ML20214W872
| ML20214W872 | |
| Person / Time | |
|---|---|
| Issue date: | 03/20/1984 |
| From: | Silberberg M NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | NRC |
| Shared Package | |
| ML20213E209 | List:
|
| References | |
| FOIA-87-113, FOIA-87-60 NUDOCS 8706160223 | |
| Download: ML20214W872 (1) | |
Text
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--:.2.w a.wiiw ~ a av y e# N UNITED STATES y* 'g NUCLEAR REGULATORY COMMisslON u
- wassiNoToN. o. c. 2 ossa e
(...../ MAR 2 01984 ,
MEMORANDUM FOR: Distribution FROM: M. Silberberg, Assistant Director for Research and Technical Support Accident Source Term Program Office, RES
SUBJECT:
IOCORRE$RTS Enclosed is a sumary of the IOCOR reports on sequence selection, fission product characteristics, hydrogen generation, hydrogen combustion, core melt phenomena, vessel failure, and debris quenching. These subjects were discussed at the NRC/IDCOR meeting held at Harpers Ferry (24 November - 31 December 1983) and at Hunt Valley (30, 31 January 1984). The summary is organized into four sections:
- 1. Reports in the sumary and an index
- 2. Level #1 sumary - general topics covered in the reports
- 3. Level #2 summary - subtopics defined under each topic in level #1
- 4. Level #3 sumary - phenomena discussed under each subtopic in level #2 The terms " topics," " subtopics," and " phenomena" are a convenient way to sumarize the IDCOR reports. The order of the topics more or less correspond to the order of the plants systems that fission products
- pass through during a severe accident. For further assistance, please contact C. Ryder, (301) 427-4737 of my staff.
) l( YY1 M. Silberberg, Assistant rec tor for Research and Techni'c Su pport Office of Nuclear Regulatory Research Accident Source Term Program Office
Enclosure:
As stated i _
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- j WASHINGTON, D. C 20555
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MAR 0 4 1986 MEMORANDUM FOR: Themis P. Speis, Director Division of Safety Review and Oversight FROM: Zoltan R. Rosztoczy, Chief Regulatory Improvements Branch Division of Safety Review and Oversight
SUBJECT:
MINUTES OF DECEMBER 10, 1985 NRC/IDCOR MEETING ON THE PEACH BOTTOM REFERENCE PLANT ANALYSIS The purpose of the December 10, 1985 NRC/IDCOR meeting was to describe preliminary results of the NRC Peach Bottom risk analyses, to discuss specific technical differences between the NRC and IDCOR analyses and to identify issues requiring further attention by NRC and IDCOR.
The NRC contractor presentations included the calculations of core melt frequency, containment response, fission product behavior and offsite consequences. IDCOR summarized some recent updates of their Peach Bottom analyses. (
Trevor Pratt of BNL presented a comparative analysis of the NRC and IDCOR calculations, and identified several specific sources of disagreement between the NRC and IDCOR results. The remainder of the meeting was devoted to discussion of each of those issues.
The principal difference between NRC and 10COR estimates of core damage sequence The IDCOR estimate, based ASEP estimateon a rebaselining of for NUREG-1150 is 8x10 GASH-1400,while the is 5x
. About one third of the ASEP estimate results from a postulated failure of the emergency service water (ESW) system.
Peach Bottom is currently cooperating with ASEP to clarify the ESW vulnerability issues.
About 40% of the ASEP estimate of core damage frequency for TB is related to assumed common cause failures. Further analysis of the common cause failure assumptions are considered to require a level of effort which is out of pro-portion to the expected gain.
A major difference between IDCOR and NRC risk assessments for Peach Bottom has been the assumed likelihood of successful wetwell venting. The NRC staff described an ongoing assessment of controlled venting, to determine the engineering feasibility human reliability and risk reduction potential. The Philadelphia Electric Company (PECO) has extended full cooperation with this effort, and both parties anticipate a satisfactory resolution of the issue by March 1986.
CONTACT: R. Barrett, X-27591 p ,' -
h 4 O/5/
s T. Speis MAR 0 4 1986 NRC contractors did not present details of the containment event tree or its quantification, because the current results are only preliminary. NRC contractors agreed to supply documentation of the unquantified event trees to IDCOR for comment by the end of January, 1986. IDCOR informed NRC of a new analysis of containment performance by the Chicago Bridge and Iron Company (CB&I), which showed more favorable response to high temperatures than previously calculated. IDCOR agreed to brief NRC on the methods, assumptions and results of the CB&I analysis, and to supply the final report by the end of January, 1986.
NRC and IDCOR estimates of fission product behavior differ in three important aspects: in-vessel revaporization of volatile fission products following vessel failure, release of non-volatiles during core-concrete interactions and fission product deposition in the secondary containment. A fourth issue is the recent evidence that significant quantities of gaseous iodine may exist in-vessel prior to vessel failure. Currently, both NRC and IDCOR assume that Iodine is chemically bound with Cesium.
Both NRC and IDCOR predict low (410%) releases of volatile fission products to the environment for most sequences. However, the mechanisms of retention differ dramatically. The NRC calculations predict permanent retention of a large fraction (as much as 85%) of volatiles in the primary system, and relatively less retention in the secondary containment. By contrast, IDCOR predicts very little long term retention in the primary system (410% for station blackout), but significant retention (380% for station bTackout) in the secondary containment. Although the overall risk predictions are comparable, it was recognized that a combination of the IDCOR revaporization results with the NRC prediction of minimal secondary containment retention would predict greatly increased releases to the environment. No specific plans were made to resolve these issues.
NRC predicts significant releases of non volatile fission products (Strontium and Lanthanum) for the leading accident sequences, whereas IDCOR predicts minimal releases. The primary difference is in the modelling of core-concrete interactions. The IDCOR assumptions of gradual release of fuel from the reactor vessel, spreading and cooling of core debris and quenching in the presence of water lead to low core debris temperatures and gas flow rates. The NRC calculations, based on coherent exit of core from the vessel, minimal upward heat transfer and relatively confined geometry, lead to higher debris temperatures, higher gas flow rates and consequently higher releases of non-volatile fission products. Given the early predicted containment fa11ure times and minimal retention of radionuclides in the secondary containment, the predicted environmental releases in the station blackout sequences are 35-50%
for Strontium and 3% for Lanthanum. No specific plans were made for resolving this issue.
o
" 'I T. Speis The issue of gaseous Iodine concentrations in vessel due to high radiation
-fields was discussed briefly. The experimental evidence is preliminary and NRC is currently conducting a peer review to determine the significance of these results to risk. NRC will keep IDCOR informed of the results.
It was tentatively decided that the next NRC/IDCOR meeting would occur in February, 1986 and will cover the subject of sensitivity and uncertainty analysis.
zA.:T272A Zoltan R. Rosztoczy, Chief Regulatory Improvements Br nch Division of Safety Review and Oversight
Enclosures:
- 1. Meeting Agenda
- 2. List of Attendees
- 3. Copies of Handouts cc: See next page t
a
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T. Speis MAk 0 41 90 cc: NRC PDR PRC System R. Bernero M. Ernst M. Cunningham J. Carter, IDCOR T. Pratt, BNL D. Houston, ACRS W/o Enclosure 3 H.-Denton F. Gillespie D. Eisenhut G. Arlotto i H. Thompson M. Silberbeg l F. Miraglia R. Curtis B. Sheron S. Niemczyk R. Houston C. Reed, IDCOR T. Dorian A. Buhl F. Coffman R. Henry L. Soffer M. Leverett J. Rosenthal D. Helwig F. Eltawila E. Warman R. Sammons P. Hill S. Sands- A. Buslik R. Palla R. Meyer J. Chen J. Mitchell i V. Leung G. Marino G. Sege D. Pyatt J. Read T. Margulies J. Lane J. Martin D. Ross H. Ashar
- 0. Bassett R. Denning, BCL
, J. Hickman, .SNL A. Benjamin A. Camp. E. Haskin F. Harper
Agenda for the December 10, 1985 NRC/IDCOR Meeting 8:30 Opening Remarks: NRC: T. Speis IDCOR: A. Buhl 8:45 Current Results of the IDCOR Analysis of Peach Bottom: J. Gabor, FAI 9:15 Summary of NRC Calculations of Peach Bottom: SNL 10:15 Break.
10:30 Preliminary Conclusions for Peach Bottom: T. Pratt, BNL 11:15 Containment Venting: J. Jenkins, NRC 12:00 Lunch 1:00 In-Vessel Issues: Fission Product Generation, Retention and Release:
T. Pratt, BNL/J. Mitchell, NRC 1:45 Ex-Vessel Fission Product Generation: D. Powers, SNL 2:30 Core / Concrete Interaction and Fission Product Chemistry: M. Plys, FAI 3:00 Break l
l 3:15 Timing and Mode of Containment Failure, Secondary Containment I
Performance: R. Denning, BCL l
3:45 Hydrogen Generation and Burn: R. Henry, FAI J. Gabor, FAI 4:15 Secondary Containment Performance:
4:30 Summary,-Follow-up Actions: NRC: Z. Rostoczy
, IOCOR: R. Henry l
1 l
NRC/IDCOR Meeting On Peach Bottom Sequence Analyses 9
Name Affiliation Telephone R.-J. Barrett. NRC/NRR/DSR0 301-492-7591 M. L. Ernst NRC/RES/DRA0 301-443-7923 T. P. Speis NRC/NRR/DSR0 301-492-7517 Farouk Eltawila NRC/NRR/DSR0 301-492-9488 Jim Jenkins NRC/RES/DRA0 301-443-7695 J. A. Murphy NRC/RES/DRA0 301-443-7921 E. T. Burns Delian Corp. 408-446-4242 A. R. Diederich PECO 215-841-4516 K. W. Holtzclaw GE 408-925-2506 0.- R. Helwig PEC0/BWROG 215-841-4542 M. C. Leverett EPRI/IDCOR 415-855-2936 M. H. Fontana IT-ENERGER (IDCOR) 615-481-3300 Tony Buhl .IDCOR/IT' 615-481-3300 R. E. Henry FAI/IDCOR 312-323-8750 J. C. Carter IT Corp /IDCOR 615-481-3300 A. S. Benjamin Sandia 505-844-5960 F. E. Haskin Sandia 505-846-0276 Bob Youngblood Brookhaven 516-282-2363 Eric Haskin Sandia 505-846-0276 Jack Hickman Sandia 505-844-3876 Martin Plys FAI 312-323-8750 Alan Kolaczkowski S.A.I.C. 505-846-8787 Fred Harper SNL 505-846-1975 David Pyatt NRC/RES/DRA0 301-443-7629
e a
Cond't Sheet NRC/IDCOR Meeting On Peach Bottom Sequence Analyses NAME AFFILIATION TELEPHONE Tsong-Lun Chu BNL 516-282-2389 Nam Cho BNL 516-282-2226 J. Yang BNL 516-282-2616 Hossein Nourbakhsh BNL 516-282-2484 M. Lee BNL 516-282-2014 Vincent Leung NRC/ RIB 301-492-9401 J. E. Rosenthal NRC/RRAB 301-492-9447 A. J. Buslik NRC 301-492-8058 Hans Ashar NRC 301-443-7892 Tim Margulies NRC/RES 301-443-7627 Dana A. Powers Sandial Nat'l Labs. 505-844-4392 Rebecca Green NJBRP 609-984-4169 Ralph Meyer NRC/RES 301-427-4461 Leonard Soffer NRC/ RIB 301-492-8298 Edward Warman Stone & Webster 617-589-6510 James Metcalf Stone & Webster 617-589-1499 F. D. Coffman NRC/ RIB 301-492-7497
- 5. J. Niemczyk 202-328-8113 John T. Chen NRC/ RIB 301-492-4921 Chris Ryder NRC/RES/FSRB 301-427-4331 Jeff R. Gabor FAI/IDCOR 312-323-8750 E. L. Fuller EPRI 415-855-2115
-P. R. Hill Pennsylvania Power & Light 215-770-7949 Harold Mitchell IT Corp (IDCOR) 615-481-3300 Stephen P. Sands- NRC/DSR0 301-492-8483 Mohsen Khatib-Rahbar BNL 516-282-2626 William J. Luckas, Jr. BNL 516-282-7562 l
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SUMMARY
OF CURRENT MAAP ANALYSES J. R. GABOR FAUSKE a ASSOCIATES, INC.
16U070 WEST 83RD STREET ,
BURR RIDGE, ILLIN0IS 60521 DECEMBER 10, 1985
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MAdP ADVANCEMENTS SINCE TASK 23.I REPORT e Core Melt Progression e Core Radiation e Improved Ex-vessel Release Model e Plugging.Model i
e Improved Aerosol Correlation e Aux Code Integration into MAAP e Lorger Structural Mass in Drywell e
Leak - Before - Break at Containnent Fnilure O
fe 4'
[
CURRENT MAAP ANALYSIS OF PEACH BOTTOM STATION BLACK 0UT SEQUENCE Fission Product Release Task 23.1 Current Noble Gas 1.0 .22 Csl .05 .03 Te0 .06 2 .01 SrS103 < 10-5 < 10-5 K2 N04 -
.001 Cs0H .05
.03 Event Timinas i
Vessel Failure (Hrs.) 12 11 Containment Failure
- 19 29 End of Calculation 80 80
- Drywell overtemperature failure for Task 23.1 drywell overpressure failure for current analysis.
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MAAP ANALYSES OF PEACH BOTTOM TC SEQUENCE Sensitivity' Studies Performed to Investigate:
- 1) Omission of Rx Building Retention
- 2) No SGTS Isolation
- 3) I and 2 Node Reactor Building
- 4) Natural Circulation Within Rx Building Node
- 5) Steady State vs. Decoy Aerosol Removal A
- 6) Drvwell Sprays Y
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Table 2 '
FISSION PRODUCT RELEASE FOR PEACH 80 TION TC-y MAAP/ Reactor Butiding Modeled SGTS Isolated (Fire Dampers) t BMI 2104 1 (Draft July 1984) 2 Node 1 Node RX Sullding RX Sullding TC-y' TC-y BMI 2104 Assumptions (Without (RX Sidg. (No SGTS No Natural Natural RX Bldg.) Natural Modeled)* Isolation)' Circulation Circulation Circulation (1) (1) (2) (2) (2) '
i (2)
+
Noble Gas s 1.0 s 1.0 1.0 1.0 1.0 1.0 l
Csl .24 .10 .21 .08 .04 .04 -
Te0 .37 .25 .23 '
2 .04 .02 .01 j SrSiO -
7 x 10 4 3
1 x 10 8 x 10 3 x 10 K Mo0 !
2 4
.17 .03 .02 .01
- l Csoll .21 .09) .21 .08 .04 .04 1 (1) Releases at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> (complete) .
(2) Releases at 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> (essentially complete) f i
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Table 3 INFLUENCE OF MAAP UNCERTAINTIES AND OPERATOR ACTIONS * .
Uncertainty Bounds Vent & Plug on Aerosol Removal in RX Butiding Size Dependent DF Steady-State A Steady-State 1 Decay A Drywell i Decay A for Pool DF (1) for Pool DF Sprays !
! (1) (2) (3) l (4)
Noble Gas 1.0 1.0 1.0 1.0 .97
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Te0 .01 2 .004 1 x 10~3
- 1 x 10-4 10-5 ;
s,s10, 3 x iO-s > x 0-s . iO-> . iO-s l K Ho0 2 4 .01 .002 4 x 10-4 3 x 10-5 10-5 Cs0H .04 .02 2 x 10-3 1 x 10'4 g x 10'4 j
- All runs with SGTS isolated. 2 node reactor building and natural circulation.
( 1) Releases at 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> fessentially completel
" t,2 Releases at 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> I;essentiallycompletej
[3 Releases at 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> tessentially completes (4) Releases at 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> { essentially complete) k J :
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i Preliminary Results 3 NUREG 1150 PRA Updates l Frontend Analyses .
l i
Methodology -- F. Harper -- SNL j Peoch Bottom Results -- A. Koloczkowski -- SAIC i t
-1 December 10, 1985 i 9
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We are doing PRA~ updates on 6 plants
- 1) Surry, Peach Bottom, Sequoyoh, Grand Gulf --
some methodology -
2-) Lo Salle, Zion -- different methodology e
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plant visfie 1
We started with a ' plant visit f
- 1) Good cooperation from plants -
- 2) Second plant visit to present preliminary results 9
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plant visits
.We took 6-10 people to each plant
- 1) 2-4 systems analysts -
- 2) i-4 human reliability specialists '
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- 3) 1 dato expert '
- 4) I containment analyat
- 5) I week visit
- . :: mv o e i l .
s a 6-
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i scope .;
The detail of our analysis is:
> RSSMAP
< RMIEP .
> < IREP _
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1
- 1) Plant analyzed "as is" (systems and procedures)
- 2) Event Tree, Fault Tree analyses
. 3) Actuation and control - abbreviated ,
- 4) Human reliability analysis - simplified analy. sis, l
detailed analysis for ATWS ~
- 5) External and special events -
c not included
- 6) Subtleties - spot checked .
I
- 7) Latest available success criteria --
some i I thermal / hydraulic calculations made ,
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- 2) Within systems - not across systems 1
1 3) Plant specific values are used if they i are higher l
l 4) Values are assumed to be upper bounds :
- 5) Conservative or non conservative???? 9 1., .
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PEACH BOTTOM ANALYSIS RESULTS I
i
{ PRESENTED BY: ALAN M. KOLACZK0WSKI '
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! SCIENCE APPLICATIONS INTERNATIONAL CORPORATION l
l CONTRACTOR T0:
l SANDIA NATIONAL Lf36RATORIES i
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PRESENTATION OUTLINE "
o STATUS OF ANALYSIS f i
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SUMMARY
OF RESULTS :
o COMPARIS0N OF RESULTS TO IDCOR AND WASil-ll100 l 1
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STATUS OF ANALYSIS 0 PRELIMINARY POINT ESTIMATE ANALYSIS NEARLY COMPLETE
~
- 75% DONE. -
V SI'0DENCE (EXPECT TilESE TO BE SMALL C0itTRIBUT10NS) .
o CORE VllLNERABLE SE00ENCES BEING ANALYZED FOR' CORE MELT CONTRIBUTION 0 RE-VISITING PORTIONS OF ANALYSIS O STARTING UNCERTAINTY / SENSITIVITY ANALYSES i .
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DOMINANT CORE MELT ACCIDENT SEQUENCES '
(CORE MELT BEFORE CONTAINMENT FAILURE)
SEQUENCE (
FREQUENCY '
% OF TOTAL T
STATION BLACK 00T
- IOSS OF C0RE COOLING IN 6 HOURS 5.lE-6 1 AFTER BATTERY DEPLLiiaN
(
77 T
STATION BLACK 0UT
- IIPC1/RCIC FAILURE CHORT TERM)
[
2.5E-6 1
~
i ATWS
- SLC FAILURE
- L.P. FAILURE (OPERATOR ERl!0R) 9.0E-7 ATWS
- IIPCI FAILURE
- L.P. FAILURE (OPERATOR ERROR) 1.0E-7 1
)
[
i T l STATION BLACK 0UT
- STUCK QPEN RELIEF VALVE
- LOSS OF 2.6E-7 ) !
CORE C00LIN6 IN 6 Il00RS
( !!
T STATION BLACK 0UT
- STUCK OPEN RELIEF VALVE
- HPCI/RCIC FAILURE 1.2E-7
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5.8E-7 6 l j
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- 9.9E-6 100 !
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DOMINANT CORE VULNERABLE ACCIDENT SEQUENCES (POTENTI L CONTAINMENT FAILURE WITH CORE COOLING INITIALLY) _
SEQUENCE FREQUENCY % OF TOTAL T2,3 CONTAINMENT COOLING FAILURE (CRD/HPSW TYPICALLY AVAILABLE) 2.1E-5 86 MEDIUM LOCA
- CONTAINMENT COOLING FAILURE (CONDENSATE /CRD/HPSW
~
1.5E-6 6 AVAILABLE)
,g ,
STUCK OPEN VALVE (WnH PCS UNAVAILABLE)
- CONT AINMENT COOLING 7.9E-7 8, 4 ' 3 i
FAILURE (CRD/IIPSW AVAILABLE) /
- SLC FAILURE
- CONTAINMENT COOLING INADEQUATE 6.4E-7 3' i
OTHER 5.6E-7 2 T0T AL
T3 = TRANSIENT Wi1H SUBSEQUEN1 LOSS w PCS ;
NOTE: EXPECT DOMINANT CORE VULNERABLE SEQUENCES TO YlELD S MID-E-8'OR LESS TO CORE MELT.
(BEING ANALYZED NOW.)
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MAJOR CONTRIBUTORS TO DOMINANT CORE MELT SEQUENCES j SEQUENCE CATEGORY APPROX. I CONTRIBUTORS CONTRIBUTION T
STATION BLACK 0UT (LONG TERM) ESW UNAVAILAB _E
- OPERATOR FAILURE 55 (EMERG. HT S NK) '
DIESEL GENERATOR FAILURES 30 (INCLUDING COMMON CAUSE) d-G1 e <W ~'/d QN T
STATION BLACK 0UT '(SHORT TERM) BATTERY COMMON CAUSE 88 ATWS SLC Fall TO START OR UNAVAILABLE 70
- 0PER. ERROR HPCI FAILURE
- OPER. ERROR 30 i
! OTilER ESW UNAVAILABLE
- OPER. FAILURE j (EMERG. HT SINK)
LPCl/HPSW COMMON CAUSE -
MISCALIBRATION OF LOW Rx PRESS SENSORS l
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MAJOR CONTRIBUTORS' TO DOMINANT CORE VULNERABLE SEQUENCES APPR0X. %
SEQUENJE CATEGORY CONTRIBUTORS CONTRIBUTION Tr o-TYPE LPCl/HPSW COMMON CAUSE 55 l ESW UNAVAILABLE
- OPER. FAILURE 43 '
(EMERG. HT. SINK)
{
MEDIUM LOCA . <A.
LPC1/HPSW COMMON CAUSE 55
. -] > '
ESW UNAVAILABLE
- OPER. FAILURE Y 43 j (EMERG. HT. SINK STUCK OPEN VALVE LPCl/HPSW COMMON CAUSE % !
ATWS SLC Fall TO START OR UNAVAILABLE 100 !
i; 0 tiler LPCl/HPSW COMMON CAUSE
, ESW UNAVAILABLE
- OPER. FAILURE (EMERG. IIT. blNK),
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i REASONS FOR MAJOR CONTRIBUTORS
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- o MULTIPLE SYSTEMS FOR CORE COOLING
,o PROCEDURE IMPROVEMENTS (EPGs + TRIPS) o PLANT SYSTEM IMPROVEMENTS i
j TilEREFORE. NOT UNREASONABLE THAT THE MOST VULNERABLE AREAS CONSIST OF THOSE FEATURES WHICH COMPROMISE THE "INDEPENDENTNESS' 0F Tile PLANT SUPPORT SYSTEMS (ESW, POWER) c COMMON CAUSE FAILURE OF COMPONENTS HUMAN i i
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. COMPARISON OF CORE I ..
' MELT SEQUENCE CONTRIBUTIONS l
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! 80 _ hC' secon l 70 .. WASH-1400 1
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HUMAN ERROR ESTIMATES
~
FOR MSIV CLOSURE.ATWS HUMAN ACTION NRC-1150llE IDCOR 112 FAILURE TO START SLC 0.05 - 0.1
'. (wfsW} f~$ B'& -
0.3 FAILURE TO CONTROL WATER LEVEL 0.1 - 0.2 .
(HIGil PRESSURE)
FAILURE TO INHIBIT ADS 0.015 - 0.03
- l l
FAILURE TO MANUALLY DEPRESSURIZE 0 . 1 11 FAILURE TO CONTROL WATER LEVEL 0.2 - 0.5 0.5 ,
(LOW PRESSURE)
. j
- NOT ADDRESSED SEPARATELY IN TASK 21.1, APP. A i
I
- a si
~
ANALYSIS FEATURES CONTRIBUTING TO DIFFERENCES IN RESULTS ll 1
j SEQUENCE IDCOR CATEGORY NRC-1150 (COMMITTED)
WASil-1100 TC o ATWS RULE o ATWS RULE o PRE-ATWS RULE o EPGs o EPGs I
o SIMPLE ANALYSIS
. . o DETAILED ANALYSIS o SEMI-DETAILED ANALYSIS l .
TB o CONSIDERATION OF LONG TERM o CONSIDERATION OF LONG o SIMPLE ANALYSIS FAILURES OF HPCI-RCIC TERM FAILURES OF HPCI-RCIC o COMMON HODE o ESW UNAVAILABILITY i
TW o CONSIDERATION OF VENTING AND l o CONSIDERATION OF VENTING o CONTAINMENT FAILURE CORE COOLING SURVIVABILITY AND CORE COOLING
= CORE MELT SURVIVABILITY o N0 VENTING OTHER o COMMON MODE o LATEST SUCCESS CRITERIA o CONSERVATIVE SUCCESS j o ESW UNAVAILABILITY CRITERIA o ONLY MAJOR COOLING o CONSIDERED MORE CORE COOLING SYSTEMS CONSIDERED SYSTEMS o LATEST SUCCESS CRITERIA I
0' *
.. o-PEACH BOTTOM CONTAINMENT AND CONSEQUENCE ANALYSIS METHODS FOR
, NUREG-1150 BY ERIC HASKIN -
MIKE GRIESMEYER DAN ALPERT CHRIS AMOS DECEMBER 10, 1985 ALLOTTED IIME: 15 MIN.
e e
- . _ , . . - _ . _ _ _ _ _ _ . -, _ , _ , _ . _ , . . _ _ _ _ _ , . . _ . , ,,.__,..,...____,,____..__.5-. _ . . _ _ , _ _ - . _ .
, RISK EDUATION RISK g - b E FRE0 7 CRMP 3,3 g W ,IJ )
I J CONS
~
SYMBOL DEFINITION SOURCE I FRE0 FREQUENCY OF CORE MELT ACCIDENT SEQUENCE I ASEP. IN COORDINATION WITH SARRP 7
CRMP y ,y PROBABILITY OF CONTAINMENT RELEASE SARRP CONTAINMENT EVENT ANALYSIS MODE J. GIVEN ACCIDENT SE00ENCE I FP FISSION PRODUCT SOURCE TERM FOR BIN TO BCL STCP CALCULATIONS BINNING g,, .
WHICH SE00ENCE I WITH RELEASE MODE J MEETINGS. AND SARRP ANALYSES IS ASSIGNED l g W ,IJ) MHWWMMNr,ma M.SENQNimm l
CONS L SOURCE TERM FP O
! IJ ;
l l e I e I *
- i. _
PEACH BOTTOM INTERFACE BETWEEN ASEP AND CET.-
FOR PEACH BOTTOM CET TEAM IS PARTICIPATING IN CORE DAMAGE SENSITIVITY / UNCERTAINTY ANALYSES USING TEMAC CODE INITIAL QUESTIO.VS ON CET USED TO TRANSFER INFORMATION REGARDING ASEP CUT SETS:
S 4
9 e
e 5
h - i
GENERALIZED CONTAINMENT EVENT TREE (SCHEMATIC) I INPUTS OUTPUTS .
PLANT DAMAGE 1. CORE MELT OR SUCCESSFUL STATE . RECOVERY (LINKED TO ASEP)
DEFINITI.ON 3. CONTAINMENT RELEASE 110DE
& FREQUENCIES FREQUENCY 2. SOURCE TERM BIN FREQUENCIES QUESTIONS ASKED ABOUT
- OPERATOR ACTIONS, E.G.
- IS THE VESSEL DEPRESSURIZED?
- ARE DRYWELL SPRAYS ACTUATED?
- IS THE CONTAINMENT VENTED?
- EVENTS & TIMING, E.G.
- CORE DAMAGE BEFORE CONTAINMENT FAILURE?
- CONTAINMENT FAILURE BEFORE, AT, OR AFTER VESSEL BREACH?
- PLANT CONDITIONS, E.G.
- WHAT IS THE CONTAINMENT LEAKAGE LEVEL?
- WHAT IS THE POOL BYPASS FLOW LEVEL?
1
. - PHENOMENA, E.G.
- CONTAINMENT FAILURE PRESSURE?
- CONTAINMENT PRESSURE RISE DUE TO H 2 BEFORE VESSEL MELTTHROUGH
+ HYDROGEN BURNS IN REACTOR BUILDING?
5 e
e 9
---,--,.--.,,-----i,,,--,-,.,.-~~-----,-y ~---.,~.---,--y-- .
-c ,-,v.- .---,--,w-. - , - .-.-,,e.,-,.,-,~.-----_-----------,y- .-v -m.-y- ,-- - -
SOURCES OF INFORMATION ~~~-
~
NRC-SPONSORED STUDIES
- CONTAINMENT LOADS WORKING GROUP (CLWG)
- CONTAINMENT PERFORMANCE WORKING GROUP (CPWG)
- SEVERE ACCIDENT SE'QUENCE ANALYSIS (SASA) PROGRAM BATTELLE ANALYSES FOR ACCIDENT SOURCE TERM PROJECT OFFICE (BMI-2104)
- REACTOR SAFETY STUDY (RSS) AND SUBSEQUENT RISK ASSESSMENTS
- STEAM EXPLOSION REVIEW GROUP (SERG)
- GENERIC SAFETY ISSUE STUDIES (TAP _A-43, A-44, A-45, ATWS)
REPORTED IN NUREG DOCUMENTS l
UTILITY i INDUSTRY-SPONSORED STUDIES l
- FULL SCOPE PRAs INDUSTRY DEGRADED CORE (IDCOR) PROGRAM
- FINAL SAFETY ANALYSIS REPORTS (FSARS)
- PLANT-SPECIFIC EMERGENCY OPERATING PROCEDURES l .
STATION BLACK 0UT DATA AND ANALYSIS (EPRI) l
[ -
- l. -
l- .
TIME STAGES ON PEACH BOTTOM CET E1 - INITIATING EVENT AND INITIAL OPERATOR / SYSTEM RESP 0NSE E2 - THROUGH DEPRESSURIZATION OR LOSS OF HIGH PRESSURE INJECTION E3 - THROUGH DETERMINATION OF CORE DAMAGE E4 - FROM ONSET OF CORE DAMAGE THROUGH VESSEL MELTTHROUGH OR IN-VESSEL RECOVERY ES - THROUGH MELT RELEASE FROM VESSEL L - AFTER E5 b
, *p -
i l
O
i 1
- WHAT IS THE INITIATING EVENT?
A S1 S2/3 T1 T2/3 TC IORV
- WHAT.IS THE INITIAL BREAK LOCATION?
- 0 0F 0FFSITE POWER?
S { RE g
- IS THERE'A STATION BLACKOUT (DIESEL GENERATORS FAIL)?
SB NSB
E1FSLC El-SLC NA
- DOES AN SRV STICK OPEN EARLY?
El-SORV EINSORV
EIFHPC El-HPC
- DOES THE CRD HYDRAULIC SYSTEM INJECT?
E1FCRD ElRCRD El-CRD
+ DO THE LPCS AND LPCI SYSTEMS FAIL?
EIFLPC ElRLPC EIALPC El-LPC
- DO THE RHR SYSTEMS FAIL?
EIFRHR ElRRHR El-RHR
- DOES THE CONDENSATE SYSTEM FAIL?
EIFCOND ElRCOND E1ACOND
+ DOES HPSW FAIL IN A MODE WHICH WOULD PRECLUDE INJECTION?
EIFHPSW ElRHPSW EIAHPSW
= IS THERE A FAILURE WHICH PRECLUDES DRYWELL COOLER OPERATION?
EIFDWC ElRDWC EIADWC
EINADS El-ADS
!
- DOES THE RCS REMAIN PRESSURIZED?
E2NDEP E2-DEP l
I
- FOR TC IS INADEQUATE LEVEL MAINTAINED?
TL TAF/OSC NA
- WHAT TYPE OF SEQUENCE IS THIS (
SUMMARY
OF PLANT DAMAGE)?
I AE AW SIE SIW TQUV TB TW TC-FDEP TC-TL TC-CV O
h
1 STAT,US OF PEACH BOTTOM CET
- TREE STRUCTURE IS ESTABLISHED ~~"
- WORKING WITH ASEP TO RESOLVE WHICH TW, AW, SIW AND TC CUT SETS PROCEED TO CORE DAMAGE
- VERY PRELIMINARY QUANTIFICATION INDICATES TB DOMINATES RISK I
- QUANTIFICATION OF VENTING, DRYWELL SPRAY, AND STUCK-OPEN i
SRV TAILPIPE VACUUM BREAKER QUESTIONS NOT YET COMPLETE I
- PEER-REVIEW CF TREE QUANTIFICATIO,N NOT YET STARTED f , ,
h l
l l
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--t- + -s-.r---r.,sp---,,..,r...-.,,,_. ,,,.,,%.., -,,,,,y,- ym,, ,,..=w.---- . ,, ,,,,, _ _ , - . . , , -. . - - mm m -*n-c. - - - -
~
CON-l Al \lM 1N l VENTlh G <
(Peach Bottom)
Peach Bottom Approach .
- 1. At 60 psig, begin venting wetwell
~
- a. 2" line, open from control room
- b. 6" line, open locally
- c. Deflate 18" valve seals locally I "d.' Open 18" torus vent, preferably from control room
- e. Open 18" tor,us supply, preferably from control room l
- 2. If wetwell venting is insufficient to reduce containment pressure, open drywell vent lines.
. . ... i. . .
- 3. In3 AMS, go directly to 1 d. l
EFFECT OF VENTING ON DOMINANT SEQUENCES ,
,N' '
Sequence Core Domoge p'\'
d' '
Class Credit Rationale -
j.1 TB None + Mitigative, not preventive . measure
. *No containment pressure indication
- High rod. and temp. in torus room
+No power / lighting'
+ Opening 18" and 2" not feasible
- Opening 6" line possible, but would require numerous local operator actions.
IC ,
S m all +For ~.1/2 of scenarious would be mitigative, not preventive measure
- Limited time available for venting
+High stress considerations - -
- Success criteria uncertain
~
i TW Large + Preventive action ,
+Long time available
+ Lower radiation levels than TB h 9
o .-
CONSEQUENCE EVALUATION
' ~
EMERGENCY RESPONSE WITHIN 10 MILE 95% MOVE AWAY AT 8.5 MI/HR STARTING 1.5 HR AFTER WARNING.
5% STAY INDOORS FOR 24 HRS
=
OUTSIDE 10 MILES ALL STAY INDOORS FOR 24 HRS CONSEQUENCE CODES CRAC2
~
MACCS
- CURRENT HEALTH EFFECTS MODEL (NUREG/CR-4214, JULY 1985)
- MULTI-PUFF RELEASE O
t e
e t
= M e ,
- O REVIEW 0F SOURCE TERM RESULTS FOR NUREG-1150 BY Cotunnus Division O
e o
e e
ll 9
RATIONALE FOR FISSION PRODUCT SOURCE IERM QuANTIFICATION
- SOURCE TERM BINNING DIFFERENT ACCIDENT SCENARIOS MAY BE BINNED TO THE SAME SOURCE TERM IF AND ONLY IF THE ,
EFFECT OF THE BINNING ON RISK IS SMALL.
. QUANTIFICATION OF CENTRAL ESTIMATES 1
CALCULATIONS ARE PERFORMED WITH A VALIDATED SOURCE TERM CODE PACKAGE FOR SOURCE TERM BINS WHICH ARE HIGHLY INDIVIDUAL. EXTRAPOLATIONS MAY BE MADE TO OBTAIN SOURCE TERMS FOR OTHER BINS IF THE AMOUNT OF EXTRAPOLATION IS REASONABLY SMALL AND IF THE PilYSICAL BASIS FOR THE EXTRAPOLATION IS SOUND.
QuANTIFICATION OF UNCERTAINTIES i KEY UNCERTAINTY PARAMETERS AND THEIR EFFECTS ON THE SOURCE TERM ARE DETERMINED BY CODE l SENSITIVITY CALCULATIONS. UNCERTAINTIES ASSOCIATED WITH UNMODELED PHENOMENA ARE I l EVALUATED BY APPLICATION OF PHYSICAL PRINC'IPLES.
l
\
.' )
1
BINNING STEPS
- 1. ASEP IDENTIFIES SEQUENCES 2.
SARRP REDUCES EVENT TREE To KEY FAILURE MODES 3.
SARRP PROVIDES PRELIMINARY ASSESSMENT OF NEEDS FOR STCP RUMS 11 .
BINNING TEAM (SNL, BCD, BNL, ORNL, NRC) MEETS TO SPECIFY STCP RUNS 5.
BCD PERFORMS ANALYSES AND PROVIDES RESULTS TO SARRP -
- 6. SNL DEVELOPS BIN CHARACTERISTICS e
O 8 6 gl e
SOURCE TERM %THODOLOGY 1
% THODOLOGY IS ESSENTIALLY THE SAME AS USED IN BMI-2104 WITH SOME IMPROVEMENTS .
SOURCE TERM CODE PACKAGE - INTERIM VERSION MARCH 3 CORS0R INTEGRATED INTO MARCH CONSISTENT TREATMENT OF FP RELEASE FROM FUEL
~
CORCON REPLACES INTER IN MARCH CONSISTENT TREATMENT OF CORE-CONCRETE ATTACK TRAP-MERGE COUPLED ANALYSIS PERMITS TREATMENT OF REVAPORIZATION IN-VESSEL CONSISTENT TREATMENT OF GAS PROPERTIES -
VANESA, NAUA, SPARC, ICEDF SAME AS BMI-2104 EXCEPT INTERFACES REQUIRE LESS USER INPUT QUALITY CONTROL l BNL REVIEW OF STCP l BNL REVIEW OF STCP ANALYSES -
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STCP LIMITATIONS THE STCP HAS LIMITED OR NO CAPABILITY TO EXAMINE SOME POTENTIALLY IMPORTANT PROCESSES:
REVAPORIZATION OF VOLATILE FISSION PRODUCTS AFTER VESSEL FAILURE THE EFFECT OF IN-VESSEL FLOW CIRCutATION PATTERNS (WHICH COULD INFLUENCE HIGH PRESSURE SEQUENCES)
TIME-DEPENDENT RELEASE OF MOLTEN CORE MATERIAL INTO THE REACTQR CAVITY AND ITS SUBSEQUENT DISPERSAL BUOYANCY DRIVEN FLOW BETWEEN CONTAINMENT COMPARTMENTS O
6
i SOURCE TERM RESULTS FOR SURRY PLANT i SCENARIOS ANALYZED
\
! ONLY ONE NEW SEQUENCE HAS BEEN ANALYZED COMPLETELY WITH I2 THE STCP - AF F ANALYSES INDICATE S2 FI 2 F AND SIFI2F WOULD NOT RESULT IN CORE MELTDOWN *AND THAT AF FI 2 IS UNLIKELY TO RESULT IN CORE MELTDOWN A NUMBER OF ADDITIONAL SEQUENCES WERE ANALYZED SUBSEQUENT T0-NUREG-0956 WI STAND-ALONE CODES S2C IS NOT A CORE MELT CASE V REANALYSIS LED TO SLIGHT REDUCTION IN CONSEQUENCES I
INSIGHTS NUREG-0956 SOURCE TERM INSIGHTS FOR SURRY BEST ESTIMATE ANALYSES OF WASH-1400 RISK DOMINANT SCENARIOS LEAD TO SMALLER TMLB's - STATION BLACKOUT WITH EARLY ?3NTAINMENT FA! LURE ENVIRONMENTAL RELEASE OF VOLATILES REDUCED BY FACTOR OF 10 PRINCIPAL ADDITIONAL CREDIT IS IN RCS RETENTION - 76% FOR I LONG TERM REVAPORIZATION COULD INCREASE RELEASES -
Y - INTERFACING SYSTEMS LOCA ENVIRONMENTAL RELEASE OF VOLATILES REDUCED BY FACTOR OF 4 FOR VOLATILES WI SCRUBBING IN SAFEGUARDS BUILDING AND BY FACTOR OF 15 WITH POOL SCRUBBING S2C6 - CONTAINMENT FAILURE INDUCES CORE MELT
, CURRENT ANALYSES INDICATE NOT A CORE MELT SEQUENCE
- NOT ONLY ARE THE SOURCE TERMS FOR SEVERE SURRY SCENARIOS LOWER, THEIR PROBABILITIES ARE j LOWER l
l l
?
' ,\
SOURCE TERM RESULTS FOR PEACH BOTTOM PLANT SCENARIOS ANALYZED -
ATWS TC1 - CONTAINMENT FAILURE' PRECEDES CORE MELT TC2 - CONTAINMENT FAILURE FOLLOWS CORE MELT TC3 - TC2, BUT WITH WETWELL VENTING STATION BLACKOUT TB1 - CONTAINMENT FAILURE FOLLOWS VESSEL FAILURE .
TB2 - CONTAINMENT FAILURE AT VESSEL FAILURE INTERFACING SYSTEMS LOCA - V SENSITIVITY STUDIES SUPPRESSION POOL DF CORE-CONCRETE RELEASE OF FP IN-VESSEL FP RELEASE COEFFICIENTS e
9 9
0 .
l INSIGHTS MODE AND LOCATION OF CONTAINMENT FAILURE HAS MAJOR IMPACT ON CONSEQUENCES CURRENT ANALYSES BASED ON DRYWELL FAILURE ,
RETENTION CAPABILITY OF REACTOR BUILDING FOR RETENTION IS LIMITED PREDICTED HYDROGEN BURNS SWEEP OUT CONTENTS TO ENVIRONMENT {
t INTEGRITY OF REFUELING BAY UNLIKELY
- I INTEGRITY OF REACTOR BUILDING IS QUESTIONABLE -
DF e 2 IN TYPICAL ANALYSES
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OINL-DWG 22-5181 ETO
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. . l
- PEACH BOTTOM TC1 23 0 REAC10R BUILDING HEFUELING BAY 22.0-4 m
in i
A 21 0 -
! Id -
20.0 -
f m
- (A N 19.0 -
i N
- i 4 I
E* 18.0 -
Z N
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M l A 16.0 -
o
' O 15.0 -
,, _ q _ __ ,_ __
l 14.0 , , , , , i i i 1 0.0 100.0 200.0 300.0 400.0 500.0 600 0 700 0 800 0 900.0 l
- TIME - (MINUTF)
I SECONDARY CONTAINMENT PRESSURE RESPONSE FOR ICl I
s
PEACH BOTTOM TC1 .
o.-e a
60 . .
co H 5.0 - '
N -
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m
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to 3.0 - ,
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0.0 = ,
0.0 i
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100.0 200.0 300.0 i
i 400.0 500.0 600 0 700 0 i
j 800 0 900.0 TIME - (MINUTE)
TOTAL VOLUME OF 6ASES LEAKED FOR ICl .
e i i i. .
INSIGHTS (CONTINUED) l i
PREDICTED RELEASES OF VOLATILE FISSION PRODUCTS ARE REDUCED FROM WASil-1400 ;
PARTLY DUE TO RCS RETENTION r PARTLY DUE TO SUPPRESSION POOL SCRUBBING l TC2 EXAMPLE FOR IODINE RELEASE TO ENVIRONMENT STCP WASH-1400 IDCOR -
1.3x10-2 0,1 0,1 -
IDCOR HIGHER THAN STCP BECAUSE OF REVAPORIZATION AND DRYWELL LEAKAGE AT HIGH TEMPE PREDICTED RELEASES OF LESS VOLABILE FISSION PRODUCTS (SR, BA, LA, CE, Pu) ARE HIGHER THAN WASil-1400 l
RESULTS FROM MORE MECHANISTIC TREATMENT OF R'EVAPORIZATION RELEASE LIMESTONE CONCRETE ENHANCES RELEASE TC2 EXAMPLE FOR STRONTIUM RELEASE TO ENVIRONMENT STCP WASil-1400 IDCOR ;
{
VAPORIZATION RELEASE OF FISSION PRODUCTS IS.A HIGH PRIORITY ISSUE - DRIVEN BY CORE-CONCRETE IsEHAVIOR 1
s
.I i
k I
~
INSIGHTS (CONTINUED)
WETWELL VENTING CAN POTENTIALLY MITIGATE THE CONSEQUENCES OF SOME SCEN TC2/TC3 EXAMPLE STCP STCP IDCOR
' NO VENT VENT VENT IODINE 1.3x10-2 6.2x10-4 6x10-4 STRONTIUM 0.30 9.2x10-3 4x10-6 .
PHENOMEN0 LOGICAL UNCERTAINTIES REGARDING THE EFFECTIVENESS OF VENTING EFFECT OF DRYWELL TEMPERATURES ON POOL BYPASS (DRYWELL SPRAYS IN COMBINATION WITH VENTING)
UNCERTAINTY IN POOL DF FOR EXPECTED CONDITIONS QUESTIONS OF TECHNICAL FEASIBILITY 9
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- PEACH BOTTOM.TC3 26.0 3
REAC1DR BUILDING REFUEI.ING DAY 4
24.0-rn i A i
[4 22.0-m .
D rn
~ Cn N 20.0-m i 4 I
s 18 0-g N
2 s
m 16.0 -
! A y , : === ; y a = -
- m & A.-e A - w w
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- o 14.0 -
12.0 - ,
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0.0 100.0 200.0 300.0 4000 5000 600 0 i
700 0 800 0 TIME - (MINUTE)
SECONDARY CONTAINMENT PRESSURE RESPONSE FOR TC3 8 8
Source TERM RESULTS FOR SEQUOYAH PLANT SCENARIOS ANALYZED SMALL BREAK WITH FAILURE OF ECC RECIRCULATION 5 SCENARIOS ANALYZED, 3 WITH FULL STCP STATION BLACKOUT TRANSIENT WITH TEMPERATURE INDUCED STEAM GENERATOR TUBE RUPTURE a
i i
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i
=
i INSIGHTS i
! RELATIVE TIMING OF ICE DEPLETION, CONTAINMENT FAILURE AND CORE MELTDOWN HAVE MAJOR INFLUENCE ON RISK - FOR MOST SCENARIOS ICE IS EFFECTIVE IN MITIGATING FISSION a PRODUCT RELEASE EXAMPLE - IB WITH PUMP SEAL LOCA i ELEMENT DF OF ICE BED -
I
~
i 40 -
SR - 13 LARGE VOLUME OF WATER IN REACTOR CAVITY IS EFFECTIVE IN MITIGATING EX-VESSEL RE OF FISSION PRODUCTS EXAMPLE - IB WITH PUMP SEAL LOCA ELEMENT DF OF CAVITY POOL
- LA 7 IN GENERAL, THE PREDICTED ENVIRONMENTAL RELEASES ARE LESS THAM IN RSSMAP EXAMPLE IB - STATION BLACKOUT WITH SEAL LOCA l
j ELEMENT STCP RSSMAP IDCOR l I S.2x10-'I 6.3x10-2 Sx10-'I
- SR 1.9x10-3 1.7x10-2 <1x10-5 PARTI ALLY BECAUSE OF INCREASED RCS RETENTION - 911% OF l POTENTIAL FOR REVAPORIZATION TO INCREASE ENVIRONMENTAL RELEASE OF VOLATILES i l SIGNIFICANTLY i
l
m .
SUMMARY
FOR SURRY AND SEQUOYAH THERE ARE NO MAJOR SOURCE TERM SURPRISES DIFFERENCES WITH INDUSTRY ON THE BACK-END ANALYSIS ARE RELATED OF THE CONTAINMENT EVENT TREE PROBABILITIES THAN THE ANALYSIS OF SOURCE TERM FOR PEACH BbTTOM THE PREDICTED EX-VESSEL RELEASES ARE LARGER THAN HIGH RELE/.SES COULD BE AN ARTIFACT OF LIMITATIONS OF ANALYSIS METH AREA FOR FURTHER METHODOLOGY DEVELOPMENT AND VALIDATION .
VENTING COULD BE EFFECTIVE IN MITIGATING THE CONSEQUENCES OF SOME SC OUTSTANDING TECHNICAL AND FEASIBILITY QUESTIONS UNCERTAINTIES.ARE LARGE.
CURRENT RESULTS MAY IN SOME AREAS BE OVER-ESTIMATES AND IN OTHER AREAS.UNDER-ESTIMATES.
i SOURCE TERM UNCERTAINTIES ARE INCLUDED EXPLICITLY IN THE SARRP METHODOLOGY.
l t i
9
0.69 9.0x10-2 5.5x10-2 Sr 0.3 0.12 1.7x10-2 1.0x10-2 6x10-2 La 3.1x10-4 2.5x10-3 1.7x10-4 4x10-3 I .
SURRY V Without Pool Scrubbing Element With Pool Scrubbine RCS RC8/ESFB Environment _RCB/ESFB Environment WASH-1400 I 0.81 2x10-2 0.17 0.15 4.1x10-2 0.7 Te 0.14 0.74 0.11 0.84 2.0x10-2 Sr 0.3 0.34 0.15 6.2x10-2 0.20 1.1x10-2 6x10-2 La 9.8x10-4 2.4x10-2 5.5x10-3 2.8x10-2 9.3x10-4 4x10-3 i
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PEACH BOTTOM TC2 SCENARIO Element RCS Pool Reactor Building Environment Wash-1400 10COR I 0.39 0.58 2.3x10-2 1.3x10-2 0.1 lx10-1 Te 0.62 8.0x10-2 9.7x10-2 8.7x10-2 0.3 lxle-I ;
Sr 6710-4 8.8x10-4 0.34 0.30 lx10-2 4x10-4 La 9x10-8 3.1x10-5 1.2x10-2 9.2x10-3 3x10-3 -
, PEACH BOTTOM TC3 SCENARIO Element RCS Pool Reactor Building Environment Wash-1400 10COR I 0.39 0.61 5.6x10-4 6.2x10-4 0.1 6x10-4 Te 0.62 0.24 4.7x10-3 6.5x10-3 0.3 4x10-4 Sr 6x10-4 0.63 6.0x10-3 9.2x10-3 lx10-2 4x10-6 La 9x10-8 2.1x10-2 2.3x10-4 3.2x10-4 3x10-3 _
SEQUOYAH lB SCENARIO Element RCS Cavity Water [IceBed Environment RSSNAP IDCOR I 0.94 2.5x10-2 2.1x10-2 5.2x10-4 Te 6.3x10-2 5x10-4 0.83 3.9x10-2 3.3x10-2 1.3x10-3 9.5x10-2 3x10-5 Sr 6.2x10-4 0.14 2.5x10-2' l.9x10-3 La 1.1x10-7 1.7x10-2 i<lx10-5 7.2x10-3 1.2x10-3 8.4x10-5 lx10-3 .
I $ h '
- =-- :
PRELIMINARY CONCLUSIONS FOR PEACH BOTTOM l
I TREVOR PRATT DEPARTMENT OF NUCLEAR ENERGv BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK 11973
, PRESENTED AT THE NRC/IDCOR MEETING ON THE PEACH BOTTOM REFERENCE PLANT ANALYSIS PHILLIPS BUILDIN.G, DECEMBER 10, 1985 L
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, , , _ . - - ~ . -
.- - , - - - - -- -- ^
OUTLINE CORE DAMAGE FREQUENCY COMPARISON OBSERVATIONS DOMINANT CONTRIBUTORS AREAS REQUIRING FURTHER WORK CONTAINMENT PERFORMANCE SOURCE TERMS CONSEQUENCES
SUMMARY
9 BROOXHAVEN NATIONAL LABORATORY l} g)l A5500ATED l!NIVER5mES, INC(1 ElI
PEACH BOTTOM CORE DAMAGE FREQUENCY ...
COMPARISON BETWEEN IDCOR AND SARP (DOMINANT CONTRIBUTORS)
ASEP/SARPA SEQUENCE CORE CORE TYPE. IDCORB DAMAGE VULNERABLE TC 7 3-6C 1 3-6 64-7 TOUV AND TQUX 4.8-7 1 6-7 ---
TB 4 5-7 7 6-6 ---
TW 1 5-7 1 6-7 2 1-5 8 5-6 9 9-6 2 4-5 ARESULTS FROM N0v/85 PRESENTATION TO THE SCG.
B PEACH BOTTOM COMMITTED RISK PROFILE.
C7 3-6 = 7 3 x 10-6 BROOKHAVEN NATIONAL LABORATOR A5500ATED UNIVERSITIES, INC. (l(II
i
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OBSERVATIONS RELATED TO CORE MELT FREQUENCY ESTIMATES -
INTERNAL EVENT CORE MELT FREQUENCY IS LOW BOTH IDCOR AND NRC CONTRACTORS HAVE IDENTIFIED SIMILAR ACCIDENT SEQUENCES AS'POTENTIALLY IMPORTANT TO CORE MELT FREQUENCY STATION BLACK 0UT FAILURE TO SCRAM THE FOLLOWING SEQUENCES ARE NOW CONSIDERED LESS IMPORTANT TO CORE MELT FREQUENCY:
LOSS OF CONTAINMENT HEAT REMOVAL SMALL BREAK LOSS OF COOLANT ACCIDENTS 0THER SEQUENCES CONSIDERED AND FOUND LESS IMPORTANT COMPLETENESS: WORK IS ONGOING BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC. (1 ElI
_ _ _ - = . . - - _._____.-.m._..m_.
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I . .
OTHER SEQUENCES WERE CONSIDERED BUT --
FOUND TO BE OF LOW FREQUENCY:
, ASEP/SARPA SEQUENCE CORE CORE TYPE IDCORB DAMAGE VULNERABLE SE3 1 2-7 1 6-7 ---
AE 1 6-8 2.6-7 ---
TPOI ---
3 8-7 7 9-7 SWi 1 5-6 AW --- ---
2 9-7 A
RESULTS FROM N0v/85 PRESENTATION TO THE SCG.
B PEACH BOTTOM COMMITTED RISK PP0 FILE.
0THER SEQUENCES MAY HAVE BEEN OVERLOOKED AND MAY HAVE TO BE ANALYZED IN THE FUTURE.
- V (INTERFACING LOCA) SEQUENCES BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(I(II
=
TC SEQUENCE MAJOR DIFFERENCE BETWEEN IDCOR AND NRC CONTRACTORS COMES FROM QUANTIFICATION OF HUMAN ERROR PROBABILITIES.
MOST IMPORTANT DIFFERENCE IS IN MANUAL ACTUATION OF SLCS AND LEVEL CONTROL.
l
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BROOKHAVEN NATIONAL LABORATORY l} gj l i A5500ATED UNIVERSITIES, INC. (Illl
HUMAN ERROR PROBABILITY (HEP) ESTIMATES PEACH BOTTOM ATWS WITH MSIV CLOSURE SEQUENCE HUMAN ACTION' 2 ASEP IDCOR' DESIGNATOR HEPs HEPS S LC 005-010' O.3s ADS O 015-0 03 N/A' HLC 01 -0 2 03 5 -
' HUMAN ACTION DESIGNATOR:
SLC INITIATE STANDBY LIQUID CONTROL SYSTEM.
ADS DEFEAT ADS INITIATION SIGNAL.
HLC (LLC)
ESTABLISH AND MAINTAIN WATER LEVEL AT TAF WHILE AT HIGH (LOW) PRESSURE.
DEP 2 MANUAL DEPRESSURIZATION.
ASEP HEP - ESTIMATE' RANGE BASED ON PREVIOUS EVENT OUTCOME.
'IDCOR HEP COMMITTED RISK PROFILE (FROM TASK 21 1 , .
APPENDIX A).
BASED ON SUPPRESSION POOL AT 110*F WITHIN 2 MIN.
AND TRIP PROCEDURE T-101 REQUIREMENT AT 110*F.
BASED ON '5-10 MIN. AVAILABLE TO ACTIVATE SLCS AND CONTROL WATER LEVEL" PROVIDING A COMBINED HEP.= 0 5, THEREFORE:
1 - (1-HEP)2 = 0 5 ; HEP - 0 3
'N/A - NOT ADDRESSED IN TASK 21 1, APPENDIX A.
BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVER5lilES, ilk. (I ElI
E . ..
TB SEQUENCE MAJOR DIFFERENCE BETWEEN IDCOR AND NRC CONTRACTORS IS DUE TO TREATMENT OF COMMON MODE FAILURES.
DOMINANT CONTRIBUTdRS TO CORE DAMAGE FREQUENCY IN NRC CONTRACTOR ANALYSIS ARE:
- COMMON MODE FAILURE OF BATTERIES
- SERVICE WATER DISCHARGE VALVE
~
BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC. (!lll
. . . _ _ _ _ . _ _ . _ . _ _ _ _ _ _ . . _ _ . _ _ - ._.z.._
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CONTAINMENT VENTING IMPACT ON CDF IDCOR C, ORE DAMAGE FREQUENCY SARP W/0 CONTAIN-MENT VENTING AND ASSUMING SEQUENCE .W/CONTAIN- W/0 CONTAIN- INJ. FAILURE CORE CORE TYPE MENT VENTING MENT VENTING W/ CONT. FAILURE DAMAGE VULNERABLE TC 7 3-6 9 2-6 1 2-5 1 3-6 6 4-7
.TW 1 5-7 1 2-6 1 1-5 ---
2 1-5 BROOKHAVEN NATIONAL LABORATOR A5500ATED UNIVERSITIES, INC.(IIlI
^
AREAS REQUIRING FilRTHER WORK IMPACT OF VENTING ON CORE MELT FREQUENCY OPERATOR ACTIONS IN TC SEQUENCE EMERGENCY SERVICE WATER COMMON MODE FAILURES l
l l
1 1
l6 -
1
- c. .. , ,
NRC/IDCOR ISSUES . _-
ISSUE 1 FP RELEASE PRIOR TO VESSEL FAILURE .,
ISSUE 2 RECIRCULATION OF COOLANT IN REACTOR VESSEL ISSUE 3 RELEASE MODEL OF CONTROL ROD MATERIALS ,
ISSUE 4 FP AND AEROSOL DEPOSITION IN RCS ISSUE 5 IN-VESSEL H2 GENERATION ISSUE 6 CORE SLUMP, CORE COLLAPSE AND RP,V FAILURE ISSUE 7 STEAM EXPLOSION FAILURE ISSUE 8 DIRECT. HEATING OF CONTAINMENT ISSUE 9 Ex-VESSEL FISSION PRODUCT RELEASE ISSUE 10 Ex-VESSEL H/T MODEL FROM MOLTEN CORE TO CONCRETE ISSUE 11 REVAPORIZATION OF FP FROM RCS ISSUE 12 FP DEPOSITION ll0 DEL IN CONTAINMENT -
BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(I ElI
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NRC/IDCOR ISSUES (CONTINUED)
ISSUE 13A SUPPRESSION POOL BYPASS (POOL SCRUBBING)
/ 13B RETENTION OF FPS IN ICE BEDS T
ISSUE 121 MODELING OF EMERGEl;CY RESPONSE ,
8: .
J f ISSUE 15 CONTAINMENT PERFORMANCE \'
I i 1 ISSUE 16 SECONDARY CONTAINMENT PERFORMANCE N
\( 1 ISSUE 17 HYDROGEN IGNITION AND BURNING t
ISSUE 18 ESSENTIAL EQUIPMENT PERFORMANCE ,
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l BROOKHAVEN NATIONAL LABORATORY l} gj )
A5500ATED UNIVER$1 TIES, INC.(I til
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CONTAINMENT PERFORMANCE INTEGRITY P'0TENTIAL FOR ISOLATION FAILURE VENTING CONSIDERATIONS:
ENGINEERING HUMAN RELIABILITY
) ,
RISK BENEFIT -
SUPPRESSION P0OL BYPASS (ISSUE 13A):
EVEN WITH VENTING LOSS OF DRYWELL INTEGRITY POSSIBLE DUE TO HIGH TEMPERATURES IMPACT OF AEROSOL PLUGGING BROOKHAVEN NATIONAL LABORATORY l} gj )
A5500ATED UNIVERSITIES, INC.(l(II Q _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ .
'=~'-
CONTAINMENT PERFORMANCE (CONT.)
CONTAINMENT FAILURE MODE (ISSUE 15):
EARLY VS LATE LARGE BLOWDOWN VS GRADUAL RELEASE LOCATION (WETWELL VS DRYWELL)
IMPACT ON FISSION PRODUCT RELEASE BROOKHAVEN NATIONAL LABORATORY)) g)l A5500ATED UNIVERSITIES, INC.(3 ElI
O 4 SOURCE TERM COMPARISON EXAMPLEt ATWS (TC) WITH N'O OPERATOR ACTIONS TAKEN.
~
EVENT IDCOR* NRC CONTRACTORS CONTAINMENT FAILURE (HR) 14 14 START OF CORE MELT (HR) 30 22 VESSEL FAILURE (HR) 39 38 FISSION PRODUCT RELEASE FRACTIONS:***
XE - KR 1.0 10 I - BR 01 0 03 Cs - Rn 01 0 03 TE - Sa 01 0 30 SR - BA 0.0004 0 43 Ru - Mo 0.001 NEG.
LA -
0 01 CE -
0 02
- IDCOR TECHNICAL REPORT 23 1PB, MARCH 1985
- INFORMAL BCL REPORT DATED NOVEMBER 18, 1985
- FRACTION OF INITIAL CORE INVENTORY . .
BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC. (IIII
~ r ,- - - - - D-r --- ,,- -, ..-- , r- n- ,,
SOURCE TERM COMPARISON EXAMPLE: STATION BLACK 0UT EVENT IDCOR* NRC CONTRACTORS **
LOSS OF INJECTION (HR) 60 60 START OF CORE MELT (HR) 11 4 10 7 VESSEL FAILURE (HR) 12 0 12 2 CONTAINMENT FAILURE (HR) 18 0 15 2 FISSION PRODUCT RELEASE FRACTIONS:***
XE - KR 10 10 I - BR 0 05 0 012 Cs - RB 0.05 0 014 TE - SB 0 06 0 22 SR - BA NEG. 0 17 Ru - Mo 0 0001 NEG.
LA -
0 03 CE -
0 05
- IDCOR TECHNICAL REPORT 23 1PB, MARCH 1985
- INFORMAL BCL REPORT DATED NOVEMBER 18, 1985
- FRACTION OF INITIAL CORE INVENTORY. -
BROOKHAVEN NAIL 0NAL LABORATORY)) g)l A5500ATED UNIVERSITIES, INC. (IllI
._. .-.-. - , - . . . . _ - - _ _ _ _ _ , _ - . . . - _ ~ - _ . . - ..
COMPARISON OF Csl DISTRIBUTION (FRACTION OF INITIAL CORE INVENTORY)
EXAMPLE: STATION BLACK 0UT EVENTS VESSEL FAILURE CONTAIMENT END OF LOCATION FAILURE CALCULATION IDCOR* NRC** IDCOR NRC IDCOR NRC (12 HR) (12 HR) (18 (HR) (15 HR) (60 HR) (22 HR)
REACTOR PRESSURE VESSEL 10 0 85 0.76 0.85 0.09 0 85 DRYWELL -
0 12 0 20 ? -
0 09 (MELT)
SUPPRESSION POOL -
? 0 04 ? 0 04 0 03 REACTOR BUILDING - - - -
0 82 0 009 ENVIRONMENT - -
0 05 0 014
- IDCOR TECHNICAL REPORT 23 1PB, MARCH 1985
' INFORMAL BCL REPORT DATED NOVEMBER 18, 1985 BROOKHAVEN NATIONAL LABORATORY l} g))
A5500ATED UNIVERSITIES, INC. (IllI
. e.
30llRCE TERM IN-VESSEL RELEASE (ISSilE 1):
- IDCOR'AND NRC CONTRACTORS SIMILAR EXCEPT FOR TE INITIAL PRIMARY SYSTEM RETENTION (ISSUE 4):
- IDCOR GENERALLY HIGHER THAN NRC CONTRACTORS REVAPORIZATION FROM PRIMARY SYSTEM (ISSUE 11)
-NRCCONTRACTORSDONOTMODELREYAPORIZATION AFTER RPV FAILURE
- IDCOR MODELS SIGNIFICANT RE-RELEASE l
IODINE:
j - RADI0 LOGICALLY PRODUCED ELEMENTAL IODINE
- ORGANIC I0 DINE l
l l
l 6k00KHAVEN Nail 0NAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC. (1 El I
,m ++ - ,--- - -
w-me - - - .- --- - --w- - - - - , - .-
- , 0-1 SOURCE TERM (CONT.)
EX-VESSEL RELEASE (ISSUE 9):
- NOT,MODELED BY IDCOR IN 23 1PB
- SIGNIFICANT. RELEASES BY NRC CONTRACTORS
- IMPACT OF SPRAYS VENTING:
EFFECTIVENESS OF P0OL SCURBBING
. POOL BYPASS (AEROSOL PLUGGING)
-SECONDARY CONTAINMENT (ISSUE 16):
- RETENTION OF FISSION PRODUCTS
- SURVIVABILITY (H2 BURNS) l i
BROOKHAVEN NAIL 0NAL LABORATORY l} g)l
, A5500ATED UNIVERSITIES,.INC.(3 ElI
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CONSEQUENCES EVACUATION MODEL (ISSUE 14)
- VELOCITY
- DELAY TIME
- NON-PARTICIPATION WARNING TIME RELEASE DURATION BROOKHAVEN NATIONAL LABORATORY l} gj )
~ ASSOCIATED UNIVERSITIES, INC. (I Ell au v- ,a '
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SUMMARY
REMAINING ISSUES FOR FURTHER DISCUSSION:
VENTING CONSIDERATIONS SUPPRESSION P0OL BYPASS (ISSUE 13A)
CONTAINMENT FAILURE MODE (ISSUE 15)
IN-VESSEL RELEASE (ISSUE 1) .
REVAPORIZATION FROM PRIMARY SYSTEM (ISSUE 11)
IODINE EX-VESSEL RELEASE (ISSUE 9)
SECONDARY CONTAINMENT (ISSUE 16)
BROOKWNEN NATIONAL LABORATORY)) g) l A5500ATED UNIVERSITIES, INC. (I(Il
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BACKGROUND:
~0 PRIOR ANALYSES BY INDUSTRY, IDCOR AND HRC 4
0 PIACH BOTTOM REVIEW BY NRC/INEL AND PECO STAFFS 0 VENTING CONSIDERATIONS AND CONCERNS e
j O
CONCLUSIONS SHARED BY NRC/INEL AND PECO STAFFS 0 DESCRIPTION OF THE ISSUES l
l l'
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TOPICAL tREA: CONTAINMENT VENTING FOR BWR, MARK I, NPP SAS THE POTENTIAL TO REDUCE RISK IN TB (STATION BLACKOUT),
TC (TRANSIENT WITH FAILURE OF REACTOR SHUTDOWN -
ATWS). TW (TRANSIENT WITH LOSS OF CONTAINMENT HEAT REMOVAL) AND TQUV ACCIDENT SEQUENCES, WHEREIN CONTAINMENT WOULD FAIL BY OVER-PRESSURIZATION.
ISSUE: ASSESSMENT OF THE CHANGE IN THE RISK OF RADIO-LOGICAL EXPOSURE THAT RESULTS FROM VENTING BWR MARK,I CONTAINMENTS, I.E., THROUGH A DETERMINATION THAT VENTING CAN BE RELIED UPON AS A MEASURE TO RELIEVE CONTAINMENT OVER-PRESSURE.
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VENTING ANALYSIS PERSPECTIVE RISK m
[ EQUIPMENT PERFORMANCE -
>=
O -
S x
- PROCEDURES AND OPERATOR ACTIONS S
0 9
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RISK ISSUES:
DETERMINE THE CONSEQUENCES OF VENTING ON RISK TO THE PLANT PERSONNEL AND RUBLIC.
PROBABILITY AND CONSEQUENCES OF VENT PATH FAILURE PROBABILITY AND CONSEQUENCES OF SGTS FAILURE DF IN REACTOR BUILDING
- SCRUBBING IN THE SUPFRESSION POOL, e.g.,
,
- SCRUBBING CF STEAM WHEN POOL IS SATURATED l
- SCRUSBING 0F FISSION PRODUCTS (NONCONDENSIBLES) l
- REINTRAINMENT OF FISSION PRODUCTS DUE TO FLASHING POTENTIAL FOR RISK MITIGATION
^ 4
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EQUIPMENT PERFORMANCE ISSUES:
DETERMINE THE CAPABILITY OF THE SYSTEMS AND HARDWARE TO PERFORM VENTING AND THE CONSEQUENCES OF VENTING ON THE SYSTEM AND _ HARDWARE.
TORUS CAPA8ILI1Y, ESPECIALLY FLUID / STEAM INTERACTION VENTING VALVES' CAPABILITY STRUCTURAL ANALYSIS OF DUCT PATH TO SGTS EFFECT OF DUCTWORK FAILURE'0N PLANT (REACTOR BUILDING, SYSTEMS, COMP 0 NESTS,RE-USABILITY,ETC.)
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9
PROCEDURES /0PERATOR ACTION ISSUES 1 DETERMINE THE ABILITY OF THE OPERATING CREW TO EFFECTIVELY, INITIATE AND TERMINATE VENTING USING " TRIPS" AND E0PS.
ADEQUACY OF ALL PROCEDURES INTEGRATION OF ALL OPERATOR ACTIONS ABILITY TO RECONFIG'URE AND JUMPER CABLES IN BACK PANELS ABILITY TO ESTABLISH LOCAL CONTROL FOR STROKING 6" AND 18" VALVES EFFECTS OF COMPETING TASKS AND STRESS EVALUATE NEED FOR POST-VENTING OPERATIONS e
e
OBJECTIVE OF THE VENTING ANALYSIS PROJECT:
- +--
PERFORM A SYSTEMATIC EVALUATION OF VENTING AS A MEANS OF MITIGATING
~
CONSEQUENCES OF OVER-PRESSURE.
~
THEREFORE. ANSWERS TO THESE QUESTIONS ARE REQUIRED:
CAN THE SYSTEM HARDWARE, PROCEDURES, AND OPERATOR TASKS PERFORM AS INSTRUCTED 7 IF S0, CAN VENTING REMOVE THE THREAT?
IF S0, WHAT IS THE GAIN AND CONSEQUENCES?
IFNOT,WiATCHANGESTOHARDWARE,PROCEDURESOROPERATORTASKS ARE NECESSARY TO ALLOW SUCCESSFUL VENTING?
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. . . . . . . . . - - - - - - - - - - ' - - - - - - - - - - - - - - - - - -' '- ~
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~
TECHNICAL APPROACH BY NRC ,1NEL AND BCL TEAM:
UTILIZE AN ACCIDENT MANAGEMENT. INTEGRATIVE APPROACH TO IDENTIFY THE RISK REDUCTION POTENTIAL ASSOCIATED WITH CONTAINMENT VENTING FOR SEVERE ACCIDENT SEQUENCES (TB AND TC) FOR PEACH BOTTOM NPP.
9 BASED ON..........
ENGINEERING ANALYSIS HUMAN RELIABILITY ANALYSIS RISK ANALYSIS i
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Venting Analysis
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Programmic Approach -
~
r A Presentation to The NRC - IDC0F1 Joint Meeting December 9-10,1985 '
'O ,
l DIVISION OF RISK ANALYSIS AND OPERATIONS i -
OFFICE OF NUCLEAR REGULATORY RESE.. ' RC H l i
I 6 6 . gl
9 e
9 Background .
o Prior analyses by industry, IDCOR and NRC -
l
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o Peach Bottom review by NRC/INEL and PECO staffs o Venting considerations and concerns o Conclusions shared by NR'C/INEL and PECO staffs .
o Description of the issues
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Outline of Uncertainty Analysis l l Section 1: Background and Statement of the issues A. Prior Analysis t
- Venting studies by GE, SASA, IDCOR !
- Peach. Bottom studies by PEC0 staff
- Peach Bottom and Browns Ferry reviews by NRC contractors
- Statement of the issues: . '
" Assessment of the change in the risk of radiological exposure that results from venting BWR Mark I contain-ments, I.E., through a determination that venting can {
be relied upon as a measure to relieve containment over-pressure" s
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\
Outline of Uncerta.inty Analysis j Section 2: Risk Importance and ' Uncertainty -
- Venting implementation, via emergency operating procedures not considered in detail by IDCOR or NRC in risk dominant ' sequences. !
I
- No independent determination that operator can successfully initiate and terminate venting.
- Uncertain that venting can be achieved with existing hardware piping, ductwork, valves, and operator tasks. l l
- Uncertain that venting will not adversely affect TORUS and its associated hardware .
- Uncertain about the applicability of prior analyses to Peach Bottom venting - -
?
s s 6'
I f
l Outline of Uncertainty Analysis
~
Section 3: Programmatic Applications
. Technical Approach: . Utilize an accMent management integrative l approach to identify the sk reduction potential ,
associated with containment venting for severe accident
, sequences (TB and TC) for Peach Bottom NPP, consist- )
ing of:
~
)
- Engineering Analysis
- Human Reliability Analysis
- Risk Analysis - j
.. a
Five Steps in Peach Bottom Venting Evaluation -
e Assessment and selection of specific secuences T
o l Definition of no venting consequences e Evaluation of plant spe~cific procecures anc ooerator actions e
Analysis of engineering performance of system l
e Assessment of consequences from venting 1
e s e 01
- Step 1: Assessment and
. Selection of Specific Sequences .
a Identification of overpressure faiure sequences.
a i Selection of ris< significant secuences using i containment event trees -
j
. j l
Identify scenarios for analysis '
4 1
i
.' l
e b
Step 2: .
.')@finition of No Venting Consequences o Define accident signatures o De ine consecuences in terms of racionuclice re ease i..c 6
Step 3: Evaluation of Plant Specific Procedures / Operator Actions o
Assessment of procedures for venting (content, form, and format) o Evaluation of operator actions to follow procecures
{ decisions, response requirements, time recuirements, specific tasks, feedback for effectiveness) t o Perform HRA for.se ectec sequences '
t
- - ' - ~
3- .. .
i i
Step 4:
l Analysis of Engineering Performance
( .
Lo Determine vent path hardware characteristics
!o Determine vent path limitations (fault tree 5? '
l je Develop MARCH system models and inputs i
- o Calculate system performance o
Determine consequences for sequences when venting
, is successful i ..
Step 5: Assessment of
! Consequences from Venting 1
J
) o Compare venting and no venting conse'quences -
o Determine effect of abiity to vent on procecures 4
l o Determine effect of abiity to vent on lardware
) .
o Assess change in risk i o Document results !
o 03tain comments and revise 8 0
- gl
1 i
l Section 4: Schedule .
l -
(task completion)
~
I
=
l~ Obtain Peach Bottom & other information 11/25/85 l
= Evaluation of procedures 12/31/85 a Analysis of engineering performance i
2/14/86 i e Analysis of operator tasks 2/14/86 o
l Assessment of consequences 2/28/86 o
) Documentation (draft) 3/28/86 a
Documentation (final) 5/30/86 l
i j ...e I '
j .
.. e IN-VESSEL ISSUES: FISSION PRODUCT .
SENERATION RETENTION AND RELEASE l
TREVOR PRATT l
DEPARTMENT OF NUCLEAR ENERGY BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK 11973 1
, PRESENTED AT THE NRC/IDCOR MEETING ON THE PEACH BOTTOM REFERENCE PLANT ANALYSIS PHILLIPS BUILDING, DECEMBER 10, 1985 w
BROOKHAVEN Nail 0NAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(llli i
IN-VESSEL ISSUES ISSUE 1: FP RELEASE PRIOR TO VESSEL FAILURE ISSUE 4: FP AND AEROSOL DEPOSITION IN RCS ISSUE 11: REVAPORIZATION OF FP FROM RCS BROOKHAVEN Nail 0NAL LABORATORY l} g} l A5500ATED UNIVERSITIES, INC.(I til A
ISSUE 1: FP RELEASE PRIOR
~
TO VESSEL FAILURE FP RELEASE FROM FUEL DURING CORE HEATUP AND DEGRADATION SIMILAR IN IBCOR AND NRC CONTRACTORS ANALYSIS
- EXCEPTION TE TREATMENT IDCOR ASSESSED IMPACT OF TE TREATMENT IN REPORT T85 2 (EXAMPLE - STATION BLACK 0UT):
UNBOUND CASE (IN-VESSEL Te RELEASE)
' - 15 xo RELEASED TO Rx. BLDG. IN
~60 HOURS
- 2 2 xs RELEASED TO ENVIRONMENT BOUND CASE (EX-VESSEL Te RELEASE)
- 15 xs RELEASED TO Rx. BLDG. IN
~6 HOURS
- 1 4 xs RELEASED TO ENVIRONMENT BROOKHAVEN Nail 0NAL LABORATORY)) g } l A5500ATED UNIVERSITIES, INC.(Illl
a
. .m l
ISSU,ES 4 AND 11: DEPOSITION AND REVAPORIZATION OF FP FROM PCS DIFFERENCES EXIST IN PREDICTED FISSION PRODUCT RETENTION IN PRIMARY SYSTEM:
NRC CONTRACTORS ~40 TO 80%
PERMANENTLY RETAINED .
~
IDCOR MODEL ~50 TO 90% ,
TEMPORARILY RETAINED DIFFERENCES EXIST IN PREDICTING REVAPORIZATION OF FISSION PRODUCTS FROM PRIMARY SYSTEM:
NRC CONTRACTORS DO NOT MODEL REVAPORIZATION AFTER VESSEL FAILURE 1
i BROOXHAVIN Nail 0NAL LABORATORY l}
A5500ATED UNIVERSITIES, INC.(llll 1
- -- -- . _ - _ . - , - . , -.m ..--.-_._-----s .,-_x._ __.. _ _ _ _ . - . , , _ . - , . . . , . , _ , . , ,_ -,,_._ ,__-.._._,,-.,,_ - -
,, --.~__,,s._ ._- . . _ _
t
, ISSUES 4 AND 11: (CONT.)
IDCOR MODELS REVAPORIZATION AFTER VESSEL FAILURE LEADING TO SIGNIFICANT RE-RELEASE 80 TO 99% DEPENDING ON THE SEQUENCE IDCOR REVAPORIZATION NAY OVERPREDICT RELEASE OF MORE VOLATILE FP I
i i
a
, BROOKHAVEN Nail 0NAL LABORATORY l} g)l A5500ATED UNIVERSITlfs, INC.(IIII
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EX-VESSE_ .
r I SS :: 0.\ 3ROJUCT RE_ EASE l
D. A. POWERS -
D. R. BRADLEY SANDIA NATIONAL LABORATORIES ALBUQUERQUE, NM .
G. A. GREENE BROOKHAVEN NATIONAL LABORATORY UPTON, NY S. B. BURSON CONTAINMENT SYSTEMS RESEARCH BRANCH 10-DEC-85
OBJECTIVES EFFECTS OF WATER ,
- ISSUES CONCERNING TM EX-VESSEL
. .BEHA VIOR OF CORE DEBRIS
- ON TW A TTACK ON CONCRETE
- EFFECIS OF WA TER COM MBRIS
- SOURCE TERM CHEMISIRY
- M TW EX-VESSEL PRODUCTION LF AEROSOLS
- COMPARISON O.' CODE PREDICTIDAS .
TO EXPERIMENTAL DA TA o SENS1TIVITY OF CALCULA TIONS .
FOR ACCIDENTS A T PEACH BOTTOM
- TO CODE 1Ar'UIS
- TO NODELS IN THh CODES -
$ 9 O
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EXPERIME;;TS '
FRAG TEST SERIES
- 45 Kg STEEL SHOT CHOSEN TO BE REAdIL Y COOLABLE BY WA TER 0:-Q
- WA TER ADDED AFTER SIGNIFICANT A TTACK ON CONCRETE
- TESTS WITH SILICEOUS CONCRETE
& LIMESTONE / SAND CONCRETE SWISS TEST SERIES -
- 45 Kg STAINLESS STEEL MEL TS
- SUSTAINED ON LIMESTONE / SAND CONCRETE
- SWISS 1 : WA TER FLOW AFTER 30 MIN. OF CONCRETE A TTACK
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- EMPIRICAL ADJUSTMENT OF THE
. HEA T FLUX MODELS IN CORCON TO MA TCH RESUL TS OF OTHER MEL T/ CONCRETE TESTS WOULD DESTROY THIS GOOD AGR$EMENT WITH THE SWISS TEST RESUL TS f
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VANESA VOJE_ Or 4 .*
J ECO \'- A V :: N A-~ :: 0.\
- PARTICLE SEDIMENTA TION
- PARTICLE DIFFUSION i
i
- INERTIAL IMPACTION .
- ALLOWS FOR NON-SPHERICAL BUBBLES
- PREDICTIONS ARE SENSITIVE TO :
- PARTICLE SIZE BUBBLE SIZE l
BUBBLE ECCENTRICITY POOL DEPTH
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i CAN WE TREAT DIFFERENCES IN IHE DESCRIPTION OF THE EFFECTS OF WATER ON MELT / CONCRETE INTERACTIONSi AS AN UNCERTAINTY 7 l I
- CORCON PREDICTIONS ARE IN GOOD
. AGREEMENT WITH THE TEST RESUL TS
- VANESA MODEL PREDICTS VERY SIGNIFICANT DECONTAMINA TION BY t\N OVERL YING WA TER' POOL
DECONTAMINA TIONS THAN THE A PRIORI PREDICTIONS
D t
Er r EC- S 0;r V E _-- .
EX?U _S :: O N RA- E
- CORCON' IS CAPABLE OF ACCEPTING TIME-DEPENDENT MEL T ADDITIONS COR COOLANT ADDITIONS.)
SEE SUBROUTINE MASS'EX SEE P.136 USER'S MANUAL .
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- ALMOST HA VE TO IF YOU DON ' T HA VE CONFIDENCE IN YOUR CORE MEL TDOWN MODELS
- CORCON COULD BE ARRANGED TO RESPOND IF THE UNCERTAINTY .
~ RANGE IS SPEC 1FIED .
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~
VANESA PREDICTION :
10 E TE=17.7 VS 17.4 %
n ,
m -
T Y l'1 . i .. . ). t -
9 _ lC -
- j y//,W.AL'% 1'- ' '
v -
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x i
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z _ _
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W I O.1 7 m :
_ m
. W
- 0. 01 50 100 150 em TIME (S)
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9
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- -- ----;- ~..m -
ARE VANESA PREDICITIONS OF ,
RADIONUCLIDE RELEASES UNCERTAIN 7
- PREDICTIONS ' OF TOTAL AEROSOL PRODUCTION ARE IN GOOD AGREEMENT
.WITH TEST DA TA
- TELLURIUM RELEASE PREDICTIONS ARE IN GOOD ACCORD WITH DA TA
- PREDICTIONS OF ALKALINE EARTH &
LANTHANIDE RELEASES SHOULD BE CONSIDERED UNCERTAIN NO TESTS HA VE BEEN DONE IN THE REGIME WHERE THESE RELEASES WOULD BE EXPECTED TO BE HIGH
- FOR PEACH BOTTOM THE UNCERTAINTY HAS NOTHING TO DO WITH SILICA TES
- PREDOMINANT CAUSE OF UNCERTAINTY IS INPUT FROM IN-VESSEL MODELS _
l
ARE CORCON & VANESA e
- i. 3ERrEC 7 ..
NC !!
SOME CURRENT ISSUES e MECHANICAL AEROSOLS WITCH / GHOST MEL T CHEMISTRY TESTS AEROSOL SCRUBBING B:.7' -
MECHANISTIC INTERLA YER MIXING SNL/BNL '
CRUSTING SNL/Bi.L
! TESTS WITH URANIA I MEL TS & MEL TS SURC WITH METALLIC Zr TESTS
. , - - --- ,.---.----,s .-----,------------.--,..,-..-,n ----,---------,,--------,-9 - ' - - - - - - - - - - - - - - - - - - - - -
i i
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l VARIATION IN ZR CONTENT I
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800 . .
i . .
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700- - -
50% IR -
\ ,.................-
95% IR Q
% 600-
- 0% IR _
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CORE-CONCRETE INTERACTION AND FISSION PRODUCT CHEMISTRY IN BWR MARK I ANALYSIS l
- Martin G Plys
. Fauske a Associates, Inc. -
16WO70 West 83rd Street Burr Ridge, IL 60521 NRC/IDCOR Meeting l December 10, 1985 me
OUTLINE
- 1. Important Factors in BWR Mark I (Peach Bottom) Analysis
- 2. EQUUS Model
- 3. Sample Sequence - TC
- 4. EQUUS and VANESA Differences O
9
^
)
IMPORTANT FACTORS IN BWR MARK I ANALYSIS ._
. 1, Prolonged Period of Debris Release from Vessel
- 2. Incomplete Debris Release from Vessel 3, Presence of Water in Pedestal During Debris Drainage
- 4. Addition of Water to Pedestal After Initial Debris Release Lower Plenum Water CRD Flow e
O
- - = - = __ - ._
I l
1 I
~
-\
DRYWELL PEDESTAL FLOW FROM
~
VESSEL FLOW TO '
V DRYWELL -
= .- .
i.
%SOLIDN hDEBRISA
] LIQUlb SOLID ,
~
~ ~ f. . .
~
DEBRIS ,/ -
i
', , . /f. ,
[, , ,. ,
ORIGINAL ~
CONCRETE.- .
SURFACE .
.. .s .
ERODED ~
I CONCRETE SURFACE N ] _-
j 1
- m. _
EVENTS FOLLOWING VESSEL FAILURE -
- 1. Initial Debris Drainage Into Pedestal 20% of Core
- $ 1 Minute
- Quenching
- 2. Initial Lower Plenum Water Drainage
- 50 MT
-
- Several Minutes ,
Quenching
- 3. Continued Drainage of Corium s 40% to 50% of Core
$ 3 Hours i 4. Continued Water Addition i -
- 180 gpm
- Several Hours
~
EVENTS FOLLOWING VESSEL FAILURE (Continued)
- 5. Debris Overflow to Drywell
- Starts During Corium Drainage Can Persist Thereafter
- Geometry Dependence 6, Core-Concrete Attack Starts During Corium' Drainage OccurE in Pedestal
- 7. Debris Solidification
' - Occurs in Drywell
- Influx fate to Drywell Low Enough to Allow Cooling e
1
l IMPACT OF THESE EVENTS -
- 1. Only Part of Core Inventory Particloates in Concrete Attack
- 2. Debris Must Reheat to Initiate Concrete Attack
- 3. CoriumlInflow, Corium Outflow, and Concrete Attack Occur Simultanecusly
- 4. All These Factors Tend to Lower the Releases O
O
< J .
- - , ~ . - - . , , _ , - - - ~ - , - - - . . , , - - - - - - . - - , - - , - - . - - - - , - , . ,
. _ . . ~ . , . ~ . . . . =
EQUUS MODEL
- 1. Equilibrium Thermodynamics
- 2. Ideal Solution Model
- 3. Compound Library can Be Varied
- 4. Gibbs Free Energy Minimization 9
f I
=~-_
Start I
i Initialize Thermal Mocal -
Initialize Chemical Mocal .
-l
' Y Determine Event Flags 7---
I
--_------_--_h------------------------ T I
Determine Rates-of-Change: DECCMP i
i i '
V
,' Heat Transfer l Gas and Slag Flow Rates i
- i l i i I i I i V No j Call EQUUS? Update Flow Rates i
i Yes i I i
' V l
Equilibrium and Flow Rates i
, U Heat of Reaction :
U Integrate State Variables Set Next Timestep y
Output Print and Plots V
Exceeded Problem Time?
q Yes Finish e
Fig. 4.3 CECCMP/ EQUUS Flowchart.
- l
1 H
" A mO LICLr3 METAL LlO U D O XiD E X I,m I X i,o g
i' A **
GAS
/
X 1,g A so 1
i Fig. 4.1 Abstract pool representation, i ~
.--.--__....o._.-_.--..--.-..-.-______,__--_-_,.-..--, - - _ . - .
O 4' f EQUILERIUM
)~-
k "
. WITH IGTALS c'on ~~
\ s N
- o2
% m _02 METALLIC VOLATILES EQUILIBRIUM STEP +1: METALS AND GASES i
EQUILIBRIUM c WITH OXIDES co j H' /
/
U j OXIDIC
! VOLATILES 1
EQUILIBRIUM STEP +2: OXIDES AND GASES rig. 4.2 Abstrac: equiitectum. c.:r.s.n u e,cn.
j .
ek we-- ' - - ,,.----em ww-,y----w.v-.-e- -,,em , ,,.,,,.-.v----
-,,-py,y---i--y,-,,--..-yy--w- -.-p , ---q --. --* --yg.,--,w- - - - - -
. . - - ._. .. - -- - _ - ~ + - - -
~.~~;.-.-- -- ~= '- ;
o: ,
l Table 4.1 STEPS IN COMPUTATION FOR COUPLED DECOMP-EQUUS RUNS Corium temperature, slag and gas flow rates from !
DECOMP 45> !,
Assume 2K 0 and Fe 230 dec mpose to X, Fe0, and 0 2 I i
4W>
l Metal phase equilibrium from EQUUS using' all concrete offgas and metallic portion of slag 9
0xide phase equilibrium from EQUUS using H,, CO, H7 0, CO2 gases from metal equilibrium and Oxidic partion of slag ,
e 9
Assemble releases (gas from results of each phase equilibrium) and retention (Itquids and solids from results of each phase equilibrium) ;
.~ _
e 4
O
[. ] -: . ; .
TC SEQUENCE BASE CASE .
e Vessel Failure 7000 Seconds '
e 20% of Core Ipput at 1700K (assumed net result of Quenching transient) 1 e 47% More of Core Added Over 9560 Seconds (2.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />) at 2500K e CRD Flow 11 kg/sec, 310 K for 3 Hours e Overflow to Drywell Allowed e 5.6% Zr Oxidation In-Core h
l-
~
, TC SEQUENCE ALTERNATIVE e 100% of Core Input at 2500K e Confined to 2 Times Pedestal Area (58'm2 )
e Other Data the Some O
e s-0 n - , - - ,y- . - --m - , - - , -- -n- - - - -.e.. - -n--,-----n--,-n,-..w-- --c - . - - - ,- c. , .n- ,-- - - - -.-,.,-n -
n .. _
Table IV.5-1 SELECTED DATA FOR EQUUS RUNS
- 1. Plant Geometry Cavity Radius 3.05 m Pedestal Sump Depth 0.20 m
Zr Mass , 64.1 MT Zr Oxidation, TQW 10.6%
Zr oxidation, TC 5.6%
- 3. Fission Products and Control Material (gram moles)
Sr0 700.
Bao 750.
La 0 350.
23 Ce0 1500.
2 l
Mo 2500.
, Ru 1700.
BC g 30000.
4 Concrete Composition CORCON-MOD 2 Limestone-Common Sand 5 t
+
1 e
4 4
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(
~
3 TC RELEASE FRACTIONS FROM DEBRIS Confined Base Full Case Drop Sr 6,0E-3 3.5E-2 Ba 7,9E-3 '
4.4E-2 La 1,IE-4 1,OE-3 Ce 3,2E-4 2,8E-3 l
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DISTANCE ERODED (M) '
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EQUUS AND VANESA DIFFERENCES: EXTERNALITIES
- l. All Integrated Analysis Inputs Which Determine Debris Tgmperature
- Debris Composition Debris Location, etc.
- 2. Data and Compound Library-IDCOR has reauested but not received parametric fits used in VANESA V J .
l 4
( } +
EQUUS AND VANESA DIFFERENCES: MODELS
- 1. EQUUS Performs a Second Eaullibrium Calculation: For Oxides l
- 2. Reduction of Major Oxide Constituents Lowers the H /H 0 2 2 and C0/C0 2 Ratios
- 3. This Mitigates the " Coking Blast" O
l i
e T
es l
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--.-..,.-,__----.-----_,,---,_-,_,,,,...,_,,,_,,,---...._,.,,,,,-,,,,,,,_-.__-,,_,....,n.--
m CONTAllK NT AND REACTOR BUILDING PERFORMANCE IN PEACH BOTTOM. SCENARIOS l
' BY r
l 4
RICHAnn S.- DENNING i
OBallelle .
CotunBus DivistoN -
l -
i i i NRC-IDCOR INronnATroN EXCHANGE MEETING '
DECEMBER 10, 1985 l i
1 '
i i
, l l l i . .' i f
CONTAINMENT AND REACTOR BUILDING PERFORMANCE IN PEACH BOTTOM SCENARIOS SCENARIOS ANALYZED ATUS TC1 - CONTAINMENT FAILURE PRECEDES CORE MELT TC2 - CORE MELT PRECEDES CONTAINMENT FAILURE .
TC3 - TC2 WITH VENTING AT 77 PSIA STATIoM nLACrour T81 - CONTAINMENT FAILURE DELAYED BEYOND VESSEL MELTTHROUGH TB2 - CONTAINMENT FAILURE OCCURS AT VESSEL MELTTHROUGH INTERFACING l_0CA -
V - BYPASS TO REACTOR BUILDING CONTAINMENT PERFORMANCE CONTAINMENT ASSUMED TO FAIL IN DRYWELL AT 132 PSIA A LARGE OPENING IS USED IN THE ANALYSIS (7 FT2 )
IIIGH TEMPERATURES ARE PREDICTED TO OCCUR IN DRYWELL (~1500 F) suT Foll IN VENTED SCENARIO TEMPERATURES RISE To ~800 F l
l .
i .
I
- e , }8
m SECONDARY CONTAINMENT PERFORMANCE REACTOR BUILDING VotuME - 1.5x106 FT3 DESIGN PRESSURE - 3 PSIG INTERCONNECTING AREA TO REFUELING BAY - 400 FT2 REFUELING BAY ~
i VOLUME - 1.0x106 rT3 -
DESIGN PRESSURE - SMALL i
BLOWOUT PANEL AREA - 240 FT2 SGTS INoeERATIvE AryEn painagy CONTAINMENT FAILURE I
I I
l 5
E i
~
d ORNL-DWG 82-5881 ETO r .
~
SL0wout m.Owout PMG5 PA#ELS
_ REFUELING BAY .
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LOCATION or Tiig Ng I pgig3gY CONTAINMENT llITHIN '
Tile SECONDARY CONTAINHENT (REACTOR llu!LDING) 9
), e
O ui .
INFtuENCE OF HYDROGEN EOMBUSTION .
EFFECT ON SECONDARY CONTAINMENT STRUCTURES NO CREDIT FOR FISSION PRODUCT RETENTION IN REFUELING BAY PREDICTED PRESSURES OF 8 - 15 FSID IN REACTOR BUILDING POTENTIAL FOR SIGNIFICANT DAMAGE EFFECT ON RELEASE FROM REACTOR BUILDING EAcn BURN EFFECTIVELY EXHAUSTS THE CONTENTS OF THE BUILDING e
e 9
9 e
e l
4 t
I i!
i
1 V
PEACH BOTTOM TB2 24 0 i
j REACNR BUILDING
........... mm g gg
~
M 220- .
1 D.
~
.\ Ed D .
m 200-M D.
g IS O -
M s
M D.
2 lec -
o O ~
y > -'w
-- ~
14 0 , , ,
00 2000 4000 8000 8000 1000 0 I200 0
- 14000 TIME - (MINUTE)
/
FIGURE 4.34 SECONDARY CONTAIGENT PRESSURE RESIWSE FOR Ill2 '
O d PEACII BOTTOM TB2 sco p
g 14.0-E-*
Ex.
g 12 0 -
M l
M a n00-m M
M 80- -
4 0 .
h 80-m p 40-a O
> 20-00 2000 4000
& l l 8000 8000 1000 0 12000 14000 TIME - (MINUTE)
FIGNtE 4.36. 10TAl_ VELINE E GASES LEAKED FOR 1R2 . .
e
$1
PEACH BOTT0ii - REACTOR BUILDliiG . .
OF OF REACTOR BUILDING SCENARIO
[ gg (g TC1 2.0 1.3 1.4 TC2 2.8 2.1 2.3 TC3 1.9 1,7 1,7 TB1 1.1 1.4 1.6 TB2 2.9 1.7 1.5 V 1.8 1.8 1.7 O
9 9
e e
4 4
r -
CHARACTERISTICS OF RB DEPOSITION t
DOMINANT RETENTION MECHANISM - DIFFUSIOPHORESIS SECONDARY RETENTION M CHANISM - GRAVITATIONAL SETTLING CHARACTERISTIC DEPOSITION TIES DuRING HIGH STEAM CONDENSATION *l7 MINUTES DURING MODERATE STEAM CONDENSATION ~2 HOURS WHAT IF NO HYDROGEN BURNS?
PEAK GAS CENERATION FROM CORE CONCRETE 250-5 M MOLES /SEC
~
h - 1 v0LuMES PER HOUR ~
l RANGE OF POSSIBLE F IS 1.5 - 8 wITH E v2 MOST LIKELY FOR THE PERIOD OF PEAK GAS GENERATION F AT LATER TIME (E.G., DURING REVAPORIZATION) COULD BE GREATER BECAUSE OF LONGER RESIDENCE TIME l
l l
l l
l t 6 e ,, -D
l l
bTSTAIIDlIBG ISSUES i
j UILL GLOBAL HYDROGEN BUR $8S DCCUR?
)j UAAT IS THE STRUCTURAL RESPOIISE OF THE REACTOR BUILDIIIG?
I WOULD A DETAILED MULTI-VOLUHE ANALYSIS WHICH INCLUDES BOOYA80CY
{ 800RE OR LESS RETENTIcel 0F FISSI001 PRODUCTS? ,
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! " Reactor Species RCS Pool Wetwell Drywell Melt Building Environment '
I 0.70 0.23 1.6 x 10-8 7.7 x 10-3 0 3.1 x 10-2 3.1 x 10-2 -
Cs 0.70 0.22 2.0 x 10-8 g,o'x 10-2 0 3.4 x 10-2 3.3 x 10-2 Te 0.29 0.I1 1.2 x 10-7 4.0 x 10-3 0,.31 5.4 x 10-2 0.26 Sr 5.4 x 10-4 0.12 3.2 x 10-8 6.8 x 10-3 0.25 0.13 0.49 Ru 7.6 x 10-7 1.2 x 10-7 2.5 x 10-12 1.6 x 10-9 1.0 1.3 x 10-7 3.5 x 10-7 {.
La 7.8 x 10-8 6.0 x 10-5 3,1 x 10-9 2.5 x 1.0-4 0.98 5.1 x 10-3 1.2 x 10-2 Ce 0 5.3 x 10-3 l'.5 x 10-9 3.4 x 10-4 0.% 7.2 x 10-3 2.2 x 10-2 .
Ba 1.0 x 10-2 8.0 x 10-2 2,4 10-8 4.4 x 10-3 0.42 8.7 x 10-2 0,39 I
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IASLE 5.14.
DISTRIENTISI 0F FISSION PGMMCTS BT GAOMP--TC2 SCEllARIO Species RCS Pool Reactor idetwell Drywell Melt -
Building Environment
-1 0.39 0.58 1.2 x 10-7 5.5 x 10-4 0 2.3 x 10-2 g,3 "x 30-2 Cs 0.55 0.42 8.6 x 10-8 5.2 x 10-4 0 2.6 x 10-2 1.4 x 10-2 le 0.62 8.0 x 10-2 7.2 x 10-7 1.0 x 10-3 0.12 9.7 x 10-2 8.7 x 10-2 Sr 5.7 x 10-4 8.8 x 10-4 2.7 x 10-7 3.0 x 10-3' 0235 0.34 0.30 Ru 9.5 x 10-7 7.7 x 10-8 6.0 x 10-11 1.4 x 10-8 1.0 4.1 x 10-7 2.0 x 10-6 La 9.2 x 10-8 3,3 x 30-5 3.0 x,jo-8 9.6 x 10-5 0.98 1.2 x 10-2 9.2 x 10-3 Ce 0 4.8 x 10-5 3,5 x 30-8 1.6 x 10-4 0.% ,
1.9 x 10-2 1.5 x 10-2 84 1.1 x 10-2 5.0 x 10-4 1.7 x 10-7 2.1 x 10-3 0.54 0.25 0.19 e
- e. . .
lAOLE 5.19. SI51RlWil8N OF FISSIM PROONCIS SV 6AguP-IC3 SCENARIO Species RCS Pool Reactor Wetwell Drywell Melt Boilding Environment 1 0.39 0.61 I.6 x 10-7 2.4 x 10-3 0 5.6 x 10-4 6.2 x 10-4 Cs 0.55 0.45 1.9 x 10-7 2.5 x 10-3 0 9.7 a 10-4 9.6 x 10-4 Te 0.62 0.24 1.8 x 10-7 1.0 x 10-2 0.12 4.7 x 10-3 6.5 x 10-3 I Sr 5.7 x 10-4 0.63 9.4 x 10-7 3.5 x 10-2 0.'34 6.0 x 10-3 9.2 x 10-3 Ru 9.5 x 10-7 6.6 x 10-7 ' I.4 x 10-10 1.3 x 10-8 1.0 2.8 x 10-7 4.1 x 10-7 La 9.2 x 10-8 2.1 x 10-2 3.5 x 10-8 1.2 x 10-3 0.98 2.3 x 10-4 3.2 x 10-4 Ce 0 3.3 x 10-2 5.1 x 10-8 1.9 x 10-3 0.% 3.3 x 10-4 4.9 x 10-4 \
Ba 1.1 x 10-2 0.41 6.6 x 10-7 2.4 x 10-2 0.54 4.2 x 10-3 6.2 x 10-3 l
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TASLE 5.26.
Bl5TRISMTION OF FISSION PRODUCTS BT GA00P-IBI SCEm4 ale Species RCS Pool Reactor Wetwell Drywell Melt Building Env irotunent I 0.85 5.0 x 10-2 8.8 x 10-7 7.8 x 10-2 0 1.4 x 10-3 1.2 x 10-2 Cs 0.85 3.0 x 10-2 7.0 x 10-7 8.8 x 10-2 0 8.9 x 10-3 1.4 x 10-2 Te 0.38 6.3 x 10-3 2.8 x 10-6 7.3 x 10-2 0.23 0.10 0.22 Sr 1.1 x 10-3 2.9 x 10-2 8.3 x 10-8 0.31 0.16 0.13 0.37 Ru 1.6 x 10-6 3.4 x 10-8 1.2 x 10-10 1.1 x 10-7 1.0 3.5 x 10-7 6.0 x 10-7 La 1.6 x 10-7 4.1 x 10-3 4.3 x 10-8 1.2 x 10-2 0.94 1.8 x 10-2 3.1 x 10-2 Ce 0 5.3 x 10-3 3.7 x 10-8 1.8 x 10-2 0.91 2.0 x 10-2 4.8 x 10-2 Ba 2.1 x 10-2 5.0 x 10-2 4.6 x 10-7 0.16 0.38 0.12 0.28 i
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I TABLE 5.34.
i DISTRIBUTION OF FISSION PRODUCIS SY GROUP--TS2 SCENARIG Species RCS Pool Reactor Wetwell Orywell Melt 1
Building Environment i
i 1 0.85 4.0 x 10-2 2.0 x 10-7
- 1.1 x 10-2 0 6.8 x 10-2 3.6 x 10-2
) Cs 0.85 3.0 x 10-2 3.1 x 10-8 1.2 x.10-2 i 0 7.6 x 10-2 4.1 x 10-2 j Te 0.38 3.1 x 10-2 2.1 x 10-6 8.4 x 10-3 0.23 0.12 1
0.23 Sr 1.1 x 10-3 6.2 x 10-2 4.7 x 10-7 8.6 x 10-3 0.1 0.30 i 0.46 j Ru 1.6 x 10-6 5.5 x 10-8 6.1 x 10-11 2.7 x 10-8 1.0 2.9 x 10-7 7.0 x 10-7
! La 1.6 x 10-7 9.4 x 10-5 2.7 x 10-8 6.8 x 10-4 0.95 1.8 x 10-2 3.5 x 10-2 j Ce 0 l.6 x 10-4 4.6 ,x 10-8 1.1 x 10-3 0.91 2.6 x 10-2
- l 5.5 x 10-2
- ~ Ba 2.1 x 10-2 4.4 x 10-2 3.5 x 10-7 6.3 x.10-3 0.38 i 0.20 0.34 I i i
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TABLE 5.39. '
DISTRIBUTION OF FISSION PRODUCTS sv Ga00P--V SCENARIO J
1 Species RCS .-
Pool . Reactor '
Orywell Melt Building Environment j ,
1 0.42 4.1 x 10-2 2.8 x 10-2 0 0.23 0.28 Cs 0.47 5.1 x 10-2 ' 3.5 x 30'2 0 0.20 0.24 Te 9.1 x 10-2 0.13 0.13 0.28 0.12 0.24 Sr 2.2 x 10-4 0.23 0.18 0.16 0.18 0.24 Ru 3.2 x 10-7 1.2 x 10-7 1.0 x 10-7 1.0 2.8 x 10-7 1.2 x 10-6 La 3.3 x 10-8 1.2 x 10-2 9,5 x 10-3 0. % 9.5 x 10-3 1.3 x 10-2 Ce 0 1.7 x 10-2 2.0 x 10-2 0.92 1.5 x 10-2 2.3 x 10-2 -
Da 4.1 x 10-3 0.15 0.14 0.38 0.12 0.20 e
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H DROGEN GENERATION AND BURN Robert E. Henry Fauske & Associates, Inc.
NRC/IDCOR Meeting December 10, 1985 1
1 44
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IN- VESSEL HYDROGEN GENERATION MODEL e Clad oxidation (Intact geometry).
1 e Clad levitation model (melt or eutectic forma tion). .
e Clad levitated - oxidation continues.
Clad slumped - oxidation cutoff.
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COMPARED TO SFD RESULTS Hg Generated (Grams)
Test Measured Calculate d SFD 1- 1 56 1to i
SFD 1-3 56 t 11 50-100 SFD 1-4 185 t 30 180-210 1
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EX~ VESSEL HYDROGEN GENERATION MODEL e Debris distributed on pedestal and drywell flo ors.
e Core-concrete attack determined by
" DECOMP" including debris quenching after vessel failure.
e Hydrogen generated by Zircaloy oxidation In the debris.
e Hydrogen generation rate limited by core-concrete thermal attack rate.
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t FEATURES CONTROLLING COMBUSTION IN THE'SECONDAR Y CONTAINMENT e Oxygen available.
e Steam in th'e atmosphere.
e Hydrogen Ignition.
e Building nodalization.
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OX YGEN/ STEAM AVAILABILITY TC e Oxygen displaced at containment venting or failure - little ingression late in time.
e High steam partial pressure at all times -
structural heat sinks quickly saturated.
, e Only upper node receives any oxygen - no l
Hg burning.
Blackout e Oxygen available.
e Moderate steam partial pressure.
e Hp at sufficient temperature to ignite as it
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e Nodalization not a major influence.
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SECONOAAY GAS: AM8 TENT Alt BETWEEN 100*F ANO 450Y l
NOZZL1 FLOWS 0.1 SCFM
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NOZZLE: 0.305 IN.10 -
$ 0 IGNITION SOURCE: HOT WIRE 5 O INCREA5tNG H2 PERCINTAGE h g ;l _
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Pt!MAAY CAS TEMPERATURE, 'F PC)
HEDL 7903-301.1 Ignition of of a Hydrogen-Nitrogen Jet (Without Sodium) as a Function Jet Temperature.
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REACTOR BUILDING NODALIZATION e Should consider circulation between compart-ments and within a compartment.
e Circulation within a compartment driven by wall-to-bulk temperature differences.
e Circulation between compartments depends on the number of parallel flow paths. -
- Two or niore prallel paths - efficient circulation, f
- One path . effectively no circulation.
, e if circulation between compartments is small -
more than a single node is required.
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( )m CONCL USIONS IDCOR MARK I ANALYSES 9
TC No burning in the secondary containment as a result of high steam and low oxygen concentrations.
TOUV Local burning due to self-ignition as the hydrogen is released from the primary containment.
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