ML20134F720
| ML20134F720 | |
| Person / Time | |
|---|---|
| Issue date: | 08/01/1985 |
| From: | Felton J NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | Pedro J NUS CORP. |
| Shared Package | |
| ML20134F729 | List: |
| References | |
| FOIA-85-465 NUDOCS 8508210231 | |
| Download: ML20134F720 (2) | |
Text
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[o,, UNITED STATES b(
8(' a NUCLEAR REGULATORY COMMISSION D j WASHINGTON, D. C. 20555
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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
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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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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
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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
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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
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- 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
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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
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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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SUMMARY
RESULTS E. FULLER
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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
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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
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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
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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
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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
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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
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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.
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SUMMARY
RESULTS E. FULLER TEC 1
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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
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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
- 3. Steam system 3. Steam system I
- 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)
i
! CoM,CoH,CnP,CN H,CNR STATION BLACKOUT 11%
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 -
- 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
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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
- 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 _
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s _
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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)
- 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)
- 5) Small LOCA with loss of injection (S,D)
1 1
The dominant sequences for ASEP and the precursor study are similar
. 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 .
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(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= DOWNCOMER FW.HPSW (OUTER SHROUD)- - r LPCl(BWR/6) ; ff . -
- JET PUMP l CORE :
C A S LC -*= FLOW LOWER PLENUM l .:: l 0.'.*. " .*.". L ...
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- + . - - .a - - - - g g
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[ 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
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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 . t
. , 3 ceg s t eest sk b 7 ,_,y, cas2.<*eecaestme<F#ecusatente Tcittaple80lceaf.tc8elen*9*lch*l#18.c0.Citt gs G '" --
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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$2EN n3 ! O O.
)
X 93 F* I'l ( 5 3 0 0 0 0 3
)
K ( 0 e0 r O W u 5 0 D t a2 w r e p o % m O e0 T 0 s 0, u0 m2 i x a M 0 0 0, 5 1 0 0 0 0 1 0 0 0 5 L$*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 i e 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
\
l
. .J
.. i
..A y
- 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
, lih- .
f...
, 1lg ..+...
, 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...
, \
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-
- m. __
.O N k
~
N _ k - LL : re m em0 m
~.
5 ! e *-s -. BW M pm M O : e m 0 I I ' f R g g g g g g g ' ' ''''I''''I O. } 2 3 4 5 TIPE HR 1 l J I i
a e 2 ION Tit _B-
. Pywa EtI2Dt .CGE W
W i
~
i R r , L. W i gO n L l 1 3 - o 1 1 , W 2.>>
,,,,,,,l f i i i i l i i i i i i > < l i_ n i iiiiiI . ,,.. o il 0 I 2 3 4 5 TIE HR l
l l t i
l e e i ZION TILB-m -.
.O .
t X : i t - e M e m 2 W l o a a
~ -
,~ u w 11t"'
wn - Qll : p M m m e g i i t t i l 1 9 I i A B t R i l l i I E t i B t i R B l i R I I - a 1 4 I I I E & & 1 1
- c. I 2 3 4 5 TIPE HR
6 5 ZION Tit _B-n g .IN-CORE . IN-CONTMT O : m : En : b
~
l 9" - d E
~
N -
- i 2 -
5 O : P- : I
,,l,,..... ,l,,.,,,. .!,,,,,,,,.I f ,,.w..
, ,,,, ,, I l
- o. s to is 2o 2s TIPE HR
- 6 ZION TM_B-a e_
O m - X _ m W
-.J c1 --
m e e W N 2 -
>m 4
U W G e l 6 .........i...... ..i.........i.........i.........i l o. d to is 20 2s { TIE HR N ) .
a ch p
- I ZION TM_B-
- 8. .
O :
>JO : -
J : g O : 2 : g R. i. O e e = J O : f .. ......l.... ....l. .... .l... ... .l ........I 0 S 10 15 20 25 TIE HR
710N TM_B-
" . CAVITY . a nLarR CGPT O .O * .
X : m kO h W O w b H : g
~
J t - 5 v : u :
>N -
4 : m 5 6 .......gu,.........i.........i.........i.........i O. b 10 15 20 25 TIE HR
i ZION TM_B-
, . CAVITY .LSER CSFT O
m e
-8 L O :
5 u8 u L 4O : W 4 : m O : A e m
%9 R. L-O :
m
/
o 3 ..y.,,.......,../......,,
. l.,,,,,...l. ,,,,..,1 O. S 10 IS 20 25 TIPE HR 1
.4 &
- s ZION TM_B-
- e. D. _
ON X :
~
O {N M O. :__ 7 e : JQ : b d iiie i e i i i li>>>>iie lii,. i i,,,j,,,,,,,,,l,,,,,,,,,l l O. b 10 15 20 25 ( TIPE HR l I l l l l
( 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 e 9
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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 l [ 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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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 I 1 i
l l
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b l l 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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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
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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
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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.
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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
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< 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
.........- - ................................._ .......... ..=--- ..- --- = =......................
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!.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
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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 I l QBattelle . ( m.-, us. ., ., y , e I
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. CORE COLLAPSE 139.8 l
l HEAD FAIL ll49.8
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! 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.
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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 l l
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 J r I
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BWR Mark l Containment Nodalization
_ = O ?
+--- PRIMARY SYSTEM
' 1 DRYWELL I 2 WETWELL PEDESTAL CAVITY
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W . lllll \/ / DY 3 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
- SHALL LOCA -- PEACH BOTT0H
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l l l
l l 1 l l 1 1 l SMALL LOCA -- PEACH BOTTON l l
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1 4 I l i I l
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l 4 SMALL LOCA -- PEACH BOTTOM sn y
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I \ o . 1 l l SMALL LOCA -- PEACH BOTTOM
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d~ y Containment Failure : g _ _ l > -
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- 1 1
I l i 1 l 1 1 1 1 1 SMALL LOCA -- PEACH BOTTON 8 444 4 3 I g ig4 444 5 ai55g45 I I I E 4 63 giiI i aaiiI giiaI 4 5 4 5 1 gI 5 Ii& I I L g8 I sae I 4 i 3g3 44 54 54II l x : : j a8 _:_ _ o : . z- : : g - 6, i E
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SMALL LOCA -- PEACH BOTTON T iaii iiiigieiiiiiiigiiiiiiei,,
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h SMALL LOCA -- PEACH BOTTOM as .. 3........ 3 ... ....ii. ..... 3........ 3 ......3........, - o. 4 x - I t.a. - 1 _ _- l 2 2 : : l
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G SMALL LOCA -- PEACH BOTTON
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I SMALL LOCA -- PEACH BOTTOM a o- ! O* : x : l t i -
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1 SHALL LOCA -- PEACH BOTTON t 4 o : _ Z :
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1 I
SMALL LOCA -- PEACH BOTTOM
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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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HLACKOUT --- PEACH HOTTOM (TZOOFF=1010) e n l
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i I BLACK 00I -- PEACH BOTTOM (NPF=10)
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BL.ACKOUT --- PE ACH BOTTOM (NPF=lO)
- 8
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e O o se 1
- BL ACKui> T -- PEACH BOTTOM (NPF=lO)
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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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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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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
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ass 00ATED UNIVERSITIES, INCU ; WTP:2,25
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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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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 l l 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 .
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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 t I
$ 9 8
~
ca o Z CO O O Z
9 8 e e . 1 AVERAGE HEAT FLUX (kW/m2)
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/2 - 6.7 '
I I
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- m l m
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L 1 100 - - 1 l l . l 50 ' ' ' ' O 2b0 4b0 Sb0 8b0 1000 1200 1400 1600 1800 Ti,me (seconds) 1 l t-. . .-. .
Comportment 7 800 , , , , , , , , . c . - .
. .) _-
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.M 600 - - -
m o . . i O L en . . n o l l o e . . . 3 1L 400 - - - o . . . Q. t! o . . . H 200 l l l l l l l l 0 200 400 600 800 1000 1200 1400 1600 1800 TLme (seconds)
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0 2b0 4b0 6b0 8b0 1000 1200 1400 1600 1800 TLme (seconds) i
i l l I Surface 27 i 1 14000 , , , , , , , , a 2 - 12000[ - n: . x . .
- 10000 - -
e . . N . . 2 . . x 8000 - - 3 . . L- . . a 6000 - - g . . z . . _3 4000 2 - - O a - , O ' g . . 2000 - - - a - . n . . z . 0- - -
-2000 l l l l l l l l 0 200 400 600 800 1000 1200 1400 1600 1800 TLme (seconds)
Surface 27 600 i i i i i i i i
- 550 - - -
C - - - Q . . . M 500 - - - u, . - . D . , e c . . . en . . . y 450-P e . . . 3 . . . Y 400-e . . . () < , . Q. ' ' ' E o - - - H 350 - - l 4 > . 300 l l l l l l l l 0 200 400 600 800 1000 1200 1400 1600 1800 TLme (seconds) i l _ - - - . . . . _ - _ _ _ . - - _ . . . . . _ . - . . . - . _ - - .-____--.-___.-.___-I
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- 0 1
rA_ g 7-y M
% D 1 x v <
O M
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l I l l i
t i i i i PRESSURE IN THE WET WELL 2s , , , , , , , , 26- - - n m 24- - - G.
- 22 - -
w a
-) 20- - -
M M w Is. - - m
- 16- - -
g4 i , , 0 200 400 600 800 1000 1200 1400 1600 1800 4 l TIME (S) i i l
J wi*. . . . - . . a A-- - _ _ - _ a- - _ _ ___A ____. m. . _ _ _ -. _.6m._. m _ .-_ _ Y MAX AND MIN TEMPERATURE IN WET WELL 1800 , , , , , , , , '1 1600 - - C 1400 - - 1200 - - W
% 1000 - -
1 D W
< 800 - .
K W 6000 o. i E 400 W - - 200 - -
^
0"
^
O 200 400 600 800 1000 1200 1400 1600 1800 TIME (S) i 1 1
l , l I i HEAT FLUX TO THE INNER WALL 1600 , , , , , , , 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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[f h -F be-f;,. Q ! h5 / l Nl 7 ///M/MM/dh//57/Y//////N/Ylll[l)//f - BWR/6 Mark 111 Safety and Other Systems
.xv.i, .
COMPARTMENT 5 PRIMARY SYSTEM COMPARTMENT A_ c\ s[
/ l l
; ORYWELL m/ '
I In WETWELL- l b g_ I PEDESTAL CAVITY V////////////////////////////1
,,, ""'l,""
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- COMPAATMENTS ; /f6
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- ORYWELL 7- == i '== -3 wETwtu.- - ,
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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,_, , _ , . _ , _ - , , - , - - - - - , - - - - - - . - ,
i 1 i i TIOUV.W ADS -- GRAND GULF l
" N
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; o. o.2o o.4o o.so o.so i TIME HR i
4
T10tN.W ADS -- GRAPO GLAS tri ........g.......;,.iii;iii..gi>>>i..ii. x L. v .
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O. 0 20 0 40 0 60 0 80 1 i TITE HR
T IDUV.W ADS -- GRAM) GULF
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i T IOUV . W ADS -- GRAPO GtA_F a e ..........g . . . ..... ... ... ...i. .i....i. o x : : u, : : m .. z : : r - l l l . 2 : : Wn p E : :. g . v : :
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f .. O. I 2 3 4 5 TIME HR x 10 ' i l s l l l l l l l
O e i l T1OUV.W ADS -- grate GULF
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- o. 1 2 3 4 s TIME HR xlO '
T 10UV . W ADS -- GRAM) GLLF
- sn OM .........i......... -
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. D . e T10UV.W ADS -- GRAM) GLAS n
X - . w Jo : : o . Z - oo :
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> :/ :
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O e T lOUV.W ADS -- GRAM) GLA_F _ in O N ''''l''''I''''l''''l''''. x - E - .
>e z
o W - m - 2 l .z. c. n
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O. I 2 3 4 5 TIME HR xlO '
9 F l i l l 1 l i I T 1OUV.W ADS -- GRAM) GLAS '
- c. m.
gn . . . . . . . . . . , . . ......,.........,.........,.......... x - N - .-- L. : : Z m. .: g - . W W - H - 8 O 2 . [ . f I,,,,,,,,,1,,,,,,,,,t,,,,,,,,,1,,,,,,,,,1,,,,,,, 0 1 2 3 4 5 TIME HR x 10
- t l
l I
T1OUV.W ADS -- GRAM) GLAS n m-O x : e - en : Jn .-
/
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i T 1OUV .W ADS -- GRA>B GLA.F a e __
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-M _
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O. 0 50 1 1.5 2 2.5 3 3.5 4 4.5 5 TIME tR l l
O . 9 9 f T 1OUV .W ADS -- GRAM) GULF c W . __ ON : . i X m W f!3 _ d - Im O - Q. 6 , . E eu. M =
=
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- o. o.so 1 1.5 2 2.5 3 3.5 4 4.5 5 TITE HR i
I l 1 i i
~
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
- t o
,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 , o - 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 { 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,- - _ _ _ , . , , _ _ , . . _ , . , _ . , _ , , , , _ , , , , , _ , _ , _ , . , _ , _ , , _ , _ _ _ ._ __ _ _ _
me .w
\
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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Lon91tudinal Section of BFNP Reactor Building
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. ,
If w 1 5 ! Fa..o o "' ant . 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 ) l ( ) l
e .*, 6 10 , RUN A-5 IC0!NE
. TROUGH AREA ELEV.. ORA!Ngo.
. $V460L LOCAft0N ft ft o L13 (-) 4.0 3400 310, "", i a LO9 (+) 29.7 20
-e L10 (+) 19.7 20 y -
- --.e LII (*) 9.9 20
- ly -- a L12 (-) 0.2 20 E h 5
w 11 ny n 4 t 30 1 :"i 5 E , Il ' I ' g10 '- g MAAP Model u . _
~U & ,
t
-*-ii yg 2 l l l I l I I l i I l !
0 2 4 6 8 10 12 14 16 18 20 22 24
- t. hr
1 AEROSOL DEPOSITION BY STEAM CONDENSATION Steam Velocity W Oc l Us = ps ,A, A s A w h gg l l Aerosol Deposition Rate W=D a a Aw Us A o Wa " p ,'a h gg l
to' .
. Rum A.2
", j CEstum ELEV.,
LOCAf!04 ft ia s r, MAAP Model a tos + 2s.1
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- t. Mr Cesium concentration in wall condensate film, Run A-2.
1
f , AEROSOL MODEL SEDIMENTATION Model
@=-Am+WI-WO dt A,Q U s
=
(O P - Og)9d V 18gg Assumption n A=cp Determine c and n from large scale experiments. Application Compare model with the spectrum of experiments available. l i
FROM THE FUNCTIONAL RELATINSHIP un
^ ~
p ~_ 90 _ n Ag (T - T g) + 1 _ This allows a large number of experiments to be combined on a universal plot, 1 l
o SODIUM OXIDE & IRON OXIDE AEROSOLS O" N i i i IiIIll i i i l I iiIl i i i i i i111 l 1 i i I i li L. gr : 1
~
go - X hg g HEDL - 7 Y g ZZ = O 60 0 TEST ABt C &W O e TEST AB5 _
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UOl ~ t O R tJ L u o a Run tot ! l O A RUN 102 _
. Y RUN 103 -
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- O RfJN 104 -
.O V PUN 106 -
7 d' O Run 107 O a X RUN 511 ([RON) _
~
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- 10 ' 10 10
- ZZs(T-To)+1 1
I I 1 o URANIUM OXIDE AEROSOLS o 4 I I I lillli i i i 1 11111 I I I 11111l I I I i 1111 t. _ _~ i a V _ I
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g ZZ = 0 60 o a
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1 i APPLICATION OF THE CORRELATION TO OTHER EXPERIMENTS AU s P d* A- C P
-G j p"
V h g Pp Assume C3a h P" Characterizes the mean settling diameter as a function of aerosol concentration [h ref h II Pp h l A = Gq ( h )(PP ref I 3 h ref ~ 5 m Ppref ~ 400 kg/m l
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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 :: :
.z 21 'Cl W~ :- h 3
' Z
'T jZ a ao: " ^J.
- s -
g Z ? MODEL PREDICTION i ;- - Cl p s _5 c- - 7 C Z O!
~
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 _ a - o - o
- 2' v
= u < ~ = c 5
00a w E -
- a a
__ a. _-
,a a
a a
=
=
g = g wa
.a i . .i g
s -- a - 7 a 3 k.-- = g E .1
-. }. = a
=
=
e a, c - , m
- 5 m E
~
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=
vi i i iin n i i .vi n i , i ,niii ...,i.
. i..,
r_O f ,Ot Ot O f .,_Ot iC++W'9:4) ,_ O3:C 103_ CEyN
~_.
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
~
10 DRY AEROSOL
~
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 - ~ _ _~ - ~ _~=_ _ :-- - - =_ - -
O a 4 i U R A
@m N
I o eA U 8 i M g 3 t h 1 O r 2 D I N ABO E O UC I 1 6 I s D RE s' . 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 : A 4 - I ~ Q.
~
(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 , CD ORNL 30! ~ h
@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 c: ,. 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 %
22 = 0 70 7 C' Ok g O 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 C) i I I IIlill i I I I I 11 I I I I t illi i i i i t ill 0 1 2 3 4 10 10 10 10 10 ZZx(T-To)+1 l 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 _ I a - ir - ! x - e .
.2 -
j $ _' ~
+ _
i e - O -
~
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 - i 0 - i n : : O ~ i e f - i u : 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' . i saasisi 5 s 5 iea aaesisiae I aiesieae 4 aaaaa6 66_'i,
- O
- : X
- ^
g
- O e m
- : o y
L : *
- O m um o
N :
- O
- W
@ y
- O
_ y
. : bb*
- O
~ O f M E
d O H e
=
m J 1 ( = 8 O I : : 3 : O N D : : .
-c: .
- o
- e. - _
S5 i O " O - c :- : *
- O m
m g
- O 7fffIfG fi! I iifff ffI ifIfff I II I fI f ff fif if ff ff e f l=
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
)
J AO SL P ONN OOE .
- LI K n
" T O o J N A Z R i
t EI B a KLN z i OAU [/ ) l RD a IIl1I} C 8DS d o
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;MU J- e e
TR MN - E E g i n g
- n
@E L - E -
j E P [/
- h i t
n o 7 s u b R m o O R C T E d A M n FEL INE RL TG OE HL
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8 O L W e _ ! ,i l IIlII ( 9 P h6 P "N A A
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t s y g s R M _ y E RU E _ Z EN P l k R 5_ j k r a RI G P E A O m U E UPL C _ i S L _ r p S E T R O
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- W A M
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E _ _ D K O R P BO NO _ UL
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,,. G }n/
- /
i I UPPER COMPARTMENT a..utAR . * " ',." co= aa?=e=v ..............., t so... Cou anTwEsef
"-"I;-" . . . . . . . .
. . - ,,.;-~ -
u n .e. . . . c"c" l
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, l'E a u at %g.a f:. j
........... , , , _ .we.ong, ,g3,3i
. .. 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
.~ .
6 4 6eiea 46 g *es 1 664 g6 a4 66 a4eg& a i4 .a ii4 gi*6s4 e ie I _ _- r
- .__ _i-
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c g _
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f m. g - 2 - - - O - k
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_. . ! .. . . ,l, ._t. . ,__r2 1 u tz ' , i. o S ir C E I 'O c OIX ISd SS38d SAS ABYWIUd I I l
i 1.. . . 1 i igiiiii.iiigi. p.i.j.isi.i.g.iiii..
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i
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- e 1_1 i ul..,. , , , . . . .il 11 .,11.,,.. ,,
o S2 E S1 I 05-0 O
, OIx ISd SS3Hd 11.J1NO3
GAS TENPERATURE DEG F x10 ' 1 3 5 7 9 11 0- _ av -gu ....
;ii,m in .i . . . - i- i
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9 M N i l n M C -
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m r.; N o - r - n-o
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l m - S' o - [ ] n l
' O l3I fM O f fk f f fff f f ff f f f f
o'A. a... ....i i.,i
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iii.1 . i i i . . . . i i: , , , ,
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tda ..., 111 .... i .- o 01 8 9 ir 2 O
, OIX A1 "1NI1Y3H del N31d dn
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m no i.6 siii6 iigi.. i4 .6i g 4. ,
- i. i . g , ii.. 64 . .iai..ii n'
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; l__ _ t i .13_ _ti d G3 E S1 I Osi- o -0 c OIx X 930 SdH31 7431d df)
I I AM \. AM GENERATOR GENERATOR PRESSURIZER s, SHELL SHELL g)I I l i s l
/ \ l /
l l
/ HOT I 3
- LEG _ .l. _
w y}j y :-- % t DOWNCOMERk Nm m - O ( ) ( ~J 3" UNBROKEN" LOOPS l' BROKEN' LOOP (NOOALIZATION SAME AS UNBROKEN LOOP) TMLB POST-VESSEL FAILURE FLOW PATTERN Y " 1
.~ c
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i X So.
' 1
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~ Q 01 S 8 L 9 S; 7
OIX N SdW3_L 3BO L3081S I l l
. s ,
s6.isiiiigiiiiiii3.gi4. .ri4..g uaiiii4igii4i iii.iji64i4...t o.
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- _ ~~. : x
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v
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n .
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,,...n. J...,..al,,,,..I,,,....l,,.,,lo,, ,e o
l u EI I 08-0 09-0 C+ - O 03-0 -O t l 0 AV330eS3SS01 Sd i
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 . p s p s S p m C C i t C v u f v d
, 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
]
. e U
c 3 l Q 0-j - - u. 2 y .
~..'
- c. -
o j . o w I ~ u . ec . e z~ (
./ _
u m a ,.
- 3. :
I' y ,/ j cc .. u E . , r
/
1 go u
* ! ... to 2 ,, u w i m o-t
& 3 @ ~
i -a. ,
.c I ./ 4 m
. 3 g
; 1 r .- ..i e !
' m. l
- - . . -....-._......_...-___.., . 1.' g c l
, . . ,1 &
e L: ,/ .5
~
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I U
.i e
, o
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.I ~
i.
.. l t 1./..L..Ll..'..LL bl1Li L. A. J-.~.[.L .LL1 L. t .1.1 .L!_L1. . .t. . l . . J. . . Lt. L _. '
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 . _ _ _ _ __ __
-o T
l o - 9 in _
- u. -
1 - N CC w - I u - 2 / _ W W [ _ o, E o , c y (B ! i ~
\ !
I s 2 ' O _ I N ; i _ -o
! ~
l i _ I,,,,,,,,,1,,,,,,,,,1,,,,,,,,,I,.,,,,,,,i,
,,,,,,,_ o ,
S & C E I O
, OIx 13 73A37 831VM 13SS3A P
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. as a
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o
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.: tn .5 L------G-..__.-...____.__ '
.1, __ D n 3
m
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D Z
> i3 -
Cr .2 o .- .. aa-
- 2. ~m .-
3 n .a u.! [ > - e, 3 :* 3 Lu 5 I"
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o Fm" 2, i la
. _ =
a U m o b N ~ o. C
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.c as as L7 "
d E 5J C L.i . _t..t...t...t.. i i t.. L_ L.t u..u._u...t._t L t_.t -. t.. . i...t.. t...t...t.w d G! I Ot; - 0 O c 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 . y- r , _ 1 r
- . _ . 3, t-
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- * * - - ;! : 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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6 6 6 i4 6 6 1 6 l4 i4 iiI ee6 g6 4 4 4 4 4 I 4 6 g6 i ei6 4 I i e g6 4& s64 i 4 e_ s e _ ; n u : : cc:
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m i _ n t _ 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 ! l i 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
- s
-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 l
s s. 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
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4 SEQUOYAH S2HF/ DRAINS OPEN .3 i i i l
s . 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
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- 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 --
i 1 I t A d I i i j SEQUOYAH S2HF/ DRAINS OPEN/ PURGE OPEN i i I I l l ( _ I
- 1
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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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4 i o l SEQUOYAH S2HF (DRAINS BLOCKED) 4
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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 . l 1 , Environment .001 1 r l l t
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. i i i j j I
e, .a f 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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. e, - p l 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 : 0 - M _ N _ o : c - __I : 2 - x !10:-- - 3 -
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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).
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