ML20214W582

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Notifies of 840828-29 Meetings W/Idcor Re Integrated Analysis of Fission Product Behavior in Primary Sys & in Containment as Function of Time.Agenda & List of Relevant Documents Encl
ML20214W582
Person / Time
Issue date: 08/03/1984
From: Bernero R, Speis T
Office of Nuclear Reactor Regulation
To:
NRC
Shared Package
ML20213E209 List:
References
FOIA-87-113, FOIA-87-60 NUDOCS 8706160089
Download: ML20214W582 (52)
v • d • e


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((sp ac j UNITED STATES NUCLEAR REGULATORY COMMISSION

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MEMORANDUM FOR: Distribution FROM: Themis P. Speis, Director Division of Safety Technology Office of Nuclear Reactor Regulation Robert M. Bernero, Director Division of Risk Analysis Office of Nuclear Regulatory Research

SUBJECT:

NRC/IDCOR MEETING ON INTEGRATED ANALYSIS OF FISSION PRODUCT BEHAVIOR Representatives of IDCOR will meet with the staff and their contractors on August 28-29, 1984, to discuss IDCOR methods and results for integrated analyses of severe accidents. This meeting is the fourth in a series of technical exchanges aimed at understanding the bases for the IDCOR methodology and comparing the NRC and IDCOR.. technical positions. The first two meetings concentrated on phenomenological models for severe accident progression, con-tainment loading and radiological releases. Although those meetings provided a great deal of insight into the IDCOR models, it was clear that a true under-standing of the IDCOR methodology requires an examination of how the models are integrated and applied to calculations of severe accident sequences at existing plants.

In May 1984, a third meeting was held to discuss the manner in which the phenom-enological models are integrated to perform accident sequence analysis. The focus of the mceting was to examine the predicted containment loads for several sequences ahd for several plant types. The sensitivity of these overall results to variations in key parameters and phenomena was also examined.

The purpose of the August 28-29 meeting will be to extend the integrated analysis to include fission product behavior. The methods and assumptions used by IDCOR for calculating fission product release, transport and deposition will be examined. The focus of the meeting will be to examine predicted fission product behavior in the primary system and in cuntainment as a function of time for selected reference plants and accioent sequences.

The agenda for the meeting is shown in Enclosure 1. The meeting will begin with a discussion of several containment loads issues which remained unresolved after the May 1984 me6 ting. Next, the IDCOR MAAP modelling of fission product Y

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2-behavior will be examined in some detail. There will also Se a discussion.

of models for fission product reevolution following deposition in the primary ~

system. -Calculations of fission product behavior for selected severe accident sequences at four reference plants will then be presented by 10COR representa-tives. NRC contractor calculations will be presented for comparison. .The status of fission product behavior will be summarized on the afternoon of August 29.

By August 6, 1984, the Technology for Energy Corporation (TEC) will provide all meeting participants with a set of 10COR documents related to the meeting topics.

Enclosure 2 contains a list of documents which will be forwarded and a list of previously provided documents which will be referenced at the meeting.

The meeting will take place at the Holiday Inn Crown Plaza in Rockville, Maryland. A number of rooms have been set aside. Please contact Linda Kriesel (301-492-7373) for more information regarding administrative matters relating to the meetings. Questions concerning the meeting agenda should be addressed to J. Mitchell (301-492-9402).

.h Themis P. Speis, Director Divi'sion of Safety Technology

. Office of Nuclear Reactor Regulation

[ w Robert M. Bernero, Director Division of Risk Analysis Office of Nuclear Regulatory Research

Enclosures:

1. Meeting Agenda
2. List of Relevant 10COR Documents 4

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'A ENCLOSURE 1 AGENDA __

NRC/IDCOR MEETING ON, .

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 A.M. -

Break Fission Product Methodology 10: 30-11: 00 A.M.c -

Correlation for Fission Product Deposition (10COR) 1 11:00-11:30 A.M. -

Reevolution of Fission Products (IOCOR) t 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 -

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',s 2-Tuesday, August 28 (Cont'd) -

PWR Large Dry 2:00-3:00 P.M. -

IDCOR Results 3:00-3:15 P.M. -

NRC Contractor Results 3:15-3:30 P.M. -

Break BWR Mark I 3:30-4:30 P.M. -

IDCOR Results 4:30-5:00 P.M. -

NRC Contactor 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 -

9:45-11:15 A.M. . -

IDCOR Results l 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 e

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Wednesday, Auaust 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. -

}DCOR Summary 3:30-4:00 P.M. -

Parting Remarks - R. Bernero, T. Speis, A. Buhl 4

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,, - ENCLOSURE 2 LIST OF IDCOR DOCUMENTS RELEVANT TO THE AUGUST 28-29, 1984 NRC/IDCOR MEETING I. Documents to be provided to Participants by TEC 11.6 - Fission Product Resuspension 23.1 - Reference. Plant Integrated Analyses II. Documents Previously Forwarded to NRC Which Will Be Referenced at the Meeting 11.1 15.2a 11.4 15.2b 11.5 15.3 11.3 16.1 14.la 16.la 14.lb 16.2 15.la 16.3 15.lb 9

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TASK 23.1 RESULTS SEQUOYAH Marc A. Kenton .

l Fauske & Associates, Inc.

16WO70 West 83rd Street l Burr Ridge, 'lllinois 60521 (312) 323-8750 NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Fission Product Behavior Rockville, Maryland i

August 28 - 29, 1984 i

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l. .- . , - - - . .

v SEQUOYAH SEQUENCES ANALYZED 1

1. Containment Failure Sequences
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d. V sequence .
2. Sequences Without Containment Failure
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m ACCIDENT

SUMMARY

Time (hrs.) Event 0 2" cold leg break

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i SEQUOYAH S2HF/ DRAINS OPEN Cal Distribution-(Fractions) at 20 Hours Deposited Airborne Primary System .93 -5 3 x 10 4

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SEQUOYAH S2HF/ DRAINS OPEN Release Fractions at 20 Hours Csl CsOH Te, Sb Sr,Ba Ru,Mo

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4. Same as 2,0772 < 1 x 10 5 x 10 2 x 10 < 1 x 10 1 x 10 releases

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5. Same as 2, degraded 2 x 10 4 x 10 5 x 10 < 1 x 10 sedimentation

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6. Same as 1 Sandia N/A 5 x 10 3 x 10 1 x 10

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SEQUOYAH S2HF/ DRAINS OPEN/ PURGE OPEN t

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  • Cs, i Te Primary System -

Deposited .92 .95

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ACCIDENT

SUMMARY

Time (hrs.) Event O Loss of feed, loss of injection, scram

.9 S/Gs dry 1.5 Sprays in recirc mode 1.6 Core uncovered 2.9 Reactor vessel failure 4.9 Ice depletion

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d SEQUOYAH T 23 Cs, I Distribution at 8 Hours

  • Deposited Airborne Primary System -3

.98 5 x 10 Pressurizer -3 -5 4 x 10 4 x 10

-5 Containment .012 1 x 10

-5 Environment <1 x 10 4

CsOH, Osl lumped, JANAF vapor pressure for CsOH.

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SEQUOYAH V

SEQUOYAH V SEOLlENCE u :D p CS. I ATE VSR ORU O ^' -

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9 fiEOUUYAH V SEQUENCE n '9

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Release history of fission products to the environment for a postulated V sequence in Sequoyah.

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9 MAAP 1.2 T/H UNCERTAINTY ANALYSIS 8

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t T/H UNCERTAINTY ANALYSIS:

Conclusions in the conte'xt of the MAAP models, essentially no important sensitivities of bottom-line results to input parameters with a few exceptions.

Category 1: Change in input value altered sequence definition, (e.g. sprays come on due to parameter change).

. Category 2: Parameter change reveals a fundamental weakness in a model b'y giving erroneous results, (e.g. core slump model).

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Table 5.6 Fisure of Merit Summarv for MAAP Uncertaints/Sensitivits Analvsis

. PWR VERSION 1.2 SEDUOYAH SEQUENCE TMLB' (FAI)

MODEL FIGURE OF MERIT CASE PARAnETER VALUE 1 2 3 4 5 6 7 8 9 10 11 12 13 (hr) (hr) (lba) (F) (Psi /he) (hr) (hr) (Iba) (hr) (hr) (hr) (lbs)

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 STnLB-2 NVP 10 18.85 15.59 86097. 0.0 0.223 4.38 8.09 1.84 459.0 2.09 3.26 9.25 1070.0 STnLD-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 Changedi Value Fisure of Merit Nomenclature 1-

= = - _ - - --------- . ----- ------------------- ___ _- .

1 HVP Humber of eenetrations failed in lower head......... 1 1 Time of containment failure

' TEXIII Ice condenser Primarv side exit temperature (F)... 100 2 Time between reactor vessel failure and containment failure 3 Integrated uall condensation (measure of 1 diffusiophoresis) between reactor vessel failure and containment failure l

' 4 Peak corstainment outer wall surface temperature l 5 Fraction of clad reacted in-vessel 6 Rate of change of containment Pressure Just erior to the time of coritainment failure 7 Time of ice deeletion (if aeelicable) i 8 Time of core uncoverv 9 Hwdrosen generated at time of vessel failure 10 Time at which elad teneerature reaches 2000 F 11 Time of vessel failure 12 Time of core melt completion 13 Hvdrogen mass at time of containment failure i t E

l

  • I li .

Table 5.7 Fisure of Merit Summarv for MAAP Uncertaintw/Sensitivitw Analvsis PWR VERSION 1.2 SE000VAH SEQUENCE S2HF (TEC)

MODEL . . F1GURE 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) (Ibe) (hr) (hr) (hr) (Ibe)

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 ~4.05 1.35 591.1 1.64 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

$2HF-E FHT 0.5 7.04 0.28 47400. 268.4 0.257 204.90 5.34 1.35 529.0 1.46 4.34 6.76 10.85 13.00 502.7*

789.B S2HF-F TZOOFF 4040 8.71 3.97 177490. 268.6 0.219 13.48 4.97 1.35 450.7 1.44 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 Changed:

__=== __ ----==__ _--

Value Fisure of Merit Nomenclature!

FCRSLU Fraction of total core mass which must melt 1 Time of containment failure to reach seePo r t e1 a t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.5 -

2 Time between reactor vessel failure and TTCSP Time to fait sueeort elate after corium Pile containment failure has reached i t (h r) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0.0333 3 Integrated wall condensation (measure of FHT Fraction of maximum heat transfer eermitted diffusioehoresis) 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 Flat Plate CHF critical velocitu coefficient..... 0.14 5 Fraction of clad reacted in-vessel 6 Rate of chanse of containment Pressure Just erior to the time of containment failure 7 Time of ice depletion (if aeelicable) 8 Time of core uncovers 9 Hudrogen senerated 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 completion 13 Hudroden mass at time of containment failure f O ()alElg43rar 2 I

i

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Table 5.8 Fisure of Herit Summarv for HAAP Uncertaintw/Sensitivits Analvsis PWR VERSION 1.2 SEQUOYAH SEDUENCE S2HF (FAI)

MODEL FIGURE OF MERIT

. CASE PARAMETER VALUE 1 2 3 4 5 6 7 8 9 10 11 12 13 (hr) (hr) (Iba) (F) (esi/hr) (hr) (hr) (lba) (hr) (hr) (hr) (lba)

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 >8.99 522.0 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 3.9S >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 4.4S >9.06 511.0 SS2HF-6 SCALH 0.5 7.62 3.09 111250. 0.0 0.178 15.22 4.71 1.33 3s8.0 1.44 4.5; 11.05 518.0 SS2HF-7 FunIN 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 IEXIII 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 Changedt Value Fisure of Merit Nosenclaturel

__-- ____________ ________==-- -_____________

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 .

CbA Flame buovanew dras coefficient in comeartment A.. 5.0 containment failure Cl:B Flame buovanew dras coefficient in comeartment 11... 5.0 3 Integrated wall condensation (measure of LbC Flame buovanew dras coef ficient in compartment C.. 5.0 diffusicehoresis) between reactor vessel CDD Flame buovancu dras coefficient in compartment D.. 5.0 failure and containment failure CLU Flame buovancs dras coefficient in comeartment U.. 5.0 4 Peak containment outer wall surface SCALU Scaling factor for all burntns velocities......... 1.0 temeerature SC6LH Scalins factor for heat transfer coefficients to -

5 Fraction of clad reacted in vessel Passive heat sinks................................ 1.0 6 Rate of change of containment eressure Just

. FUMIH hintsum burn velocits (must be in SI units)........ 1.0 error to the time of containment failure

' IEXIII Ice condenser erinars side e>:it teneerature (F).... 100  ? Time of ice deeletion (if aerlicable) 8 Time of core uncovers 9 Hudrosen generated 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.

-o (3 Ell EHg O r y 2 13 Hudrosen mass at time of containment failure I

4

' * . s\

CONTAINMENT LOADS ISSUES FROM THE MAY 15th - 16th IDCOR/NRC INTERACTION MEETING Robert E. Henry Fauske & Associates, Inc.

16WO70 West 83rd Street Burr Ridge, lilinois 60521 (312) 323-8750 NRC/IDCOR Meeting on Integrated Analysis of Severe Accident Fission Product Behavior Rockville, Maryland '

August 28 - 29, 1984 J

l i

I

~n IN-VESSEL HYDROGEN PRODUCTION i

e Consider b5unds of outer cladding surface and outer plus inner surface available for oxidation.

Upper bound results in about a 50% increase in H generated before core slump.

2 e Upper bound (twice cladding surface area) results in about twice the H2 generated following core slump.

e Water swell covering damaged regions can also increase the H2 generated before vessel f ailure.

l I

e The IDCOR results are insensitive to these bounds for in-vessel hydrogen generation.

N

~

i i e,

~

HYDROGEN COMBUSTION BEHAVIOR Considered issues related to H2 reaction at high temperatures and high H2O partial pressures, flame

_ temperature and H 2 fl w path to and from the RC. _

e While other reactions may occur at high temperatures, l there is no apparent cutoff due to high steam partial pressures.

e incorporated the Westinghouse flame temperature correlation based upon measurements with high steam partial pressures. Containment responses are insensitive to large variations in the flame .

temperature.

e incorporated natural circulation between the primary system and containment after vessel failure including H-2 e Containment analyses do not demonstrate H 2 concentrations sufficient for detonation in any compartments.

e s .,

y IN-VESSEL FUEL-COOLANT INTERACTIONS (HYDROGEN GENERATION) e Calculate the steam generation by debris slumping into the lower plenum.

o Calculate the level swell of lower plenum water due to steam generation.

4 i

e Calculate the overheated regions which can be covered by the level swell.

e Calculate additional oxidation due to the level swell.

1 k I

,, 0 h

b

~

[ __

CORE-CONCRETE INTERACTION IN MARK I BWRs e Considered'Several Different Debris Configurations

- uniformly distributed within the pedestal

and drywell .

l - distributed over the pedestal floor and one-fourth of the drywell floor

- restricted to the pedestal l

  1. No Significant Changes for Time to Containment Failure or Calculated Releases to the l Environment l

m

9 ,

m SENSITIVITY AND UNCERTAINTY ANALYSES

  • Uncertainty analyses were performed with MAAP 1.2 on. individual parameters (models) for thermal-hydraulics. Results presented at Meeting 3A.
  1. .No large influences found with respect to whether containment f ailure occurs and if so the time of occurrence. '

k

  1. Additional uncertainty analyses have been performed with MAAP 2.0 - integrated fission product release and deposition.

m Uncertainties considered Sequence definition /

Vapor pressures of volatile fission products Aerosol sedimentation J

-~

y .. ,.

d' DISTRIBUTION

_-\

NRC/NRR NRC/RES ACRS H. R. Denton R. Minogue -

R. Tripathi E. Case 0. Bassett S. Seth R.'Mattson G. Arlotto R. Cushman R. Houston R. Curtis J. Hulman T. Walker R. Vollmer G. Marino Z. Rosztoczy , J. Glyr.n J. Rosonthal W. Morrison J. Shapaker J. Han C. Tinkler J, .. R. Wright R. Palla N W. Lyon ~

M. Sfiberberg TEC

~~

A. El-Bassioni T. Lee R. Barrett C. Peabo'dy A. Buhl P. Niyogi M. Fontana T. Eng E. Fuller Other NRC B. Agrawal J. Carter R. Meyer H. Mitchell J. Conran, DEDR0GR B. Burson S. Asselin M. Taylor DEDR0GR J. Larkins K. Meyer J. Telford P. Baranowsky EPRI DOE FAI M. Leverett F. Witmer R. Sehgal H. Fauske D. Squarer l R. Henry

! J. Gabor Bechtel M. Kenton NUS S. Blazo P. O'Reilly P. Fulford l J. Siegel, AIF PDR l

l .

l Ol3V

'. ~ s .. ,,

~

Sandia Nat'l Lab. EGG-Idaho D. Moore, EI D. Dahlgren S. Behling -

J. Young, El M. Berman R. Gottula S. Thompson M. Lloyd, Middle South R. Cole ,

Services J. McGlaun Battelle Columbus D. Paddleford, W J. Hickman P. Nakayama, Jaycor K. Bergeron P. Cybulskis L. Azarello, Duke Power D. Kunsman R. Denning W. Mims, TVA *D. Aldrich M. Cosella, Coned J. Sprung Oak Ridge Nat'l Lab J. Meincke, CPC0 J. Walker W. Iyer, NYPA J. Griesmeyer S. Hodge A. Marie, PECO F. Harper G. Krueger, PECO D. Powers

'R. Smith, Scandpower V. Behr L. Engstrom, OKGAB Sweden J. Linebarger C. Ader, Stone & Webster S. Dingman L. Rib, LNR Associates A. Camp J. Metcalf, Stone & Webster A. Benjamin Brookhaven Nat'l Lab. "

W. T. Pratt M. Khatib-Rahbar R. Newton T. Ginsberg G. Greens M. Corradini, Univ. of Wisc.

I. Spiewak, American Physical Society S. Niemczyk, UCS T. Theofanous, Purdue University ,

l

.. . _ . __ . _._

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