ML20028C776
| ML20028C776 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 01/07/1983 |
| From: | Wilcove M NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Sinclair M SINCLAIR, M.P. |
| References | |
| NUDOCS 8301140089 | |
| Download: ML20028C776 (29) | |
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- January 7,1983 -
. UNITED STATES OF AMERICA >
' NUCLEAR REGULATORY' COMMISSION-BEFORE THE ATOMIC' SAFETY AND LICENSING BOARD-In the Matter of.
CONSUMERS POWER COMPANY, Docket Nos. 50-329 50-330 (Midland Plant,. Units l' and 2)
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NRC STAFF RESPONSE TO THE FIRST INTERROGATORY SUBMITTED BY MARY SINCLAIR'ON AUGUST 25, 1982 I.
INTRODUCTION On August 25, 1982, Intervenor Mary Sinclair filed " Discovery Questions to the Nuclear ~ Regulatory Connission on New Contentions Accepted by Board Order, August 14, 1982." By letter dated September 2, 1982, the Staff advised the Board that the Staff would voluntarily-respond to.those interrogatories. The~ Staff now provides its response to the first of the interrogatories.
Responses to other interrogatories contained in Ms. Sinclair's August 25, 1982 submittal.will be-forthcoming.-
Similarly, affidavits in support of the responses,below will be filed shortly.
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II.
INTERROGATORY INTERR0GATORY-I Contention 3 questions the adequacy of the methodology in the DES for determining the possibility of severe accidents at the Midland nuclear plants, and recommends NUREG/CR/2497, as a better basis.
Question 1 of Interrogatory I:
-The.FES includes an extensive discussion of the uncertainties asso-ciated with the numerical estimates of the likelihood, as well as the 0RIGIN DESIGNAT 3
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consequences, of severe reactor accidents that the DES did not carry.
For example, the FES states tha.t it is the judgment of the staff that "the uncertainty bounds could be well over a factor of.10, but not likely to be so large as a factor of 100"(5-48).
NUREG/CR/2497 estimates that the Rasmussen study-(relied upon in DES 5-46-66) underestimates the risk by a factor of 20. To what extent did the new uncertainty bounds in FES depend on NUREG/CR/2497 for their new uncertainty estimates?'
Response
The same. uncertainty bounds govern the probabilistic risk assess-ments in both the DES and the FES. Section 5.9.4.6(7) of the FES is merely an explanation of uncertainty. bounds inherent in the environmental assessment of severe accident conditions.
Contrary to this question, it is not a' discussion of new uncertainty bounds.
The uncertainty bounds of the calculation are not based in any way onNUREG(CR)2497.
Question 2 of Interrogatory I:
Provide copies of any studies (including NUREG/CR/2497) which dddress risk assessment of accidents subsequent to the Rasmussen study.
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Response
Plant NUREG #
Program Status
.Surry 1, 2
~ Published-Sequoyah 1 NUREG/CR-1659/1 of 4 RSSMAP
. Published Oconee 3 NUREG/CR-1659/2 of.4 RSSMAP Published Calvert Cliffs 1 NUREG/CR-1659/3 of 4' RSSMAP Published Grand Gulf 1 NUREG/CR-1659/4 of 4 RSSMAP Published Crystal River 3 NUREG/CR-2515 IREP Published Arkansas One 1 NUREG/CR-2787 IREP Draft Available Browns Ferry 1 NUREG/CR-2802 IREP Published Precursors to NUREG/CR-2497 Published Potential Severe Damage Accidents:
1969-1979 A Status Report l
These documents have been placed in the loca' public document room.
Docket #
Probabalistic Safety Studies j
Zion 1, 2 50-295, 50-304 ZIP Published I,
Indian Point 2 50-247 ZIP Published l
Indian Point 3 50-286 ZIP Published Probabalistic Risk Assessments Limerick 1,2(Rev4)50-352,50-353 PECO Published Big Rock Point 1 50-155 CPC0 Published Pursuant to agreement with counsel for Ms. Sinclair, these documer.ts, due to their volume, are being listed only. They are, however, available Jn the public document room in Washington,-D.C.
Question 3 of Interrogatory I:
What other studies were done after the DES was issued to prompt this new evaluation of us. certainties in the FES?
Response
No additional studies have been made..The text has been revised from that presented in the DES solely to clarify the subject matter.
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c Question 4 of Interrogatory I:-
Provide corresponoence, me'mos and any other appropriate documents that deal with this new evaluation of uncertainties identified in the FES (5-48) in the following areas:
- 1) oversimplified analysis of the magni-tude and timing of the fission product releases; 2) uncertainties in calculated energy release; 3) radionuclide transport from the core to the receptor; 4) lack of precise dosimetry; 5) statistical variations of-health effects.
Response
The 5 factors' listed are quoted from the text of the FES. They are meant to describe areas of potential uncertainty and to be all-inclusive of the processes of generation of airborne radioactive material in the plant to observation of a particular health effect in an individual or the population (or of an a?fect on the environment). ' The uncertainties in these factors have not been evaluated, either individually or in combination, for Midland.
However, an overview of the subject of uncertainties is given in
" Summary of Committee's Findings" from a technial committee of the International Atomic Energy Agency (Attachment 1). Another document, r
" Assumptions, Sensitivities and Uncertainties," Section 9.5 of "PRA Procedures Guide," NUREG/CR-2300, Vol. 2, Rev. 1, April 1982, is currently being printed. When it becomes available, it will be placed in.
the local public document room in Midland, Michigan.
" Technical Bases for Estimating Fission Product Behavior during LWR accidents," June 1981, contains a discussion of certain aspects of the uncertainty in the source term. The latter document will be placed in the local public document room in Midland, Michigan and is also available in the public document room in Washington, D.C.
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.gestion5of-InterrogatoryI:
The study, NUREG/CR/2497, bases the probabilities of severe accidents on the basis of actual accident consequences and significant events. Using this methodology:
a.
What percertage of ~ accidents were initiated by operator error?'
List them, b.
What percentage were initiated by non-safety system's that failed' and impacted on safety-systems? List them.
c.
What percentage were due to disbelief of actual instrument readings which were not safety grade equipment? List them.
d.
What percentage were due to' instruments actually giving a false reading to operators? List them.
e.
What percentage were due to maintenance during plant operation that disrupted the safety systems? List them.
f.
What percentage were due.to minor mishaps that disrupted larger systems? List them.
g.
What percentage of accidents were due to failure of safety systems? List them.
h.
What percentage were due to poor QA procedures during operation?
1.
What percentage took place when the plant was at less than full power?
- j. Hc,w many of the accidents studied took place at the Applicant's Palisades and Big Rock plants? Describe them.
Response
At the outset, several qualifications should be made concerning the answers to this question. The categorizations requested in the interrogatory can be misleading. Several of the accidents or incidents listed in NUREG/CR/2479 could fit more than one of the specified interrogatory categories.
In other cases, the accidents or incidents do not really fit any of the specified categories.
Considerable judgment was used in-the LER categorization provided in this response.
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example, with respect _to one event, documented by LER accession No. 91676 (seeattachment:2)i two of three' pump's failed due to tight packing (i.e.,
. poor or incorrect maintenance) while the third pump failed due to a mechanical problem.
In this case all of the incident failures are placed in the " failure of safety systems" category although part of the failures could also be classified as poor or incorrect maintenance - if there were such a category.
There is also a problem with some of the terminology that is used-in the question. Not all the occurrences listed in NUREG/CR/2497 which contribute to a severe core damage are " accidents" as implied in the interrogatory. Some contribution to core damage probability are simply incidents which challenge the plant safety systems. However, in answering the questions, all significant core damage contributors were, in this response, considered to be " accidents" to allow placing them in the specified categories.
An attempt was made to categorize the " root cause' of each significant LER, i.e., to determine and classify the basic cause of the accident or incident rather than any subsequent failure of safety systems which might have tended to exacerbate the accident or incident. However,.
where the initiating event was an anticipated transient (i.e., loss of offsite power, loss of feedwater, etc.) then the concurrent failure of any significant and required safety system was considered as the initiating event. Also, "non-safety" system failures impacting safety systems were interpreted as failures adversely affecting the safety i
systems, and not non-safety system initiating events which merely called upon or required the safety systems to work.
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In_NUREG/CR/2497;(page 4-23 bottom) it is noted that only those
- accidents orEincidents of significance category 30 or less contribute-significantly to the severe core ~ damage calculation made in the~ report.
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Hencel only these significant accidents or' incidents are included in the tabulations and categorizations provided in this' response.
With the above qualifications the precursors can be categorized as
'shown be, low.
Refer to attachment 2 for the listing of the specifici precursors which were grouped in;each_ category.
.a) Approximately 19 percent of accidents were initiated by operator-
- error, b) Approximately 14 percent were initiated by non-safety _ systems that failed and impacted on safety systems.
c) Approximately 0 percent were_due to disbelief of actual instrument readings which were not safety grade equipment.
d) Approximately 0 percent were due to instruments actually giving a false reading to operators, e) Approximately 0 percent were due to maintenance during plant operations that disrupted the safety systems.
f) Approximately 0 percent were due to minor mishaps that disrupted.
Targersystems.1/
g) Approximately 60 percent were due to failure of safety systems.
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Note: The incident at Rancho Seco involving a dropped light bulb 1
(LERAccessionNo. 138830,dtd3/20/78)could,dependingon interpretation, be classified as a " minor mishap." In this submittal, the LER was classified as "non-safety system failure impacting safety system."
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h) Approximatley 8 percent were due to poor QA procedures during operation.
- 1) Approximately 65 percent took place when the plant was at less
-than full power.
j) 'Approximately 2 percent:-(one incident) took place at the apr'icant's Palisades and Big Rock plants.
The Licensee Event Report Jescribing this incident is attached (Attachment 3).
Question 6 of-Interrogatory I:
Based on the new information in the FES of the probability of accidents that has increased by a factor of 100, what is the worst case probability of accidents at the Midland reactors and the U.S.?-
Response
Contary to this question, it is not stated in the FES that the probability of accidents "has increased by a factor of 100.
"Rather, it is stated, as part of a general discussion of uncertainties inherent in a probabilistic risk analysis, that the uncertainty bounds in such an analysis may be greater than a factor of 10, but are not likely-to be as large as a factor of 100.
(FES,p.5-48)
Furthermore, the Staff objects to this question on the grounds that it is vague. The Staff is unable to determine what Ms. Sinclair means by a " worst case probability of accidents."
Question 7 of Interrogatory 1:
What is the new probability of risk of accidents calculated for Unit I that has the bad weld?
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. Response No ; evaluation of a' plant specific prob ~ ability.'of any accident at-Midland Unit.1 has been; performed.
- Respectfully submitted,.
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Michael'N. Wilcove Counsel for NRC Staff Dated in Bethesda, Maryland this 7th-day of Jan:ary 1983 6
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4 l-UNITED STATES OF AMERICA'
' NUCLEAR ' REGULATORY ~'COPWISSION 1 BEFORE:THEATOMIdSAFETYAND~LICENSINGBOARD.
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'In the MatterJof=
CONSUMERS POWER COMPANY-
. Docket Nos.;50-329:
50-330 (MidlandPlant, Units 1and2)
'(OperatingLicenseProceeding)'
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSE-TO THE FIRST. INTERP.0GATORY SUBMITTED BY MARY SINCLAIR ON' AUGUST 25, 1982" in the above-captioned proceeding have been served on the following by: deposit in the United. States mail, first class,-
or, as indicated by an asterisk through deposit in the Nuclear Regulatory Commission's-internal mail' system, this 7th day'of January,1983:
- Charles Bechhoefer, Esq.
Frank J. Kelley
' Administrative Judge
. Attorney General'of the State Atomic Safety and Licensing Board
- of Michigan U.S. Nuclear Regulatory Comission Steward H.. Freeman Washington, D.C.
20555 Assistant Attorney General Environmental Protection Division
- Dr. Jerry Harbour -
525 W. Ottawa'St., 720 Law Bldg.
Administrative Judge Lansing, Michigan 48913 Atomic Safety and Licensing Board-
- U.S. Nuclear Regulatory Comission Ms. Mary Sinclair Washington, D.C.
20555 5711 Summerset Street Midland, Michigan 48640 Dr. Frederick P. Cowan Administrative Judge 6152 N. Verde Trail Michael I. Miller, Esq.
Apt. B-125 Ronald G. Zamarin, Esq.
Boca Raton, Florida 33433 Alan S. Farnell, Esq.
Isham, Lincoln & Beale James E. - Brunner, Esq.
Three First National Pla:a C2msumers Power Company 52nd Floor 212 West Michigan Avenue Chicago, Illinois 60602 Jackson,' Michigan 49201 y
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-Ms.- Barbara Stamiris
- Atomic Safety and Licensing Board 5795 N. River U.S. Nuclear Regulatory Comission-Freeland, Michigan 48623.
Washington, D.C. !20555-
-James R. Kates
'* Atomic Safety and Licensing' Appeal 203-S. Washington Avenue Panel-
,Saginaw, Michigan 48605 U.S. Nuclear Regulatory Comission Washington, D.C.
205551 Wendell H.- Marshall,' President Mapleton Intervenors
- Docketing and Service Sec. tion RFD 10 U.S. Nuclear Regulatory Commission ~
Midland, Michigan 48640 Washington, D.C. -20555-Wayne Hearn Ste've' J. Gadler, P.E..
Bay City Times 2120 Carter Avenue-311 Fifth Street St. Paul, MN. 55108 Bay. City, Michigan 48706 Frederick C. Williams-Paul C. Rau Isham, Lincoln &-Beale-Midland Daily News 1120 Connecticut Avenue, NW-124 Mcdonald Street-Midland, Michigan 48640 -
Washington, D.C.
20036 Lee L. Bishop Myron M. Cherry, p.c.
Harmon & Weiss Peter Flynn, p.c.
1725 I Street, N.W..
.L Cherry & Flynn
' Suite 306 Three First National Plaza Washington, D.C.
20006 Suite 3700 Chicago, IL 60602 T. J. Creswell Michigan. Division Legal Department Dow Chemical Company Midland, Michigan 48640 i,
Michael N. Wilcove-Counsel for NRC Staff 4
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-,y INTERNATIONAL ATOMIC ENERGY AGENCY DIVISION OF NUCLEAR SAFETY o
TECHNICAL COMMITTEE MEETING on AIRBORNE FISSION PRODUCT RELEASE FOLLOWING EXTENSIVE CORE DAMAGE ACCIDENTS 12-16 October 1981 s
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RECOMMENDATIONS TO IAEA o
The Technical Committee agreed that the first meeting had
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of the topic in many countries and that research and development
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programmes were being organized in many centres.'It was also noted
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likely that reductions in the conservatism in current source terms
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many phenomena, e.g. aerosol behaviour in both primary circuit and
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,s 9.$hj fission products, from release onwards. It was emphasised that the C ';j presence-of water and/or steam at various stages of the accident
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The Technical Committee recommends that a second meeting be
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held in about one' year's time. Icess which should be included in the L
Agenda icelude :
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Review of research programmes and results to date, g,'
includingpatafromTMI.
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Update of current licensing procedures and regulatory perspectives on siting and emergency planning as specifically related to source term.
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Review of the NEA aerosol work if activity 3, below,
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Report from the Consultants Meeting (or Advisory 3
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Croup) on accident experience if activity 1, below, is e
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The Agency was also asked to undertake the following activities :
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Request Member States to review in their countries
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events (accidents, destructive tests, etc.) involving fission f'
product release to see if these contribute to the understanding of 1
behaviour in the cast of water reactor accidents. If support from i
3 the Member countries justifies it, the Agency should convene a t
Consultants' Meeting (or a small Advisory Croup) in the spring of
)fy 1982 which should report to the next Technical Committee meeting.
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Delegates were requested.co make suitable information available on planned and on going experimental work for preparation T
of a report by the Secretariat for distribution to Connaittee d,
Members. The Secretary would circulate proposals for the form of the contributed items. Typical headin;s include (a) Fission Product Release, (b)' Fission Product Chemistry, (c) Aerosol Behaviour (d)
,h Integral-Experiments.
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The Connaittee noted that the current uncertainties in k
aerosol behaviour justified further expert discussion and suggested
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.ex si ting Expert Group on this topic. They cautioned that the terms of reference should not be set so narrowly as to exclude other aspects important for aerosol behaviour.
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INTRODUCTION
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ile f/je During the 'weeki12-16 octobe'r 1981 the IAEA convened a 1,q meeting on Airborne Fiasion Product Release Tollowing Extensive Core Damage Accidents. The meeting was attended by experts from 14 Member
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States and one International Organization. The major purpose was'to I
review a central and current issue of the nuclear reactor safety R
coassunity - the matter of the size of the " source tern" -in view of
,.j recent developments questioning the validity of traditionally used
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values - in a broad international setting. A second purpose was to
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so, what it is and when to implement it.
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The subject matter of the meeting is not only controversial
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4j between member states but also within the member states themselves gi,. gI and it is furthermore very broad in its need for expertise in many.
...,'l disciplines. Yet an attempt to reach a consensus on certain aspects
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d't was made. In areas where no consensus seemed possible the h
diff erences of opinion were isolated and highlighted.
l The more formal presentations made by the participants are
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.given in the following section. In this section the Secretariat has
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j week and, hopefully, to retain the flavor of the meeting. The A
j distillation of many hours of technical discussion into a few readable pages necessarily requires that many details be lost. It is l
hoped that the inclusion of the summary presentations in the next j
section will make up for that loss. The participants represented a 22h variety of viewpoints. On the one hand it was recognized that
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whereas one source term might be appropriate for siting, for i
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considerations or in the design of engineered safety features. As an
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alternativethe[rrobabilisticriskassessmentapproachYouldbeused
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the impact of plant design on source ~ters. The inherent differences'.
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between PWRs and BWRs'will lead to significant dif ferences in source
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term for many,. accident sequences.
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b.m The coalescencg of the week-long' discussions into a finite-
-4 framework as presented in this section was achieved through the G'j.a-f guidance of the chairman, Mr. Gilby, and with the cooperation of the 9
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. experts in all of the ' scientific disciplines from the various member states. It is from this summary of the week's discussions that'the 3_;
reconsnandations to the Agency in the previous section have been 3@.:s gleaned. The summary statements by the Technical Committee given in this section are presented in five parts :
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. Accident Scenarios
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. Release of Fission Products from Fuel c f..
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' Chemistry of Fission Products
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. Aetosol Behaviour 3
. Magnitude of the Source Term - Relation to Fission Product A
iMd Chemistry and Aerosol Behaviour.
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~ ACCIDENT SCENARIOS
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The general understanding of the timing and progress of the
'T major stages of' core meltdown scenarios in LWRs is good although ~
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there are uncertainties associated with human factors. The release' and transport of radioactive sacerial is, however, closely n'
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. influenced by the detailp of the-th'ernal and hydraulic behaviour of-
.j the scenarios. The current ability to predict the timing of the
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release of fission products from the fuel, the formation of 3
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aerosols, the transport of radioactive sacerials 'through the ' reactor 4
coolant system and the containment, the effect of possible retention J
.a mechanisms and ultimately the magnitude of release to the :
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environment is limited by the accuracy of thermal, hydraulic and' y
.j structural analyses. Although the description of the physical'
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processes of reactor _ accidents does not fall within the primary purpose of the meeting, the Coassittee feels that the following areas
<gg of deficiency that have 'a direct influence on the magnitude of the
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The estimation of the core tcaperature profile versus 1.
time di
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The prediction of surface temperatures, flows, water l
content and condensation in the reactor coolant system j
and in the containment system
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The prediction of the mode and timing of containment s.
failure and the release pathways which result.
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i RELEASE OF FISSION PRODUCTS FROM FUEL
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The release of fission products from defected fuel rods is l
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well understood to fuel temperatures of about 1200 C. Release data
,7b from 1200 C to the appearance of molten m'acerial (about 1800*C) however, are sparse ; nor has an adequate mechanistic description
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been developed. This temperature range is important for the release i ge]
$ g; of the noble gas,es and the semi-volatile fission produE s, such as
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molten material appears to be adequate, but is based upon
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Little information is available on the later stages of the
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and leached by water, release during interaction of the molten core
'.M.>. '.. I with the concrete basenac, and release from dispersed fuel material
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ru chenium, releases in the latter two situation will impact on the
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source term only through their effect on aerosol behaviour.
.:Y L Osl 7,f fj The Committee recognizes the lack o^ experimental data in e g.
g4.g some areas, particularly of the release of the semi-volatile species 4j 0$.v$:5]
over the approximate temperature range 1200 - 1800*C. These data
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several radiologically important fission product nuclides (such as
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cesium, iodine, and cellurium) are associated with particulates.
~
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Also, it has been noted elsewhere that aerosol characteristics 3
1.3 depend to some extent on the chemical forms and rates of fission
- J
.3 product release from the fuel. Other than these areas, however, the
.. :f p
Committer notes that only a limited effect on source terms could be expected by additional studies of fission product. release.
.'lJ CHEMISTRY OF FISSION PRODUCTS
.i Recent studies have let to a better understanding of iodine behaviour under accident conditions. These studies may explain the small iodine releases described in the paper reviewing accidents-in
~E h Section 3 of this report.
V4
. On;
)hff.
.There are major uncertainties associated with organic
,. O,j iodide formation and al'though total iodine partitioning into' the gas phase is small, organic iodides may be the dominant airborne
'g.'
chemical forms.
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Zquilibrium thermodynamic calculations for reducing steam
~$
conditions in the primary system indicate that CsI'is the dominant iodine chemical form. Similar calculations for cesium show that Cs0H r
and Cs1 are the most stable species. Tellurium is predicted to appear as elemental Te at temperature below 5.00*C and as Te2e I*
and H Te at higher temperatures. Since reaction rates are fast at l
2 e,... '
high temperatures, the equilibrium calculations may give a e
~
reasonably accurate' desc~ription of primary system chemistry. Below about 500*C, the vapour' pressures of the I, Cs, an'd Te species-will be low and very little will be in the gaseous state.
[j If water is contacted in the primary system, iodine and J.
cesium will be converted to non-volatile I and Cs+.
Cs does not - form volatile aqueous species and, will not be. transported to the
,j gaseous state. Iodine, in dilute solutions, exists primarily as g,j-non-volatile I orIO}overabroadrangeofconditionsand-Mi little partitioning into the gas phase is expected if proper water '
chemistry conditions are established. In this regard, it was noted
~?
that dissolved impurities were unlikely to lead to oxidizing
- ..gl }
conditions.
,.a In dilute solution radiation effects are not expected to be
.}
j important but for the high concentrations of cesium todide which may 4
be expected in aerosols there is a possibility of formation of some 2
]
elemental iodine.
J!l j
In most reactor accident scenarios, particularly those
'j involving aqueous conditions, other fission products (except for the j
noble gases) will not be volatile a'nd will not be expected to occur in a gaseous form.
]
-l The Committee recognizes that there are some major
.c 4 f
conservatisms in the treatment of fi,ssion product chemistry in
'"d existing risk studies and believes that much of the necessary work m'
3
. d;8 has already been carried out to enable a more realistic assessment
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source term for all scenarios involvint non-oxidizing ano aqueous
'l conditions. Some further work is still necessary, however in ' the
~~
.a 7,4 following areas :
9;. M ;g 11.,. ;
1.
Organic iodide production
- h. F i Cn;'4 j;j]'j 2.
Reaction rate measurements to assess the relevancy of
, 'l thermodynamic prediction at low temperatures
.,'44 I [,(
3.
Radiation effects in Cs! solutions R..L.
4
'Jacer chemistry between about 200'C and the critical
~ d.l point
,f..&;a
'p.9.).l 5.
Chemistry of other fission products such as Te.
.1
, t :i:
ri
.)
AEROSOL BEHAVIOUR
'i j
There is a good general understanding of aerosol physics
]
and quanticative calculations using existing codes have been I.'. j validated in a range of conditions including initial concentrations
~'I 3
j op to some tens of grams / meter. However, aerosol production and
.]
behaviour can be very dependent on the particular scenario and may
" y:)
well vary with variation in temperature, flow race, and the various
,i
- 7. g parameters associated with the thermalhydraulic conditions in the
.q various event sequences.
,,4 Recent review papers have suggested that for short times in
'4 a fuel melting accident there may be extraordinarily high density 3
4' serosols (kilograms / meter ) in and near the damaged core. Some i.~5f recent papers have suggested that large decreases in these aerosol
,,}
concentrations must occur if due account is taken of the very large aerosol den'ities, the extensive surface area available and the j
s
( ?.,h.
presence of water and steam in many scenarios.
.gj 1*
It has been suggested that the aerosols would carry much of
- il the airborne fission product inventory in fuel melting accidents.
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"Jhile the physics of aerosol beliaviour, once airborne. is fea'sonably
.).
well understood, other matters of importance are -in need 'of
'i definition. These' include :
4 i
1.
The rate of production of aerosols as an function of 1
temperature, heating rate, pressure, steam flow and
[:
time y
2.
The relative timing of the escape of fission products -
l (especially cesium, iodine, tellurium) and the 3
production of aerosols
. l' 3.
The use of t'hermalhydraulic computer programmes as 4
input for the prediction of movement and behaviour of
-bl aerosols
~ s.):
[s 4
Experimental validation of aerosol code predictions for appropriate conditions.
}
~4 1
The possibility of existence of very dense aerosols in 5.
.)
]
specified accident sequences and their subsequent
)
stability.
I
,1 6.
An investigation of a number of specific effects
.i associated with GR accidents which have not yet been
. g adequately studied, including :
.i
+
=
- 3 I
- a. effect of hydrogen combustion on aerosol benaviour j
- b. possible resuspension of aerosols and fission 1
products at various stages of the accident 1
}
- c. effects of steam condensation on aerosol and 1
i fission product behaviour
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1
- d. effect of small steam explosions on aerosol behaviour
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.j MACNITUDE OF THE SOURCE TERM - REI.ATION TO FISSION
- [
PRODUCT CHEMISTRY AND AEROSOL BEHAVIOUR
~
l The incorporation of realistic descriptions of aerosol
')
j behaviour and fission. product chemistry into p' articular water
}y, '.
reactor ' accident sequences is still in an early stage. There is a?/4Ad general agressent that there will be reductions in the source terms i
, developed in earlier US ind German Risk Studies but there is also uncertainty in the amount of these reductions. For sequences in which water or wet steam is present in che' flow path, o'r in which
?
the containment remains intact, or in which failure is delayed for some hours or more the reduction in s'ource term,could well be several orders of magnitude. These reductions could be important ie consideration of emergency planni,ng and siting, but possibly would
- -.14 not significantly affect overall risks as currently evaluated. For 9
4*'t ']
sequences involving early containment failure it is less clear that I
such significant reductions in source term will result from naw i
1 assessments. However, even a breached containment maj delay the s
'A release of aerosols for many cens of minutes and so give an order of 3
magnitude attenuation factor ; this is especial'ly so if condensing 3
1 conditions are predominant. Major uncertainties exist in the current I,
ability to predict the mode and tining of containment failure.
Another point of major difficulty concerns the behaviour in the i
j primary circuit where aerosol densities, steam temperatures 2nd flow rates are very different from those in which aerosol codes are
.i usually applied.
The Committee recognizes the great importance of developing j
a full understanding of the source term and reducing unnecessary conservatisms. To achieve this desirable objective the following two
,.i items arose from the work of the Committee.
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1.
A better understanding of thermal hydraulic conditions
~9j fer an appropriate range of accident scenarios is required.
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Application of aerosol codes to conditions J.
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experimental prograamme designed to validate such
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predictior.s is required.
I These studies should allow for the appropriate chemistry and water and steam conditions.
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i l' a ACCESSION 06!565 J N.IN[.Y M?i EMe'!i d o.i; LOSS OF POWER AND DIESEL GENERATOR h ! d < *y CONSUMERS POWER COMPANY F}.;;.'
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DOCKT-502SS-53 +. 12.PAGES. LETTER - CONSUMERS POWER COMPANY TO DIVISION p
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OF REACTOR L ICEhSING (AECB - SEPTEMBER
- 9. 1971. DOCKET 50-255 P
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T Y PE-P WR's ' M FG --C O MB'..N AE-- EECHT Et;~
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s THE 345 KV ARGENTA NO 2 L INE TRIPPED AT 0706 ONSSEPTEMBER 2.
1971.
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GREAKERS 29H9 AND 2SR8. CLEARING THE 345 KV BUS AhD RESULTING IN THE
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LOSS OF POWER TO THE ST ART-UP TRANSFORMERS.
THE EMERGENCY -DIESEL O
4
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GENERATORS BOTH STARTED AND QU ICKLY ACHIEVED R ATE!D SPEED WITH THE I-I 9
UNIT PROPERLY'.CLOSIhG,IN CN THE DEAD 2400IV BUS 4. AND P ICK ING UP LOAD AS T.O (
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FAULTY OREAKER FAILURE RELAY ON THE 27R8 BREAK ER WHICH CAUSED THE DUS (sf'b'U?%
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AVAILABILITY: AV AILADILITY Q.yg!N)j
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