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j J OCT1 2 E4 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 i. 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.
Contact:
J. Mitchell, 49-29402 C. Peabody, 42-74632 Y./_gd2h b gg
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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.
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Themis P. Speis, Director Division of Safety Technology Office of Nuclear Reactor Regulation
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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 N
af DISTRIBUTION NRC/NRR NRC/RES ACRS H. R. Denton R. Minogue R. Tripathi F. 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 W. Morrison B. Sheron -
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 R. 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, OEDR0GR 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
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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 D. Paddleford, W~
J. Hickman L. A. Wooten, W 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, PECO V. Behr G. Krueger, PECO 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 i
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B g C Enclosure 1
SUMMARY
OF THE PRINCIPAL TECHNICAL RESULTS 10COR/NRC MEETING AUGUST 28-29, 1984 The August 28-29, 1984 meeting between IDCOR and NRC focused on the integra- 'J
. 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 discussions, there is a better understanding of the technical differences and their bases between IDCOR and the NRC contractors and consultants.
IDCOR 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 (FPRAT 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
J 3o 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 concentration.
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Two fitted constants have been determined from experiments and compared with other experiments. Water absorption by hygros.copic 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.
IDCOR 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 TRAPMELT 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 IDCOR 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 IDCOR and NRC contractors in their presentations and summaries.
For details, see the viewgraphs of sequence definitions and ultimate distribu-
D ,UI O
'd' tions in Enclosure 5.
Zion For Zion, IDCOR 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 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> for the TMLB and was assumed from the beginning
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for the open purge valve and V sequence cases. IDCOR found a sensitivity to .
assumed 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 CsI 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).
t During the summary discussions, the NRC contractors stated the belief that the uncertainties were "much larger than the IDCOR 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
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~4 of CsI, Cs0H, and Te of order 10 or less. For isolation failure (comparable to an open purge valve) the releases of Cs and I were of order 10 -2 . 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 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> (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 i
size failure of the containment. In contrast, previous calculations have assumed that any failure of the containment for this sequence would cause l
injection failure. Cs and I release fractions were calculated to be of the l-same order as given above.
l l
l The station blackout event was also analyzed for different assumptions of the l
area over which core / concrete reactions could take place. Although minor ,
details varied, containment failure time (about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />) and Cs and I release fractions (0.07) were unchanged. A sensitivity analysis for thermal hydraulic
.o e o parameters similar to that for Zion was performed with similar results. It was noted following this presentation and reiterated in the summary presen-tation by the NRC contractors that extraordinary operator actions'were 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. -
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.
For the large break LOCA, containment failure was calculated to be at about 30 minutes, with release to the environment of about 30% of the Cs and I and about 60% of the Te. The releases for a;. 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 about 5% and for Te is about 20%.
i Sequoyah l
, 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
e s +
0 about 0.5% of the Te. For other sequences, the release fractions are 10 -3 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, and 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 (delayed) 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. 10COR
! analyzed four sequences and three were reported from BMI-2104. Both analyses l
l included significant credit for suppression pool fission product scrubbing.
i
[ The techniques for taking scrubbing into account were different. IDCOR 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.
l- 10COR assumed a nominal pool bypass until vessel failure (affecting only the ,
large break LOCA sequences). Subsequent to vessel failure, the bypass path has j
been previously evaluated to be blocked by aerosol particles. For all sequences, t
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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.
e
- t a ENCLOSURE 2 AGENDA l
-l NRC/IOCOR MEETING E!!
INTEGRATED ANALYSIS OF SEVERE ACCIDENT FISSION PRODUCT BEHAVIOR AUGUST 28-29, 1984 Tuesday, August 28 Introduction ,
8:15-8: 45 A.M. -
Welcome, Purpose, Ground Rules, Schedule - T. Speis, R. Bernero (NRC)
- Introduction - A. Buhl, M. Fontana (IDCOR) 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~ .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 Fission 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 ,
i Enclosure 2
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 3:15-3:30 P.M. -
Break BWR Mark I _
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3:30-4:30 P.M. -
IDCOR Results 4:30-5:00 P.M. -
NRC Conractor Results 5:00 P.M. -
Adjourn 7:00- -
NRC Working Group Meetings Wednesday, August 29 PWR Ice Condenser 8:00-9:00 A.M. -
IDCOR Results' 9:00-9:30 A.M. -
NRC Contractor Results 9:30-9:45 A.M. -
Break
, BWR Mark III l' 9:45-11:15 A.M. -
IDCOR 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 '
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I Enclosure 2
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- 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. -
10COR Summary 3:30-4:00 P.M. -
Parting Remarks - R. Bernero, T. Speis, A. Buhl t
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Enclosure 2
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- a 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
- 8. Burson Wen-Shi Yu L. Chan J. Conran EPRI M. Cunningham R. Curtis M. Everett
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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 R. 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 3
.. = .
l 4 Enclosure 3 Other Affiliations ,
K. Araj, Harvard University S. Asselin, TEC-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 S. 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 O
Enclosure 3
.o o a .
Enclosure 4 NRC CONTRACTOR AND IOCOR
SUMMARY
VIEWGRAPHS* ,
'*VIEWGRAPHS AND PORTIONS OF VIEWGRAPHS IN THIS TYPEFACE WERE HAN0 WRITTEN AT THE MEETING AND RETYPE 0 FOR CLARITY.
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Enclosure 4 ,
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IOCOR AND BMI-2104 RESULTS NOT GREATLY DIFFERENT e PROBLEM IS VERY COMPLEX :
i e SOME DETAILS DIFFER, BUT OVERALL THE SIMILARITIES PREDOMINATE ',: ;
+
e EXTENT OF AGREEMENT GRATIFYING ,
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10COR l l
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_..-.__.--,--n.--,----- - - . . , -
.
- o BOTH IDCOR AND BMI-2104 ANALYSES ARE ADVANCES FROM WASH-1400 0 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 10 DINE AND CESIUM -
e SIGNIFICANT DEPOSITION OF FISSION PRODUCTS IN PRIMARY SYSTEMS
. CALCULATED WITH MODELS ACCOUNTING FOR AEROSOL FORMATION AND SETTLING 6
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HYDROGEN PRODUCTION e IDCOR UNCERTAINTY AND SENSITIVITY ANALYSES SHOW NO EFFECT ON CONTAINMENT FAILURE AND FISSION PRODUCT RELEASE
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b e IDCOR BENCHMARKING OF TMI SHOWS MAGNITUDES OF H GENERATION 2
THAT AGREE WELL t
A 9
IDCOR
-o r o TE RELEASE e BMI-2104 TREATS TE RELEASE BOTH IN-VESSEL AND IN CORE DEBRIS-CONCRETE ATTACK e IDCOR CALCULATES COMPLETE TE RELEASE EARLY _
o EXPERIMENTAL EVIDENCE SEEMS TO SUPPORT SOME TE RELEASE Dur1NG CORE-CONCRETE ATTACK e IDCOR WILL CARRY OUT A SENSITIVITY ANALYSIS TO DETERMINE EFFECTS OF TE RELEASE DURING CONCRETE ATTACK e SHOULD ONLY AFFECT CASES WHERE CONTAINMENT FAILS EARLY h
IDCOR
O-* A
+-
O REVAPORIZATION 4
j e IDCOR CONSIDERS UNCERTAINTIES IN VAPOR PRESSURE i
e VAPOR PRESSURE MAY BE LOWERED ONCE EXPERIMENTAL DATA ARE OBTAINED AND EVALUATED q-
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' IDCOR e
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EARLY DRYWELL FAILURE FROM DRYWELL LINER ABLATION e NRC POINTED OUT POSSIBILITY BUT THEY ASSUMED HIGH DEBRIS INVENTORY e IDCOR CALCULATES THAT DEBRIS INITIALLY QUENCHED BEFORE DRYOUT AND REHEATING e IDCOR SENSITIVITY CASE SHOWED LITTLE EFFECT 9
6 IDCOR
.. o SGTS IN PEACH BOTTOM e PEACH BOTTOM TEAM HAS EVALUATED THE SYSTEM e 10COR WILL LOOK AT THE DESIGN AND OPERATION AGAIN 0
0 IKolt t_
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CONTAINMENT LOADS
SUMMARY
IN-VESSEL HYDROGEN PRODUCTION HYDROGEN COMBUSTION BEHAVIOR MODE OF MELT EJECTION CORE / CONCRETE INTERACTION STEAM EXPLOSIONS f DIRECT HEATING OF CONTAINMENT ATMOSPHERE SEQUENCE DEFINITION SENSITIVITY UNCERTAINTY ANALYSIS t
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NRC CONTRACTORS
2 a r
IN-VESSEL HYDR 0 GEN PRODUCTION DIFFERENCE HYDROGEN GENERATION DIFFERS BY FACTORS OF 1.4 TO 10,0 j
BECAUSE OF BLOCKAGE ASSUMPTIONS, LACK OF DOWNWARD RADIATION HEAT TRANSFER AND THE SHORT TIME BETWEEN _
- SLUMP AND VESSEL FAILURE j'
i SIGNIFICANCE AFFECTS PROBABILITY AND TIMING OF CONTAINMENT FAILURE _
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NRC CONTRACTORS l
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HYDROGEN COMBUSTION BEHAVIOR DIFFERENCE ,
FLAME TEMPERATURE CRITERIA USED PREVENTS HIGH HYDROGEN CONCENTRATION,
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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),
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l NRC CONTRACTORS t
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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.
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NRC CONTRACTORS
r s 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 MOLTEN, 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 P0OL NO FURTHER OXIDATION OF METAL COMPONENT OF MELT l
NRC CONTRACTORS -
< 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 l-
. r DIRECT HEATING OF 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, l.
SIGNIFICANCE POSSIBILITY OF EARLY CONTAINMENT FAILURE EITHER BY DIRECT HEATING ALONE OR IN CONCERT WITH HYDROGEN, r 1 NRC CONTRACTORS I
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SEQUENCES DEFINITION DIFFERENCE THE IMPORTANCE OF SEQUENCE DEFINITION IS RECOGNIZED BY BOTH NRC AND IDCOR.
SIGNIFICANCE THIS IS AN EXTREMELY IMPORTANT ISSUE AND CAN INFLUENCE PROGRESSION OF CORE MELT AND MODE AND TIMING 0F CON-TAINMENT FAILURE, NRC CONTRACTORS
+. ,
SENSITIVITY - UNCERTAINTY ANALYSIS DIFFERENCES IDCOR STUDY BASED UPON INCOMPLETE SET OF PARAMETERS DOES NOT INCLUDE VARIATION OF PARAMETERS BASED UPON ENGINEERING JUDGMENT EFFECTS OF VARYING 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, s
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1 e NRC CONTMCTORS i
l
a 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 EFFECTS 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 s
COMPARIS0N WITH EXPERIMENTS AT DIFFERENT SCALES OR WITH MORE RIGOROUS MODELS WOULD HELP TO DISPEL SKEPTICISM NRC CONTLK; TORS
- 2. REVAPORIZATION MODELING 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 l _
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 TO IGNORE RESUSPENSION FROM THE RCS AND CONTAINMENT WITH THE CURRENT LEVEL OF UNDERSTANDING.
O NRC CONUMC10RS
4 TELLURIUM RELEASE THE IDCOR MODELING 0F 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 10DIDE RAD 10 LYSIS 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, O
hRC CONTRACTORS
..' o
. ,e '
/6, CORE-CONCRETE RELEASE F56MOURANALSESWEBELIEVETHERELEASEOFFISSION
't- PRODUCTS AND AEROSOLS DURING CORE-CONCRETE ATTACK W IS IMPORTANT, IDCOR'S APPROACH NEEDS' CLARIFICATION,
/ 7,' POOL SCRUBBING WE DISAtiREE WITH THE APPROACll, THE DF'S SELECTED ARE' CONJECTURAL. THESTATE-0F(THE-ARTOFDATAAND
, MODELS PERMITS A BETTER TREATMENT, Y~ , ;,8. . !CE BED DECONTAMINATION ~
/ ,1
, , f WE BASICAl.LY AGREE THAT.THE TWO MECHANISMS BEING INVESTIGATED BY IDCOR ARE'Td MOST IMPORTANT BUT ARE CONCERNED ABOUT THE ABILITY TO MODEL SE0iMENTATION WITHOUT AN EXPLICIT TREATMENT OF PARTICLE
" Sl ZE . . - , ,
- 9. FISSiONPRODUcfDECAYCHAINS /
TliE DCK 3 OF EXPLICIT TREATMENT OF FISSION PRODUCT DECAY CHAINS (E.G.,Te132 ,, g 132) COULD AFFECT THE TRA$ SPORT OF FISSION PRODUCTS
~
AND,TH71R PARTICULAR RELEASE TO THE ENVIRONMENT, PARTICULARY OVER LONG TIM 7. ?EAICDS OR FOR SMALL SOURCE TERMS.
/
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/
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NRC CONTRACIORS
. 1 i
O m e s
- 10. AEROSOL GENERATION MECHANISMS THE POTENTIAL FOR ADDITIONAL AEROSOL GENERATION MECHANISMS (STEAM EXPLOSIONS, HIGH PRESSURE BLOWDOWN) SHOULD BE RECOGNIZED AS CONTRIBUTING TO THE UNCERTAINTY IN kNVIRONMENTAL RELEASES.
O 4
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O NRC CONTRACTORS l
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e 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.
l l.
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l l-NRC CON 1RACIORS l
l
._ =
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 CASES WHERE THERE IS APPARENT AGREEMENT, IT MAY BE FORTUITOUS, 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.
FOLLOWING COMPLETION OF MARCH-MAAP COMPARISON CALCULATIONS 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 POOL BYPASS IN THE MARK III CASES IS INADEQUATE.
l t .
NRC CONTRACTORS
Enclosure 5 9
VIEWGRAPHS FROM TECHNICAL PRESENTATIONS
. Enclosure 5
OVERVIEW 0F MEETING TOPICS EDWARD L. FULLER EPRI/IDCOR IDCOR/NRC MEETING ON INTEGRATED ANALYSES OF SEVERE
- ACCIDENT FISSION PRODUCT BEHAVIOR i
.5 u
ROCKVILLE, MD
(, .
AUGUST 28-29, 1984
e .
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
.1 0 INTEGRATED ANALYSES OF THERMAL-HYDRAULIC BEHAVIOR _
DISCUSSED AT ROCKVILLE, MD IN MAY 1984 9
ELF /Is 4
2 1
, s' .
THE MAAP 2.0 CODE IS NOW USED TO PERFORM THE COMPLETE ANALYSIS 0 PROPER MODELING OF EFFECTS OF REVAPORIZATION REQUIRES TREATING j NATURAL CIRCULATION IN THE PRIMARY SYSTEM. THE CIRC CODES WERE DEVELOPED FOR THIS O 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 IS MAAP 2.C h
a
FOCUS IS ON FISSION PRODUCT BEHAVIOR DURING POSTULATED .
SEVERE ACCIDENTS IN THE IDCOR REFERENCE PLANTS 2
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 ELF /Is e
. _ . . _ _.. m 7,
O #
SEVERAL ISSUES DISCUSSED AT PREVIOUS MEETINGS WILL BE BRIEFLY REVISITED 0 IN-VESSEL HYDROGEN PROUCTION 0 HYDR 0 GEN COMBUSTION BEHAVIOR 0 CORE DEBRIS-CONCRETE INTERACTIONS .
O IN-VESSEL FUEL / COOLANT INTERACTIONS 0 SENSITIVITY'AND UNCERTAINTY ANALYSIS ELF /Is i
i
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 1
l ELF /Is
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 c 0 GRAND GULF: T230W, T23C, AE, AND Il0VV SEQUENCES ELF /Is
J f -
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 i Rockville, Maryland l
August 28 - 29, 1984 l
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I IN-VESSEL HYDROGEN PRODUCTION e Consider bounds of outer cladding surface and . _
outer plus inner surf ace available for oxidation.
Upper bound results in about a 50% increase in H generated before core slump.
2 4
l e Upper bound (twice cladding surf ace 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.
ls e The IDCOR results are insensitive to these
" bounds for in-vessel hydrogen generation.
k -
D 0 i,
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.
l 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 Hg-s Containment analyses do not demonstrate H2 concentrations sufficient for detonation in any I compartments.
l Q ] .
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l IN-VESSEL FUEL-COOLANT INTERACTIONS (HYDROGEN GENERATION) 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.
l e Calculate the overheated regions which can be covered by the level swell.
1 e Calculate additional oxidation due to the level swell.
i a
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CORE-CONCRETE INTERACTION IN MARK I BWRs e Considered Several Different Debris .
< Configurations
- uniformly distributed within the pedestal and drywell
- distributed over the pedestal floor and one-fourth of the drywell floor restricted to the pedestal
- No Significant Changes for Time to Containment Failure or Calculated Releases to the Environment N Y .
i .
4 SENSITIVITY AND UNCERTAINTY ANALYSES e 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.
e Additional uncertainty analyses have been performed with MAAP 2.0 - integrated fission product release and deposition.
- Uncertainties considered
- - Sequence definition
- Vapor pressures of volatile fission products
- Aerosol sedimentation k ..
l I
k
IN VESSEL HYDROGEN PRODUCTION
- A COMPARISON OF NRC CONTRACTOR -'
, AND IDCOR ESTIMATIONS.
JAMES T. HAN, NRC-RES NRC/IDCOR ttETING ON INTEGPATED ANALYSIS OF FISSION PRODUCT BEHAVIOR ROCKVILLE, MARYLAND AUGUST 28-29,1984
o ,
THE PRESENTATION IS BASED ON THE FOLLOWING REFERENCES e NRC CONTRACTOR CALCULATIONS USING THE MARCH CODE:
- 1. GIESEKE, J. A., ET AL., "RADIONUCLIDE RELEASE UNDER SPECIFIC LWR ACCIDENT CONDITIONS," BMI - 2104, VOLS, 1 - 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 VEETING (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 ABOVE.
- 3. IDCOR TECHNICAL REPORT 12.1, " HYDROGEN GENERATION DURING SEVERE CORE DAMAGE SEQUENCES," (JULY 1983),
- 4. IDCOR TECHNICAL REPORT 15.1A, "IN-VESSEL CORE ltLT PROGRESSION PHENOMENA," (JULY 1983).
- IDCOR TECHNICAL REPORTS 23,1 SUPERSEDE OTHER IDCOR RESULTS SHOULD
~
ANY DIFFERENCES ExlST.
F o .
QUANTITATIVE COMPARISON OF IN-VESSEL HYDROGEN PRODUCTIONS .
CALCULATED FOR VARIOUS ACCIDENT SEQUENCES e NLM3ERS GIVEN ARE THE IN-VESSEL HYDROGEN PRODUCTIONS EQUIVALENT TO THE FRACTION OF ALL ZIRCALOY IN THE CORE OXIDIZED, o HYDROGEN PRODUCTION FOR ALL ZIRCALOY IN THE CORE OXIDIZED:
~
I ZION - 1950 LBs SEQUOYAH - 2230 LBs
-~
, GRAND GULF - 7660 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 ratio
- 3,4 1.4
- RATIO = BCL RESULT /IDCOR RESULT e PWR, ICE CONDENSER CONTAINMENT - S HF 2
PLANT SEQUOYAH BCL 0.66 (0.57 BEFORE CORE SuEP) i IDCOR 0.39 - 0.40 RATIO 1.7 3
e PWR - S 2D PLANT ZION OR SEQUOYAH BCL 0.85 (0.73 BEFORE CORE SLUMP) FOR ZION .
IDCOR 0.30 FOR SEQUOYAH
. o QUANTITATIVE CCFPARISON OF IN-VESSEL HYDROGEN. PRODUCTIONS CALCULATED FOR VARIOUS ACCIDENT SEQUENCES (CONT'D) e BhR/6, VARK III CONTAINMENT - TOUV PLANT GRAND GULF BCL 0.40 (0.18 BEFORE CORE SLUMP)
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 SLWP)
IDCOR 0.068 RATIO 9.0 e BWR/4, PARK I CONTAINMENT - IC PLANT PEACH BOTTOM BCL 0.47 (0.26 BEFORE CORE SLUMP)
IDCOR 0.047 RATIO' 10.
s
, ,, - -- , ..------.n- - - ..
o .
e IDCOR RESULTS ARE OBTAINED BY ASSUMING (1) AS A CUT-0FF TEMPERATURE OF 2300 K (OR USER INPUT) IS FEACHED, BLOCKAGE F0PFS IN THE CHANNEL AND SHUTS OFF STEAM SUPPLY AND STOPS HYDROGEN PRODUCTION, AND (2) HYDROGEN PRODUCTION DUE TO
. CORIUM - WATER INTERACTION IN THE VESSEL LO ER PLENUM IS NEGLIGIBLE, THEREFORE, LOWER HYDROGEN PRODUCTIONS ARE CALCULAED, '.
e ON THE CONTRARY, BCL RESULTS ARE OBTAINED BY ASSUMING (1)
NO BLOCKAGE FORMS AND MELTING NODES REMAIN IN THE CORE FOR 0XIDATION UNTIL COPE SLUMPS, (2) AS CORE SLlFPS, MORE STEAM IS GENERATED TO FURTHER 0XIDIZE THE REFAINING 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 ABOUNT 4
0F HYDROGEN E0UIVALENT TO AB0lfT 50% OF ALL ZIRCALOY IN THE CORE OXIDIZED, REPEATED PROCESSES OF CORE UNC0VERY - WATER i BOILOFF - CORE REFLOOD ARE ELIEVED TO PLAY AN IVFORTANT
ROLE IN HYDROGEN PRODUCTION, e
RECENT RESULTS FROM HYDROGEN RESEARCH PROGRAMS o MULTIPLE BURNS APPEAR LESS LIKELY WITH CONTINU0US RELEASE OF HYDROGEN 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 HYDR 0 GEN AT CONCENTRATIONS AB0VE 9% CAN THREATEN EQUIPMENT PERFORMANCE o TEMPERATURE HAS A SIGNIFICNAT IMPACT ON H2/ STEAM DETONATION LIMITS AND REDUCES THE EFFECTIVENESS OF STEAM TO PREVENT DETONATIONS o DETONATIONS HAVE BEEN OBSERVED IN CONCENTRATION OF 13% HYDROGEN o TRMISITION FROM DEFLAGRATION TO DETONATION CAN BE INITIATED WITH WEAK SOURCES l AND INDUCED BY SIMPLE OBSTACLES
"NEEDED HYDR 0 GEN-SAFETY RESEARCH FOR LARGE DRY FWRs" 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 TRANSITI6NFROMDEFLAGRATIONTODETONATION o ASSESSMENT OF CONSEQUENCES FROM GLOBAL OR LOCAL DETONATIONS o ASSESSMENT OF GLOBAL AND DIFFUSION FLAME BURNING ON EQUIPMENT SURVIVAL o ASSESSMENT OF THE P0TENTIAL FOR DETONATION COINCIDENTAL WITH REACTOR VESSEL FAILURE 1
i .
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\
"NEEDED HYDROGEN-SAFETY RESEARCH FOR ICE-CONDENSER PLANTS" o ASSESSMENT OF H2 / STEAM TRANSPORT AND P0TENTIAL 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 DETONATI'ON o ASSESSMENT OF THE STRUCTURAL CONSEQUENCES FROM A GLOBAL OR LOCAL DETONATION o CONFIRMATION OF THE EFFECTS OF HYDR 0 GEN BURNING ON EQUIPMENT SURVIVAL 4
o ASSESSMENT OF THE POTENTIAL AND CONSEQUENCES OF AUT0 IGNITION OF HYDR 0 GEN t
AT RELEASE / BREAK LOCATION l
l .
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- . J. . .
NEEDED HYDROGEN 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 l
l l
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1
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i C0RE-C0NCRETE INTERACTI0N IN MARK I B W R's BY 1
S. B. BURSON CONTAINMENT SYSTEMS RESEARCH BRANCH U.S. NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 i
AUGUST 28, 1984 i
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O CLWGOB,JEkTIVES 0 ,f J
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. O MODES OF CONTAINMENT FAILURE
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4 PLANT SPECIFIC DESIGN FOR MARK I PROBLEM
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8 IESULTS OF CLWG/CPWG DRYWELL FAILURE TIMES VS FAILURE MODE ,_
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8 CONCLUSIONS ,
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CLWG OBJECTIVES 3 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.
3 PROVIDE CONSENSUS VIEWS fN AREAS WHERE CALCULATIONS CAN BE PERFORMED WITH CONFIDENCE.
I O IDENTIFY MODELING UNCERTAINTIES AND VARIANCES IN METHODOLOGICAL ASSUMPTIONS THAT SIGNIFICANTLY IMPACT THE TIMING AND MODES OF CONTAINMENT FAILURE.
O
0 O 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 ~
0 OVERTEf.PERATURE FAILURE OF SEAL MATERIALS
- HIGH TEMPERATURE DEGRADATION OF POLYMER SEALS l
- LEAKAGE THROUGH SEALS PRIOR TO GROSS FAILURE !
- CONSIDERED BY CLWG/CPWG
- RESULTS REPORTED: CPWG DRAFT REPORT NUREG-1037 ;
i 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 1
L' 0 ABLATION FAILURE OF DRYWELL LINER
- HIGH TEMPERATURE CORIUM FLOWS TO STEEL LINER
- ABLATIVE ATTACK CREATES FLOW PATH INTO REACTOR
- . BUILDING l - CONSIDERED BY CLWG l - RESULTS NOT PREVIOUSLY REPORTED L
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4 COW ARIS0N OF APPR0XIMATE ORYWLL FAILURE TINES BY OVER PRE 55URE, DVER TEMPERATURE,~AND LINER MELT-THROUGH J
CLWG EPWG LINER
- CLWG DERRIS TEMPERATIME MAXIMtst MVWLL* OVERPRE55tNit DVER TEMPERATURE WLT-THROUGH i CASE CONCRETE COMPOSITION P AND T FAILURE (NIN) FAILHRE(MIN) FAILifRE(MIN) f _ _ _
l 2550 K, 145 psia 133 62 3.5 l Limestone 622K(660F) 2 1755 K. BR psia 500* 329 45 Limestone 533K(500F) 3 2550 K, 108 psla 460* No Leakage 5.5 Basal t 477K(400F) Calculated 4 1756 K, 65 psia 950' No Leakage No Basalt 411K(780F) Failure Calculated Melt-through j
Unlikely Calculated i
j
- Entrapolated value.
l
+ Mesimum during five hours of core / concrete lateraction.
I l
i
CONCLUSIONS 8 FOUR MAJOR DRYWELL FAILURE MODES
- STRUCTURAL OVERPRESSURIZATION
- OVERTEMPERATURE FAILURE OF PENETRATION SEALS
- OVERTEMPERATURE CREEP OF 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 8
ALL FOUR FAILURE MODES ARE POSSIBLE AND MUST BE CONSIDERED VARIATIONS IN CONTAINMENT MODELING ASSUMPTIONS MAY HAVE SUBSTANTAL IMPACl 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.
t b
l l
1 l
d CORRELATION FOR FISSION PRODUCT DEPOSITION BY AEROSOL SEDIMENTATION i
Robert E. Henry Fauske & Associates, Inc.
16WO70 West 83rd Street Burr Ridge, Illinois 60521 (312) 323-8750 l> NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Fission Product Behavior 4
Rockville, Maryland August 28 - 29, 1984 i
i
o . i
[ ~
FISSION PRODUCT REMOVAL MECHANISMS MODELED IN MAAP l
- Vapor Deposition (Diffusion)
- Steam Condensation e Sedimentation il n
9
VAPOR DEPOSITION DAM wt [ P. - P, h -
D" RT ( 6 /
1' 4
J I
e *
~
100 AV4 A.5 100tME
- T40uGM AREA ELtv., ORAIMto.
$YM80( LOCAftom ft ft' g
w 0 L13 (-) 4.0 3400
}30 "- 9 i - --a Los (+) 29.7 20
-e L10 (+) 19.7 20 y ; ,
- - -e L11 (+) 9.9 20 2 ly .a Liz (-) 0.2 20 2
8 f8 d
. 10 4 ---
a ;
\> ,
= a g '"g -
2 s s 3
0=
0 I0 3 "*
%5g--
/---MAAP L
Model
'f R 'TM'M"r=@-Jh. n 0 2 4 8 8 to 12 14 16 18 23 22 24
- t. Mr
AEROSOL DEPOSITION BY STEAM CONDENSATION Steam Velocity _
W, Oc s = ps A, A s A w h gg Aerosol Deposition Rate W*N a a A, U s l
l AO c Wa " p,ah gg
- ., , - ~ , - - - -
6 O
( 3 -
lo" .
. Rum A.2
- y casrun I , (Ltv.,
"dl LOCAff04 ft ia s e MAAP Model 2 Los . as.1
~
[ ------e uo . is.
I L11 . s.9 6 L12 (.) o.2
- a l 2
3 102 ,
S a 4
- A
= ;; s a y a tal fl N A E 6 l
- .. s's N
- N , N .a
^ "%'"
.,-4.( .N.
s.
- l. 6 3
to -
% ,, 'N, N
,- N , ----- * ,,--- -8 .
, i i *
,,.i i i e 3 2 4 6 8 to 12 14
- t. Mr l
l Cesium concentration in wall condensate film, Run A-2.
m l
I l
l l
I
AEROSOL MODEL SEDIMENTATION Model
@=-Am+W1-W0 dt I P - Oc}9d A,AUgE U =
s V 18gg Assumption n
h=cp Determine c and n from large scale experiments.
ADolication Compare model with the spectrum of experiments l
available, 5
f FROM THE FUNCTIONAL RELATINSHIP p
i 1/n Ma _
n Ag (T - Tg) + 1 _
T This allows a large number of experiments to be combined on a universal plot, i
l
o 500IUM OXIDE & IRON OXIDE AEROSOLS o m_
's; i 6 6 644118 i i i l i lill 6 i i l i a isi i i i l i l ii 1 . oE :
1 go -
-X g g HEDL - -
v 22 = 0 60 7
g.C C rest est C &g *
-. rest ass _
5 a _~
Q :
.: y o O
~
- ~
e c,
~
0 -
es 0 C' 0RNL VCl ~
CD y o a Rui4 tot E O A RUN 102 [
D . y RUN 103 -
O 8 9 R'JN 104 -
TJ PUN 106 ~
7 h .O4. O RuN 107 C a X RUN S L t ( (RON ) _
r 1
o i ! t ititir i i e itfin i ! t titui i i !ifin lc 10
- ZZxCT-To)+1 10 ' 10 ' 10
- 1 o o l
)
i i
o URANIUM OXIDE AEROSOLS o
% i i I lilill I I i 1 11111 I I I l i llli l i i i lig sa :
6 V -
A 22 s 0.60 O_ , N _
- = y ~
0 $ Y 3
g ? -
m 0RNL
, UO_
N- -
y :
~
E NM Mt ANNN3 [
- p MuN 304 -
- e nun as -
- 0 WM 30s _
7 A M M7 O_ v aun = _
I
- 4 -
t O I I I 111111 1 I I t illit i i i 111111 1 I I 11111 0 1 2 3 4 l 10 10 10 10 10 ZZx(T-To)+1 l
I I
4.-_a r._J i_>_ ____a--.-_.L.- . + . ._ _ _ . - _ao _a .u -a_m a l . .
1 N
I I k
. - k i
[
, l 4
- - o "E s
! (lP 1
C 2
.O o
5 7 A 9 5 U
.e N U
} - -
.o B
a 4
m.
g i
I i T.O '
N M f C
'Q 'Q 'Q 'Q !
i
- i. oes'eloW loAoWoW
!' ( .
I 1
O
- APPLICATION OF THE CORRELATION TO OTHER EXPERIMENTS AU s p d* -
A- C =C 3 P"
V hH9 Pp Assume C3a l h P" Characterizes the mean settling diameter as a function of aerosol concentration 1
h Pp A-C
~ ref \ l
( h )(PP ref I l
h,., . . .P , . , - 9, e
4 . ,
l o
lll l l 1111l1 l l l lillil l I I tillllI i i llilliI I I lillII e I i o z
- o -
@ l C -
= w ,,
u = 0 -
5 03 a
c w
5
~
E e -
. a. -
- J -
n w
= 0 I
= O
= = -
J -
\ -
v
@ W a
g a
' w W
< =
=
=
=
O*
y a : e : r n +
,E -
2 - -
n
= = 0
~
!s
~
5 5 3 O i
t
- ~
. lilliI i ! tillit i I I tilli II If!!!t i I I ff!!!!! I I ffillf I ! t O
t 0[ -
c( C o *W/ D' J ) D N O ]s 0 2 V N
( .
. +
e, mis t i i ino 6 61 i iiiin i i i iiiiii i i i iiis o i l i iiiiiil l i O
~
g ,
- o. ,o
. I w
= u '>
= O .-
= e : -
Q O a _ .
w a 8
_ c ,
_ c. _
w n a
= a
=
0
.a -
i un _
o - _
s' E" .
e u
< = m a o-5 w a,
se s -
n *
- afm
.s r -
=a -
O
~
=_
m .
5 5_
O litie t i t i tn ii t sto ttt ito!ii i t it"o e i i sin i e i i _
t.
_O! ,_Of
, ,_ O C ,_ O C .OC ,_ U C ,.. O ' -
l 1 C 6 +W, S .I 1 ?flO.g CDf 4 ~ e
~
l l
h e
~ . - - . - . . - , , - - - - - , - , . , _ - _ . _ , - . . - - . _ _ - . ,
l e.
iiiu t e i i sinis t i l i itaa t i i i O gli t i i ilialli s i it,iti s i i e 2 -
a g }
o o
w v
v e. -
c =_ ._.
_= w - w a m a
_ c. -
w "
c
= a
= m
= o
= _- -
N a - -
o -
u m a -
w o -
a r m e
w
< = = = o' 5 E u o - -
r n _ _
2 _ _
E E O
= = -
- g :
i .
- a s E O utn e ! ! ' nit i t i v in i t t i rnn i i i i .nin i i i in n i , i
(_*
t OI ,Ot
,C++H/DM) t J N Q~.g )G C V N I
i i
l
P C>
4 7 NA20 AEROSOL C
~
i i l itilli f i illeill i i i 6611ll 6 6 6 664446 i e i lielli i i ill
~
N -
s ; ~
C'
- 1e
- =
E LEGENO :
_ c, c1 ' ., a near aus is -
.v s -
=
- ~;a:.
u
=
s ;
.s ,-.
, :.cI -
- C 5
W~ M -
=_
5 l l-g
.e.8 -
m oc a _
^1 s ~ 's, :
E
" 8 d MODEL PREDICTION f
<9 \- -
c:
=
i 5-n, r -
O' i 8 i finiti i i ! riti,i ,iii,,, , , , ,,,,,, ,,,,,,,,, , , , , , ,
2 #
IC ' 10 10 ' 10 10 ' 1O
- 10 7
TIME CSEC)
Comparison between the correlation and JAERI Run 13 for lean concentrations.
l O
+ .
m,ii i i ,inii , i i ,iniiii i iiini i i i naisi i i iin i u e i = '_.
_ a
_ - 0 Q 2' u
.o _
.
a _
o
- u u : -
a e -
- a
_. a. _-
.a -
.a. ..
g
$ g :-
~.,
N
.a;
=*
5 7 m
% g *j
- u. _. m .-
= a -
= -
a.,
n -
a _
E a =
= -
=_ _-
1,... , , iiii n , , , ....i,,. , , , , , , , , , . , ,
g.O ! OC - OI -
nr I C + + W ' >. D :i O tot 103.O C?vN g
)
O
- - - - - - - - - - . - ,,,-.--e- - .,,.n..n-------,._r..- - - - - -. -
. o l I
~
(a)
DRY WATER s
t
' w SOLIO (b)
WATER ACCUMULATION
(
+ .
S
( 3 -
AVERAGE DENSITY INCREASE '
P p = P,gg + /1pPeff 3 W th )
/
Influence of Condensation A, U, 6, Ac y = p,y i
i 1
l
e +
~
i i i i s i i
,oo DRY AEROSOL e
a
- * ~
o 10" WITH WATER u
~
ACCUMULATION e
e e
e WATER ACCUMULATION PLUS CONDENSATION .
a A
10'3 - -
WATER ACCUMULATION PLUS S X CONDENSATION a 10 'j 10 20 0 100 200 00 1000 TIME, min.
Comparison with NSPP experiments for wet ( A Run 612) 0 /U 0 and dry (OThe aerosols. Rur.influence 631) atmospheres using Fe$1dn 8is ,j,,
of wall condensa illustrated.
l
- i i i i i i i 0 ~
10 DRY MROSOL "w
! u 10
- a*
, WITH WATER ACCUWULATION 10-2 . g _
E WATER ACCUMULATION PLUS CONDENSATION O
, o i
O lo-3 -
o _
Q l
l
- 4 i f f I t t l 1 10 4 10 20 40 100 200 400 1000 1 TIME, min..
- Comparison with NSPP experiments for a wet atmosphere (O Run 611 and O Run 613) using Fe 0 0 aerosols.
The influence of wall condensation i$ $ 11ustrated.
\
l
. . +
e, s w
^
w l
R e
5 10 i i i I I i i h I
% DRY AEROSOL z PLUS CONDESATION -
U z O 8 idl - O -
I d
m O- O E O O w
O 2
10
~
j - O O WATER ACCUMULATION
~
E g PLUS CONDENSATION O O h
Q l63 - -
g O CONCRETE COMPONENT M O Fe23 O COMPONENT W O O
M o
u 4 i i i i i i ; i j 10 4 10 20 40 100 200 400 1000 2000 TIME,(min) g N (TimeSinceTerminationoftheAerosolSource)
Comparison of the model with materials which do not coagglomerate.
I t
~
]
WATER ABSORBED BY HYGROSCOPIC AEROSOLS
. MW, = 2 Z [ Mw )[ R \
' M1 \ M,9 / \ 1'- R )
Pst R=
P sat Water Mass Added to the Aerosol Mass Mass i
f l
m , .- - - _ , , - . _ . - - -. . , _ _ , - - _ . -,._.,.,,,_,,,._,,.-,_,y-_.,m,,--_,,,_--__.---.--__._-.-_~---,-,,,-,_.----.m
MARVIKEN TEST COMPARISON
> Quasi-Steady State W i=Ws+Wo PQ=C g
VE~+pQ 9
P g = C, E P .6 _}_p Q
l t
\
o r
s 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) Cs! (1.7)
Relief Tank < 100 < 100 Te (1.6) e i
4 4
gj
_ _ _ - - __ ____._ -, _ p - .,, x - __-
COMPARISON WITH THE MARVIKEN EXPERIMENTS Pressurizer and Relief Tank Deposition for Test 1 9
Proportion of Mass injected
', Cs i Te
. Pressurizer 4
Model Prediction 70% 70% 70%
Experiment 34% 26% 35%
Relief Tank Model Prediction 23% 23% 23%
Experiment 23% 23% 16%
Pressurizer Deposition for Test 2b I
Proportion of Mass injection l
I' Cs i Te Model Prediction 75% 75% 75%
- Experiment 43.9% 44.2% 43.6%
k
~
i l
o .
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
,' Total Surface Area, m 2 571 571 571 2 4jg 4)g Wall Area, m 4 79 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 i Avg. Steam 0.082 0.037 0 Makeup, kg/sec
- Initial condition.
o ~
10'8,
- RUN A-2
- 1 E C O CESIUM "b10
~
i O URANIUM l
6 IODINE 2 .
i 9 -
H -
t 4
% 10-7 2 :
W :
O .
2 O -
O w 10'O r m :
4 -
I
- c. 4 m
O 1
$. 10 r .
8 h g LODINE, CESIUM 8 URANIUM A
go i e i i i O 4 8 12 16 20 24 t,hr ,
i Comparison of the predicted removal rates due to steam l condensation and sedimentation to the measured data from Run A-2.
4
O o(4(
n .OI4(w n 8 zouz % 4POz 1 Em 1 1 1 1 1 I
O~
0- 0- 0- 0- 0-i C 8 8 7 8 5 O
W __ - =
1:-
O r a-4 '
U R
A
@s N
I e OA U
8 '
M
- @ g t
h 1 O r 2 '
D I
N AOO E
O UC I
1 6 ' s D RE N. a I NN U ASI R
EI U C UM N 2 E M 0
S A I -
U 5 M
2 q 4
1
( 3 -
10-5 , .y RUN A-Il
~
f 10-6
~
> - E 4 O G
% 10-7 --
a z :
w - G O
~
A URANIUM '
8 - & CESIUM O 0 w 10 <n :
A 4 -
I ~
CL
~
(O O- l0 ?~
m O -
3 A
O 4 8 12 16 20 24 48 i t,hr N Y
UNCERTAINTY ANALYSES FOR THE SEDIMENTATION CORRELATION i
n 3
A=CP Vary n Sufficiently to Bound the Data Example n = 0.6 0.5 < n < 0.7 4
k .
- --- - _ _ _ _ _ . _ _m___ . - - - - _ . _ _-- ,x ---,-,--.--,,-,-.-w c - , - , - - - - - , . , - - , , - . - - --------,--.-.-,,e---,
o SODIUM OXIDE & IRON OXIDE AEROSOLS oa t i i i ijiiIi i i i i i illl l l l 1 l l lll l l i i i i II
- j._a 1 :
a s*
g HEDL 3
4 v 0 5o
- - O , 22 = o TEST Aat C = D .1 TEST AS5
,6 =
.E vo E O
A _-
@4<
v dg _
e m a. A O -
j'g _
p CD 0RNL
! "!! 0 03 m sun tot E
~
O A RUN 102 07 y aun soa _
3 gceg d e sun low -
i ,,
i ga 7 .
v sus ics O RUN 107 O 3 x RuN Stt (IRON) _
S-E 0
C',, I l l 1 1l111 I l l l !II i l i l l lIII I l f l l llI
~
> L A li- 10 10 ~ 10 ' 10 ZZ 3 (T-To)+1
1 I
o SODIUM OXIDE & IRON OXIDE AEROSOLS C6 #
- I i i iiilli i l 4 i i Illi a i 6 a ilisi i i i i i i ll
'. e -
0 HEDL
- . ouT%
22 = 0 70 7 ' Og , 0 TEST Aet C, &p,a A TEST ABS y _
X Ooq Z x VJ , -
~
C' $g ORNL so v4a _
j~ j a auN tot E
. O A FUN 102 [
07 Y AUN 103 -
O, Q e RUN 109 -
0 g 7 RUN 106 _
7 7"A O RUN 107 C 3 X RUN 5 L 1 ( IRON )
^
c E 1
1 C i l f 1IIIlI l l l l f l IlI l ! ! ! 1l l!l I I I I ! l Il
- , ' 3 1
li ' 10 10 - 10 10
- ZZx(T-To)+1
- w- - - - - , . . , -.-_ _ - _ __ ~ --
o URANIUM OXIDE AEROSOLS o.
l 41 I I I 1111ll l 1 I 1 11111 1 1 1 111111 1 I I t i l l2 i c I a ,v _
a v zz = o.so -
O = e 077
,_, - =
v _
A V I O 7 -
08 0RNL uO _ y _
B- E: .-m A mu ma E
- y RUN 204 -
e mm m -
_ O MN Z6 _
? h %M 307 o
_ v nua =
l C). I I I IIIIII I I I f Iil I I I I I t ill I I I I IIII
~
0 1 2 3 4 10 10 10 10 10 ZZxCT-To)+1
~ w ----_ - - - - _ _ - _ _ - - - _ - - -
O O 4
o URANIUM OXIDE AEROSOLS o,
W .,
i I I I lilli i i i l i lill i I I I lilla i i i i iig i
s) '
~
7
~
g Yv ZZ = 0 70 2 5
- y E v
5 a .o E
a V _
g 08 y 0RNL uO _ _
b" E ~
.-w A m M3 E
Z
_ y m ao4 _
, e m 20s -
_ O M 206 _
7 A M "JC7 O _ v m 2cs _
2 Z 1
O I I I i18111 1 I I I t illi i I ii lill i I I I 1111
" 2 3 4 0 1 10 10 10 10 10 ZZxCT-To3+1
'=W w Twew+ = -
e- -ee- e- spr- e- - - - - - - - - - - - -n -,w -
- , o
SUMMARY
i e MAAP calculates aerosol removal due to vapor deposition, steam condensation and sedimentation. _
l e 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 l
Hygroscopic aerosols i
1
[ .
I l
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 l
August 28 - 29, 1984
- , . - . - . _ . _ _ ,,,.-y_. - _ . _ . , _. _ _ , _, - , , . , , _ . , , _ _ . - - . _ _ _ . . , , , , . , , . , , , , _ ..-,__,.w. __ _ , ._ _ . . - . _7__ . ,. y-- - ,. . - ._ _ ..- __
a o RESUSPENSION e 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.
l e
Debris beds accumulated in either the primary system or the containment would not be sufficiently deep to be dispersed by the depressurizati~on followl'ng f ailure.
o Debris beds accumul~ated near a f ailed penetration l would not experience velocities sufficient for L entrainment.
l l
l Q ] -
l
[
REEVOLUTION OF FISSION PRODUCTS e Considered resuspension of particulate l'ayers 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 -
i also a conservatism in the current analyses.
4 .
e I
j
o .
d J
CALCULATIONAL PROCEDURES FOR REEVOLUTION OF FISSION PRODUCTS ,
by Jarnes A. Gieseke i
Preser.wd at the NRC/10COR E ETING August 28-29, 1984 4
OBattelle Columbus Laboratories
4 4 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 i
- OBalfelle Columbus Laboratories
9 PREVIOUS TECHNIQUE
- = (Continuous Time) (20 Time Intervals) ==
Fluid & Structure Core Outlet Thermal-Hydraul ic MARCH :- MERGE = TRAP-MELT
- Conditions Conditions Down-
- stream of Core 4
ified Thermal Conditions of Transport & Deposi-
- Structures Includ- ~- MERGE
' tion of Fission l=:
l ing Fission Produc '
Product $pecies and Heating Aerosols l
1 9 '. e
' NEW TECHNIQUE
= MARCH OUTPUT FILE MARCH Read Core Problem 3
Outlet Conditions Time y
Primary Coolant DRIVER MERGE & System States &
ROUTINE 5 Structure Temps h
s, f - '
Fluid & Structure Fission Product i ,
' Thennal-Hydraulic Transport and 1 MARCH Deposition lTimeStepl i
Conditions Down-
\ ,e stream of Core TRAP-MELT +
7 ission Product Distributions 4
e e,i .
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 SM0OTHER AND MORE RELIABLE I
e GENERAL CLEANUP /0RGANIZATION AND DESCRIPTIVE COMMENTS ADDED e UPGRADED TO FORTRAN 77 -
E i
OBattelle Columbus Laboratones
- ---- - - -. - , . ,- ._ --,.r , --, --- --,.- - -
,,v-,
, ._. -- --,.--.-9,-
MERGE FISSION PRODUCT HEATING MODEL -
, Structure W
e 0 0 Gas w/ airborne F.P.
\ h\
~
h ' 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
-- S ENERGY ALLOCATED BETWEEN GAS AND STRUCTURES DEPENDING UPON VOLUME DIMENSION AND 8 TRACK LENGTH e ENERGY ADDED HAS THE FORM: E = Eg *T*R*G where En= INITIAL (TIME = 0) ENERGY T = FRACTION REMAINING DUE TO DECAY
! R = FRACTION RESIDENT WITHIN VOLUME G = GE0 METRIC RELATIONSHIPS l
i O
e
> e IMPROVEMENTS TO TRAP-MELT 2 I
e USE OF GAS PROPERTIES APPROPRI ATE FOR STEAM MIXTURES
/H2 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
. OBatteile Columbus Laborstories
L 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 l of Severe Accident Fission Product Behavior Rockville, Maryland August 28 - 29, 1984 k .
t 1
MAAP FlSSION PRODUCT RELEASE l
e FPRAT - Steam oxidation model for release of fission ' products from the fuel.
l e MAAP module - FPRATP (PWR) and FPRATB -
(BWR).
e Gas flow controls the release from the core -
Minimum flow allowed is that required to remove volatile fission products.
i e Aerosols formed 'in the upper plenum.
e FPRATP compared to FPRAT for release rates and magnitudes.
! e NUREG-0772 models for fission product release are included and can be specified by the user.
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f '
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.
e REFINS - 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.
s l
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GAS FLOW j.
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TRANSPORT SEPARATORS 2 '
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e-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 Rockville, Maryland August 28 - 29, 1984 l
e ,
ZION SEQUENCES ANALYZED
- 1. Containment Failure Sequences
' ~
- c. TMLB'/ containment purge open
- d. V sequence
- 2. Sequences Without Containment Failure
- a. TMLB' with recovery l , b. SLFC with and without recovery
- c. ALFC with and without recovery I
...._ ,, -._., y - , ,---
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1 I i COLD LEG 1 7 HOT LEG A 1 4
STEAM GENERATOR Q ,
STEAM GENERATOR PRESSURIZER % i SHELL , SHELL g
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i % ) Y J 3" UNBROKEN" LOOPS l'8ROKEN" LOOP (NOOALIZATION SAME AS UN8ROKEN LOOP) i' j MAAP-WESTINGHOUSE PWR PRIMARY SYSTEM NODALIZATION i
8 '
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STEAM GENERATORS l '
IN BROKEN AND - O '
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o ,
ZION TMLB'/ Seal LOCA I
e
. I l
( 3 -
ACCIDENT
SUMMARY
Time (hrs.) Event 0 Loss of all AC, DC, AFW -
.75 Seal LOCAs 1.8 S/Gs dry 2.2 Core uncovered 3.9 RV fails 32 Containment fai!a l
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ET I 08 0 09 0 O& 0 03 0 0 -
0 AV330eS3SS01 Sd i
ZION TMLB' Csl Distribution (Fractions) at 30 hrs.
Settled Airborne I. Primary system .98 .009
-5 Pressurizer .001 3 x 10 Containment .003 8 x 10 -4 Environment 0 0 l
I l
( ) .
_ _ _m FISSION PRODUCT UNCERTAINTY ANALYSES M AAP 2.0 Y -
- -. n r - - . .. . --~~, - , , , , . , . - . - _ - , . - , - - - , - , - . - - . , --,.- . . _ , . . . - . . . , - . , - - . . . ...,-,,,.m .
o o l
ZION TMLB' inspection of results indicates potential sensitivity to:
j
- 1. Mass of OsOH and Csi vapor (assumed vapor pressure, number of groups, temperature).
- 2. Sedimentation rates (correlation parameters).
- 3. Containment hole size.
Y Y .
, e '
0 5 5 o 0 0 0 M 1 1 1 u
x x x R 1 1 1 5 5 5 a - - -
B 0 0 0 1 1 1 .
r x .
S x x
. 1 1 1 s - < <
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m H e m e r n u p ; t e o s O p n m a wi t s s m u . o o ta A C l v c S l 4
5
- O f
ZION TMLB'/ CONTAINMENT PURGE OPEN e
i k ..
P
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C 8
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9 5.; k c c' t .
, OIx I&l SG3ld IN'lWNIV.l.NO]
l t
i,
( 3 -
ZION TMLB'/ CONTAINMENT PURGE OPEN Csl Distribution (Fractions) at 9 Hours I , Settled Airborne Primary system .96 .005 1 a'10 -5 5 Pressurizer < 1 x 10 Containment .02 .001 Environment N/A .01 k
9 e
\
+
, _ - - -v,---e-,--,.----- - - - - - - - . - -,- ----- - ~ , ,, - - - - . - - . -
, - - - - - , - - ,n-- ,,-. - - - , - - , - , , , -
( ) -
ZION TMLB'/ CONTAINMENT PURGE OPEN Release fractions at 8 hrs.
Nobles .89 l Os, I .01 i
-4 Te 3 x 10 ,
6 x 10 -4 Sr,Ba
-5 Ru,Mo 6 x 10 l
i l
4 1
t I
4 e l
i l
i e.
1 ZION V SEQUENCE I
\
[ -
ACCIDENT
SUMMARY
, Time (hrs.) Event e
O .1 f t equiv. hot leg break, ,
scram, steam dumps open 5.8 RWST depleted 21.6 Core uncovered 26.3 RV failed 1
e l.'
l l
l
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T o -
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gi t (E 0 'O i z Ol x 9M 9018 XnV O.L 3SV:1 liRI cl.1 t
( 3 -
ZION V SEQUENCE Cs, I 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 i
. 4
- LO .
m f
c l
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G /. G C l I --
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I f-l
ZION V SEQUENCE Release fractions at 31 hrs.
Nobles ~ .9 ._
Os, i -4 1 x 10
-5 Te 7 x 10
-5 Sr,Ba 6 x 10 Ru,Mo -4 1.5 x 10 1
i l.
k .
- - e --
e e
4 5
6 e -
e h
ZION ALFC I 9 I
1 1 1
I t
I
s .
f .
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 i
e i
l M
........g.................i........,.........
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f l S & C E I O
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i 1
l
1 1
1 i
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w ..
n U - _
- u. -
_ m J - _
< : : w g I z _ _ -
o H
N m
i -
n 1, _
l _ _
,,,,,,,,,I,,,,,,..,' , , , , , I , , , , , , , , , ! , , , , , , , , ,- f 9 5 & C E I
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, OIx ISd SS3Bd IWINO3
t 4
[
ZION ALFC Csl Distribution (Fractions) at 8 Hours Deposited Airborne ,
-5 Primary System .84 < 1 x 10
-5 Pressurizer .004 = 1 x 10
-4 Containment .16 4 x 10 l
1 I
J n
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b a
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T/H UNCERTAINTY ANALYSIS:
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).
9 Category 2: Parameter change reveals a fundamental weakness in a model by giving erroneous i results, (e.g. core slump model).
b
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o Toble 5.1 Figure of Merit Summars for HAAP Uncertaintv/Sesisitivity Anaissis PWR VERSION 1.2 ZION SEQUENCE TMLB' (W)
MODEL FIGURE OF MFRIT CASE PARAMETER VALUE 1 2 3 4 5 6 7 8 9 10 . 11 12 13 (hr) (br) (Ibe) ' (F) (Psi /hr) (hr) (hr) (Iba) (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 0.00 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 3.95 6.20 365.5 11 FHT 0.5 22.33 18.59 199870. 469.0 0.050 3.28 N/A 2.03 113.0 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 TZOOFF 4040 22.12 18.42 200910. 468.0 0.003 3.52 N/A 2.03 142.0 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.?0 434.1 Default Model Parameters Changed: Value Figure of Merit Homenclature:
FRCOEF Friction coefficient for corium in VFAIL........ 0.005 1 Time of cur tainment f ailure FCRSLU Fraction of total core mass which must melt 2 Time between reactor vessel failure asd to reach sueeort plate............................ 0.5 contairment feilure TTCSP Time to fail sueport elate after corium eile 3 Integrated wall ces.desisatiori (measure of has reached it (hr)............................ 0.0333 diffusioehoresis) between reactos vessel FHT Fraction of ma :imum heat transfer permitted f ailure and containmerit failure between clad and sas stream........................ 1.0 4 feak containees.t outer uall surface TZOOFF Zircalov oxidation cut-of f and channel blocking temperature temperature (F)................................... 3680 5 Fractino of cled seacted in-vessel XCNREF Corium reference thermal boundars laser 6 Rate of change of containment eressure Just thickness (ft)................................... 0.320 prior to the time of contair.mer t f ailure FCHF Flat plate CHF critical velocitu coefficient..... 0.14 7 Time of ice deeletion (if crelii;ble) 8 Tinie of cure uricoverv 9 H9.trosen der.creted at time of vessel failure 10 Time at whi.i .124 temeer: Lure reache; 7000 F 11 Time of venel f;iluie 12 Time of cose .ncit ..meletion.
13 Hu.iroder. mar.s , t t iue of cont ..inment f ailure
. . i .
,, 1, ..'. . .--
Table 5.2 Fisure of Merit Summarv for MAAP Uncertaintv/Sensitivits Analysis PWR VERSION 1.2 ZION SEQUENCE THLB' (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) (Ibm) (hr) (hr) (hr) (lba)
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 ZTMLB-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 ZIMLB-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 Chansed: Value Fisure of Merit Nomenclature 1
____ .=. _____---------- _-
FCRSLU Fraction of total core mass which must melt 1 Time of containment failure to reach suerort plate............................ 0.5 2 Time between reactor vessel failure and-TTENTR Entrainment effective enPtMins time (hr)....... 1.39E-4 containment failure FENTR Multieller for Kutateladze criterion for cavits 3 Integrated wall condensation (measure of blowout (GT 1.0 = difficultiLT 1.0 = easier)..... 0.33 diffusioehoresis) between reactor vessel NVP Number of Penetrations railed in lower head.......... I failure and containment failure SCALH Scaling f actor for heat transfer coefficients to 4 Peat containment outer wall surface eassive heat sinks................................ 1.0 temperature 5 Fraction or clad reacted in-vessel 6 Rate of change of containment pressure Just Prior to the time of containment f ailure 7 Time of ice depletion (if arelicable) 8 Time of care uncovers 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 Hwdrosen mass at time of containment failure 4488 This case was not run to containment failure. Bv extrapolating the containment Pressures it is estimated that the failure criterion would have been exceeded at 44.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
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o Table 5.3 Fisure of Merit Summarv for MAAP Uncertaintv/Sensitivits Analvsis PWR VERSION 1.2 ZIDN SEQUENCE S2HF (W)
MODEL FIGURE OF MERIT CASE PARAMETER VALUE 1 2 3 4 5 6 7 8 9 10 ~11 12 13 (hr) (hr) (lba) (F) (esi/hr) (hr) (hr) (lba) (hr) (hr) (hr) (lba) 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. 6/. 288580. 347.0 0.153 6.60 N/A 3.78 200.0 4.22 5.38 s2.66 327.0 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 IfCSP 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 3tl.0 4.22 7.50 10.74 643.0 Default l Model Parameters Chansed: Value Fisure of Merit Nomenclature:
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 reactor vessel failure and to reach support elate............................ 0.5 coritainment failure
, 11 CSP Time to fail suerort elate efter corium eile 3 Intesrated wall condensation (measure of 1 has reached it (hr)............................. 0.0333 dif fusioehoresis) between reactor vessel 1 FCHF Flat elate CHF critical velocitu coef ficient. .... 0.14 failure and containment failure j 4 Peak containment outer wall surface temperature 5 Frection of clad reacted in-vessel
, 6 Rate of chense of containment pressure Just ersor to the time of containment failure 7 Time of ice deeletion (if aeelicable)
- CategOFy 2 8 Time of core uncoverv 9 Hvdrosen sereerated at time of vessel failure 10 Time at which clad temperature reaches 2000 F 11 Time of vessel failure i
12 Time of core melt comeletion 13 Hvdrogen mass at time of containment failure 4 8 0o
Table 5.4 Fisure of Herit Summars for MAAP Uncertaintv/Sensitivits Analvsis PWR VERSION 1.2 ZION SEQUENCE S2HF (FAI)
MODEL FIGURE 0F HERIT CASE PARAMETER VALUE 1 2 3 4 5 6 7 8 9 10 11 12 13 (hr) (hr) (Iba) (F) (esi/hr) (hr) (br) (Ibe) (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 7S2HF-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 6S5.0 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 emetving time (br)....... 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 containmen; failure NVP Number of Penetrations failed in lower head......... 1 3 ' Integrated vall condensation (measure 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 temeerature l
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 aFellcable) 8 Time of core uncovers 9 Hvdrosen 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 Hvdrosen mass at time of containment failure
- Category 1 S
0 'e e
i RADIONUCLIDE RELEASE PWR -- LARGE, DRY CONTAINMENT (ZION) _
BMI-2104 RESULTS by James A. Gieseke Presented at the 1
-NRC/IDCOR MEETING August 28-29, 1984 OBattelle Columbus Laboratones
. =
f
( .
. 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
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0Ballelle Columbus Laborarones
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3
- ACCIDENT 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 3
Vessel Head Dry 179.9
, Head Fail 187.7 Concrete Attack
- 187.7 End Calculation 788.2
- Assuming debris uncoolable.
4
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OBallelle Columbus Laborator.cs l
l DISTRIBUTION SPECIES AFTER ACCIDENT, S 0-c 2
l t
.I Fraction of Core Inventory Species RCS Containment Environment
]
t 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 I 1
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OBallelle Columbus Laborarones i
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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 MUCH 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 POTENTIALLY RESULT IN A LARGE RELEASE OF FISSION PRODUCTS OBallelle Columbus taboratorie$
s 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 1-l i OBallelle Columbus taboratories t
DISTRIBUTION OF SPECIES AFTER ACCIDENT, TMLB'-c Fraction of Core Inventory Species RCS Containment Environment 4
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.
OBallelle Columbus Laboratories
,A+=n - 1,A---- -- A - r5 h-s - A +k6------- 4-+ J-4,---=- - - - * + - - A xm-amaEA 4 ,-ama-- A a- -A *Mb- AA-AL
'l 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 8.7 x 10 ~7 8.6 x 10 -7 7 9 8.5 x 10-7 4 1.3 x 10-6 2.2 x 10 2.5 x 10 1.6 x 10 -6 7 7 1.6 x 10-6 2.7 x 10-6 1.9 x 10-6 4.4 x 10 1.0 x 10-8 10 1.9 x 10-6 j .9 x 10 -6 2.7 x 10-5 1.8 x 10 -5 5.2 x 10-7 9.8 x 10-7 6
l 15 2.0 x 10 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 10-0 7.8 x 10-5 3.0 x 10-5 5.8 x 10-7 1.7 x 10-6 2.1 x 10-6 6
50 2.1 x 10-6 8.4 x 10-5 3.1 x 10-5 5.9 x 10-7 1.8 x 10
- 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
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- 4 OBattelle Colun*>us Laboratones I
i i
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TOTAL LEAK AREA ESTIMATED AS A FUNCTION OF CONTAINMENT PRESSURE ContainAent Low Medium High Pressure Leakgrea LeakAgea Leakgrea (Psig) (in. ) (in. ) (in. )
Nonnal 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 134l 0.1 5.33 23.72 g."
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1 OBattelle Columbus Laboratories
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s DISTRIBUTION OF FISSION PRODUCTS FOR VARIOUS FAILURES MODES, TMLB' SEQUENCE RCS Contairment Environment _
Design Leak Cs! 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 Cs! 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 Cs! 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 l Isolation Failure r
- 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 1
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OBattelle Columbus Laboratones
TASK 23.1 RESULTS PEACH BOTTOM Jeff R. Gabor Fauske & Associates, Inc.
16WO70 West 83rd Street Burr Ridge, Illinois 60521 (312) 323-8750 m
NRC/IDCOR Meeting on Integrated Analysis 4
of Severe Accident Fission Product Behavior 4 Rockville, Maryland August 28 - 29,1984
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- -- __b. + _ . - - . . , , . . . , , - . ._ _ _ , _ . . . . - - . _ _ , - v-- - -
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 on (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 CR0 flow ceases 12 hr Drywell failure 12 hr Begin release to environment Ns_ _s/
PEACH BOTTOM - Sj E EVENT
SUMMARY
Time Event ._
2 O 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 CR0 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 w
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H GAS FLOW _L .L A#e MATERIAL ~% -- --
TRANSPORT SEPARATORS -
& 6 ORYERS OOWNCOMER A;, 'C CORE 'O 1
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u BWR Primary System Circulation
METAL DECK
'NNN \ \~/////
METAL "
SIDWG '
REFUELING FLOOR _
REACTOR $ e i BUILDING M*NE')
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/ s REACTOR .
Q ( , i s BUILDING L (...J L- l l ,
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q -DRYWELL EQUIPMENT b a-
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- e STANDBY GAS TREATMENT SYSTEM Peach Bottom
. e Fire dampers close at temperature = 165'F. ,.
o Fans have automatic trip on low flow.
e High aerosol loading on filters.
e Due to above restrictions and detailed sequence analyses; SGTS assumed not operational.
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- DISTRIBUTI0t4 0F Csl Ifl PLAT 4T AteD EtiVIR0t.MEili (FRACTI0ti 0F CORE ItiVEt4 TORY)
At Vessel Failure TC SEj TQVW RPV .50 .78 .99 Orywell 0 .03 0
~
Suppression Pool .50 .19 .01 ;
Secondary Containment 1.2 x 10'4 0 0 Environment 3.7 x 10 0 0 At Containment Failure TC SEj TQVW RPV 1.0 .48 .67 Orywell 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.)l RPV 0 0 .07 Drywell 0 0 0 Suppression Pool .66 .49 .06 l
j t Secondary Containment .31 .50 .80
- Environment .03 .009 .07 l- .
i k--
SUMMARY
OFFISSIONPRODUCTRELEASEFRACTIONS(a)
Sequence WASH-1400 -
F.P. Group TC(b) SlE TQVW BWR2 IC) 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 10'# 7 x 10-5 2 x 10'# 0.03 0.02 (a) Fraction of core inventory released to the environment.
(b) Wetwell vent area - 1.98 f t2 ,
(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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PEACH BOTTOM TW SEQUENCE o WASH-1400 assumed core injection systems f ailed at containment f ailure.
e Recent IDCOR Task 23.1 report analyzed TW with a .
- small containment failure area with same assumption -
on pump f ailure.
e Recent GE qualification tests indicate that pumps will operate in a steam environment.
e IDCOR case would indicate that with a small containment failure area the pumps would not fall 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.
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PEACH BOTTOM - TW EVENT
SUMMARY
Containment Break Area = 1 f t 2 Time Event ,'
O 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 To hr Start of core melt 37 hr Vessel failure
[ -
PEACH BOTTOM - TW Containment Failure Size Uncertainty /Senaitivity 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.
s
f RESULTS FROM BWR-MAAP THERMAL-HYDRAULIC UNCERTAINTY / SENSITIVITY ANALYSIS
- From a large set of single parameter variations the _
following was demonstrated.
Peach Bottom - S 3E
! e There is sensitivity to the core blockage phenomena on core recovery with limited CRD flow.
Grand Gulf - T jQUV 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.
o The absence of core blockage allows increased Zr
- oxidation and increased in-vessel hydrogen k .
Table 2.1 MAAP-BWR Uncertaints/Sensitivits Analvsis Parameter Set . .
5' /
g P5ra.etert Default Miniaun Minimon Husber Name Parameter Deserietion ,
Units Olue Value Va1ue 1 FRCOEF Friction coefficient for corium in VFAIL / 5.000E-03 1.000E- 03 1.000E-02 Fraction of total core mass which must selt 2 FMAXCP , ,
to fail the core Plate 4 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 boiling heat transfer coefficient Elbr/ft**2/F 52.8 17.6 70.4
. 5 FBLOCK - Fuel channel blockage model switch ,
. 0=use blockase modeli1= turn blockase model off 0.000E400 0.000Ef00< 1.00 6 TZOOFF 0xidation cut-off temperature . F , 3680 3310 4040 7 FACPF Fraction of area of core Plate failed 0.300 1.000E-02 1.00 8 CDBPD Flame buovancy dras coefficient in the redestal 4 . 5 00 0.500 10.0 9 CDPDW Flame buovanes dras coefficientein the druwell , 5.00 0.500 10.0 5.00 10.0 l 10 CD3'IJ Flame buovar.cv dras.cce!ficient in the wetwell 0.500 '
l 11 CDBCA Flame buovancs dras coeffsetent in coseartzept A 5 00 0.500 10.0 12 CDBCB Flame buovancs dras coefficient i':s coneartment B 5.00 0.500 10.0 1 13 XCHKEF Corium reference thermal boundary laver thickness feet 0.328 0.320 3.28 14 HTCMCR Corium-crust heat transfer coefficient in DECOMP B/hr/fts*2/F 176 88. 880.
15 XCMX Miniaun corium thickness on drvwell floor aird eedestal floor (Mark II onlv) feet 0.161 1.64E-02 0.328 16 XCCMSP Particle st:e (diameter) for corius as it falls 0.328 into sueeression ecol fMarkII onlv) , feet 3.281E-02 3.281E-02 17 TCFLAd Critical flase tenPerdture F 1310 1160 1700 18 FCHTUR Churn-turbulent critical flow earameter 1.53 1.00 5.00 19 FDFOP Droelet critieri flow Parameter 3.70 3.00 5.00
' ?G FFLOOD Floodins flow Pare:eter 3.00 2.00 4.00 21 FSPAR -Parameter for bottow-sparsed steam void fraction 1.00 1.00 4.00
~
2? FUOL farameter for volume source void f raction model 2.00 1.00 4.00 23 IfENTR Entrainment effective ehPtvins time hour 1.39E-04 2.78E-05 2.78E-05 24 EU Emissivity of water 0.900 0.800 1.00 25 EWL Emissivity of wall 0.850 0.700 1.00 26 ECM Emissivity of corium 0.850 0.700 1.00 27 EG Entssivits of sas 0.600 0.500 1.00 28 EED Emissivity of eauiPaent 0.850 0.700 1.00 29 F0VER Fraction of core seras flow allowed to bveass core 0.500 0.000Et00 1.00 30 NPF Number of eenetration failed in lower head 1.00 1.00 10.0 31 FCDCDU Downconer Perimeter eer meter from eedestal door (Mark II ontv) 2.00 1.00 5.00 32 FCHF Coefficient for CHF correlation in PLSTM 0.140 0.120 0.300 33 FCDPRK Discharse coefficient for pipe break 0.750 0.100 1.00 34 FENTR Multiplier for Kutatelad:e criteriori for cavity blowout (GT 1.0 = dif ficultiLT 1.0 = easier) 0.330 0.200 100.
35 SCALU Scalins f actor for all burnins velocittes 1.00 1.00 100.
36 SCALH Scalins factor for heat transfer coefficients to Passive heat sinks 1.00 0.500 10.0 37 FUMIN Minimum burn velocity n/s. 1.00 0.500 10.0 38 ACVENT Containment failure vent area ft**2 0.1 5.0 0.01
Table 4.1 Figure of Merit Summars for MAAP Uncertaintv/Sensitivits Analvsis BUR VERSION 1.2 PEACH BOTTOM SEQUENCE SIE (PECO)
M01(L FIGURE OF MERIT CASE PARAP.ETER VALUE 1 2 3 4 5 6 7 8 9 to 11 12 13 (hr) (hr) (Iba) (F) (Psi /hr) (br) (hr) (lba) (hr) (hr) (hr) (lba)
SIE6 None N/A 30.71 7.04 33779. 1200.6 0.061 1.27 N/A 17.85 672.0 1.84 23.67 42.31 672.0 SIE1 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 5103 FCDBRK 0.6 21.29 17.74 107231. 1200.6 0.014 0.44 N/A 1.26 205.0 1.94 4.16 23.07 205.0 51E4 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 Chansed! Value Figure of Nerit Nomenclature 1 FCHF Coefficient for CHF correlation in PLSTM......... 0.14 1 Time of containment failure FC0BRK Discharse coefficient for Pipe break............. 0.75 2 Time between reactor vessel failure and ACVENT Containment failure vent area (ft*42)............. 0.1 containment failure 3 Integrated uall condensation (measure of diffusiornoresis) between reactor vessel failure and containment failure 4 Peak containment outer uali surface temperature 5 Fraction of clad reacted in vessel 6 Rate of change of enntainment pressure Just prior to the time of containment failure 7 Time of ice deeletion (if applicable) 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 comeletion 13 Hwdrogen mass at time of containment failure 4
. * . . e
~
e Table 4.2 Fisure of Merit Summarv for MAAP Uncertaintv/Sensitivitw Analvsis BWR VERSION 1.2 PEACH BOTTOM SEQUENCE TGVW (PECO)
, MODEL FIGURE OF MERIT CASE PARAhETER VALUE 1 2 3 4 5 6 7 8 9 10 11 12 13 (hr) (hr) (Iba) (F) (Psi /hr) (hr) (hr) (Iba) (hr) (br) (hr) (lba)
TGVUO 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 T0Vul FRCOEF 0.001 16.49 4.13 28570. 1200.6 0.061 2.86 N/A 8.40 690.0 9.87 12.36 >20.00 990.0 TGVU2 FMAXCP 0.4 16.63 3.30 22110. 1200.6 0.061 2.14 N/A 6.40 750.0 9.87 13.33 >20.00 070.0 TOVu3 FBLOCK 1 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 10Vu4 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 TuVuS XCNREF 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 10VuS HTChCR 80 16.59 4.23 29710. 1200.6 0.062 3.08 N/A 8.40 690.0 9.07 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 0.062 2.31 N/A 8.40 690.0 9.87 12.36 >20.00 960.0 TOVu8 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 TGVu? HPF 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 TGVull 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 8.27 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
- - _ =- _____ .-
FRCOEF Friction coefficient for corium in VFAIL........ 0.005 1 Time of containment failure FdAXCP Fractior of total core mass which must melt 2 Time between reactor vessel failure and to f ail the co re Pl ate. . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.2 contairinent f ailure FBLOCK- Fuel channel blockage model switch 3 Integrated wall condensatio.i (measure of 0=use blockage modeli1=turri blockase model of f....... O diffusioPhoresis) between reactor vessel TZOOFF 0xidation cut-of f tempe rature (F) . . . . . . . . . . . . . . . . . 3680 failure and containment failure XCHREF Cortum reference thermal boundarv laver 4 Peak containment outer uall surface thickness (ft).................................. 0.328 temperature HTCnCR Cortum-crust heat transfer coefficient 5 Fraction of clad reacted in-vessel (Btu /hr-ft**2-F)................................... 176 6 Rate of chande of containment Pressure Just TTENTR Entrainment effective emetving time (hr)....... 1.39E-4 Prior to the time of containment failure ECM L a issivi ts of co rium. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.85 7 Time of ice depletion (if applicable) 4 HPF Humber of penetrations failed in lower head......... 1 8 Time of core uncovers FEHTR Hultiplier for Kutateladze criterion for cavits 9 Hvdrosen generated 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 coef ficients to 11 Time of vessel failure 4 passive heat sinks................................ 1.0 12 Time of core melt completiors 13 Hvdrogen mass at time of containment failure
. ...t .
?
Table 4.3 Fisure of Merit Summarv for MAAP Uncertaints/Sensitivits Analvsis i
BWR VERSIDH 1.2 GRAND GULF SEQUENCE T100V (TEC)
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) (Iba)
T100VO 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 0.64 417.2 1.62 3.74 8.94 3190.0 T100V2 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 i
T10UV3 FBLOCK 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 TIGUV5- All CDs 0.5 44.63 40.89 171800. 194.6 0.047 1.57 N/A 0.63 417.2 1.62 3.74 8.94 3229.0 T100V6 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 1103VB 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 I 110UV10 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 Chansed:
Value Fisure of Merit Nomenclature:
FRC0EF Friction coefficient for corium in VFAIL........ 0.005 1 Time of containment failure FhAXCP Fraction of total core mass which must selt 2 Time between reactor vessel failure and to fail the core Plate............................ 0.2 containment failure FBLOCK Fuel channel blockase model switch 3 Integrated wall condensation (measo e of
- 0
- use blockase modeli1= turn blockage model off....... O diffusiophoresis) between reactor vessel TZOOFF 0xidation cut-off tenPerature (F)................. 3680 failure and containment failure
- CDbFD Flame buovancs dras coefficient in the Pedestal... 5.0 4 Peak containment outer wall surface Cl4Du Flame buovancs dras coefficient in the drvuell.... 5.0 temperature CubJu Flame buovancs dras coefficient in the wetwell.... 5.0 5 Fraction of clad reacted in-vessel UtilfA Flame buosancs dras coefficient in compartment A.. 5.0 6 Rate of chanse of containment Pressure Just Cbl.Cb Flame buosancs dras coefficient in compartment B.. 5.0 prior to the time of containment failure XCNEEF Corium reference thermal boundarv laver 7 Time of ice depletion (if applicable) thickness (ft).................................. 0.328 8 Time of core uncovers '
HTCMCR Corium-crust heat transfer coefficient 9 H9drogen generated at time of vessel failure (Btu /hr-ft**2-F)................................... 174 10 Time at which clad tenPerature reaches 2000 F ECM Emissivits of co r1um. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.85 11 Time of vessel failure NFF Humber of Penetrations failed in lower head......... 1 12 Time of core melt completion 13 Hsdrosen mass at taae of containment failure 1
. ..,i .
e Table 4.4 Fisure of Merit Summarv for HAAP Uncertaints/Sensitivits Analysis BWR VERSION 1.2 GRAND GULF SEQUENCE T100V (FAI)
NODEL FIGURE OF MERIT CASE PARAMETER VALUE 1 2 3 4 5 6 7 8 9 10 11 12 13 (hr) (hr) (lba) (F) (esi/hr) (hr) (hr) (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 T100V12 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 3220.0 T100V14 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 1.44 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 1.0 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 Changed: Value Figure of Merit Nomenclature:
FENTR Multiplier for Kutatelad:e criterion for cavits 1 Time of containment failure blowout (GT 1.0 m difficultiLT 1.0 = easier)...... 0.33 2 Time between reactor vessel failure and SCALU Scalins factor for all burntns velocities......... 1.0 containment failure LCALH Scalins factor for heat transfer coefficients to 3 Integrated wall condensation (measure of passive heat sinks................................ 1.0 diffusiophoresis) between reactor vessel FUMIN Hinimum burn velocits (reouired to be n/s)......... 1.0 failure and containment failure ACVENT Containment failure vent area (ft**2)............. 0.1 4 Peak containment outer wall surface temperature 5 Fraction of clad reacted in-vessel 6 Rate of change of containment Pressure dust erior to the time of containment failure 7 Time of ice depletion (if applicable) 8 Time of core uncovers 9 Hvdrosen generated at time of vessel failure 10 Time at which clad temperature reaciies 2000 F 11 Time of vessel failure 12 Time of core melt completion 13 Hvdrogen mass at time of containment failure s
. .,i .
Table 4.5 Figure of Merit Summarv for MAAP Uncertaintv/Sensitivits Analvsis BWR VERSION 1.2 GRAND GULF SEQUENCE AE (TEC)
MODEL FIGURE OF MERIT CASE PARANETER VALUE 1 2 3 4 5 6- 7 8 9 10 11 12 13 (hr) (hr) (Iba) (F) (Psi /hr) (hr) (br) (lba) (hr) (hr) (hr) (1ba)
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 6El 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 Discharge coefficient 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 Hydrogen 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 Hydrosen mass at time of containment failure O
b t o ,i ,
i # e 1
4
.I RESULTS FROM BWR-MAAP FISSION PRODUCT UNCERTAINTY /SENSITI'VITY ANALYSIS PEACH BOTTOM -
t .
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PEACH BOTTOM - STATION BLACKOUT Uncertainty / Sensitivity Analysis Base Case -
Case' 1 Drywell structure mass increased to 6
2 x 10 kg.
I Case 2 Csl vapor pressure.
Case 3 NbREG 0772 fission product releases.
Case 4 No core blockage, Temp. ZR oxidation cut-off = 3100 K.
Case 5 Sandia CsOH vapor pressure.
=t
-Case 6 1/4.of drywell floor available for corium.
. Case 7 Drywell f ailure temp =600'F
o PEACil BOTT0li - STATION BLACK 0UT Uncertainty / Sensitivity Analysis DW Csl 0772 No Sandla 1/4 DW DW Base Vapor Structure Release Blk. Cs0il Floor Falls Case (1) Pressure (3) (4) (5) (6) at (2) 600*F Containnent failure, hr* 18 29 18 18 18 18 18 14,5 Cs a I release fraction .07 .07 .07 .07 .07 .07 .06 ,oS at 60 hrs.
Cs a 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 (OF)
Drywell pressure at con- 114 128 121 120 90 114 122 100 tainment failure * (psla)
- Fraction of clad reacted .08 .08 .08 .08 .18 .08 .08 .08
, in-vessel "Contalanent failure due to drywell gas temperature > 1200 0F.
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PEACH BOTTOM - STATION BLACKOUT l 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.
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Uncertainty / Sensitivity Analysis Base Case 1 Case 2 -
a, Containment failure (hr) 18* 18.5+ 20*
Cs & I release fraction .07 .05 .05 at 60 hrs.
Cs & 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.
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PEACH BOTTOM - STATION BLACKOUT AEROSOL MODEL Uncertainty / Sensitivity Analysis Gravitatica 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 r- , . . _ _ , . - , -
& P
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l RADIONUCLIDE RELEASE BWR -- MARK I DESIGN (PEACH BOTTOM) -
BMI-2104 RESULTS by James A. Gieseke Presented at the NRC/IDCOR MEETING August 28-29, 1984 OBattelle
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AE SEQUENCE e LARGE BREAK LOSS-0F-COOLANT ACCIDENT (LOCA), FAILURE OF EMERGENCY CORE COOLING SYSTEM T.
o BREAK OCCURS IN A RECIRCULATION LINE e SUPPRESSION POOL REMAINS SUBC00 LED THROUGHOUT THE ACCIDENT e CONTAINMENT IS ASSUMED TO FAIL BY OVERPRESSURIZATION FROM NONCONDENSIBLE GASES PRODUCED BY STEAM-CLADDING REACTIONS AND CORE-CONCRETE INTERACTIONS
- 1 9
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OBattelle Columbus Laboratorres i
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ACCIDENT EVENT TIMES Event Time, minutes ',
Peach Bottom AEY Core Uncover 1.5 Suppression Pool Cooling On 10.0 Start 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 i
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OBallelle Columbus Laboratones l
DECONTAMINATION FACTORS CALCULATED AS A FUNCTION OF PARTICLE SIZE AND.0F 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 10 5
1504 5
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 33.3 1.3 5.4 69 10 10 4,)
(a) A decontamination factor larger,than 105is assumed to be 105 ,
Pool depth: 4 ft Bubble diameter: 0.75 cm Aspect ratio: 1:3 l
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DISTRIBUTION OF SPECIES AT 20 HOURS AFTER ACCIDENT, AE Fraction of Core Inventory .
Species RCS Pool Drywell Wetwell Environment Cs1 0.19 0.35 0.12 0 0.34 Cs0H 0.19 0.34 0.14 0 0.33 Te 2.9 x 10-2 3.2 x 10-3 0.32 0 0.65
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OBattelle c , _ e., <. e. ,. ..,
FRACTION OF CORE INVENTORY RELEASED TO THE ATMOSPHERE FOR GROUPS OF 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 1 0.19 0.19 3.6 x 10 -2 1.2 x 10 -2 2.7 x 10 9.9 x 10-5
-4 2 0.25 0.24 6.6 x 10 -2 1.3 x 10 -2 3.6 x 10 -3 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 15 0.34 0.33 0.65 0.68 4.6 x 10 -3 0.49 h
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OBattelle Columbus Laboratones
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TC SEQUENCE e TRANSIENT, FAILURE OF CONTROL ROD INSERTION (FAILURE TO SCRAM) e EMERGENCY CORE COOLING SYSTEMS OPERATE e CONTAINMENT FAILURE RESULTS FROM THE IMBALANCE IN HEAT GENERATION AND HEAT REMOVAL DUE TO THE CONTINUED HIGH POWER LEVEL OF THE REACTOR 4
OBallelle Columbus Laboratones
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ACCIDENT EVENT TIMES I ME II E I I I I I 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 1' s
}
i OBallelle
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Columbus Laboratones t
DECONTAMINATION FACTORS CALCULATED AS A FUNCTION OF PARTICLE SIZE AND OF TIME FOR TC SEQUENCE
~
Time Particle Diameter, t.m DF Based on (min) 0.1 0.7 1.2 5.1 8.4 Total Mass 5 5 96.2 1.3 1.08 x 103 105 (a) 10 10 3690 5 5 5 99.2 5.2 98 10 10 10 2850 5 5 5 104 3.0 45 10 10 10 2166 3 5 5 121.7 1.1 15.8 1.87 x 10 10 10 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 10 5is assumed to be 105, Pool depth: 6.5 ft (198 cm)
Bubble diameter: 0.75 cm Aspect ratio: 1 :3 l
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OBattelle Columbus Laboratories
. 2.. : . :..
DISTRIBUTION OF SPECIES AT 20 HOURS AFTER ACCIDENT, TC-Y Fraction of Core Inventory Reactor Species RCS Pool Drywell Wetwell Bldg SGTS Environment Cs! 0.06 0.69
-2 1.5 x 10 - 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
- This includes a fraction of 0.13 for Te which is found not to be released from the core-concrete interaction.
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- o DISTRIBUTION OF SPECIES AT 20 HOURS AFTER ACCIDENT, TC-y' 1
Fraction of Core Inventory -
Species RCS Pool Drywell Wetwell Environment i
t Cs! 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.
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OBattelle Columbui taboratones
e TW SE0VENCE-s i e TRANSIENT, LOSS OF DECAY HEAT REMOVAL e EMERGENCY CORE COOLING SYSTEMS OPERATE o CONTAINMENT FAILURE BY OVERPRESSURIZATION PRECEDES CORE MELTING e IT IS ASSUMED THAT OPERATORS WILL DEPRESSURIZE THE PRIMARY COOLANT SYSTEM BEFORE CORE MELTING OCCURS i'
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ACCIDENT EVENT TIMES Event Time, minutes Peach Bottom TWY _
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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 1
OBattelle Columbus Laboratories
DISTRIBUTION FACTORS CALCULATED AS A FUNCTION OF PARTICLE SIZE AND CF TIME FOR TW SEQUENCE
- Time Particle Diameter, un DF Based on I (min) 0.2 05 1.0 4 10 Total Mass 4 5 2756 1.9 18 1.17 x 10 105 (a) 10 257 5
2777 1.4 -9.4 2.51 x 10 3 10 5
10 576 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 2815 1.1 11.7 3.5 x 10 3 10 5
10 5
352 3 5 5 2818 10.3 60 9.9 x 10 10 10 326 3 4 5 5 2820 2.3 x 10 3 7.2 x 10 1.7 x 10 110 10 1336 5 5 5 5 5 5 2827 10 10 10 10 10 10 5
(a) A decontamination factor larger than 105is assumed to be 10 ,
Pool depth: 6.5 ft (198 cm)
Bubble diameter: 0.75 cm Aspect ratio: 1:3 OBattelle Columbus taboratones
a .
DISTRIBUTION OF SPECIES AT 60 HOURS AFTER ACCIDENT, TWy' k v
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Fraction of Core Inventory
- Species RCS Pool Drywell Wetwell Environcent f
CsI 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.
i i
OBattelle Columbus taboratories
J 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 l
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- S,EOUOIAH SEQUENCES ANALYZED
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- a. S2HF ~.-
(1) Drains open/ cavity wet (2) Drains closed / cavity dry
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SEQUOYAH S2HF/ DRAINS OPEN Csl Distribution (Fractions) at 20 Hours Deposited Airborne -
-5 Primary System .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
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SEQUOYAH S2HF/ DRAINS OPEN/ PURGE OPEN I.
9
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i SEQUOYAH S2HF/ PURGE OPEN Fission Product Distribution (Fractions) at 12 Hours f
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 l, Environment .02 .005
- Csi, OsOH lumped, JANAF v.p. for CsOH. I L
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SEQUOYAH S2HF
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, SEQUOYAH TMLB' Csl Distribution (Fractions) at 40 Hours
. > Deposited Airborne
~
Primary System
. .92 .05 P
-3 -4 Pressurizer ,
2 x 10 2 x 10 Containment .02 .001 Environment .001 i .
i l
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e SEQUOYAH TMLB' RELEASE FRACTIONS AT 40 HOURS Csl CsOH Te, Sb Sr,Ba Ru,Mo
.001 -5 -5 -5
- 1. CsOH, Csl not .001 4 x 10 < 1 x 10 < 1 x 10 lumped; Sandia CsOH v.p.
- 2. CsOH,Csl N/A .002 5 x 10- < 1 x 10 -5 < 1 x 10
~0 lumped; JANAF CsOH v.p.
4 *
( 3 -
O.
SEQUOYAH T 23 4
i e
4
[
ACCIDENT
SUMMARY
Time (hrs.) Event
. O Loss of feed, loss of injection, scram ;
.9 S/Gs dry 5
1.5 Sprays in recirc mode 1.6 Core uncovered 2.9 Reactor vessel failure 4.9 Ice depletion l
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SEQUOYAH T 23 Cs,1 Distribution at 8 Hours *
! Deposited Airborne
-3 Primary System .98 5 x 10
-3 -5 Pressurizer 4 x 10 4 x 10
-5 Containment .012 1 x 10
-5 Environment <1 x 10
- CsOH, Osl lumped, JANAF vapor pressure for CsOH.
4 9
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SEQUOYAH V 1
i -
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SEQUOYAH V SEQUENCE
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SEQUOYAH V SEQUENCE u - TE SR
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O MAAP 1.2 T/H UNCERTAINTY ANALYSIS 1 J l .
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T/H UNCERTAINTY ANALYSIS:
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
-3 weakness in a model by giving erroneous results, (e.g. core slump model).
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e Table 5.6 Fisure of Merit Summarv for MAAP Uncertaintv/Sensitivits Analysis PWR VERSION 1.2' SEQUOYAH SEQUENCE ThLB' (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)
STHLB-1 None N/A 18.88 15,62 83103. 0.0 0.223 4.25 8.04 1.84 459.0 2.09 3.26 9.27 1025.0 SinLb-2 HVP 10 18.85 15.59 86097. 0.0 0.223 4.38 0.09 1.84 459.0 2.09 3.26 9.25 1020.0 SinLD-3 TEXITI 60 F 18.96 15.69 101390. 0.0 0.224 4.18 4.78 1.85 461'.0 2.09 3.26 9.28 1026.0 Default Model Parameters Chansed:
Value Fisure of Merit Nomenclature:
_= - ..------
NVP Number of Penetrations failed in lower head......... 1 1 Time of containment failure TEXITI Ice condenser Primarv side exit temperature (F).... 100 2 Time between reactor vessel failure and containment failure 3 Integrated wall condensation (measure of diffusioehoresis) between reactor vessel feilure and containment failure 4 Peak cuntainment outer wall surface temperature 5 Fraction of clad reacted in-vessel 6 Rate of chanse of containment Pressure Just erior to the tise of containment failure 7 Time of ice depletion (if applicable) 8 Time of core uncovers 9 Hvdrosen 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 Hvdrosen mass at time of containment failure
. .i. .
Table 5.7 Fisure of Merit Summarv for MAAP Uncertaintv/Sensitivits Analvsis PWR VERSION 1.2 SEQUOYAH SEQUENCE S2HF (TEC) 1 1
MODEL FIGURE OF MERIT VALUE 1 2 3 4 5 6 7 8 9 -10 11 12 13 CASE PARAMETER (hr) (hr) (lba) (F) (Psi /hr) (hr) (hr) (lba) (br) (hr) (hr) (lba)
- S2HF-A None N/A 9.13 5.03 192520. 270.3 0.208 10.97 4.38 1.35 428.3 1.64 4.10 10.57 517.1 S2HF-K FCRSLU .8 15.67 0.06 5836. 370.4 0.287 380.00 6.05 1.35 591.1 1.66 15.61 >24.00 1006.4
- S2HF-B FCRSLU .2 9.69 6.63 199600. 270.4 0.208 11.48 4.97 1.35 427.7 1.64 3.06 9.15 528.0 S2HF-D TTCSP .27 9.21 4.87 192290. 269.3 0.208 11.25 4.53 1.35 428.3 1.64 4.34 10.85 502.7 6.76 13.80 789.B*
4 S2HF-E FHT 0.5 7.04 0.28 47400. 268.4 0.257 204.90 5.34 1.35 529.0 1.66 S2HF-F TZOOFF 4040 8.71 3.97 177490. 268.6 0.219 13.48 4.97 1.35 450.7 1.64 4.74 11.41 616.5 S2HF-P FCHF .05 8.65 4.54 176880. 270.2 0.208 12.60 4.67 1.35 428.2 1.64 4.11 10.58 669.9 S2HF-0 FCHF .3 9.46 5.36 197030. 270.5 0.208 10.01 4.23 1.35 428.4 1.64 4.10 10.56 478.6 Default Model Parameters Chansedt Value Figure o. Merit Homenclaturel FCRSLU Fraction of total core mass which must melt 1 Time of containment failure to reach suPPo r t Pl a te . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.5 2 Time between reactor vessel failure and TTCSP Time to fail support Plate after corium Pile containment failure h a s r e ached i t ( b r ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0. 0333 3 Intesrated wall condensation (measure of FHT Fraction of maximum heat transfer Permitted diffusiophoresis) between reactor vessel between clad and sas streza....................... 1.0 failure and containment failure TZOOFF Zircalow oxidation cut-off and channel blockins 4 Peak containment outer wall surface temperature-(F)................................... .3680 temperature FCHF F'at Plate CHF critical velocitw coefficient..... 0.14 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 aePlacable) 8 Time of core uncovers 9 Hvdrosen senerated at time of vessel failure 10 Time at which clad temperature reaches 2000 F 11 Time of vessel failure 12 Time of core cielt completion
- ()alErgory 2 4
e e .I ,
e Table 5.8 Fisure of Merit Summarv for MAAP Uncertaintv/Sensitivits Anaissis , ,
PWR VERSION 1.2 SEQUOYAH SEQUENCE S2HF (FAI)
NODEL FIGURE OF MERIT CASE PARAMETER VALUE 1 2 3 4 5 6 7 8 9 10 11 12 13 (hr) (hr) (1ba) (F) (esi/hr) (hr) (hr) (Ibm) (hr) (hr) (hr) (Iba)
SS2HF-1 None N/A 9.00 4.51 198630. 0.0 0.180 8.90 4.72 1.35 372.0 1.63 4.49 10.99 521.0 SS2HF-9 FCRSLU 0.8 19.09 0.06 5875. 0.0 0.312 309.00 8.79 1.35 638.0 1.63 19.03 >19.09 638.0
- SS2HF-11 FCRSLU 0.2 10.18 7.40 219940. 0.0 0.173 7.24 4.79 1.35 357.0 1.63 2.70 8.69 386.0 SS2HF-4 All CDs 0.5 8.99 4.50 198510. 0.0 0.180 9.04 4.72 1.29 372.0 1.63 4.49 >0.99 522.0 SS2HF-10 All CDs 0.1 9.21 5.25 194280. 0.0 0.197 8.54 4.29 1.33 406.0 1.64 3.95 >9.21 494.0 SS2HF-5 SCALU 10 9.06 4.61 199980. 0.0 0.181 9.91 4.70 1.29 372.0 1.63 4.45 >9.06 511.0 SS2HF-6 SCALH 0.5 7.62 3.09 111250. 0.0 0.178 15.22 4.71 1.33 368.0 1.64 4.5; 11.05 518.0 SS2HF-7 FUMIN 0.2 9.00 4.51 198630. 0.0 0.180 9.85 4.72 1.35 372.0 1.63 4.49 >9.00 521.0 SS2HF-2 TEXITI 150 F 10.47 6.21 169110. 0.0 0.189 8.47 6.25 1.35 390.0 1.63 4.26 >10.47 505.0 Default Model Parameters Changed: Value Fisure of Merit Noienclature 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 CDA Flame buovancs dras coefficient in compartment A.. 5.0 containment failure Cl:B Flame buovancs dras coefficient in compartment B.. 5.0 3 Integrated wall condensation (measure of CDC Flame buovaricw dras coefficient in compartment C.. 5.0 diffusiophoresis) between reactor vessel
- CDD Flame buovanew dras coefficient in ComParlment D.. 5.0 failure and containment failure CLU Flame buovanew dras coefficient in comeartment U.. 5.0 4 Peak containment outer wall surface SCALU Scaling factor for all burnans velocities......... 1.0 temeerature SCALH Scalans factor for heat transfer coeffsetents to 5 Fraction of clad reacted in-vessel eassive heat sinks................................ 1.0 6 Rate of change of containment Pressure Just FUMIN Minimum burn velocits (must be in SI units)........ 1.0 error to the time of containment failure IEXIII Ice condenser erinarv side exit temperature (F).... 100 7 Time of ice deeletion (if apelicable) 8 Time of core uncovers V Hwdrosen senerated at time of vessel failure 10 Time at which clad temeereture reaches 2000 F '
11 Time of vessel failure 12 Time of core melt comeletion o (3ateg40fl/ 2 13 Hwdrogen mass at time of containment failure 1
e a
.I .
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~
l RADIONUCLIDE RELEASE
~
PWR -- ICE CONDENSER (SEQUOYAH) ,
BMI-2104 RESULTS by James A. Gieseke Presented at the NRC/IDCOR MEETING August 28-29, 1984 1
~
OBattelle Columbus Laboratories e -+ - ~-= =v N
4 0
- TMLB' SEQUENCE e TRANSIENT, COMPLETE LOSS OF ELECTRIC POWER, LEADING TO LOSS OF
'0WER CONVERSION, AUXILIARY FEEDWATER, AND PRIMARY COOLANT 4AKEUP SYSTEM e CONTAINMENT SAFETY FEATURES (CONTAINMENT SPRAYS, CONTAINMENT COOLING SYSTEMS) ARE NOT AVAILABLE
~
OBaHelle Columbus Laboratones
f =
ACCIDENT EVENT TIMES
~
Event Time, minutes ;
Sequoyah TMLB'-Y Steam Generator Dry 62.0 Core Uncover 97.8 Start Melt 121.5 Start Slump 143.5 Core Collapse 145.0 Vessel Head Dry 149.2 Bottom Head Fail 157.8 Containment Fail 157.8 Concrete Attack 158.9 Ice Melt Complete 428.5 End Calculation 761.2 e
OBallelle Columbus Laboratones
4 ACCIDENT EVENT TIMES Event Time, minutes 4
Sequoyah TMLB'-6 Steam Generator Dry 62.0
, Core Uncover 97.8 Start Melt 121.5 Start Slump 143.5 Core Collapse 145.0 Vessel Head Dry 149.2 Bottom Head Fail 157.8 Concrete Attack 161.5 Containment Fail 552.5 Ice Melt Complete 556.0 End Calculation 761.6 l
l l
l l
~
OBallelle Columbus Laboratories
2 t
4 5
DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED AS CALCULATED BY THE NAVA CODE, TMLB' Containment Fraction of Core Inventory .
~
Failure Lower Ice Upper Mode Species RCS Cont Bed Cont Environment
-2 Cs! 0.82 6.1 x 10 -2 0.10 1.5 x 10 -3 1.7 x 10 y Cs0H 0.83 3.9 x 10 -2 0.12 2.9 x 10 -3 2.3 x 10 -2 Te. 0.25 2.4 x 10-2 3.7 x 10 -2 6.2 x 10 -4 1.4 x 10 -2 Cs! 0.82 8.6 x 10 -3 0.17 5.5 x 10 -3 - 3.9 x 10
-4 6 Cs0H 0.83 5.4 x 10-2 0.13 5.0 x 10 -3 4.5 x 10-4 Te 0.25 4.0 x 10 -2 3.1 x 10 -2 3.8 x 10-3 2.0 x 10 -3 l
l 9
~
QBattelle Columbus Laboratories l
4
4 O
l TML SEQUENCE r __
. e TRANSIENT, LOSS OF POWER CONVERSION SYSTEM, AUXILI ARY FEEDWATER, AND PRIMARY COOLANT MAKEUP e CONTAINMENT SAFETY FEATURES (AIR RECIRCULATION FANS, CONTAINMENT SPRAYS, HYDROGEN IGNITERS) ARE OPERABLE J
t
~
OBaHelle Columbus Laboratories
o-r e ACCIDENT CVENT TIMES Event Time, minutes Il Sequoyah TML-Y Steam Generator Dry 62.0 Spray and Fan On 62.0 ,
Core Uncover 97.2 Start Melt 121.0 Spray Recirculation On 126.0 5 tart Slump 143.0 Core Collapse 144.5 Vessel Head Dry 148.8 Bottom Head Fail 157.2 Hydrogen Burn 157.2 Containment Fail 157.2 Hydrogen Burn 159.7 End Calculation 263.1
~
OBalfelle Columbus Laboratories l
,,-e
, a l
l l
l ACCIDENT EVENT TIMES Event Time, minutes c I
Seouoyah TML-6/c Steam Generator Dry 62.0 l Spray and Fan On 62.0 Core Uncover 97.2 Start Melt 121.0 Spray Recirculation On 126.0 Start Slump 143.0 Core Collapse 144.5 l: Vessel Head Dry 148.8 Hydrogen Burn 150.1 Hydrogen Burn 155.2 Bottom Head Fail 157.3 Concrete Attack 157.3 Hydrogen Burn 198.7 Hydrogen Burn 264.7 I Hydrogen Burn 301.5 Hydrogen Burn 335.5 Hydrogen Burn 364.3 i
Ice Melted 531.1 1
End Calculation 760.4 l
1 OBaffelle Columbus taboratones I -i t.
i l
DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED AS CALCULATED BY.THE NAUA CODE, TML Containment Fraction of Core Inventory Failure Lower Ice Upper ~
Mode Species RCS Cont Eed Cont Environment *
-2 -3 Cs1 0.82 4.2 x 10 -2 ' 9.4 x 10
-2 4.6 x.10 1.3 x 10 y Cs0H 0.83 3.1 x 10 -2 -0.11 3.4 x 10 -2 7.0 x 10-3
-4 -3 Te 0.25 3.1 x 10 1.2 x 10 -2 8.6 x 10 5.5 x 10-4 Csl 0.82 5.7 x 10-2 8.5 x 10 -2 3.4 x 10-2 6.9 x 10-9 6 Cs0H 0.83 5.8 x 10 -2 9.4 x 10-2 3.5 x 10-2 7.4 x 10 -9 Te 0.25 8.1 x 10 -3 1.0 x 10 -2 9.5 x 10 -3 1.6 x 10-8
- All iodine is assumed to be present as Csl and formation of volatile iodides in the containment was not considered. If production of volatile iodides is consid the iodide release to the environment is estinated to be of the order of 10'gred, .
i l
l l
l
~
l 0Ballelle Columbus Laboratones l
1
- - , , ,- -. ,, , ------w r, - -s--- - -.--w
O 9 S2HF SEQUENCE
(
e SMALL-BREAK LOSS-0F-COOLANT ACCIDENT (LOCA) e DELAYED FAILURE OF EMERGENCY CORE COOLING AND CONTAINMENT SPRAY RECIRCULATION SYSTEMS l
e AIR RETURN FANS AND HYDR 0 GEN IGNITERS ARE OPERABLE t
e a
I
~
OBattelle Columbus Laborarones l.
ACCIDENT EVENT TIMES Event Tine, minutes Sequoyah S HF-Y 2 ,
Fan On 0.6 Spray On 1.1 ECCS Recirculation 40.9 Spray Recirculation 52.7 ECC and CS Fail 107.7 Core Uncover 162.9 Start Melt 195.9 Ice Melt Complete 197.7 Hydrogen Burn 201.7 Start Core Slump 207.9 Hydrogen Burn 208.1 Core Collapse 208.7 Hydrogen Burn 208.9 Vessel Head Dry 216.0 Bottom Head Fail' 259.6 i Concrete Attack 259.7 Hydrogen Burn 352.1
! Containment Fail 352.2 End Calculation 860.3 l
OBattelle Columbus laboratories
O DISTRIBUTION OF SPECIES AFTER ACCIDENT IS COMPLETED AS CALCULATED BY THE NAUA CODE, 5 HF-y 2
Fraction of Core Inventory -,i Lower Ice Upper Species RCS Volume Bed Volume Environment Cs! 0.73 4.5 x 10-2 8.3 x 10-2 0.109 3.3 x 10 -2 3.2 x 10 -2
~
, Cs0H- 0.75 4.2 x 10-2 7.6 x 10-2 0.100 Te 0.69 6.4 x 10 -2 6.5 x 10-2 0.045 5.5 x 10-2 1
4 1
OBattelle Columbus taboratones I
l
[
TASK 23.1 RESULTS GRAND GULF _
Jeff R. Gabor !
Fauske & Associates, Inc.
16WO70 West 83rd Street i 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
) .
GRAND GULF NUCLEAR STATION T jQUV ACCIDEtiT CHR0t10 LOGY Time Event 7 0.0 sec Initiating Event: Loss of off-site power; Loss of main feedwater; TSV/TBV closures 7.8 sec Reactor scram completed 95 sec RPV Level 2 LOCA setpoint reached i 26.0 min RPV Level 1 LOCA setpoint reached; Vessel depressurization manually initiated 26.5 min DW purge system actuates 28.0 min Core begins to uncover
, 57.0 min SPMU actuates 2.0 hr Fuel melting begins
! 2.35 hr High DW pressure LOCA setpoint reached l 2.35 hr Core plate failure followed by vessel l
failure 47.0 hr Containment failure i
1 I
9 m GRAND GULF NUCLEAR STATION AE ACCIDENT CHRON0 LOGY i
Time Event -
0.0 see Initiating Event: A large break in suction
'l side of a recirculation loop
! l 0.2 sec High CW pressure LOCA setpoints reached 3.9 sec Reactor scram completed 5.2 sec RPV Level 2 LOCA setpoint reached 6.5 sec RPV Level 1 LOCA setpoint reached; MSIVs close; Main feedwater pumps trip off 45.0 sec Core begins to uncover 11.6 min DW purge system actuates 30.4 min SPMU actuates
- l.1 hr Fuel melting begins 1.4 hr Core plate failure followed by vessel failure 1 ,
22.3 hr CST drained and CRD flow to vessel ceases j 58.0 hr Containment failure l
l l'
GRAND GULF NUCLEAR STATION N .
[#
T230N ACCIDENT CHRONOLOGY Time Event 0 sec Initiating event: MSIV closures; Loss of main
, feedwater _
3.7 sec Reactor scram completed 28 sec RPV Level 2 LOCA setpoint reached 1.0 min HPCS and RCIC systems begin operating 1.1 hr HPCS and RCIC systems transfer suction from CST to SP 2.35 hr Suppression pool temperature exceeds 145*C, manual ADS 2.49 hr RCIC pump fails on low RPV pressure
- 4. 1 h,- High DW pressure LOCA setpoint reached; DW purge system actuates; LPCS and LPCI actuate 4.6 hr SPMU actuates 22.4 hr CST empties l 23.5 hr High wetwell pressure setpoint reached; Contain-ment sprays actuate 40.0 hr Containment failure; All ECCS assumed to fail
( 48.8 hr Core begirs to uncover 54.1 hr Fuel melting begins 56.2 hr Core plate failure followed by vessel failure s
Ns ./
1 I
L
GRAND GULF NUCLEAR STATION T23C ACCIDENT CHRONOLOGY
, Time Event O sec Initiating events: MSIV closures; Failure to scram; Loss of main feedwater 33 sec RPV Level 2 LOCA setpoint reached 49 sec HPCS negins operating 52 sec RCIC begins operating 4.5 min HPCS/RCIC systems transfer suction from CST to SP 8 min ADS manually initiated 18.3 min RCIC pump fails on h'igh suction temperature 23.0 min High DW pressure LOCA setpoint reached; Post-LOCA DW vacuum breakers open 23.6 min Drywell purge system actuates 23.8 min LPCS and LPCI actuate 26.2 min High wetwell pressure setpoint reached 33.8 min Containment sprays actuate 53.1 min SPMU actuates 1.0 hr Containment failure and subsequent ECCS failure 1.3 hr Core begins to uncover
. 3.0 hr Fuel melting begins 3.8 hr Core plate failure followed by vessel failure
DISTRIBUTION OF Csl IN PLANT AND ENVIRONMENT (FRACTION OF CORE INVENTORY)
At Vessel Failure
. T23 QW T23C AE T jQUV RPV .90 .68 .98 .98 Drywell 0.0 0.0 .02 0.0
~
Suppression Pool .10 .32 0.0 .02 .-
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Base Case Case 1 Case 2 Containment Failure (hr) 42 48 47 Cs&l Fraction Released -s from Containment at 60 < 10 -5 3 x 10 3 x 10-s hrs.
Drywell Temperature at 670 1000 1000 60 hrs. ( F)
Cs&l Fraction in Drywell .59 .53 0 at 60 hrs.
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GRAND GULF - T,QUV (CORE BLOCKAGE)
UNCERTAINTY / SENSITIVITY ANALYSIS Base Case - No ADS Case 1 - No Core Blockage, Temperature for Zr Oxidation Cut-off = 3100K Base Case Case 1 Vessel Failure (hr) 3.4 3.1 in-Vossel Zr Oxidation Fraction .04 .21 Containment Failure (hrs) 57 49
-5 = 10 -5 l Cs&l Release Fraction at 60 < 10 r hrs.
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RADIONUCLIDE RELEASE BWR -- MARK III DESIGN (GRAND GULF) _
BMI-2104 RESULTS by James A. Gieseke 1
Presented it the NRC/IDCOR MEETING August 28-29, 1984 i
OBattelle -
( c. . . ,<. s ,. ,.., y
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TC SEQUENCE e TRANSIENT, FAILURE OF CONTROL R00 INSERTION (FAILURE TO -
~-
SCRAM) e EMERGENCY CORE COOLING SYSTEMS OPERATE f
e CONTAINMENT FAILURE RESULTS FROM IMBALANCE IN POWER GENERATION AND CONTAINMENT HEAT REMOVAL DUE TO THE CONTINUED HIGH POWER LEVEL OF THE REACTOR m
OBallelle Columbus Laboratones
- =
s ACCIDENT EVENT TIMES Event Time, minutes -
Grand Gulf TCY i
Containment Heat Removal On 10.0 ECC Recirculation On 76.9 Containment Fail 80.6 ECC Off 80.8 Core Uncover 88.2 Start Melt 117.5 Core Slump 168.1 Core Collapse 175.1 Bottom Head Dry 195.3 Bottom Head Fail 197.6 Reactor Cavity Dry 197.6 Start Concrete Attack 197.6 End Calculation 798.8 1
e f
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Time Particle Diameter (um) DF Based on
, (min) 0.1 0.7 1.2 5.1 8. 4 - Total Mass 118 1.3 105 (a) 10 5
10 5
10 5
1400 5 5 5 5 158 1.7 10 10 10 10 81500 5 5 5 5 207 6.9 10 10 10 10 47 258 1.02 8.6 1.63 x 10 3 10 5
10 5
43 300 l'.04 10.6 2.35 x 10 3 10 5
10 5
59 400 1.1 17.3 3.95 x 10 3 10 5
10 5
60 61 8 1.15 23.3 5.86 x 10 3 10 5
10 5
27 (a) A decontamination factor larger than 105is assumed to be 105 ,
Pool depth: 17' 10" (543 cm) up to vessel failure and 12' 10" (390 cm) thereaf ter.
Bubble diameter: 0.75 cm Axes ratio: 1:3 l OBattelle CMumbus taboratories l
l
l DISTRIBUTION OF SPECIES AT 25 HOURS AFTER ACCIDENT, TC I E E E I ,-
Fraction of Core Inventory
- . Species RCS Drywell Pool Containment Environment Cs1 0.19 3.6 x 10 -2 0.77 1.9 x 10-4 6.8 x 10 -3 Cs0H 0. 51 1.4 x 10 -3 0.49 9.2 x 10 -6 3.5 x 10 -4 Te 0.22 0.32I ") 0.45 4.3 x 10 -4 8.8 x 10 -3 (a) This number includes a fraction of 0.26 for Te not being released from the core-concrete interaction.
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FRACTION OF CORE INVENTORY RELEASED TO THE ATMOSPHERE FOR GROUPS OF REACTOR SAFETY STUDY, TC Time I . Cs . Te Sr Ru La (br) Group 2 Group 3 Group 4 Group 5 Group 6 Group 7 l.5 0 0 0 0 0 0
-8 2 4.4 x 10 6.3 x 10 -8 4.2 x 10 ~I4 8.1 x 10 -10 2.8 x 10-10 1.0 x 10 -12 4 1.5 x 10 -3 2.5 x 10 -4 3.0 x 10-4 5.6 x 10 -4 3.8 x 10-5 8.3 x 10 -5 7 6.6 x 10 -3 6.1 x 10-4 3.5 x 10 -3 9.4 x 10-3 7.6 x 10 -5 1.8 x 10 -3
-4 10 6.8 x 10-3 8.3 x 10 5.5 x 10 -3 1.2 x 10 -2 7.7 x 10-5 2.3 x 10 -3 15 6.8 x 10 -3 8.3 x 10-4 8.6 x 10 -3 1.3 x 10-2 7.7 x 10 -5 2.4 x 10 -3 20 6.8 x 10 -3 8.3 x 10~4 8.8 x 10 -3 1.3 x 10 -2 7.7 x 10-5 2.4 x 10-3
,I
- The release of volatile iodides is not included in these release fractions.
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o e TPI SEQUENCE o TRANSIENT, STUCK OPEN RELIEF VALVE, LOSS OF CONTAINMENT HEAT _
REMOVAL s EMERGENCY CORE COOLING SYSTEM SUPPLIES MAKEUP WATER TO REACTOR VESSEL e CONTAINMENT FAILS BEFORE CORE MELTS e EMERGENCY CORE COOLING PUMPS ARE ASSUMED TO FAIL AFTER CONTAIN-MENT FAILURE
~
OBa!!elle Columbus taboratories
S ACCIDENT EVENT TIMES q
Event Time, minutes
{ .
Grand Gulf TpIY ECC Recirculation On 1278.8 Containment Fail 1331.8 ECC Off 1332.9 Core Uncover 1535.7 Start Melt 1645.7 Core Slump 1729.5 Core Collapse 1765.2 Bottom Head Dry 1801.0
'm Head Fail 1978.0
..or Cavity Dry 1978.0 Start Concrete Attack 1978.0 End Calculation 2579.0 OBallelle Columbus Laboratories
o f.
DECONTAMINT. TION FACTORS CALCULATED AS A FUNCTION OF PARTICLE SIZE AND OF TIE FOR TPI SEQUENCE Time Particle Diameter (um) DF Based on 'I (min) 0.1 0.4 1.2 5.1 8.4 Total Mass 5
105 (a) 5 1650 1.95 37.7 10 10 23300 5 5 5 1750 11.2 6.69 x 10 3 10 10 10 14000 5 5 5 1850 34.5 141 10 10 10 8080 5 5 5 2004 7.3 15.8 10 10 10 284 5
1.61 x 10 3 5
2100 -1.19 1.73 10 - 10 45.1 3 5 5 2200 1.08 1.66 1.82 x 10 10 10 77,4 5 5 2400 1.14 2.18 3.36 x 10 3 10 10 73.0 5
(a) A decontamination factor larger than 10 5is assumed to be 10 ,
Pool depth: 17' 10" (543 cm) up to' vessel failure and 12' 10" (390 cm) thereafter.
Bubble diameter: 0.75 cm Axes ratio: 1:3
'0Ballelle Columbus Laboratories
a 4
DISTRIBUTION OF SPECIES AT 50 HOURS AFTER ACCIDENT, TPI l
~
Fraction of Core Inventory .
Species RCS Drywell Fool Contai nment Envi ronment
)
i
-2 -3 -7 -4 8.4 x 10 3.9 x 10 Csl 0.91 7.5 x 10 2.4 x 10 Cs0H 0.24 3.7 x 10 -3 0.76 9.0 x 10
-7 3.1 x 10 -4 Te 0.45 0.41 (a) 0.14 1.4 x 10 -5 1.3 x 10 -3 I
t (a) This number includes a fraction of 0.35 for Te not being released from the core-concrete interaction.
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1 OBattelle Columbus taborar ....,
a ..
TQUV SEQUENCE e TRANSIENT, LOSS OF ALL PRIMARY SYSTEM MAKEUP WATER .
e OPERATORS DEPRESSURIZE PRIMARY SYSTEM PRIOR TO CORE MELTING o HYDROGEN IGNITERS ARE ASSUMED TO BE OPERABLE e CONTAINMENT FAILURE RESULTS FROM BUILDUP OF NONCONDENSIBLE
' GASES OBattelle Columbus Laboratories
h ACCIDENT EVENT TIMES i
Event Tine, minutes I Grand Gulf TOUVY Containment Heat Removal On 11.0 Core Uncover 47.0 Activate ADS 79.0 Start Melt 103.0 Hydrogen-Burn 115.5 Hydrogen Burn 136.4 Core Slump 142.0 Hydrogen Burn 149.3 Hydrogen Burn 154.2
.I Core Collapse 155.0 Bottom Head Dry 170.6 Bottom Head Fail 216.6
- Reactor Cavity Dry 216.6 Start Concrete Attack 216.6
' ! Containment Fail 834.3
.; End Calculation 1416.7 4
4 OBattelle Columbus laboratories
f 1
1 DISTRIBUTION OF SPECIES AT 30 HOURS AFTER ACCIDENT, TQUV Fraction of Core Inventory ;
Species RCS Drywell Pool Containment Environment i
Csl 6.3 x 10-2 3.8 x 10-6 0.94 6.8 x 10-4 8.4 x 10-4 Cs0H 0.54 2.8 x 10-6 0.46 3.5 x 10-4 4.4 x 10-4
-3 Te .40 0.38(a) 0.21 1.6 x 10 -3 2.1 x 10 (a) This number includes a fraction of 0.30 for Te not being released from the core-concrete interaction.
~
OBallelle Columbus Laboratones
,\'
S2E SEQUENCE e SMALL-BREAK LOSS-OF-COOLANT ACCIDENT (LOCA), FAILURE OF -
EMERGENCY CORE COOLING INJECTION SYSTEM e HYDROGEN IGNITERS ARE OPERABLE e SUPPRESSION POOL HEAT REMOVAL IS AVAILABLE e SOME LEAKAGE, INCLUDING FISSION 'RODUCTS, FROM THE DRYWELL TO THE MAIN CONTAINMENT, BYPASSING THE SUPPRESSION POOL IS ASSUMED OBallelle Columbus Laboratories
a ACCIDENT EVENT TIMES Event Time, minutes ,
Grand Gulf S2E Core Uncover 5.6 Start Melt 27.8 Hydrogen Burn 41.4 Start Slump 45.6 Hydrogen Burn 54.4 Containment Fail 54.4 Core Collapse 57.6 Hydrogen Burn 57.6 Hydrogen Burn 59.6 Vessel Head Dry 80.7 Head Fail 119.2 Concrete Attack 119.2 Hydrogen Burn 125.6 Cavity Dry 213.7 Hydrogen Burn 245.5 End Calculation 720.3 OBallelle Cniumbus Laboratories
s 1
DISTRIBUTION OF SPECIES AT 20 HOURS AFTER ACCIDENT, S E WITH NOMINAL POOL BYPASS 2 ._
Fraction of Core Inventory Species RCS Drywell Pool Containment Environment Csl 9.1 x 10-2 1.4 x 10~2 0.89 9.6 x 10~4 7.0 x 10-3
-2 Cs0H 0.16 1.3 x 10 0.82 8.6 x 10~4 6.3 x 10-3 Te 0.26 0.39(a) 0.32 6.5 x 10 -3 2.4 x 10 -2 (a) This number includes a fraction of 0.355 for Te not being released from the core-concrete interaction.
OBattelle Columbus laboratones
e ..
DISTRIBUTION OF SPECIES AT 24 HOURS AFTER ACCIDENT, S E WITH STUCK OPEN VACUUM BREAKER 2
Fraction of Core Inventory Species RCS Drywell Pool Containment Environment Cs! 9.1 x 10
-2 1.2 x 10-2 0.89 2.0 x 10 -3 3.3 x 10 -3 Cs0H 0.16 1.1 x 10-2 0.82 1.9 x 10-3 3.2 x 10-3 Te 0.26 0.39(a) 0.11 0.10 0.14 (a) This number includes a fraction of 0.355 for Te not being released from the core-concrete interaction.
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Enclosure 6 m an.y
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' - ;; d ' .
TECHNOLOGY for ENERGY CORPORATION September 7,1984 Mr. Robert Bernero, Director Division of Systems Integrations U. S. Nuclear Regulatory Commission Washington, D. C. 20555 I
Dear Bob:
Comments on NRC Comments at IDCOR/NRC Meeting 38-Integrated Analysis of Plant Response -
Rockville, Maryland, August 28-29, 1984
- The comments made by the two NRC summary teams at the end of meeting 3B were not available to us prior to their presentations. Therefore, 3 we could only comment on them orally.
We feel that the record, which includes the viewgraphs used at the meeting, should contain our comments on the NRC comments.
Enclosed are our item-by-item comments. The NRC comments are copied from the sumary teams' viewgraphs; the IDCOR comments follow each item.
Please include these in the meeting minutes.
Sincerely,
!'1 ; ,
/ /ft !li ,
- M. H. Fontana, Director National IDCOR Program MHF/cb ID0 984-001
, Enclosures cc: C. Ree'd -
A. Buhl G. Edgar M. Leverett J. Siegel
/
9 Enclosure 6 ONE ENERGY CENTER PELLISSIPPI PARKWAY KNOXVfLLE, TN 37922 . PHONE (615) 986-$456 TELEX 810-6701770
_ , , _ ~ . _ . . - . . . _ .
a -
IDCOR COMMENTS ON NRC C0pmENTS IDCOR/NRC TECHNICAL EXCHANGE E ETING 38-INTEGRATED ANALYSIS OF FISSION PRODUCT BEHAVIOR - ROCKVILLE, M , AUGUST 28-29, 1984 ,
- 1. Conta'inment Loads Summary 1.1 In-Vessel Hydrogen Production NRC:
Difference -
Hydrogen generation differs by factors of 1.4 to 10.0 because of blockage i assumptions, lack of downward radiation heat transfer and the short ;
time between slump and vessel failure Significance Affects probability and timing of containment failure IDCOR:
IDCOR analyzed downward radiant heat transfer to the pool and found it to be insignificant.
IDCOR feels that their analysis of vessel melt though, taking into consideration the effects of welds in vessel penetrations, is correct.
The time during which the debris is in the lower head, generating steam, affects the amount of hydrogen produced by zircaloy oxidation in the core remaining above. IDCOR feels that when molten debris collects and refreezes in the interstices between the fuel rods, a blockage effect exists where some of the steam arising from the debris (which had fallen into the pool of water in the lower head) is prevented from reaching some of the unreacted zircaloy in the remaining sections of the core above. The exact configuration of the blockage and the flow patterns around it are very difficult to predict, but we feel that this effect should not be ignored.
1.2 Hydrogen Combustion Behavior NRC:
Difference -
Flame temperature criteria used prevents high hydrogen 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). Important for ice condensors.
T Page 2 IDCOR:
- IDCOR feels that high temperatures are a valid ignition source for hydrogen, and that natural circulation of gases from cavity to containment will occur. These effects combine to prevent high hydrogen concentrations from being attained, followed by globa.1 burns. This is a more accurate representation of the physical facts than that of models where~these effects are not recognized.
1.3 Mode of Melt Ejection NRC: -
Differences CLWG uses a coherent 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.
IDCOR:
IDCOR models of sequential core melt-through through penetrations, which have much shallower welds than the vessel head itself, appear to be more realistic tnan a coherent release of all core melt available at vessel failure. Direct containment heating in the IDCOR reference plants is not of concern because of structural configurations which prevent material from being ejected into the large containment volume.
Other plants may require individual analyses for this effect.
Liner failure generally is a small leak, which in some cases would be self sealing; for cases where it may not, IDCOR's impaired containment calculations represent the cases of early containment failure.
1.4 Core / concrete Interactions NRC:
Differences IDCOR model results in thick insulating crust between melt and concrete. .
Corcon model results in thinner, les3 stable crusts.
W i
Page 3 1.4 contd.
Significance Solid, not molten, debris attack which affects Upward . heat flux into conta'inment
, Gas generation from concrete I
Fission product release from melt
~
Degradation of overhead structures including degassing of .
unlined concrete
! Interaction with overlying water pool t '
No further oxidation of metal component of melt IDCOR:
Although IOCOR's core concrete interaction model is quite simple, we feel it is adequate in the context of the overall accident analysis.
In particular, upward heat flux is taken into consideration. In fact, IDCOR first discovered the direct heating effect for debris that could fail the drywell in BWR Mark I's by overtemperature.
1.5 Steam Explosions NRC: '
Positions unchanged since Harper's Ferry meeting NRC will organize an experts group to review the issue. Meeting is l planned for October 1984.
l.l IDCOR:
l NRC held a steam explosion experts' meeting about three years ago under the chairmanship of Dr. L. S. Tong. EPRI had a task force study the issue and a report was issued in 1982. IDCOR Task 14 report was published in 1983. The Germans and French have evaluated the problem.
Further, the issue was discussed at Harper's Ferry, and the agreement reached that although differences of opinion exist about detail, steam explosions sufficiently energetic to fail containment are so unlikely that they need not be considered in severe accident analyses.
4-
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Page 4 1.6 Direct Heating of Containment .
NRC:
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 along or in concert with hydrogen.
IDCOR:
See previous discussion under melt ejection.
1.7 Sequences Definition NRC:
Difference
! The importance of sequence definition is recognized by both NRC and IDCOR.
Significance This is an extremely important issue and can influence progression of core melt and mode and timing of containment failure.
IDCOR:
We agree.
1.8 Sensitivity - Uncertainty Analysis NRC:
Differences
~
- IDCOR study based upon incomplete set of parameters
- Does not include variation of parameters based upon engineering judgment
- Effects of varying more than one parameter per case not examined.
_ _ _ _ _ _ _ _ - _ _ _ - _ _ - _ . _ - _ _ _ - _ _ _ _ _ ___-__-_-__A
Page 5 .
Significance Conclusions of the study are difficult to accept without a more complete examination of input parameters and underlying modeling assumptions and their range.
IDCOR:
IDCOR feels that uncertainty and sensitivity analysis are important and should be done properly. IDCOR did indeed vary parameters based on engineering judgment. We do not think that it is either feasible or necessary to vary all possible combinations of parameters. This ._
would require an astronomical, number of computations. In fact, we J used engineering judgment to reduce the number of parameters to be varied. We also feel that it is inappropriate to allow parameters
, to vary beyond the limits of physical reality.
4 9
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Page 6
- 2. Severe Accident Analysis Models Summary ,
2.1. Aerosol Model NRC:
NRC contractors and staff have significant reservations about the FAI aerosol model, particularly for use within the RCS:
- The correlation does not include effects and parameters known to be important from other studies (e.g., an explicit treatment of~ size distribution)
~
1
- The agreement shown with experimental data is not compelling .
because of the similarity of these experiments to the correlation -
{e data base and, for the wet experiments, the need to set tunable parameters.
3 Comparison with experiments at different scales or with more rigorous models would help to dispe,1 skepticism.
IDCOR:
IDCOR's assessment of detailed aerosol models convinced us not to use them. Therefore, we developed empirical correlations based on relevant experiments. In fact, the detailed models have more tunable parameters than the IDCOR empirical correlation which has none.
We feel that the IDCOR model is a valid engineering approach and adequate for our purposes.
2.2 Revaporization Modeling NRC:
The NRC contractors and staff agree that revaporization 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.
i Although the environmental release terms obtained by the IDCOR analyses l were found to be small for the accident sequences analyzed, the simplified j, treatment of RCS behavior may be unacceptable when more realistic
.! assessment is made of retention external to the RCS or to RCS thermal-j1 hydraulic behavior.
IDCOR:
IDCOR first discovered revaporization in the context of severe accidents.
- Although our model is simple, it probably is conservative. In the l
- absence of further experimental data involving, e.g., possible chemical .
i reaction of Cs0H with structural surfaces, we feel that adding sophistication to the model is not warranted.
+< , , .
~
Page 7 2.3 Resuspension Modeling NRC:
Although we feel that there is some potential for resuspension within the RCS that should be resolved, we agree with the IDCOR approach to ignore resuspension from the RCS and containment with the current level of understanding.
IDCOR:
We agree. However, we did not simply ignore the issue. IDCOR analyzed the phenomenon of resuspension and performed some experiments to support our conclusions.
.. 2.4 Tellurium Release i
NRC:
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.
10COR:
The IDCOR approach is consistent with the state of knowledge that l- existed at the time the Task 11.1 .4 .5 reports were written.
However, IOCOR will perform a sensitivity analysis of the effect on ultimate fission product release of differing as'sumptions with respect to tellurium behavior. The question concerns whether it is released at the time of fuel failure as 10COR supposes, or whether it reacts I with zircaloy and is released later at the time of core debris interaction with concrete. We feel that the difference is not particularly significant for cases where containment failure occurs late in time.
2.5 Chemical Reactions of Fission Products
, NRC:
The 10COR 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
4.
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Page 8 Although some of these effects may not be very large, they may dominate -
sequences with small release fractions as analyzed by IDCOR.
IDCOR:
IDCOR recognizes that insufficient experimental data exists to make detailed analyses of these effects, which, as stated by-the NRC spokesman', have not been od'ne by NRC, either. We do not think that these effects'would have sufficiently large impacts to affect conclusions with respect to risk. In other words, small releases would remain small, although their values might be different.
~
2.6 Core Concrete Release i NRC:
From our analyses we believe the release of fission products and aerosols during core-concrete attack is important. IDCOR's approach needs clarification.
IDCOR:
IDCOR will clarify.
't 2.7 Pool Scrubbing NRC:
We disagree with the approach. The DF's selected are conjectural.
The state-of-the-art of data and models permits a better treatment.
IDCOR:
The pool scrubbing decontamination factors used by IDCOR are based on conservative interpretations of GE and EPRI experimental results.
They are not conjectural. NRC should integrate the DF's over the spectrum of particle sizes; we believe that this would result in DF's greater than those used by IDCOR.
2.8 Ice Bed Decontamination NRC:
We basically agree with that the mechanistic effects used by IDCOR are correct. We are concerned about ability to model sedimentation without explicit treatment of particle sizes (Paraphrased from discussion .
notes.)
2" ,,
Page 9 IDCOR:
IDCOR believes that the empirical model covers this. Validation of the model may be appropriate for confirmatory research.
2.9 Fission Product Decay Chains NRC:
For example, the Tel32 __> I132 transition could affect transport behavior, especially over long times. We believe this effect is _
small. .
IDCOR We believe that the decay chains should be taken into consideration for very long times after the accident, but we agree that the effect should not be major for the time periods of interest in the IDCOR analyses. .
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c Page 10 2.10 Aerosol Generation Mechanisms l 1.
NRC:
Potential for additional aerosol generation due to steam explosions and high pressure blowdowns, for example, should be recognized as a component of uncertainty. [ Paraphrased from discussion notes.,
f' IDCOR:
! IDCOR assessments indicate that neither steam explosions or high pressure 4 '
, blowdowns would be sufficiently energetic to cause early containment failure. ~.
~
Additional aerosols ~ generated by these events should have agglomerated and fallen out of the containment atmospheres over the long time periods l' prior to containment failure.
i i
2.11 Uncertainty / Sensitivity Analyses
- NRC: '
I, i Based on other studies, we believe that the uncertainties are much larger i 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.
The NRC group thinks that uncertainty must be larger than indicated in the IOCOR presentations. Maybe IDCOR models do not include most sensitive parameters or didn't allow them to be varied sufficiently. The QUEST study indicates that undertainties are dominated by differences between i alternative models, rather than in variations in parameters. [Last three sentences paraphrased from discussion notes.]
IDCOR:
IDCOR has chosen what it considers to be the best appropriate models l for inclusion in the analyses codes and methods. Also, we feel that parameters and models should not be varied beyond the range
- of physical reasonableness. We don't think that investigating the effects of models, if they are inappropriate, is reasonable. Investigaticn 2
of the validity of models within the context of the ultimate " bottom
- line" effects may be appropriate for confirmatory research.
2.12 Plant Analyses NRC: .!
There are many differences in results. It is not clear yet to what extent j these arise from differences in modeling versus accident sequences defi-i nition. In those cases where there is apparent agreement, it may be
- fortuitous.
- .-,,.~.---,- _._ ,..,_., _ _._.- ,,. _.,_ ,-_- - _ . - . , _ _ . , _ _ , _ _ _ _ _ - - , , , .
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Page 11 .
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.
Following completion of MARCH-MAAP comparison calculations at BCL, interac-tions with IDCOR staff would help to understand differences.
The IDCOR Mark I results depend heavily on the retention capability of the reactor building. We are concerned about our understanding of the failure modes of the primary containment and the reactor building. .
Similarly, the PWR "V" sequence results rely heavily on retention in the I auxiliary building. Also, we believe that the treatment of the drywell leakage and suppression pool bypass in the BWR Mark III cases may be inade-quate. [Last two sentences paraphrased from discussion notes.]
IDCOR:
There are many areas of agreement on results, in that most fission product releases are low. For cases where NRC obtained early containment failure, such as in ice condensers, we feel that the NRC analyses do not take into consideration physical realities such as continuous hydrogen burning as aided by natural circulation in the ice beds and upper compartment. Also, NRC does not take into consideration the potential for fission product attenuation in the reactor building for BWRs Mark I's and the auxiliary buildings for PWR "V" sequences.
For areas where there is agreement, we do not think it is necessarily fortuitous, but rather the natural consequence of working in the same world, which is governed by physical laws.
Accident sequence definition indeed makes a big difference and we so stated repeatedly.
As we stated during the meeting, detailed consideration of operator actions will be covered in meeting 4. IDCOR's assumptions with respect to operator action are based on NRC-approved symptom-oriented emergency procedure guidelines. In fact, if these guidelines are followed, the progression of an accident into severe conditions will not occur. Therefore, IDCOR.
had to assure operator inaction in many cases to allow the accidents to proceed to their end points.
We agree that the MAAP/ MARCH comparison calculations should be discussed in detail.
We agree that the detailed understanding of the failure made of primary containment and reactor buildings could be improved. However, when failure
- occurs late in the accident sequence, these modes of failure become less important.
The exact amount of drywell leakage and suppression pool bypass in BWR Mark III's appear to be plant-specific. In any case, a containment still exists, outside the drywell and the suppression pool, which has fission I product retention capability.
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