES REMAIN IN THE CORE FOR OXIDATION UNTil CORE SLulPS, (2) AS CORE SLUMPS, MORE STEAM IS GENERATED TO FURTHER OXIDIZE THE REMAINING FUEL RODS IN THE CORE, AND (3) HYDROGEN PRODUCTION DUE TO CORIUM - WATER INTERACTION IN THE VESSEL LOWER PLENUM IS NOT NEGLIGIBLE.
e   IT SHOULD BE NOTED THAT THE TMI-2 ACCIDENT PRODUCED AN AMOUhT OF HYDR 0 GEN EQUIVALENT TO ABOUT 50% OF ALL ZIRCALOY IN THE CORE OXIDIZED. REPEATED PROCESSES OF CORE UNC0VERY - WATER B0ILOFF - CORE REFLOOD ARE RELIEVED TO PLAY AN IMPORTANT ROLE IN HYDROGEN PRODUCTION, i
l

v RECENT RESULTS FROM HYDROGEN RESEARCH PROGRAMS o MULTIPLE BURNS APPEAR LESS LIKELY WITH CONTINUOUS RELEASE OF HYDR 0 GEN WITH IGNITION SOURCE AVAILABLE (DEPENDS ON RELEASE RATE AND H2/ STEAM RATIO) o AIR AND WALL TEMPERATURES SUBSTANTIALLY HIGHER IN UPPER VESSEL THAN LOWER FOR PREMIXED AND CONTINUOUS INJECTION BURNS o TURBULENCE SIGNIFICANTLY ENHANCES BURNS PROPAGATION, RATE AND COMPLETENESS o BURNING HYDROGEN AT CONCENTRATIONS AB0VE 9% CAN THREATEN EQUIPMENT PERFORMANCE o TEMPERATURE HAS A SIGNIFICN.tT IMPACT ON H 2/ STEAM DET0 NATION LIMITS AND REDUCES THE EFFECTIVENESS OF STEAM TO PREVENT DETONATIONS o DETONATIONS HAVE BEEN OBSERVED IN CONCENTRATION OF 13% HYDROGEN o TRANSITION FROM DEFLAGRATION TO DETONATION CAN BE INITIATED WITH WEAK SOURCES AND INDUCED BY SIMPLE OBSTACLES

           "NEEDED HYDR 0 GEN-SAFETY RESEARCH FOR LARGE DRY PWRs" o ASSESSMENT OF H / STEAM TRANSPORT, MIXING AND CONDENSATION FOR SELECTED 2
ACCIDENT SEQUENCES o ASSESSMENT OF POTENTIAL DETONATION IGNITION SOURCES o ASSESSMENT OF GE0 METRIC CONSIDERATIONS FOR FLAME ACCELERATION AND TRANSITION FROM DEFLAGRATION TO DETONATION o ASSESSMENT OF CONSEQUENCES FROM GLOBAL OR LOCAL DETONATIONS o ASSESSMENT OF GLOBAL AND DIFFUSION FLAME BURNING ON EQUIPMENT SURVIVAL o ASSESSMENT OF THE POTENTIAL FOR DETONATION COINCIDENTAL WITH REACTOR VESSEL FAILURE

5 "NEEDED HYDROGEN-SAFETY RESEARCH FOR ICE-CONDENSER PLANTS" o ASSESSMENT OF H2 / STEAM TRANSPORT AND POTENTIAL FOR DETONATION IN THE ABSENCE OF FORCE CIRCULATION o CONFIRMATION OF THE EFFICACY OF IGNITERS IN THE PRESENCES OF SPRAYS, FANS AND CONDENSATION EFFECTS o ASSESSMENT OF GE0 METRIC CONSIDERATIONS FOR FLAME ACCELERATION AND LIKELIHOOD OF TRANSITION FROM DEFLAGRATION TO DETONATION o ASSESSMENT OF THE STRUCTURAL CONSEQUENCES FROM A GLOBAL OR LOCAL DETONATION o CONFIRMATION OF THE EFFECTS OF HYDROGEN BURNING ON EQUIPMENT SURVIVAL o ASSESSMENT OF THE POTENTIAL AND CONSEQUENCES OF AUT0 IGNITION OF HYDROGEN AT RELEASE / BREAK LOCATION

NEEDED HYDR 0 GEN SAFETY RESEARCH FOR MARK III CONTAINMENTS o CONFIRMATION OF CONDITIONS LEADING TO DIFFUSION FLAMES IN WET-WELL REGION o ASSESSMENT OF LOCALIZED THERMAL LOADS FROM HYDROGEN DIFFUSION FLAMES o ASSESSMENT OF THE CONSEQUENCES OF THERMAL LOADS ON SAFETY RELATED EQUIPMENT
                                             ~
o

C0RE-CONCRETE               INTERACT 10N
    ,                           IN   MARK      I  B W R's BY S. B. BURSON CONTAINMENT SYSTEMS RESEARCH BRANCH U.S. NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 AUGUST 28, 1984 i
s

    ,                                                                                                                                                                                                                 a                       ***
f.
OUTLINE l
0               CLWG OBJECTIVES <
i 8              MODES OF CONTAINMENT FAILURE O              PLANT SPECIFIC DESIGN FOR MARK 1 PROBLEM j.
;                                     9               RESULTS OF CLWG/CPWG DRYWELL FAILURE TIMES
:                                                  -VS FAILURE MODE 4               CONCLUSIONS I
4 4
4 1
4
      .- _ . . - --       ~. _ . _ _ . _ _ _ . . . . _ _ _ _ , . . _ _ _ . . _ - _ . . , - , _ .       , . - . . - . . _ _ - . _ . . . . _ - _ - . _ . _ _ _ . _ . _ . _ . _ . _ _ _ . . . _ . _ _                    . . _ . . _ . . _ . .

CLWG OBJECTIVES 8 TO MECHANISTICALLY MODEL CONTAINMENT BEHAVIOR OF VARIOUS REACTOR DESIGNS UNDER A VARIETY OF " STANDARD PROBLEM" ACCIDENT SEQUENCES.
8 IN PARTICULAR, TO PROVIDE A STANDARD METHODOLOGY FOR BWR MARK I CONTAINMENT ANALYSES.
8 PROVIDE CONSENSUS VIEWS'IN AREAS WHERE CALCULATIONS CAN BE PERFORMED WITH CONFIDENCE.
8 IDENTIFY MODELING UNCERTAINTIES AND VARIANCES IN METHODOLOGICAL ASSUMPTIONS THAT SIGNIFICANTLY IMPACT THE TIMING AND MODES OF CONTAINMENT FAILURE.
l

                                                        . s.
MODES OF CONTAINMENT FAILURE 8   OVERPRESSURIZATION OF DRYWELL
    - PRESSURE EXCEEDS 132 PSIA CAPACITY
    - GROSS STRUCTURAL FAILURE OF LINER ENSUES
    - CONSIDERED BY CLWG AND IDCOR
    - RESULTS REPORTED: NRC/IDCOR MEETING 5/17/84 8   OVERTEMPERATURE FAILURE OF SEAL MATERIALS
    - HIGH TEMPERATURE DEGRADATION OF POLYMER SEALS
    - LEAKAGE THROUGH SEALS PRIOR TO GROSS FAILURE
    - CONSIDERED BY CLWG/CPWG
    - RESULTS REPORTED: CPWG DRAFT REPORT NUREG-1037 0   OVERTEMPERATURE FAILURE OF LINER
    - CONTAINMENT ATMOSPHERE TEMPERATURE EXCEEDS 1200 F
    - STEEL LOSES STRENGTH AND CREEPS
    - CONSIDERED BY IDCOR
    - RESULTS REPORTED:    IDCOR REPORT 23 1 0   ABLATION FAILURE OF DRYWELL LINER
    - HIGH TEMPERATURE CORIUM FLOWS TO STEEL LINER
    - ABLATIVE ATTACK CREATES FLOW PATH INTO REACTOR BUILDING
    - CONSIDERED BY CLWG
    - RESULTS NOT PREVIOUSLY REPORTED

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4 .a.
                                       !                  g DRYWELL LINER g
                                                   )
f                 ] 2-1/4 n POLVESTER FOAM i
f  i 3 2-3/8 a FIDERGLASE
:  a f  ;         ,
                                                            $; FIDERGLA&s LAMINATE FORM
                                            -                      SAND j               l   .y     CONCRETE
                                                       . g.      ,
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N CORE DEBRIS .                                           ELO jhh a
Drywell Liner - Concrete Shield Wall Gap Geometry

                                                                                                                                                                                                                                                                               'r C0W ARISON OF APPR0XIMATE DRYWLL FAILURE TIMES SY OVER PRE 5SURE. OVER TEMPERATURE. AND LINER MELT-THROUGH CLtd6                                 CPWG      LINER ClifG DERRIS TEMPERATURE      MAXIMUM M YW LL+ DVERPRESSURE                   OVER TEMPERATURE    ELT-THROUGH CASE CONCREft CMPOSITION          P AND T        FAILURE (MIN)                  FAILINIE(MIN)     FAllllRE(MIN) l          2550 K,             145 psia             133                                  62              3.5 Limestone            622K(660F) 2          1755 K,              88 psia             500*                                329           45 Limestone            533K(500F) 3          2550 K,             108 psia             460*                             Mo Leakage          5.5 Basalt             477K(400f)                                             Calculated 4          1755 K,              65 psia             950*                             Mo Leakage          Nn Basalt             411K(280F)           Failure                           Calculated Melt through Unlikely                                       Calculated
* Entrapolated value.
* Menimum during five hours of core / concrete interaction.

CONCLUSIONS' I   FOUR MAJOR DRYWELL FAILURE MODES
      - STRUCTURAL OVERPRESSURIZATION
      - OVERTEMPERATURE FAILURE OF PENETRATION SEALS
      - OVERTEMPERATURE CREEP 0F DRYWELL LINER
      - LINER ABLATION BY MOLTEN CORIUM 8
MAJOR DIFFERENCES IN CALCULATED FAILURE TIMES
    - OVERPRESSURE:        2-10  HOURS
    - SEAL FAILURE:        1-5   HOURS
    - LINER CREEP:         > 10  HOURS
    - LINER ABLATION:      MINUTES I
ALL FOUR FAILURE MODES ARE POSSIBLE AND MUST BE CONSIDERED VARIATIONS IN CONTAINMENT MODELING ASSUMPTIONS MAY HAVE SUBSTANTAL IMPACT ON CALCULATED FAILURE MODES AND TIMING 8
ANALYSIS MUST BE REACTOR AND ACCIDENT SEQUENCE SPECIFIC WITH PARTICULAR ATTENTION TO THE IMPACT OF MODELING ASSUMPTIONS ON EACH FAILURE MODE.
l

1 f
CORRELATION FOR FISSION PRODUCT DEPOSITION BY AEROSOL SEDIMENTATION Robert E. Henry Fauske & Associates, Inc.
16WO70 West 83rd Street Burr Ridge, Illinois 60521 (312) 323-8750 NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Fission Product Behavior Rockville, Maryland August 28 - 29, 1984 k
I r                                    -

FISSION PRODUCT REMOVAL MECHANISMS MODELED IN MAAP
                   # Vapor Deposition (Diffusion) i
* Steam Condensation i
l
* Sedimentation

VAPOR DEPOSITION DAM wt   [ P. - P, )
D  RT      ( 6      )
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: t. hr

1 AEROSOL DEPOSITION BY STEAM CONDENSATION Steam Velocity W            Oc l
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This allows a large number of experiments to be combined on a universal plot, 1
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                                  ~
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NA20 AEROSOL o
I- i i i lillit        i i i lilill    I i i lilill    I i i ililli    l i i lilill    i l i liiE o, =                                                                                                 =
             !=                                                                                              =
i
               !                                                                        LEGEND q?
l-                                                                           a #ERI RUN 13        -
+
c     =                                                                                             _
       ~\~                                                                                                  E 7         ::                                                                                                :
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'Cl W~        :-
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MODEL PREDICTION i     ;-                                                                                               -
Cl p
s                                                                                             _5 c-                                                                                               -
7      C                                                                                                Z O!
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I I I illf!!    I I f i lflil   i I I Illill   i I 1 1I1111    I I I 181111     I I t illil 2                3              4               5                6                7 10 :             10               10             10              10              10              10 TIME (SEC)
Comparison between the correlation and JAERI Run 13 for lean concentrations.

m.ii i i   iino i s i       oittiii i       iiiiiii i i   aittii i i i!n n i i i e E
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                                                                                                 ~_.

t i
l                  I 1
(a)
DRY WATER y
sSOLID (b)
WATER ACCUMULATION

AVERAGE DENSITY INCREASE t
i P
                                 /       eff 3 Pp=Pggg + 1    1 p      p, influence of Condensation i
Aw U,       hs A c*    y        p, y

e .<,
:               I      i i
                                                                              ~
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  ~
w 0
           -                                  *                              ~
o   10"                                                    WITH WATER U                                                          ACCUMULATION 7                                                 .
E                                                   .
8                                                       .
a 8                                                          .
  @ 10-2   _                                                                  _
g           WATER ACCUMULATION PLUS CONDENSATION                                  .
a a
10'3 -                                                                  -
WATER ACCUMULATION PLUS 5 X CONDENSATION                 a 10'#                            40       100 200 400          1000 4        10    20 TIME, min.
Comparison with NSPP experiments for wet ( A Run 612) 0 /U and  dry (SThe aerosols. Runinfluence 631) atmospheres           using Fe$i$n }0s8also of wall condensa illustrated.

s          I    8       I     I    I     i
                                                                     ~
10 ORY AEROSOL
      ^
w 10''
o o
7                                              WITH WATER y                                              ACCUMULATION 8
_a 8
O 10-2   _                        q                           _
N          WATER ACCUMULATION PLUS CONDENSATION        O o
O 10~3 -                               o                    -
O o
            -4      I       I   I     I      I     I   t      i 10 4       10  20   40      100 200 400     1000 TIME, min.
Comparison with NSPP experiments for a wet atmosphere (O Run 611 and O Run 613) using Fe 0 00 aerosols.
The influence of wall condensation i$ 2 s llustrated.

 ^
w E
N e
 $                                                                       i
(     10                  I   I    I      I  I   I    I    I z
DRY AEROSOL PLUS CONDESATION U
z                            O 8      ggi -                     O                            -
O E                                     O         O w
0 i
O          O R            WATER ACCUMULATION j            PLUS CONDENSATION z                                                        O O
N    163   -                                                  -
g              O CONCRETE COMPONENT
$              0 Fe23O   COMPONENT                  g 0                                                        0 o                   ,     ,   ,    ,      ,  ,    ,   ,    ,
10
_4 4     10 20 40       100 200 400 1000 2000 g                                TIME,(min) w 4            (Time Since Termination of the Aerosol Source)
Comparison of the model with materials-which do not coagglomerate.
l

l WATER ABSORBED BY HYGROSCOPIC AEROSOLS MWe  = 2 Z[M w     [    R M1         \ M,9 / \ 1'- R /
Pst R=
P sat Water Mass Added to the Aerosol Mass Mass k

i            MARVIKEN TEST COMPARISON Quasi-Steady State W, = Ws+Wo p Q = C, VF'+pQ g
p; = Cy QY   p 'O f9 l

J CONDITIONS FOR MARVIKEN TEST 2b Nominal Temp.        Flow Rates of Injected Materials
(*C)                        (10-3 kg/s)
Geometry                        Gas       Surface            Gas                        Fissium Pressurizer                      400         350          Steam (42)                   Cs0H(11.4)
Piping                       100-300     100-200  Noncondensables (10.4)              Csl (1.7)
Relief Tank                     < 100       < 100                                       Te (1.6) f
('                                                                                                  )

O '' 8 COMPARISON WITH THE MARVIKEN EXPERIMENTS Pressurizer and Relief Tank Deposition for Test 1 Proportion of Mass injected Cs            i        Te Pressurizer Model Prediction       70%          70%       70%
Experiment            34%            26%      35%
Relief Tank l       Model Prediction       23%           23%      23%
1 Experiment             23%          23%       16%
1 Pressurizer Deposition for Test 2b Proportion of Mass injection Cs             i       Te Model Prediction      75%           75%       75%
Experiment           43.9%          44.2%    43.6%
k

i e. ,

==SUMMARY==
OF CSE CONDITIONS Parameter            Run A-2        Run A-5 Run A-ll Vessel Volume, m 3              596            596     596 Vessel Diameter, m              7.6            7.6     7.6 2
Total Surface Area, m           571            571     571 Wall Area, m 2                  4)g            4)g     4jg Settling Area, m 2            45.5            45.5   45.5 Temperature, K                  358            396   396*
Pressure, MPa                 0.16            0.32   0.34*
Vessel Height, m               12.2           12.2    12.2 Isothermal                      Yes            Yes     No Thermal Insulation              No            Yes*   Yes*
on Walls Previous Steam                  114            586    1500 Exposure, hr Aerosol Release                 10             10      10 Duration, min Avg. Steam                    0.082           0.037     0 Makeup, kg/sec
* Initial condition.
1

       ,0-s.
:                           RUN A-2 E        C                          O CESIUM
 *h10-6[
O URANIUM l
o IODINE Z                  i 9             -
i
>             -                                                            1 4-0: [0-7 :
I-Z             :
              ~
W O             _
Z O             -
1 O
w 10-8 g         _
m             :                                                            l 4             -
I Q.
:         4 m
O 1       -
$. 10         5 8            _-
g LODINE, CESIUM 8 URANIUM
             -                                   A
[g-lO           '    '     '       '    I O          4     8    12      16   20    24 t,hr Comperison of the predicted removal rates due to steam condensation and sedimentation to the measured data from Run A-2.

1lllllll                                                    'l l     ll         lll          ll             -                  lIillill            l1lll N
O O e4@ Qy4@w                       OZoWzf4pO                              m
                                                     .                                                 *xoNE g                   1                   1                      1                l             1 g-I 0-                  0-                     0-               g-            0-C                   8                      8                    7                 6           .5 0          - ~ _ _~ -     ~       _~=_             _            :--   - -        =_      - -
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a                                 I NN ASI R
EI    U  U C                                           UM   N 2                                             E                                          M 0
i S                                                A-I U                                                5 M
2                            q                   N 4
           ]

10-5                                          ,,g RUN A-ll IODINE O CESIUM to(         y\A                                                  ;
E 10-6       A                    O URANIUM
         ~
G                      IODINE F-         -      E 4                               O G
f    10-7  --
a z         :
w         -                                  G o         -
A    URANIUM           '
      $         -                        & CESIUM o                                            A       O w    10-a _
m         :
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                ~
(n
(.o  10 1  --
num o         -
B B
0      4 8     12      16     20    24       48 t,hr k                                                           j

4 UNCERTAINTY ANALYSES FOR THE SEDIMENTATION CORRELATION A = C 9" Vary n Sufficiently to Bound the Data Example n = 0.6 0.5 < n < 0.7 i
l l

o                                   SODIUM OXIDE & IRON OXIDE AEROSOLS ou
     -        t            I i i 181111       1    1 1 1 1 l l 11    I   l I 11 l l11      l  1 I i 1 111
:'3:_a E '
                                                                                                         ~
              '_^         OW                                                                             _
HEDL              -
Eg                                                                        ~
T                         OY     g               ZZ :      0 50                    0 TEST ABt O                              D                                                 4 TEST AB5         _
n           -
                   .               T U                                                                  b
            .._                         o                                                                _
m-dA 7A g                                                                 _
L.                                                                 -
A O 74       ,
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                                          @0                                          m sun sot          E
_.                                          03                                        A RUN 102          [
O D     j                                  y RUN 103          -
(f"e 9*O                             O RUN 109 V RUN 106 7                                                  hbd                              O RUN 107 O                                                      g                         X RUN Sit (IRON)   _
                                                               ^                                        E L                                       _
r I
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i   r it!ini        i    i ; i iii         i    i iiiini          i  ! i iiin 4
     "_1; J
10 L
10 ~                10 "                  10 ZZx(T-To)+1 l
1 l
1

  .~ .
o                                   SODIUM OXIDE &. IRON OXIOE AEEJSOLS CL   '
           -                 i   i i itilli        i   i l i l illi      i  i i i l illi       l  i i i l i li g
O
                        <                                                                 HEDL               -
                          .O      %
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0 TEST A91 A TEST ABS
                          &% 9      ~                                                                       =
v                                                                       r X    ogg                                                                     :
X   Vg ,                                                                  -
        ?
O
                                          $E A 7
ORNL
                                                                                                            ~
       '"                                       0 3                                           m Run tot          3 O                                 A PUN 10~_         [
07                                        y RUN 103          -
O      8                               e RuN 109           -
n g{
7 RUN 106           _
i                                              7"A                              O RUN 1Q7 C                                                  g                          X RUN 511 (!RON) _ _
a                                            :
a                                          :
1 t
o" u           i     i ii!!in         i    i i iirin        i  i i iirin          i   i i i sin 1                       2                    3                   4 1F ~i                       10                    10                    10                   10 ZZx(T-TO)+1 l
i l

o                             URANIUM OXIDE AEROSOLS o.
4             1  I i 11lll1       1 1 I 1 11111       1 1 1 111111       I   I I I l l it c -
                         -6           7                                                                      _
7            A        V                 22 =    0 50 O =                e OVv                                                                            '
                     - -                                                                                     =
v                                                               _
2          A         V O   7                                                           -
g O'                                                                         ORNL uo s .-. -
_                             y                                                     _
u         :                                                               a su 20t           :
Z A RUN 203
                         -                                                                y NN M4            _
e mn 2os           -
_                                                                o RUN 206          _            )
7                                                                       A NH 207                        l O_                                                                   V mN 209            _
l 1
1 1
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l

o                                                         URANIUM OXIDE AEROSOLS o.
W       _,
I        i l iilill                                                    i I I I I lill    l     I i l iIlli                              l   I i i i 112 el     '                                                                                                                                                                2 7         -a kV                                                                           ZZ =    0 70 O_          ,0                                                                                                                                                             _
             -:                      y                                                                                                                                                  :
Y E O 90                                                                                                                                                                3 O                Y                                                                                                                                   -
O'                                                                                                                                                    ORNL uO_                                                                y                                                                                                             _
nE :                                                                                                                                                 . - =
A WN M3 E
[
_                                                                                                                                         y e n 2o4                   -
e RuN 20s                  -
_                                                                                                                                          O MN 206                   _
7                                                                                                                                                   A NH 207 O_
v =N 2o.                   _
2 I
O           I       I I IIllil                                                     i I I I l illi    i     l ii lill                                i   I I I l ill O                                                               1                          2                                             3                            4 10                                         10                                               10                                             10                       10 ZZxCT-To)+1 l
t I
l

==SUMMARY==

e MAAP calculates aerosol removal due to vapor deposition, steam condensation and sedimentation.
* MAAP models for vapor and steam condensation removal processes are in agreement with experimental results.
e Sedimentation model compared to:
Dry aerosols
                                           -  Limited water affinity aerosols High density, liquid aerosols Hygroscopic aerosols I
e

   ... .                                                                             l l
REEVOLUTION OF FISSION PRODUCTS
]
Robert E. Henry Fauske & Associates, Inc.
16WO70 West 83rd Street Burr Ridge, Illinois    60521 (312) 323-8750
;                NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Fission Product Behavior Rockville, Maryland
.                          August 28 - 29, 1984

RESUSPENSION o Material layers deposited on vertical surf aces would be very thin. Typical gas velocities associated with primary system or containment failure would be insufficient to overcome attractive forces.
e  Materials dissolved in water and subsequently dried would be caked on the surface and would not be resuspended.
e Liquids or hygroscopic materials would stick to the surface and not be resuspended.
* Debris beds accumulated in either the primary system or the containment would not be sufficientl'y deep to be dispersed by the depressurizati~on f ollowi'ng f ailure.
e  Debris beds accumulated near a f ailed penetration would not experience velocities sufficient for entrainment.
  - - - - - - ~   m       ,    e      -                - - .- -

       .= 0 REEVOLUTION OF FISSION PRODUCTS 4
o Considered resuspension of particulate layers deposited within the primary system and containment.
e Calculate revaporization of deposited materials due to partial pressure differences and the convective flows.
e inert materials are not revaporized - a conservatism in the analyses.
e Solution chemistry is currently not credited -
also a conservatism in the current analyses.
I I

  .~   .
CALCULATIONAL PROCEDURES FOR REEVOLUTION OF FISSION PRODUCTS by James A. Gieseke Presented at the NRC/IDCOR MEETING
;                           August 28-29, 1984 i
l 0Ballelle Columbus Laboratories t

o o, COMBINED TRAP-MELT / MERGE CODE PURPOSE:      DETERMINE THE EFFECT OF INTERNAL HEAT SOURCES ON STRUCTURE SURFACE TEMPERATURES AND FISSION PRODUCT REEVOLUTION METHOD:       DIRECTLY COUPLE TRAP-MELT AND MERGE e ELIMINATE ROUND-ROBIN TECHNIQUE AND AVERAGING INTERVAL CONCEPT e TRANSFER UPDATED INFORMATION BETWEEN CODES AT EVERY MARCH TIME STEP 1
l                                                    OBallelle Columbus Laboratories l          . --

                                                                                                     }
PREVIOUS TECHNIQUE l
1 i
:= (Continuous Time)                     (20 Time Intervals)                 ==
l l                                                         Fluid & Structure Core Outlet                       Thermal-Hydraul ic MARCH                                 :-   MERGE                            :-   TRAP-MELT Conditions                        Conditions Down-stream of Core odified Thermal                                                  Transport & Deposi-Conditions of                                 -
tion of Fission Structures Includ-                     MERGE l
i:
Product species and ing Fission Produc                                                Aerosols Heating I

NEW TECHNIQUE
              =     MARCH OUTPUT FILE MARCH Read                    Core              g  Problem Outlet                  Conditions           Time y
Primary Coolant DRIVER                   =     MERGE             &     System States &
ROUTINE                                                Structure Temps h
s, f           's            Fluid & Structure Fission Product                      i
                                                            $          Themal-Hydraulic Transport and                           1 MARCH Deposition lTimeStepl i
Conditions Down-
                                          \'              ,e           stream of Core TRAP-MELT       e l
ission Product Distributions

.m o IMPROVEMENTS TO MERGE e ADDITION OF A MORE PHYSICAL FISSION PRODUCT HEATING MODEL e ELIMINATION OF TIME-AVERAGED THERMODYNAMIC QUANTITIES AS INPUT TO TRAP-MELT e TIME STEP CONTROL AND STEAM PROPERTY ITERATIVE SCHEME SMOOTHER AND MORE RELIABLE s
e GENERAL CLEANUP /0RGANIZATION AND DESCRIPTIVE COMMENTS ADDED e UPGRADED TO FORTRAN 77 OBallelle Columbus Laboratories

MERGE FISSION PRODUCT HEATING MODEL
                                            .              Structure a        '
o O
   ,                          g,               e           Gas w/airbcrne F.P.
i    I O
1
                              \
                                             .             Structure w/ deposited F.P.
e  EMITTED FISSION PRODUCT ENERGY IS ALLOCATED AMONG STRUCTURES AND GAS BASED UPON SIMPLE GE0 METRIC RELATIONS e  s AND y ENERGY TREATED SEPARATELY
     -- ALL y ENERGY DEPOSITED IN STRUCTURES a ENERGY ALLOCATED BETilEEN GAS AND STRUCTURES DEPENDING UPON VOLUME DIMENSION AND s TRACK LENGTH e  ENERGY ADDED HAS THE FORM: E = Eg *T*R*G where Eg = INITIAL (TIME = 0) ENERGY T = FRACTION REMAINING DUE TO DECAY R = FRACTION RESIDENT WITHIN VOLUME G = GE0 METRIC RELATIONSHIPS l
l 1
l

I IMPROVEMENTS TO TRAP-MELT 2 e USE OF GAS PROPERTIES APPROPRI ATE FOR STEAM /H2 MIXTURES PROVIDED BY MERGE e ADDITION OF MULTIPLE SURFACE ORIENTATIONS WITHIN EACH VOLUME HORIZONTAL FACING UP HORIZONTAL FACING DOWN VERTICAL e VAPORIZATION OF DEPOSITED SPECIES USING RA0VLT'S LAW i
OBatteile Columbus Laboratones

MAAP MODELING OF FISSION PRODUCT RELEASE AND TRANSPORT IN THE PRIMARY SYSTEM AND CONTAINMENT Robert E. Henry Fauske & Associates, Inc.
16WO70 West 83rd Street Burr Ridge, Illinois                                                   60521 (312) 323-8750 NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Fission Product Behavior Rockville, Maryland August 28 - 29, 1984 l

MAAP FISSION PRODUCT RELEASE o          FPRAT - Steam oxidation model for release of fission products from the fuel.
e         MAAP module - FPRATP (PWR) and FPRATB (BWR).
o         Gas flow controls the release from the core -
Minimum flow allowed is that required to remove i             volatile fission products.
o        Aerosols formed in the upper plenum.
e          FPRATP compared to FPRAT for release rates and magnitudes.
o         NUREG-0772 models for fission product release are included and can be specified by the user, k

i 9 _.........g.........g ......      g......     . j.. .. ..g iir.....g ........g.........g.........g......... ,
:        a FPRAT                                                                                         -
                                                                                                                              ~
:        A FPRATP                                                                                        ~
                                                                                                                              ~
y NUREG-0772                                                                                    _
;          e         .
i       _                                       i i
1 Sq        _
                                                                  ).
1 x       _
e    .
f s
s                                                                                                                   _
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ir                                                                   -
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e       .
         .2        -
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            +                                                                                                                _
i e       -
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l                  _
                       ,,,,,,,,i........,1,.......
o                                    ,,,,,,,,l.........I.........I,........i.........          ,,,,i.........-
O.        O 10 0 20 0. 30 0.40                       0.50 0 60 0-70 0-80 0 90                             1-0 TIME       SECONDS                                   x10
* Y

1 i           ;     -
I
: o.        I""""I"'"""I'""""I"'"""I'""""I"'"'"'I'""""I'""""I"'"""I'""'"i X         :
                         -    A FPRATP l
i
               .         E    v NUREG-0772                                                                             ;
<              N      .  -
!              vi o      :    a FPRAT                                                                              :
M          -
1                                                                                                                 .
j                                                                                                               _
o :   -
1                       -
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i              n        :                                                                                         :
O        ~
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i 4
o N
(f)      .  -
2 O :   -
1 f 5........I.........l........<l....... 11..... ........l.........I,........I.........!........9.
O. 0 10 0 20 0 30 0.40 0 50 0 60 0-70 0-80 0-90                                        1.0 1
1 TINE SECONDS                                    x10
* 4

O' .
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:         O
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I              08"O          09-0             OVO           0E 0             O 1
l l
I

MAAP SUBROUTINES INFLUENCING FISSION PRODUCT TRANSPORT IN THE PRIMARY SYSTEM e FPRATP or FPRATB - fission product release from the fuel.
e CIRC - forced and natural circulation through the primary system.
o P.EFINS - heat loss through the insulation -
reflective insulation on the IDCOR reference plants.
e FPTRANS - fission product deposition by vapor condensation, steam condensation and sedimentation
            - also calculates revaporization when applicable.
k

l
(                                                       h l
M GAS FLOW          _L
                              \"/         _L Me WATERIAL~%     --             --
TRAMsPORT            SEPARATORS        2 a   =                6 ORYERS NODE 2
                       =
DOWNCOMER g                 =$
NODE 3 A;,  "O            CORE         C NODEI i                   ;    i 3
4 MAAP-BWR primary system nodalization.

g               EP MO
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1 TASK 23.1 RESULTS ZION Marc A. Kenton Fauske & Associates, Inc.
.                            16WO70 West 83rd Street Burr Ridge, Illinois 60521 (312) 323-8750 NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Fission Product Behavior I
Rockville, Maryland August 28 - 29, 1984 l

l
{                                          D ZION SEQUENCES ANALYZED i
: 1. Containment Failure Sequences
: a. TMLB'/ seal LOCA
: b. TMLB'/no LOCA
: c. TMLB'/ containment purge open
: d. V sequence
: 2. Sequences Without Containment Failure
: a. TMLB' with recovery
: b. SLFC with and without recovery
: c. ALFC with and without recovery k

[                                                                                  )
* I                                           I COLD LEG            i       g HOT LEG                    A      I STEAM GENERATOR Q,
STEAM GENERATOR     PRESSURIZER- %
g SHELL                                                           i
                      ,         SHELL                             i s   l
                            /                                \    l    /
l                                           l
                    / l
                          ~
HOT                             I 8           LEG                          _3    _
ea"     /f                                                    ^%
t     (j                                         ()
( e=                          c      9 3"UNBROMEN" LOOPS                                      1"BROMEN" LOOP (NOOALIZATION SAME AS UNBROKEN LOOP)
MAAP-WESTINGHOUSE PWR PRIMARY SYSTEM NODAllZATION k

                                                                                                                                          . s.
                                                                                                  'i
                                                                                                 /
                                         <                          _                                        UPPER COMPARTMENT
                                         /                                                      /
STEAM GENERATORS l IN BROKEN AND                       .
O' UNBROKEN LEGS                                                           q^               "
h''                                                                  PRESSURIZER
                                        -          r     m              -
W           r
                                        , :a ca r._      u              /_ 6 g g'v
                                                           <                                               LOWER ANNULAR p' [
(      -
                                                                        /.                    '
                                              -                                                            COMPARTMENT COMPARTMENT                                                                   f QUENCH TANK PRIMARY SYSTEM                              c                            @
                                                                             ,v/mdn V'        '.W CAVITY                          \
hi,,.
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i I
UPPER COMPARTMENT a..utAR               . * "                            ',."
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t so...
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                       .                                      ..                                                  i,e. 2 .....-....,
(            PWR Large, Dry Containment Nodalization

ZION TMLB'/ Seal LOCA k
l l
l l

a ...
ACCIDENT 

==SUMMARY==

Time (hrs.)              Event 0            Loss of all AC, DC, AFW
   .75          Seal LOCAs 1.8          S/Gs dry 2.2          Core uncovered l
3.9          RV fails 32           Containment fails
                                              )
i

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TMLB POST-VESSEL FAILURE FLOW PATTERN Y "
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i l
l ZION TMLB' Csl Distribution (Fractions) at 30 hrs.
I Settled       Airborne         l Primary system          .98            .009
                                              -5 Pressurizer            .001          3 x 10 Containment           .003           8 x 10 -4 Environment               0              O

*%   y f
FISSION PRODUCT UNCERTAINTY ANALYSES M A AP 2.0

ZION TMLB' Inspection of results indicates potential sensitivity to:
: 1. Mass of CsOH and Csl vapor (assumed vapor pressure, number of groups, temperature).
: 2. Sedimentation rates (correlation parameters).
: 3. Containment hole size.
w

i 5                 5                    5 o        0                 0                    0 M         1                1                    1 u
x                  x                   x R         1                 1                   1 0                 5                    5 a       -                 -                   -
0 B         0                 0                    1 1                 1 r
S           x                 x                    x 1                 1                   1 s
r u         5                 5                    5 o            -                 -                    -
H    b          0                  0                   0 S          1                  1                   1 0        ,
5     e          x                  x                    x t    T
    'B  a               3                 5                    2 L   s M   n         4                                      4 T i o            -                                      -
t                                -
N   c  H           0 1                 0                   0 1
O   a  O                             1 I
Z r   s           x                                      x F    C                              x e             6                 1                    6 s
a l
e                                                4 e                                                   -
R   l A                  A 0
1 s         /                 /
C           N                 N                      x 5
                          ;                F d       p A
e       .
N p    v                                           a A                    t     i n          m u   H           J                     o    d o         l O             o                    n         n i
t          l s         t                    l            a     .
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                            ,    a                                 ,
s         H i
e H            H             e   H s               d                                              p O A          O       n           mO                 O s           m   s s     a           a     s                          u C     S            S      C             C         l      C
[
1                 2                    3 ll    1l             llll!lll        l

ZION TMLB' Release Fractions at 50 Hours Assumption        Csl     CsOH         Te, Sb       Sr,Ba        Ru,Mo
: 4. CsOH, Csl not                .027                 1 x 10
                                                                    -3           -5            -0
                                                  .037                  < 1 x 10      < 1 x 10 lumped; Sandia v.p., .5 f t contmt. failure
                                                                    -5            -5
: 5. Same as 3 but                .001      .001       7 x 10      < 1 x 10    < 1 x 10 -5 lowered sedimen-tation rate a

ZION TMLB'/ CONTAINMENT PURGE OPEN 1
ar-     t i
l l
l                   .

                                                                                                                                            -- m m
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9                       5.;           y                         c                     c:                                   t
                    , 01 x                   'tSd SS:Vid             I N:MN I V.I.NO 3

o I
ZION TMLB'/ CONTAINMENT PURGE OPEN Csl Distribution (Fractions) at 9 Hours Settled       Airborne Primary system          .96           .005
                                 -5             -5 Pressurizer           1 x 10       c 1 x 10 Containment              .02          .001 Environment             N/A            .01

ZION TMLB'/ CONTAINMENT PURGE OPEN Release fractions at 8 hrs.
Nobles                          .89 Cs, I                           .01 Te                                         -4 3 x 10 Sr,Ba                         6 x 10 -4            l Ru,Mo                                      -5 6 x 10 l
l l
1 f

e Ce p ZION V SEQUENCE i
1

                                                                         ,    o.
ACCIDENT 

==SUMMARY==

Time (hrs.)                     Event O                 .1 f t equiv. hot leg break, scram, steam dumps open 5.8               RWST depleted 21.6              Core uncovered 26.3              RV failed I
b risi- si a                                                  . _ _ _ _   __    __

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I,,,,,,,,,1,,,,,,,,,1,,,,,,,,,I,.,,,,,,,i,
                                                                      ,,,,,,,_ o                             ,
S         &          C         E                     I                                    O
              , OIx      13 73A37 831VM 13SS3A P

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0Ix9>               9018 XDV O.L 3SV:1~l:DH ti.i

.s        ,
l 1
ZION V SEQUENCE Cs,1 Distr. (kg) at 30 hrs.
Settled          Airborne
                                                                          -6 Primary system                   42            3 x 10
                                                                          -5 Pressurizer                    .77            3 x 10
                                                                          -3 Containment                  .22            1 x 10 Released to aux.                               126

if I.LLIN V SE.UUENCE                                                                                !
r,     ,
aH00E 1                 ,3 MODE 2                   --NODE 3                                                                      4 A          5 I                                                                                                                                     .
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1
                               -           *   *   -  - ;!      : i-   ,?       *,.-                     t-    ,t ; r   i  l   *,: ,t t t l l
0              5           10             15             20                      25                         30            35       !
i t
i IMi_         1-1R                                                                  4 Aerosol deposition in the lower compartments of the auxiliary building.                                           [
t

                   .~ ,
ZION V SEQUENCE Release fractions at 31 hrs.
Nobles                          ~ .9
                                                             -4 Cs, i                         1 x 10 l
                                                             -5 Te                            7 x 10 Sr,Ba                                -5 6 x 10 Ru,Mo                                 -4 1.5 x 10 l
l l
l

D #e e
ZION ALFC

., .                                                      l
{                                                    D i
ZION ALFC Accident Summary Time (hrs)                 Event O          Double-ended cold leg break, reactor scram
         .5         RWST low-level, injection off 1.3        RWST dry, sprays off 2.2        RV f ailure

o   o.
           .....                           ii                                                   O
                 .. ii.       4 .i.iiii-       iiiiiii...i.....
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S:.              v-         C            E              I                     'O
     ,     OIx        L-!    13A31 83.LVM         13SS3A

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m i          _
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1              _
2,,,,,,,,,l.,,,,,...'                                    I , , , , , , , , , l , , , , , , , , ,:
d 9                    5             &                   C                      E                   I 1
i                  i OIx                    ISd SS3Bd 1 WINO 3 I
i l

ZION ALFC Csl Distribution (Fractions) at 8 Hours Airborne Deposited
                                                          -5 Primary System                .84           < 1 x 10
                                                          -5 Pressurizer                  .004            < 1 x 10
                                                          -4 Containment                   .16              4 x 10
                                      ~
i l
l Q                                                          ]
: a.     -                  a 4
e% g T/H UNCERTAINTY ANALYSES      -
M AAP 1.2 f

                                                               , o.
l T/H UNCERTAINTY ANALYSIS:                            I Conclusions in the context of the MAAP models, essentially no important sensitivities of bottom-line results to input parameters with a few exceptions:
Category 1:   Change in input value altered sequence definition, (e.g. sprays come on due to parameter change).
Category 2:   Parameter change reveals a fundamental weakness in a model by giving erroneous results, (e.g. core slump model).

i r
s Table 5.1 Fisure of Herit Summaru for MAAP Uncertaints/Sensitivits Analysis PWR VERSION 1.2    ZION      SEQUENCE THLB' (W)
MODEL                                             FIGURE             OF      MFRIT CASE    PARAMETER      VALUE          1       2        3      4          5       6             7               8       9   10   11   12     13 (hr)    (hr)    (lba)    (F)             (Psi /hr) (hr)                (hr)     (lba) (hr) (hr) (hr) (lba) 2        None             N/A       22.31   18.60 202260. 469.0      0.075    3.16         N/A          2.03      146.6 2.30 3.70 6.20 381.2 3        FRCOEF           0.001     21.64   17.93 195380. 465.0      0.075    4.14         N/A          2.03      146.6 2.30 3.70 6.20 438.0 5        FCRSLU             0.2     24.96   21.72 125180. 428.0      0.068    1.72         N/A          2.03      133.7 2.30 3.33 9.07 733.8 7        FCRSLU             0.8     21.75   17.15 116840. 481.0      0.075    0.80         N/A          2.03      146.4 2.30 4.59 6.24 448.4 9        TTCSP         0.278        21.26   17.31 176610. 478.0      0.075    1.88         N/A          2.03      146.6 2.30 3.?5 6.20 365.5 11       FHT                0.5     22.33   18.59 199870. 469.0      0.058    3.20         N/A          2.03      113.8 2.33 3.74 6.26 345.0 12       TZOOFF           3500      22.51   18.81 205010. 472.0      0.070    1.52         N/A          2.03      137.6 2.30 3.71 6.38 337.2 26       T200FF           4040      22.12   18.42 200910. 468.0      0.083    3.52         N/A          2.03      162.0 2.30 3.71 5.92 363.0 13       XCNREF           3.28      19.81   16.10 200650. 469.0      0.075    2.16         N/A          2.03      146.6 2.30 3.70 6.20 331.6 14       FCHF               0.3     22.40   18.69 202200. 468.0      0.075    3.56         N/A          2.03      146.8 2.30 3.70 6.40 347.3 15       FCHF             0.05      22.15   18.44 198570. 466.0      0.075    2.72         N/A          2.03      146.5 2.30 3.70 6.20 434.1 Default Model Parameters Chansed:                                 Value                Fisure of Herit flomenclature:
FRCOEF  Friction coef ficient for corium in VFAIL.. . . . . . . 0.005               1 Time of containment failure FCRSLU  Fraction of total core mass which must melt                                 2 Time between reactoi vessel failure and to reach support Plate............................. r>.5                        containment feilure TTCSP   Time to fail support Plate after Corius Pile                                3 Integrated Wall condenSalibn (heasure Of has reached it (br)............................. 0.0333                         diffusiophoresis) between re.scloi vessel FHT     Fraction of maximusi heat transfer Permitted                                     failure and Containment failure between clad and sas stream........................ 1.0                     4 Peak containment outer uall surface TZOOFF  Zircalov oxidation cut-off and channel blockins                                 temperature temperature (F)................................... 3680                     5 Fraction of cl,d reacted in-vessel l XCNREF  Corium reference thermal boundary laver                                     6 Rate of change of containment pressure Just t         thickness (ft)................................... 0.320                         prior to the time of contairimer.t f ailure FCHF    Flat Plate CHF critical velocitu coefficient..... 0.14                      7 Time of ice depletion (if arPlit:.ble) 8 Time of cure uncoverv 9 Hvdrosen senerated at time of vessel failure 10 Time at whi. b cl.:4 temeer: Lure reaches ''000 F 11 Time of vessel failure 12 Time of cos e .hel t _..wr letion 13 Hudiosen mass , l I me of cunt e.inment failure

Table 5.2 Fisure of Merit Summarv for MAAP Uncertaintv/Sensitivits Analvsis FWR VERSION 1.2    ZION                                          SEQUENCE TMLB' (FAI)
MODEL                                             FIGURE                                                                  OF                        MERIT CASE          PARAMETER    VALUE           1       2       3          4                                                            5         6                           7         8                   9                          10     11   12     13 (hr)      (hr)    (Ibe)       (F)                                                                 (Psi /hr) (hr)                          (hr)            (Iba)                           (hr)   (br) (hr) (Ibe)
ZTMLB-1          None          N/A     21.95     18.24 198360.        0.0                                0.069                              1.59                        N/A      2.03          135.0                             2.30   3.71 6.32 363.0 ZTMLB-6          FCRSLU      0.2       25.01     21.76 127200.        0.0                                 0.064                             1.78                        N/A      2.03          125.0                             2.30   3.25 9.10 727.0 7TMLB-2          TTENTR   2.778E-3     23.34     19.36 157740.        0.0                                 0.069                             1.63                        N/A      2.03          135.0                             2.30   3.71 6.32 607.0 ZTMLB-3          FENTR         99         ****    **** 102550.        0.0                                 0.069                             2.29                        N/A      2.03           135.0                            2.30   3.71 6.32 777.0 21MLB-5          NVP           5       22.46     18.75 203340.        0.0                                0.069                              1.69                        N/A      2.03          135.0                             2.30   3.71 6.32 355.0 ZTMLB-4          SCALH       0.5       20.79     17.09 216680.        0.0                                 0.073                             4.23                        N/A      2.01          143.0                             2.29   3.70 6.30 410.0 Default Model Parameters Chansedt                                              Value                                                                            Figure of Merit Nomenclature 1 FCRSLU        Fraction of total core mass which must melt                                                                                             1 Time of containment failure to reach support Plate............................ 0.5                                                                                2 Time between reactor vessel failure and-TTENTR        Entrainment effective enPtVins time (hr)....... 1.39E-4                                                                                                 containment failure FENTR         Multiplier for Kutateladze criterion for cavits                                                                                        3 Integrated wall condensation (measure of blowout (GT 1.0 = difficultiLT 1.0 = easier).....                             0.33                                                                      diffusiophoresis) between reactor vessel NVP           Number of Penetrations railed in lower head. . . . . . . . . 1                                                                                          railure and containment failure SCALH         Scalins factor for heat transfer coefficients to                                                                                       4 Feat containment outer wc11 surface Passive heat sinks................................                                       1.0                                                            temperature 5 Fraction of clad reacted in-vessel 6 Rate of change of containment Pressure Just Prior to the time of Containment failure 7              Time of ice depletion (if applicable)                              '
8              Time of core uncoverv 9              Hudrosen senerated at time of vessel failure 10                   Time at which clad temperature reaches 2000 F 11                    Time of vessel failure 12                    Time of core melt completion 13 Hudrosen mass at time of containment failure 44 s      This case was not run to containment failure. Bw extrapolating the containment Pressurer i
it is estimated that the failure criterion would have been exceeded at 44.1 hours.
l 4

                                                                                                                                              ?
Table 5.3 Fisure of Merit Summarv for MAAP Uncertaints/Sensitivits Anaissis PWR VERSION 1.2  ZION     SEQUENCE S2HF (W)
MODEL                                             FIGURE- 0F             MERIT CASE      PARAMETER        VALUE         1       2        3     4        5      6         7            8            9  10   11     12    13 (br)    (br)     (lba)  (F)          (Psi /hr) (hr)          (hr)        (Ibm) (hr) (hr)   (hr) (Ibe) 19         None                N/A     19.98   11.98 234040. 347.0 0.220       6.64     N/A        3.78         431.0 4.22 8.01 13.45 640.0 20         FRCOEF            0.001     19.99   11.99 233820. 347.0 0.220       6.64     N/A         3.78        431.0 4.22 8.01 13.45 639.0 21         FCRSLU              0.2     21.05   15.66 288580. 347.0 0.153       6.60      N/A        3.78        300.0 4.22 5.38 22.66 327.0 22         FCRSLU              0.8     19.14    9.60 225540. 346.0 0.231       6.92     N/A         3.78        451.0 4.21 9.54 17.82 831.0
* 23         IICSP             0.278     19.79   11.54 232790. 347.0 0.221       6.64     N/A         3.78        433.0 4.22 8.25 14.03 663.0 24         FCHF                0.3     20.44   12.46 234110. 347.0 0.227       6.60      N/A        3.78        443.0 4.22 7.99 13.70 583.0 25         FCHF              0.05      19.73   12.23 236480. 347.0 0.185       6.68      N/A        3.78        3(1.0 4.22 7.50 10.74 643.0 Default Model Parameters Chansed:                                 Value             Figure of Merit Nomenclature:
FRCOEF    Friction coefficient for corium in VFAIL........         0.005           1 Time of containment failure FCRSLU    Fraction of total core mass which must melt                              2 Time between reactor vessel failure and to reach support Plate............................         0.5              containment failure T1 CSP    Time to fail sueeort elate af ter corium Pile                            3 Integrated Wall condensation (measure of has reached it (hr)............................. 0.0333                     diffusioehoresis) betueen reactor vessel FCHF      Flat Plate CHF critical velocite coefficient..... 0.14                      failure and containment failure 4 Peak containment outer wall surface temeerature 5 Frcetion of clad reacted in vessel 6 Rate of chcose of containment Pressure Just ertor to the time of containment failure 7 Time of ice deeletion (if aePlicable)
     * (3ategOry 2                                                                 8 Time of core uncoveru 9 Hsdrogen senerated at time of vessel failure 10 Time at which clad temperature reaches 2000 F 11 Time of vessel failure 12 Time of core melt completion 13 Hsdrosen mass at time of containment failure

Table 5.4 Fisure of Merit Summars for MAAP Uncertaints/Sensitivits Analvsis PWR VERSION 1.2   ZION         SEQUENCE S2HF (FAI)
MODEL                                                FIGURE              OF     MERIT CASE      PARAMETER          VALUE           1       2       3        4            5     6          7         8       9   10    11     12     13 (hr)    (hr)    (lba)     (F)             (Psi /hr) (hr)       (hr)  (lba)   (hr)  (hr)    (hr) (lba)
ZS2HF-1     None               N/A         20.01   12.01 231980.      0.0      0.229    6.60      N/A     3.78  447.0   4.22  8.00  13.25  651.0 ZS2HF-3      TTENTR         2.778E-3       20.02   12.02 235140.      0.0      0.229    6.59      N/A     3.78  447.0   4.22  8.00  13.25  643.0 ZS2HF-4     FENTR               99         19.59   11.59 344320.      0.0      0.229    6.64      N/A     3.78  447.0   4.22  8.00  13.25  628.0 ZS2HF-5      NVP                 5         19.99   11.99 221790.      0.0      0.229    6.55      N/A     3.78  447.0   4.22  8.00  14.94  655.0 ZS2HF-2     SCALH             0.25         18.12   11.04 198860.      0.0      0.209    6.84      N/A     3.34  409.0   3.75  7.07 13.25 599.0*
Default Model Parameters Chansed!                                        Value                Fisure of Merit Nomenclature 1
          - _ _ _ _ _ _ _ _ -         .-----                               -----                      --  .__=_--    -_
ITENTR    Entrainment effective enPthins time (hr)....... 1.39E-4                           1 Time of containment failure FENTR     Multiplier for Kutateladze criterion for cavits                                   2 Time between reactor vessel failure and blowout (GT 1.0 = difficultiLT 1.0 = easier)..... 0.33                                containment failure NVP       Number of Penetrations failed in lower head......... 1                            3 Intesfaled Wall Condensation (mea 5ure of SCALH     Scalins factor for heat transfer coefficients to                                      diffusiophoresis) between reactor vessel Passive heat        sinks................................            1.0              failure and containment failure 4 Peak containment outer wall surface temperature 5 Fraction of clad reacted in-vessel 6 Rate of change of containment Pressure Just Prior to the time of containment failure 7 Time of ice depletion (if applicable) 8 Time of core uncovers 9 H9drosen senerated at time of vessel failure 10 Time at which clad temperature reaches 2000   F 11 Time of vessel failure 12 Time of core melt completion 13 Hydrogen mass at time of containment failure
* Category 1

  .s o RADIONUCLIDE RELEASE PWR -- LARGE, DRY CONTAINMENT (ZION)
BMI-2104 RESULTS by James A. Gieseke Presented at the NRC/IDCOR MEETING August 28-29, 1984
{
l l
l OBallelle Columbus Laboratones

                                       .. .-     ~
o .
S D SEQUENCE 2
e   SMALL PIPE BREAK, FAILURE OF EMERGENCY CORE COOLING SYSTEM e   INTERMEDIATE FLOW AND PRESSURE OCCUR IN THE REACTOR COOLANT SYSTEM DURING THE MELTDOWN PERIOD e   CONTAINMENT SAFETY FEATURES (CONTAINMENT SPRAY AND AIR COOLERS) ARE OPERABLE 4
OBallelle Columbus Laboratones

ACCIDEbE EVENT TIMES Event                                     Time, minutes Zion S2D Containment Spray Injection On                21.8 Initial Core Uncovery                        33.9 Cont. Spray Recirculation On                  71.3 Final Core Uncovery                         112.5 Start Melt                                   150.6 Start Slump                                  163.8 Core Collapse                                167.5 Vessel Head Dry                              179.9 Head Fail                                    187.7 Concrete Attack
* 187.7 End Calculation                              788.2
* Assuming debris uncoolable.
OBallelle    Columbus laboratories

DISTRIBUTION SPECIES AFTER ACCIDENT, S D-c 2
Fraction of Core Inventory Species                     RCS          Containment           Environment Csl                   .
0.34                 0.66          2.5 x 10 -8, Cs0H                        0.42                 0.58          2.3 x 10-8 Te                          0.93            1.7 x 10-2         3.6 x 10-8
*The release of more volatile iodine chemical forms has not been included in this release fraction'.
4 OBallelle Columbus Laboratories
          . .                            - . = _         .---                . .-          - -

  . .. +
l l
TMLB' SEQUENCE e TRANSIENT, LOSS OF PRIMARY SYSTEM HEAT REMOVAL e SYSTEM PRESSURE REMAINS HIGH DURING CORE HEATUP (I.E.,
CORE UNC0VERY IS DELAYED FOR A FEW HOURS) e HIGH PRESSURE, ESSENTI ALLY STAGNANT FLOW, A MUCd LONGER RELEASE PATH, AND INTERACTIONS BETWEEN CORE MATERIALS AND WATER IN THE REACTOR CAVITY e CONTAINMENT SAFETY FEATURES (CONTAINMENT SPRAYS, CONTAIN-MENT COOLING SYSTEMS) ARE NOT AVAILABLE e RAPID PRESSURE RISES IN THE CONTAINMENT BUILDING FOLLOWING VESSEL FAILURE COULD THREATEN CONTAINMENT INTEGRITY AND P0TENTIALLY RESULT IN A LARGE RELEASE OF FISSION PRODUCTS OBallelle Columbus Laboratones 4

s    o.
ACCIDENT EVENT TIMES Event                                Time, minutes Zion TMLB' Steam Generator Dry                    82.5 Core Uncover                          109.8 Start Melt                            130.5 Start Slump                           158.5 Core Collapse                         159.8 Vessel Head Dry                       169.2 Head Fail                             169.5 Cavity Dry                            316.4                              ,
Concrete Attack                       389.1 End Calculation                      1001.8 OBallelle Columbus Laboratones

DISTRIBUTION OF SPECIES AFTER ACCIDENT, TMLB'-c Fraction of Core Inventory Species                RCS          Containment           Environment Cs1                    0.98          2.5 x 10-2           1.9 x 10-6 Cs0H                   0.98          2.5 x 10-2           1.9 x 10-6 Te                     0.28             0.64              7.8 x 10-5
    *The release of volatile iodides has not been included in this release fraction.
OBaneHe Columbus Laboratones

                                                                                                             .    +.
FRACTION OF CORE INVENTORY RELEASED TO THE ATMOSPHERE FOR GROUPS OF REACTOR SAFETY STUDY, TMLB'-c Time          I              Cs                      Te                     Sr        Ru                  La (br)       Group 2*       Group 3                 Group 4        Group 5          Group 6            Group 7 2           0              0                        0                    0          0                   0 4     8.5 x 10~7       8.7 x 10 -7             1.3 x 10 -6    8.6 x 10 ~7    2.2 x 10-7           2.5 x 10 -9 7     1.6 x 10-6       1.6 x 10 -6             2.7 x 10~0     1.9 x 10-6     4.4 x 10 -7          1.0 x 10 -8 10     1.9 x 10 -6      1.9 x 10-6              2.7 x 10-5     1.8 x 10 -5    5.2 x 10-7           9.8 x 10 -7 15     2.0 x 10-6       2.1 x 10-6              6.6 x 10-5     2.8 x 10-5     5.7 x 10-7           1.6 x 10-6 20     2.0 x 10-6       2.1 x 10-6              7.8 x 10-5     3.0 x 10-5     5.8 x 10~7           1.7 x 10-6 50     2.1 x 10-6       2.1 x 10 -6             8.4 x 10 -5    3.1 x 10-5     5.9 x 10 -7         1.8 x 10 -6
  *The release of volatile iodides is not included in these release fractions. Their inclusion would be expected to increase the total release fraction for iodine up to perhaps 10-3 i
OBallelle Columbus Laboratorms

TOTAL LEAK AREA ESTIMATED AS A FUNCTION OF CONTAINMENT PRESSURE                           i Containment           Low               Medium         High Pressure           Leakgrea           LeakAgea      Leakgrea (Psig)              (in. )              (in. )       (in.)
Normal Operating     0.1               -0.5            1.0
;        23                   0.1                0.62           1.48 47                   0.1                0.62           1.84
,        105                  0.1                2.13          10.96
        ~ 134                 0.1                5.33          23.72 OBallelle Columbus Laboratories

DISTRIBUTION 0F FISSION PRODUCTS FOR
                    .VARIOUS FAILURES MODES, TMLB' SEQUENCE I E     Il RCS          Containment    Environment Design Leak CsI       0.98         2.5 x 10 -2    1.9 x 10 -6 Cs0H      0.98         2.5 x 10 -2    1.9 x 10-6 Te       0.29             0.63       7.8 x 10-5 Medium Leak CsI-      0.98         2.5 x 10 -2    7.1 x 10 -5.
Cs0H      0.98         2.5 x 10 -2    7.2 x 10 -5 Te       0.29             0.63       5.5 x 10 -3 High Leak Cs1       0.98         2.5 x 10-2     9.0 x 10-5
                                                              -5 Cs0H      0.98         2.5 x(10-2     9.1 x 10 Te       0.29             0.62       1.6 x 10-2 Isolation Failure Csl       0.98         1.8 x 10-2     7.0 x 10-3 Cs0H      0.98         1.8 x 10-2     7.1 x 10 -3 Te       0.29             0.42       2.2 x 10 -I OBallelle Columbus Laboratones

    .~   ,
TASK 23.1 RESULTS PEACH BOTTOM Jeff R. Gabor Fauske & Associates, Inc.
16WO70 West 83rd Street q                                               Burr Ridge, Illinois                     60521 (312) 323-8750 NRC/IDCOR Meeting on integrated Analysis of Severe Accident Fission Product Behavior Rockville, Maryland August 28 - 29, 1984

s c.
PEACH BOTTOM - TC EVENT 

==SUMMARY==

Time                       Event 0     Transient (MSIV closure) 1.5 min   HPCI, RCIC on 27 min    HPCI assumed lost (SP at 200 F) 38 min    ADS on 40 min    LPCI, LPCS en (reduced flow) 54 min    RCIC lost 1.2 hr    ADS valves close 1.3 hr    Top of core uncovered 1.3 hr    Wetwell vent open 1.3 hr    LPCI, LPCS assumed lost 1.4 hr    ADS valves reopen 3.2 hr   Start of core melting 3.8 hr   Vessel failure 6.0 hr .
CRD flow ceases 12 hr     Drywell failure 12 hr     Begin release to environment

O% g)
PEACH BOTTOM - Sj E EVENT 

==SUMMARY==

Time                         Event 2
0      Break in steam line (0.1 ft )
6.8 sec   Reactor scrammed 84 sec    MSIVs closed, feedwater tripped 10 min    Suppression pool cooling on 1.0.hr    Automatic depressurization on (ADS) 1.1 hr    Top of core uncovered 2.6 hr    Start of core melt 3.6 hr    Vessel failure 15 hr    CRP flow ceases 23 hr    Containment failure (overtemperature) 23 hr    Begin release to environment

PEACH BOTTOM - TQVW EVENT 

==SUMMARY==

Time                    Event 0    Loss of off-site and on-site AC power 4 sec' Reactor scrammed 4 min  High pressure injection on (HPCI, RCIC) 6 hr   HPCI, RCIC off (loss of DC power) 8.5.hr Top of core uncovered 11.5 hr Start of Core melt 12.4 hr Vessel failure 18 hr  Containment failure (overtemperature) 18 hr  Begin release to environment

o 6
H GAS FLOW                _L
                                 \"/        _l.
AM) MATERIAL ~ %        --            --
TRANSPORT           F      SEPAR TORS     1 ORYERS DOWNCOMER
            - A;,      =C           CORE        'O
()
i   I                   J  A x
s o
BWR Primary System Circulation

                                                                                   ,             s.
i METAL DECK METAL "
NNNNN////7 SIDWG                                                         '
REFUELING FLOOR                                                             !
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REACTOR   I    e                                                   5 BUILDING  N N '\')                                               I FAN    ;                 l v                 ,
i
                                                          /       l' REACTOR ROOu    :                 :         L.._...%
y                 s                                                                     '
REACTOR             Q'              ['.               (s ''    1   s
                                        ~
s -..J L-BUILDING  h                                           N            l l            ,
l' VENTILATION h R    s{a F"] I" s -,
s
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                                                   -s MDRYWELL EQUIPMENT b ' '                      s              ;           3; tg D
h   ( ".j                      l g'
l s%n     .    ,
a m           sl       ;
: s. n, l
b q s            s YK suxuuuusuus L'       !UU         hts N  V Mismwi<i N                             MYM M m m t %s S
PEDESTAL REGION               SUPPRESSION POOL BWR Mark l Reactor Building

O' % p STANDBY GAS TREATMENT SYSTEM Peach Bottom o    Fire dampers close at temperature = 165 F.
f e    Fans have automatic trip on low flow, o    High aerosol loading on filters.
e    Due to above restrictions and detailed sequence analyses; SGTS assumed not operational.
i 4
9 1

r-g .
I DISTRIBUTION OF Csl IN PLANT AND ENVIRONMENT (FRACTION OF CORE INVENTORY)
At Vessel Failure TC          SE j   TQVW RPV                                               .50          .78     .99 Drywell                                              0          .03      0 Suppression Pool                                  .50          .19      .01 Secondary Containment                          1.2 x 10 -4        0      0
                                                            -4 Environment                                     3.7 x 10          0      0 At Containment Failure TC           SEj   TQVW RPV                                                 1.0          .48    .67 Drywell                                              0          .03     .27 Suppression Pool                                     0          .49     .06 Secondary Containment                                0            0      0 Environment                                          0            0       0 Ultimate Distribution TC          SEj   TQVW (57 hrs.)    (60 hrs.)(60 hrs.)
RPV                                 4 0            0     .07 Drywell                                  e            0           0       0 Suppression Pool                                   .66        .49       .06 Secondary Containment                               .31        .50      .80 Environment                                         .03       .009      .07

o e. j

==SUMMARY==
OF FISSION PRODUCT RELEASE FRACTIONS IA)
Sequence                     WASH-1400 F.P. Group                  TC(b)     SlE     TQVW      BWR2(c)    BWR3(d)
Cesium, Iodine                 .03       .01     .07    0.50, 0.90    0.10 Tellurium                      .09      .016     .08       0.30       0.30
                                               -4       -5      -5 Strontium                    2 x 10    2 x 10 6 x 10       0.10       0.01 Ruthenium                    2 x ;0~4 7 x 10-52 x 10'#     0.03       0.02 (a) Fraction of core inventory released to the environment.
2 (b) Wetwell vent area = 1.98 f t .
(c) Containment failure prior to vessel failure; can be compared with (TW, TC).
(d) Failure to scram or remove decay heat; can be compared with (TC, SlE, TQVW).

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i PEACH BOTTOM TW SEQUENCE                        I l                                                                         4 i
e WASH-1400 assumed core injection systems f ailed at containment f ailure.
o Recent IDCOR Task 23.1 report analyzed TW with a 4
small containment failure area with same assumption on pump f ailure.
i e Recent GE qualification tests indicate that pumps will operate in a steam environment.
i       e IDCOR case would indicate that with a small
;         containment failure area the pumps would not fail due
 !        to steam environment and therefore not lead to core degradation.
e IDCOR has performed analyses on TW assuming pump f ailure associated with containment f ailure areas sufficiently large to potentially challenge core injection via dynamic response of the drywell or suppression pool.
k

PEACH BOTTOM - TW EVENT 

==SUMMARY==

Containment Break Area = 1 ft 2 Time                            Event 0    Transient (MSIV closure) 4 sec  Reactor scrammed 4.5 min HPCI, RCIC on 8.0 hr  High SP temperature failure assumed for HPCI (200 F) 10 hr  RCIC lost 14 hr  CRD flow ceases 15 hr  ADS on, LPCI and LPCS injecting 25 hr  ADS valves close 32 hr  Containment failure (overpressurization); LPCI and LPCS lost 32 hr  Top of core uncovered 32 hr  ADS valves open 36 hr  Start of core melt 37 hr  Vessel failure

0%                4
                                              , PEACH BOTTOM - TW Containment Failure Size Uncertainty / Sensitivity Analysis Area of Containment Failure 2
1 ft     10 f t Cs & I release fraction                          .04      .06 at 100 hrs.
Os & I fraction in                               .17      .01 suppression pool at 100 hrs.
l l
l l
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RESULTS FROM BWR-MAAP THERMAL-HYDRAULIC UNCERTAINTY / SENSITIVITY ANALYSIS From a large set of single parameter variations the l
i      following was demonstrated.
l Peach Bottom - S3E e   There is sensitivity to the core blockage phenomena on core recovery with limited CRD flow.
Grand Gulf - T 1QUV w/o ADS e   if corium is dispersed into drywell then there is less concrete ablation in pedestal and a longer calculated time to containment failure.
e The absence of core blockage allows increased Zr oxidation and increased in-vessel hydrogen
,  k

Table 2.1 MAAP-BWR Uncertaints/Sensitivitw Analvsis Parameter Set                          s Parameter                                                                         Default      Mininua      Minimun Nuaber Nanc              Parameter Description                          Units           Value        Value       Value
   ....-- ----                       - ==== _ .__                          -----           .....        .....       .....
1     FRCOEF Friction coefficient for corium in VFAIL                           5.000E-03    1.000E-03    1.000E-02 2     FMAXCP Fraction of total core mass which must melt to fail the core Plate                                             0.200        0.100        0.400 3     HTBLAD Fuel channel to control ~ blade heat transfer coefficient                                          B/hr/ft**2/F      8.8        0.88          88.
4     HTFB   Film boilins heat transfer coefficient               B/hr/ft**2/F    52.8         17.6        70.4 l        5     FBLOCK Fuel channel blockage model switch 0=use blockase modeli1= turn blockase model off                    0.000Ef00    0 000Ef00     1.00 6     TZOOFF 0xidation cut-off temperature                        F               3680         3310        4040 7     FACPF  Fraction of area 6f core elate failed                              0 300        1 000E-02     1.00 8     CDBPD  Flame buovanew dras coefficient in the Pedestal                      5.00       0.500         10.0 9     CDBDW  Flame buovancs dras coefficient in the druwell                       5.00       0.500         10.0 10      CD3WW  Flame buovancs dras coefficient in the wetwell                       5.00       0.500         10.0 11      CDBCA  Flame buovancs dras coefficient in conPartment A                     5.00       0.500         10.0 12      CDBCB  Flame buovanes dras coefficient in compartment B                     5 00       0.500         10.0 13      XCNREF Corium reference thermal boundars laser ~ thickness  feet           0.328       0.328         3.28 14      HTCMCR Corium-crust heat transfer coefficient in DECOMP     B/hr/ft**2/F       176         88.       880.
15      XCMX   Minimum corium thickness on drwwell floor and pedestal floor (Mark II onlv)                        feet           0.164          1.64E-02  0.328 16      XDCHSP Particle si:e (diameter) for corium as it falls                                              0.328 into suppression Pool (MarkII on1v)                  feet           3.281E-02    3 281E-02 17      TCFLAM Critical flame tetPerdture                           F                1310         1160        1700 16      FCHTUR Churn-turbulent critical flow Parameter                               1.53         1.00       5.00 19      FDROP  Droplet Crit 1 Cal flow Parameter                                    3.70          3.00       5.00 20      FFLOOD Flooding flow earameter                                              3.00         2.00        4.00 21      FSPAR  Parameter for botton-sParsed steam void fraction                      1.00         1.00       4.00 22      ~VOL   Parameter for volume source void fraction model                      2.00          1.00       4.00 23      TTENTR Entrainment effective etPtwins time                  hour             1.39E-04     2.78E-05   2.78E-05 24      EW     Emissivits of water                                                 0.900        0.800         1.00 25      EWL    Emissivitw of wall                                                  0.850        0.700         1.00 26      ECM    Emissivits of corium                                               0.850         0.700         1.00 27      EG     Emissivits of sas                                                   0.600        0.500         1.00 28      EE0    Emissivity of eauiPaent                                             0.850        0.700         1.00 29      F0VER  Fraction of core sPrav flow allowed to DuPass core                  0.500        0 000Et00     1.00 30      NPF    Number of Penetration failed in lower head                            1.00         1.00        10.0 31      FCDCDW Downcomer Perimeter Per meter from Pedestal door (Mark II onlv)                                                       2.00          1.00        5.00 32      FCHF   Coefficient for CHF correlation in PLSTM                            0.140        0.120       0.300 33      FCDBRK Discharse coefficient for Pipe break                                0.750        0.100         1.00 34      FENTR  Mult1 Plier for Kutatelad:e criterion for cavits                                               100.
blowout (GT 1.0 = difficultiLT 1.0 = easier)                        0.330        0.200 35      SCALU  Scalins factor for all burning velocities                             1.00         1.00        100.
36      SCALH  Scaling factor for heat transfer coefficients to rassive heat sinks                                                    1.00       0.500         10.0 37      FUMIN  Minimum burn velocits                                 e/s             1.00       0.500         10.0 38      ACVENT Containment failure vent area                         ft**2             0.1          5.0      0.01 wmy                                             .g            . ; y             3;9                w            3 y.

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Table 4.1 Figure of Merit Summarv for MAAP Uncertaintv/Sensitivits Analvsis BWR VERSION 1.2             PEACH BOTTOM     SEQUENCE S1E (PECO)
MODEL                                                      -FIGURE          OF       MERIT                                                 ~
CASE    PARAMCTER    VALUE           1           2             3        4        5    '6           7      8         9   10     11    12    13 (hr)         (hr)          (Iba)     (F)         (Psi /hr) (br)       (br)     (lba) (hr)   (br)  (br) (lba)         i i
SIE6     None         N/A        30.71          7.04        33779.-1200.6 0.061       1.27        N/A 17.85      672.0 1.84 23.'7 6  42.31 672.0         l G1E1     FCHF        0.05        30.45          6.78        29273. 1200.6 0.061.      1.77        N/A 17.85      720.0 1.84 23.67 42.33 720.0 SIE2     FCHF         0.3        30.57          6.90        33294. 1200.6 0.061       1.18        N/A 17.85      644.0 1.84 23.67 42.30 644.0 SIE3     FCDBRK       0.6        21.89         17.74 107231. 1200.6 0.014             0.44        N/A   1.26    205.0  1.84  4.16 23.07 205.0 31E4     ACVENT       1.0        30.70          7.04        33779. 1200.6 0.061       1.52        N/A 17.85      672.0 1.84 23.67 42.30 672.0 Default Model Parameters Changed 1                                          Value            Fisure of Merit Nomenclature 1.
j FCHF    Coefficient for CHF co relation in' PLSTM.. ... .. . . 0.14                       1 Time of containment failure FCDBRK  Discharge coefficient for Pipe break............. 0.75                            2 Time between reactor vessel failure and ACVENT  Containment failure vent area (ft**2).............. 0.1                                containment failure 3 Integrated wall condensation (measure of diffusioPhoresis) DetWeen reactor Vessel failure and containment failure 4 Peak containment outer wall surface temperature 5 Fraction of clad reacted in-vessel 6 Rate of change of containment Pressure Just Prior to the time Of Containment failure 7 Time of ice deeletion (if aPelicablel 8 Time of core uncoverv 9 Hwdrogen senerated at time of vessel failure 10 Time at which clad temperature reaches 2000 F 11 Time of vessel failure .
12 Time of core melt completion 13 Hvdrogen mass at time of containment failure 1

Table 4.2 Fisure of Merit Sunnarv for MAAP Uncertaintw/Sensitivits Analvsis BWR VERSION 1.2                PEACH BOTTOM        SEQUENCE TGVW (PECO)
I h0 DEL                                                                 FIGURE          OF     MERIT PARAMETER       VALUE                          2             3             4        5     6                7   8     9     10    11     12     13 CASE                                      1 (br)          (hr)          (1ba)           (F)          (Psi /hr) (hr)         (br) (1ba)  (hr)   (hr)   (hr) (lba)
TGVu0     None            N/A         16.45           4.09        29290. 1200.6 0.063            '2.31      N/A       8.40 690.0  9.87 12.36 >20.00 960.0 l                                                                                                                              690.0 TQVul     FRCOEF         0.001        16.49           4.13        28570. 1200.6 0.061             2.86      N/A       8.40        9.87 12.36 >20.00 990.0 10Vu2     FHAXCP          0.4         16.63           3.30        22110. 1200.6 0.061             2.14      N/A       8.40 750.0  9.87 13.33 >20.00 870.0 TOVW3     FBLOCK             I        16,70           4.27        31070. 1200.6 0.069             2.31      N/A       8.40 780.0  9 87 12.43 >20.00 990.0 10VW4     TZOOFF        .4040         16.44           4.08        29500. 1200.6 0.066             2.14      N/A       8.40 750.0  9.87 12.36 >20.00 990.0 T0VuS     XCHREF         3.28         16.27           3.91        28840. 1200.6 0.063'            2.14      N/A       8.40 690.0  9.87 12.36 >20.00 960.0 l
TOVu6     HTChCR           88         16.59           4.23        29710. 1200.6 0.062             3.08      N/A       8.40 690.0  9.87 12.36 >20.00 900.0 TGVu7     HTCMCR          880         16.23           3.87        28020. 1200.6 ,0.062            2.86      N/A       8.40 720.0  9.87 12.36 >20.00 990.0 10Vul0    TTENTR         2.78E-3      16.51           4.15        29240. 1200.-6' O.062           2.31      N/A       8.40 690.0  9.87 12.36 >20.00 960.0 TGVUB     ECM             0.7         16.41           4.05        29170. 1200.6 0.060             2.14      N/A       8.40 690.0  9.87 12.36 >20.00 960.0 TGVu9     NPF              10         16.43           4.08        29070. 1200.6 0.063             2.14      N/A       8.40 720.0  9.87 12.36 >20.00 990.0 10Vu11    FENTR           100-        16.70           4.34        31390. 1200.6 0.063             3.08      N/A       8.40 720.0  9.87 12.36 >20.00 930.0 10Vu12    SCALH           0.5         16.47           4.06        26360. 1200.6 0.061             2.86      N/A       8.43 690.0  9.90 12.40 >20.00 930.0 T0Vu13    SCALH             10        20.88               j,
: 8. '! 7 ; 89250. 1200.6 0.064               2.78      N/A       8.57 720.0 10.06 12.61 >23.24 1280.0
                                                   ''                                     Default Model Parameters Chansed:                                                      Value            Figure of Merit Nomenclature 1
  . FRC0EF   Friction coefficient for corium in VFAIL........                              0.005          1 Time of containment failure FnAXCP   Fraction of total core mass which must melt                                                  2 Time between reactor vessel failure and to fail the core P1 ate............................                              0.2            containment failure FBLOCK-  Fuel channel blockase modal switch                                                           3 Integrated wall condensation (neasure of 0=use blockase modelit= turn blockase model                       off....... O                  diffusiophoresis) between reactor vessel TZOOFF   0xidation cut-off temperature (F)................. 3680                                          failure and containment failure XCNREF   Corium reference thermal boundarv laver                                                      4 Peak containment outer wall surface thickness (ft)..................................                              0.328             temperature HTCnCR   Corium-crust heat transfer coefficient                                                       5 Fraction of clad reacted in-vessel (Btu /hr-ft**2-F)................................... 176                                     6 Rate of chanse of containment Pressure Just TTENTR   Entrainment effective emetwins time (hr)....... 1.39E-4                                          Prior to the time of containment failure ECn      Laissivi ts of co rius. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.85         7 Time of ice depletion (if applicable)
HPF      Number of Penetrations failed in lower head......... 1                                       8 Time of core uncovers FENTR    Multiplier for Kutateladze criterion for cavits                                              9 Hvdrosen senerated at time of vessel failure blowout (GT 1.0 = difficultiLT 1.0 = easier)...... 0.33                                     10 Time at which clad temperature reaches 2000 F SCALH    Scalins factor for heat transfer coefficients to                                            11 . Time of vessel failure Passive heat sinks................................                               1.0        12 Time of core melt comeletion 13 Hwdrosen mass at time of containment failure

Table 4.3 Fisure of Merit Susaarv for MAAP Uncertaintv/Sensitivits Analvsis BWR VERSION 1.2             GRAND GULF          SEQUENCE T10VV.(TEC)
MODEL                                                                 FIGURE            OF        MERIT CASE          PARAMETER          VALUE               1        2             3             4          5      6             7            8       9   10    11     12     13 (hr)     (hr)          (Iba)           (F)            (Psi /hr) (hr)               (hr)   (1ba) (hr)  (br)   (hr) (Ibe)
T10UVO             None              N/A          44.71   40.97 172050. 196.7 0.047 1.04                                N/A          0.66   417.2 1.62  3.74   8.94 3231.0 T100V1          FRCOEF          0.001            46.10   42.37 ,172720. 197.5 0.047                      0.86          N/A          0.66   417.2 1.62  3.74   8.94 3190.0 TaCUV2           FMAXCP              0.4          62.37. 58.10 288160. 217.9 0.047                       0.80          N/A          0.66   417.2 1.62  4.27   8.94 2714 0 T10UV3          Fbt0CK              1.0          40.81   37.19 255770. 213.0 0.164                        1.02         N/A          0.66 1443.0. 1 62  3.62 10.68 3878.0 T100V4           TZOOFF             4040          44.78   41.05 171890. 197.0 0.051                       0.96          N/A          0.66   449.6 1.62  3.73   8.96 3258.0 T100V5           All CDs             0.5          44.63   40.89 171800. 196.6 0.047                        1.57         N/A          0.63   417.2 1.62  3.74   8.94 3229.0 T10UV6           XCNREF          0.328            44.64   40.90 171800. 196.6 0.047                       0.89          N/A          0.66   417.2 1.62  3.74   8.94 3231.0 T10UV7         HTCMCR               88          45.55   41.81- 176000. 197.6 0.047                      0.88          N/A          0.66   417.2 1.62  3.74   8.94 3234.0 110'JVU          HTCMCR              880          44.38   40.64 168840. 197.2 0.047                       0.96          N/A          0.66   417.2 1.62  3.74   8.94 3249.0 T10UV9          ECM                 0.7          44.71   40.97 172040. 196.6 0.047                       0.93          N/A          0.66   417.2 1.62  3.74   8.94 3232.0 116UV10        NPF                  10          44.46   40.72 171940. 196.1 0.047                       1.03          N/A- 0.66           417.1 1.62  3.74   8.94 3226.0 Default Model Parameters Changed
* Value               Figure of Merit Nomenclature:
FRC0EF         Friction coefficient for corium in VFAIL........                               0.005             1 Time of containment failure FMAXCP          Fraction of total core mass which must selt                                                      2 Time between reactor vessel failure and to f ail the co re Pl ate . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.2               containment failure FBLOCK          Fuel channel blockage model switch                                                               3 Integrated wall condensation (neasure of 0=use blockase modeli1= turn blockage model off....... O                                             diffusiophoresis) between reactor vessel T200FF         0xidation cut-off temperature (F)................. 3680                                              failure and containment failure CDbFD          Flame buovancs dras coefficient in the pedestal... 5.0                                           4 Peak containment outer wall surface CD0DU           Flame buovanew dras coefficient in the drwwell.... 5.0                                               temperature CubJu         Flame buovancs dras coefficient in the wetwell.... 5.0                                           5 Fraction of clad reacted in-vessel UtilfA          Flasie buovanes dras coefficient in compartment A..                                5.0           6 Rate of chanse of containment Pressure just CbLCB          Flame buovancs dras coefficient in compartment B..                                 5.0               Prior to the time of containment failure XCNREF          Corium reference thermal boundars laver                                                          7 Time of ice deeletion (if applicable) thickness (ft).................................. 0.328                                           8 Tinie of core uncovers HVCMCR          Corium-crust heat transfer coefficient                                                           9 Hwdrogen senerated at time of vessel failure (8tu/hr-ft**2-F)................................... 176                                        10 Time at which clad temperature reaches 2000 F ECM             Emissivits of carium............................. 0.85                                          11 Time of vessel failure NFF             Number of Penetrations failed in lower head.......... I                                         12 Time of core melt completion 13 Hwdrogen mass at time of containment failure s
f

A Table 4.4 Fisure of Merit Summarv for MAAP Uncertaintv/Sensitivitw Anaissis BWR VERSION 1.2   GRAND GULF    SEQUENCE T100V (FAI)
MODEL                                         FIGURE          OF     MERIT CASE      PARAMETER    VALUE         1       2       3       4       5       6                  7     8        9     10        11     12      13 (hr)     (br)    (Iba)    (F)          (esi/hr) (hr)             (br)     (lba)   (hr)      (hr)   (hr) (lba)
T10UV11      TTENTR    2.78E-3    52.89   49.14  260580. 182.2    0.047    1.15               N/A 0.66     412.0   1.64      3.76 10.94 3081.0 T10UV12 .FENTR         100.0      42.85   39.10  188280. 196.3    0.047    1.58               N/A 0.66     412.0   1.64      3.76 11.03 3432.0 T100V13      SCALU     100.0      51.57   47.82  247540. 182.3    0.047    1.26               N/A 0.66     412.0   1.64      3.75 11.03 3228.0 T10UV14      SCALU      20.0      51.36   47.60  245630. 182.3    0.047    0.79               N/A 0.66     479.8   1.64      3.75 11.00 3230.0 T10UV15      SCALH        0.5     55.84   52.08  132020. 162.4    0.046    0.86               N/A 0.65     407.3   1.63      3.76 11.02 3214.0 T10UV17      FUMIN      0.01      51.78   48.03  249680. 182.5    0.043    1.03               N/A 0.66     380.7   1.64      3.74 10.95 3194.0 T100V18      FUMIN      20.0      51.87   48.13  250640. 182.6    0.043    0.99               N/A 0.66     380.7   1.64      3.74 10.94 3189.0 T10UV19      ACVENT        10     51.71   47.96  249060. 205.3    0.043    0.84               N/A 0.66     380.7   1.64      3.74 10.97 3194.0 Default Model Parameters Chansedl                               Value             Fisure of Merit Homenclaturel FENTR     Multiplier for Kutateladze criterion for cavits                       1 Time of containment failure blowout (GT 1.0 =.difficultiLT 1.0 = easier)..... 0.33                2 Time between reactor vessel failure and Scalins factor for all burnins velocities......... 1.0
                                               ~
SCALU                                                                              containment failure SCALH     Scalins factor for heat transfer coefficients to                      3 Integrated wall condensation (measure of Passive heat sinks................................ 1.0                   diffusiophoresis) between reactor vessel FUMIN     Minimum burn velocitw.(reouired to be n/s)......... 1.0                   failure and containment failure 6CVENT    Containment failure vent area (ft**2)............. 0.1                4 Peak containment outer wall surface temperature                             e 5 Fraction of clad reacted in-vessel 6 Rate of change of containment Pressure dust prior to the time of containment failure 7 Time of ice depletion (if suplicable) 8 Time of core uncovers 9' Hydrosen senerated at time of vessel failure 10 Time at which clad temperature reeciies 2000 F            '
11 Time of vessel failure 12 Time of core melt Completion 13 Hwdrogen mass at time of containment failure

Table 4.5            Figure of Merit Summarv for MAAP Uncertaintw/Sensitivits Analvsis                               .
BWR VERSION 1.2    GRAND GULF    SEQUENCE AE (TEC)
MODEL                                                 FIGURE         OF      MERIT CASE     PARAMETER      VALUE          1            2         3      4        5     6'           7    8       9     10      11   12      13 (hr)          (hr)    . (Ibe)   (F)         (Psi /hr) (hr)      (hr)   (lba)   (hr)   (br)  (br) .(Iba)
AE0         None          N/A       50.62         49.06  854310. 256.9  0.004    1.61         N/A 0.01    33.3   0.50   1.57  5.81 2235.6 AE1        FCHF          0.05       50.63-        49.07  801200. 251.8  0.004    1.28         N/A 0.01    33.3   0.50   1.57  5.81 2598.3 AE2         FCHF          0.3       50.63         49.06  858760. 257.0  0.004    2.00         N/A 0.01    33.3   0.50   1.54  5.81 2236.9 AE3        FCDBRK         0.6       50.60         49.01  855280. 257.0  0.004    2.67         N/A 0.01    36.3   0.51   1.59  5.93 2237.2 Default Model Parameters Changed
* Value            Figure of Merit Nomenclature
* FCHF     Coefficient for CHF correlation in PLSTM.........               0.14          1 Time of containment failure FCDBRK   Discha rge coef ficient for Pipe break. . . . . . . . . . . . . 0.75          2 Time between reactor vessel failure and containment failtre 3 Integrated wall condensation (measure of diffusiophoresis) between reactor vessel failure and containment' failure 4 Peak containment outer wall surface temperature 5 Fraction of clad reacted in-vessel 6 Rate of change of containment. Pressure Just Prior to the time of containment failure 7 Time of ice depletion (if applicable) 8 Time of core uncoverv 9 Hwdrogen senerated at time of vessel failure 10 Time at which clad temeerature reaches 2000 F 11 Time of vessel failure 12 Time of core melt completion 13 Hwdrosen mass at time of containment failure 1
1 e

  * % 4 RESULTS FROM BWR-MAAP FISSION PRODUCT UNCERTAINTY / SENSITIVITY ANALYSIS PEACH BOTTOM l
4
        . _. _-  -_.          _ . _ i     ._  ._ . -- ---

s         s.
PEACH BOTTOM - STATION BLACKOUT Uncertainty / Sensitivity Analysis Base Case Case 1    Drywell structure mass increased to 0
2 x 10     kg.
Case 2    Osl vapor pressure.
Case 3     NUREG 0772 fission product releases.
Case 4     No core blockage, Temp. ZR oxidation cut-off = 3100 K.
Case 5     Sandia CsOH vapor pressure.
Case 6     1/4 of drywell floor available for corium.
Case 7    Drywell f ailure temp =600'F
   \
v

[
PEACH BOTT0li - STATION BLACK 0UT Uncertainty / Sensitivity Analysis DW Csl     0772    No- Sandia 1/4 DW  DW Base                 vapor Structure             Release Blk. Cs0H   Floor Falls Case      (1)      Pressure    (3)    (4)   (5)     (6)  at (2) 600 F Containment failure, hr*     18       29          18       18     18    18      18  14.5 Cs & I release fraction      .07      .07         .07      .07    .07   .07     .06  ,05 at 60 hrs.
Cs & I fraction in sup-      .06      .06         .01      .07    .07   .08     .07  ,03 pression pool at 60 hrs Drywell gas temperature     1875     1620        1900     1875   1850  1870   1520  1890 at 60 hrs (op)
Dr>well pressure _at con-    114      128         121      120    90    114     122  100 toinment failure" (psla)
Fraction of clad reacted     .08      .0d         .08      .08    .18   .08     .08  .08 in-vesse!
    ' Containment failure due to drywell 90s temperature = 1200 0F.

g ,e .
PEACH BOTTOM - STATION BLACKOUT Uncertainty / Sensitivity Analysis Base Case Case 1        All corium remains in-pedestal.
Case 2        All corium remains in pedestal.
Drywell structure mass     2 x 10 k g.

    -o                                                                      ;
PEACH BOTTOM - STATION BLACKOUT (Corium Distribution)
Uncertainty / Sensitivity Analysis Ba        Case 1 Case 2 ase Containment failure (hr)         18*      18.5+   20*
Cs & I release fraction         .07       .05    .05 at 60 hrs.
Os & I fraction in              .06       .05    .03 suppression pool at 60 hrs.
Drywell gas temperature        1875      1140   1080 at 60 hrs.-("F)
* Containment failure due to overtemperature.
* Containment failure due to overpressure.
i l
l i

o s.
f PEACH BOTTOM - STATION BLACKOUT AEROSOL MODEL Uncertainty / Sensitivity Analysis Gravitation Settling Proportional to p N N = .6        N = .5 N = .7 Cs & I fraction released         .07          .07  .06 to environment Cs & I fraction in RPV at        .67          .72  .60 containment failure Cs & I fraction in drywell      . 27           .22 .31 at containment failure Containment failure (hr)          18            18  18 l
N                                                           ]
l

or s RADIONUCLIDE RELEASE BWR -- MARK I DESRN (PEACH BOTTOM)
BMI-2104 RESULTS by James A. Gieseke Presented at the                                ,
NRC/IDCOR MEET'ING August 28-29, 1984 l
l OBattelle Columbus Laboratories

                                                                                 ., s.
l AE SEQUENCE e    LARGE BREAK LOSS-0F-COOLANT ACCIDENT (LOCA), FAILURE OF EERGENCY CORE COOLING SYSTEM
                                                                               ~
o    BREAK OCCURS IN A RECIRCULATION LINE e'   SUPPRESSION POOL REMAINS SUBC00 LED THROUGHOUT THE ACCIDENT e    C0tRAINMENT IS ASSUMED TO FAIL BY OVERPRESSURIZATION FROM NONCONDENSIBLE GASES PRODUCED BY STEAM-CLADDING REACTIONS AND CORE-CONCRETE INTERACTIONS OBallelle Columbus tahoratones

*%   4 I
ACCIDENT EVENT TIMES Event                                     Time, minutes Peach Bottom AEY Core Uncover                                 1.5 Suppression Pool Cooling On                 10.0 5 tart Melt                                 11.5 Core Slump                                  26.8 Containment Fail                           33.9 Bottom Head Dry                             40.0 Core Collapse                               65.2 Bottom Head Fail                           126.2 Reactor Cavity Dry                         126.3 Start Concrete Attack                      126.3 End Calculation                            727.0 OBattelle Columbus Laboratories

g c0 DECONTAMINATION FACTORS CALCULATED AS A FUNCTION OF PARTICLE SIZE AND OF TIME FOR AE SEQUENCE Time                 Particle Diameter, um               DF Based on (min)    0.1        0.7          1.2        5     8.4   Total Mass 5
14.3     1.2     3.3 x 10 2    105 (a)     10 5
10        1504 18.9     1.2     2.9 x 10 2      10 5
10 5
10        1400 4      5      5 27.4     1.2        51        2.5 x 10     10     10         25 5      5 33.2     1.3        5.4          69        10     10        4,)
(a) A decontamination factor larger than 10 5is assumed to be 105, Pool depth: 4 ft Bubble diameter:   0.75 cm Aspect ratio:  1:3 OBattelle Columbus Laboratories

.s o DISTRIBUTION OF SPECIES AT 20 HOURS AFTER ACCIDENT, AE-Fraction of Core Inventory Species         RCS          Pool       Drywell    Wetwell         Environment CsI           '0.19         0.35        0.12          0               0.34 Cs0H           0.19         3.34        0.14          0               0.33 Te         2.9 x 10-2    3.2 x 10 ~3    0.32          0               0.65 QBaltelle Columbus Laboratories

s  s.
FRACTION OF CORE INVENTORY RELEASED TO THE ATMOSPHERE FOR GROUPS 0F REACTOR SAFETY STUDY, AE Time                I              Cs           Te          Sr                  Ru                La (br)              Group 2        Group 3   Group 4       Group 5             Group 6          Group 7 0.5                   0             0           0           0                   0                 0
                                                                                       -3                 -5 1                 0.19           0.19    3.6 x 10 -2   1.2 x 10 -2         2.7 x 10         9.9 x 10
                                                      -2                               -3                 -4 2                 0.25           0.24    6.'6 x 10     1.3 x 10 -2         3.6 x 10         1.0 x 10 4-                0.34           0.33         0.51        0.64             4.6 x 10 -3          0.44 7                 0.34           0.33         0.64        0.68             4.6 x 10 -3          0.49 10                0.34           0.33         0.65        0.68             4.6 x 10 -3          0.49
                                                                                       -3 15                0.34           0.33         0.65        0.68             4.6 x 10             0.49
,                                                                                  OBallelle Columbus Laoorranes J

i O t. p TC SEQUENCE o TRANSIENT, FAILURE OF CONTROL R0D INSERTION (FAILURE TO SCRAM) e EMERGENCY CORE C00 LING SYSTEMS ' OPERATE e CONTAINMENT FAILURE RESULTS FROM THE IMBALANCE IN HEAT GENERATION AND HEAT REMOVAL DUE TO THE CONTINUED HIGH POWER LEVEL 0F THE REACTOR OBattelle Columbus taboratones

                                                                                 . .; s.
ACCIDENT EVENT TIMES Event                                     Time, minutes Peach Bottom TCY Containment Heat Removal On                 10.0 Containment Fail                            58.1 ECC Recirculation On                        72.4 ECC Off                                     72.6 Core Uncover                                73.0 Start Melt                                  93.6
  ~ Core Slump                                124.6 Bottom Head Dry                            136.6 Core Collapse                              178.9
- Bottom Head Fail                            216.6 Reactor Cavity Dry                         216.7 Start Concrete Attack                      216.7 End Calculation                            816.9 OBattelle Columbus Laboratories

sa DECONTAMINATION FACTORS CALCULATED AS A FUNCTION OF PARTICLE SIZE AND OF TIME FOR TC SEQUENCE Time      _
Particle Diameter, um                 DF Based on (min)    0.1        0.7            1.2       5.1   8.4      Total Mass 96.2    1.3    1.08 x 10 3         5 10 (a)     10 5
10 5
3690 5         5     5 99.2    5.2         98            10        10    10          2850 5         5     5 104      3.0         45            10        10    10          2166 121.7    1.1        15.8        1.87 x 10 3  10 5
10 5
7,7 5     5 131.5    1.2         4.5            41       10    10          298 5     5 156.3    1.2         4.0            32       10    10          600 (a) A decontamination factor larger than 105is assumed to be 105 ,
Pool depth: 6.5 ft (198 cm)
Bubble diameter: 0.15 cm Aspect ratio:  1:3
                                                                                   \
QBallelle Columbus Laboratories

DISTRIBUTION OF SPECIES AT 20 HOURS AFTER ACCIDENT,' TC y Fraction of Core Inventory Reactor Species      RCS           Pool           Drywell          Wetwell         Bldg          SGTS      Environment Csl        0.06          0.69          1.5 x 10 -2          0         6.9 x 10 -2   6.8 x 10 -2      0.10 Cs0H       0.22          0.56          1.4 x 10 -2          0         6.1 x 10 -2   5.8 x 10 -2   9.1 x 10 -2 Te         0.34       7.9 x 10 -3        0.29*              0            0.11       1.3 x 10 -2      0.25
   *Tnis includes a fraction of 0.13 for Te which is found not to be released from the core-concrete interaction.
L I
w m

i c. j DISTRIBUTION OF SPECIES AT 20 HOURS AFTER ACCIDENT, TC-y' Fraction of Core Inventory Species    RCS         Pool         Drywell        Wetwell Environment Csl       0.06        0.69        1.5 x 10 -2       0        0.24 Cs0H      0.22        0.56       1.4 x 10 -2        0        0.21 Te        0.34    7.9 x 10 -3         0.29*         0        0.37
         *This includes a fraction of 0.13 for Te that is found not to be released from the core-concrete interaction.
OBaneHe Columbus laboratones

                                                                              .,   e<
4 TW SEQUENCE e TRANSIENT, LOSS OF DECAY HEAT REMOVAL -
e EMERGENCY CORE COOLING SYSTEMS OPERATE e CONTAINMENT FAILURE BY OVERPRESSURIZATION-PRECEDES CORE MELTING e IT IS ASSUMED THAT OPERATORS WILL DEPRESSURIZE THE PRIMARY COOLANT SYSTEM BEFORE CORE MELTING OCCURS OBallelle Columbus Laboratones

ACCIDENT EVENT TIMES Event                                Time, minutes Peach Bottom TWY Containment Fail                     1756.2 Core Uncover                         2619.6 Start Melt                           2747.9 Start Slump                          2817.1 Core Collapse                        2818.9 Bottom Head Dry                      2829.3 Bottom Head Fail                     3055.2 Reactor Cavity Dry                   3055.2 Start Concrete Attack                3055.2                      ,
End Calculation                      3655.4 OBattelle Columbus Laboratories

                                                                                        ,, a.
DISTRIBUTION FACTORS CALCULATED AS A FUNCTION OF PARTICLE SIZE AND OF TIME FOR TW SEQUENCE Time                      Particle Diameter, um                        DF Based on (min)       0.2           05            1.0          4      10         Total Mass 4                5 2756        1.9           18        1.17 x 10      105 (a)  10             257 2777        1.4           9.4       2.51 x 10 3     10 5
10 5
576 5
2801        1.2          10.5        3.8 x 10 3     10 5
10             408, 5       5 2811        1.1          10.4        2.2 x 10 3     10      10             865 5
2815        1.1          11.7        3.5 x 10 3     10 5
10             352 3        5       5 2818       10.3          -60         9.9 x 10       10      10             326 2820     2.3 x 10 3   7.2 x 10 3     1.7 x 10 4    110 5
10 5
1336 5             5             5           5       5              5 2827        10            10            10          10      10             10 (a) A decontamination factor larger than 10 5is assumed to be 105, Pool depth: 6.5 ft (198 cm)
Bubble diameter: 0.75 cm Aspect ratio: 1:3 OBallelle Columbus Laboratones

DISTRIBUTION OF SPECIES AT 60 HOURS AFTER ACCIDENT, TWy' Fraction of Core Inventory Species    RCS        Pool         Drywell      Wetwell    Environment Cs!       0.14       0.80        5.4 x 10-3       0      4.8 x 10 -2 Cs0H      0.15       0.79        5.0 x 10 -3      0       4.5 x 10 -2 Te        0.40    8.6 x 10 -3      0.40*          0          0.19
    *This includes a fraction of 0.20 for Te that is found not to be released from the core-concrete interaction.
l l
l                                                             OBallelle Columbus Laboratones i

.s TASK 23.1 RESULTS SEQUOYAH Marc A. Kenton Fauske &. Associates, Inc.
16WO70 West 83rd Street Burr Ridge, Illinois 60521 (312) 323-8750 NRC/IDCOR Meeting on integrated Analysis of Severe Accident Fission Product Behavior Rockville, Maryland August 28 - 29, 1984

                                        <,              s e-f                                 '
SEQUOYAH SEQUENCES ANALYZED
: 1. Containment Failure Sequences a.. S 2 H F (1) Drains open/ cavity wet (2) Drains closed / cavity dry-
: b. TMLB'/ seal LOCA
: c. S2HF/ drains open/ containment purge open
: d. V sequence
: 2. Sequences Without Containment Failure
: a. S2D
                       ^
: b. S2H
: c. T     ML                ,
: d. AD i....

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4 SEQUOYAH S2HF/ DRAINS OPEN
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ACCIDENT 

==SUMMARY==

Time (hrs.)                Event 0               2" cold leg break
        .37             RWST low-level 1.2             Core uncovered 2.8             RV falls 4.2             lce depleted 10.0             Containment fails

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  ' 8   p SEQUOYAH S2HF/ DRAINS OPEN Cal Distribution (Fractions) at 20 Hours Deposited                      Airborne Primary System                                                                 -5
                                                                     .93                     3 x 10
                                                                         -3                             -5 Pressurizer                             1 x 10              = 1 x 10
                                                                                                        -5 Containment                                .07                     2 x 10
                                                                                                        -5 Environment                                                 = 1 x 10 t
i

SEQUOYAH S2HF/ DRAINS OPEN Release Fractions at 20 Hours Csl          CsOH      Te, Sb        Sr Ba         Ru,Mo
(
I
<                                                                                                  -5             -5           -5           -5
: 1. CsOH, Csl lumped;                                 N/A       9 x 10      1 x 10       < 1 x 10      < 1 x 10 JANAF v.p.
1
                                                                                                                              -5            -5
: 2. CsOH, Csl not                                 <1 x 10      6 x 10    4 x 10         < 1 x 10        1 x 10 lumped Sandia v.p.
                                                                                       -4          -4          -
                                                                                                                                -5 < 1 x 10-
: 3. Same as 2 pump                                 3 x 10      4 x 10    6 x 10         < 1 x 10 l                                               bowls don't clear
                                                                                         -5        -            -             -
                                                                                                                                            -5
: 4. Same as 2,0772                              < 1 x 10     5 x 10     2 x 10        < 1 x 10       1 x 10 releases
                                                                                       -5          -4        -5               -5    2 x 10
                                                                                                                                           -5
: 5. Same as 2, degraded                          2 x 10      4 x 10    5 x 10         < 1 x 10 sedimentation
                                                                                                   -5           -5 < 1 x 10 -5             -5
: 6. Same as 1. Sandia                               N/A      5 x 10     3 x 10                       1 x 10 v.p.

                           +.                 m    4.-     A--          -_*m        --._Ea--r_                _-- * ..- _ - , . ,    -   --. _ _ *_J ____4 --e_A._>                  ._ - - -                   -A.u _ a.--4 m      --
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s .
SEQUOYAH S2HF/ PURGE OPEN Fission Product Distribution (Fractions) at 12 Hours
* Cs, I                Te Primary System Deposited                       .92              .95
                                        -5                -5 Airborne                 6 x 10              4 x 10 Containment Deposited                       .07               .05
                                         -5                -5 Airborne                < 1 x 10            < 1 x 10 Environment                       .02              .005
  *Cs,1, CsOH lumped, JANAF v.p. for CsOH.
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SEQUOYAH TMLB' Csl Distribution (Fractions) at 40 Hours s
Deposited        Airborne I
Primary System                .92              .05         l
                                      -3               -4 Pressurizer                2 x 10           2 x 10 i
l Containment                   .02             .001         .
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o SEQUOYAH TMLB'
;                                                                      RELEASE FRACTIONS AT 40 HOURS Csl              CsOH   Te, Sb      Sr,Ba               Ru,Mo
                                                                                                      -5           -5                  -5 4
: 1. CsOH, Csl not   .001              .001 4 x 10     < 1 x 10            < 1 x 10 lumped; Sandia CsOH v.p.
                                                                                                       -5          -5                  -5
: 2. CsOH,Csl        N/A               .002 5 x 10     < 1 x 10            < 1 x 10 lumped; JANAF i
CsOH v.p.
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SEQUOYAH T 23 I

ACCIDENT 

==SUMMARY==

Time (hrs.)                  Event 0        Loss of feed, loss of injection, scram
             .9       S/Gs dry 1.5       Sprays in recirc mode 1.6       Core uncovered 2.9       Reactor vessel failure 4.9       Ice depletion
                                                             ~
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1 SEQUOYAH T        ML~
Cs,1 Distribution at 8 Hours
* Deposited          Airborne
                                                                  -3 Primary System                .98           ~ 5 x 10
                                              -3                -5 Pressurizer                4 x 10           4 x 10 Containment                  .012            1 x 10
                                                                 -5 Environment                               <1 x 10
* CsOH, Osl lumped, JANAF vapor pressure for CsOH.
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* SEQUOYAH V i
l

O SEQUOYAH V SEQUENCE n                q0C5.1                               ATE                 vSR                        ORU oN X                  :
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                                                                 . . . . ,                . i m                 m      m       m     m O.               5                   10         15                   20          25         30 TIME      HR Release history of fission products to the Sequoyah auxiliary building for a postulated V sequence.

SEQUOYAH V SEQUENCE
~
y0r C5.1                 cTE                       SR                            ,,-) RU oNt                                                                           /E       "
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e V

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MAAP 1.2 T/H UNCERTAINTY ANALYSIS 0
4

x T/H UNCERTAINTY ANALYSIS:
Conclusions in the context of the MAAP models, essentially no important sensitivities of bottom-!ine results to input parameters with a few exceptions:
1 Category 1:    Change in input value altered sequence definition, (e.g. sprays come on due to parameter change).
Category 2:    Parameter change reveals a fundamental
:                       weakness in a model by giving erroneous results, (e.g. core slump model).
'l 1

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