Containment Event Analysis for Postulated Severe Accidents

487 Pages·2012·19.34 MB·English

XA04NO370
NUREG/CR-4700 DRAFT REPORT FOR COMMENT
SAND86-1135
Volume 3
Printed May 1987 INIS-XA-N--074
Containment Event Analysis for Postulated
Severe Accidents: Peach Bottom Atomic
Power Stationr Unit 2
C. N. Amos, J. M. Griesmeyer,
A. M. Kolaczkowski
Prepared by
Sandia National Laboratories
Albuquerque, New Mexico 87185 and Livermore, California 94550
for the United States Department of Energy
under Contract DE-AC04-76DPOO789
Prepared for
U.S. NUCLEAR REGULATORY COMMISSION
NOTICE
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DRAFT FOR COMMENT
NUREG/CR-4700
SAND86-1135
Volume 3
CONTAINMENT EVENT ANALYSIS FOR POSTULATED SEVERE
ACCIDENTS: PEACH BOTTOM ATOMIC POWER STATION, UNIT 2
C. N. Amos*
J. M. Griesmeyer
A. M. Kolaczkowski**
Printed May 1987
Sandia National Laboratories
Albuquerque, NM 87185
Operated by
Sandia Corporation
for the
U. S. Department of Energy
Prepared for
Division of Reactor System Safety
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555
Under Memorandum of Understanding DOE 40-550-75
NRC FIN A1322
*Technadyne Engineering Consultants, Inc., Albuq. NM
**Science Applications International Corporation, Albuq. NM
REQUEST FOR COMMENT
This report, NUREG/CR-4700, "Containment Event Analysis for Postulated
Severe Accidents," was prepared for the U.S. Nuclear ReqUlatory
Commission by the Sandia National Laboratories and its subcontractors.
The methods and results set forth in this report are being used by the
NRC to support the development of the Reactor Risk Reference Document
(NUREG-1150) and will be used in areas of broad public interest such as
probabilistic risk analyses, emergency response planning, siting, NPC
safety goal applications, and cost/risk/benefit analyses--indeed,
wherever risks to public health need to be considered in regulatory
applications. Thus, it is considered imperative that an opportunity for
public comment on the results as presented in the report be provided.
Comments should be sent to the U. S. Nuclear Regulatory Commission,
Washington, D.C. 20555, Attention: Joseph urphy, Division of Reactor
System Safety. These comments will be most useful to the staff if they
are received by August 1, 1987.
Of particular interest to us is the receipt of comments on the
methodology, and results, related to uncertainty analysis. One
criticism voiced with respect to the Reactor Safety Study was its lack
of an uncertainty analysis. We have included an uncertainty analysis,
but we are sure that its nature will be the subject of lively debate.
We welcome this, and solicit constructive advice and criticism.
The NRC hereby expresses its great appreciation to all participants in
this study for their considerable efforts, as well as to all who will
take the time and effort to provide it with comments on this report.
_(cid:1)4
D. F. Ross, Deputy Director
Office of Nuclear Regulatory Research
iii
ABSTRACT
A study has been performed as part of the Severe Accident Risk
Reduction Program (SARRP) to investigate the response of a par-
ticular boiling water reactor with a Mark I containment (Peach
Bottom Unit 2 to postulated severe accidents. A detailed con-
tainment event tree for the Peach Bottom plant has been devel-
oped to describe the various possible accident pathways that can
lead to radioactive releases from containment. Data and analy-
ses from a large number of NRC and industry-sponsored pograms
have been reviewed and used as a basis for quantifying the event
tree, i.e., determining the likelihood of the pathways at each
branch point for a variety of accident sequence initiators. A
generalized containment event tree code, called EVNTRE, has been
developed to facilitate the quantification. The uncertainty in
the results has been examined by performing the quantification
three times, using a different set of input each time to repre-
sent the variation of opinion in the reactor safety community.
In the so-called "central" estimate, the likelihood of early
containment failure (occurring before or within a short time
after reactor vessel breach) was found to be significant because
of the possible occurrence of the following phenomena that can
threaten containment integrity: (1) meltthrough of the drywell
shell caused by thermal attack from core debris, and 2 drywell
overpressurization caused by rapid depressurization of the reac-
tor vessel in combination with other events such as direct heat-
ing. However, uncertainties surrounding these issues could
cause the early failure likelihood to be significantly lower
than in the central estimate. This work supports NRC's assess-
ment of severe accident risks to be published in NUREG-1150.
iv
PREFACE
Because of the time constraints imposed on this work to meet the
Nuclear Regulatory Commission's schedule for publication of
NUREG-1150, this draft report has not yet received the full
level of peer and management review customarily accorded to
reports issued by Sandia National Laboratories. The review will
be completed and corrections made if necessary prior to the
final publication.
v

TABLE OF CONTENTS
SECTION PAGE
1. INTRODUCTION .................................... 1
2. METHOD .......................................... 4
2.1 General Description and Observations ....... 4
2.2 Specific Example: Peach Bottom TBUX ........ 8
2.3 Summary of Plant Features and Accident
Sequences .................................. 8
3. RESULTS ......................................... 14
3.1 Cut Set Groups TBUX and TBU ................ 15
3.2 Cut Set Group TB ........................... 18
3.3 Cut Set Group TCUX ......................... 20
3.4 Cut Set Group TCSX ......................... 21
3.5 Cut Set Group TCSRX2 ....................... 23
3.6 Cut Set Group TCSRV23 ...................... 24
4. SUMMARY AND CONCLUSIONS ......................... 26
5. REFERENCES ...................................... 50
App A DETAILED RESULTS OF THE CONTAINMENT EVENT TREE
ANALYSIS ........................................ A.1
App. B PEACH BOTTOM CONTAINMENT EVENT TREE
DESCRIPTION ..................................... B.1
App. C SUPPORTING CALCULATIONS FOR THE SARRP PEACH
BOTTOM EVENT TREE ............................... C.1
App. D SAMPLE EVENT TREE PROBLEM ....................... D.1
App. E EVNTRE INPUT FOR PEACH BOTTOM ................... E.1
vii
LIST OF FIGURES
FIGURE PAGE
3.1 Dominant Accident Progression Bins, Short-
Term Station Blackout ......................... 46
3.2 Dominant Accident Progression Bins, Long-
Term Station Blackout ......................... 47
3.3 Dominant Accident Progression Bins, ATWS
Followed by No Injection ...................... 48
3.4 Dominant Accident Progression Bins, ATWS
With Initial Injection ........................ 49
viii
LIST OF TABLES
TABLE PAGE
2.1 Summary of CET Events for Peach Bottom ........ 28
2.2 Nomenclature Used in Plant Damage State Names. 32
2.3 Mean Frequency of Plant Damage States ........ 33
3.1 Central Evaluation of Peach Bottom
Containment Performance ....................... 34
3.2 Optimistic Evaluation of Peach Bottom
Containment Performance ....................... 42
3.3 Pessimistic Evaluation of Peach Bottom
Containment Performance ....................... 44
ix