Evaluation of severe accident risks and the potential for risk reduction: Peach Bottom, Unit 2. Main

245 Pages·2015·12.22 MB·English

NUREG/CR-4551 DRAFT FOR COMMENT
SAND86-1309
Volume 3 1 of 2)
Printed April 1987 XA04NO364
INIS-XA-N--068
Evaluation of Severe Accident Risks
and the Potential for Risk Reduction:
Peach Bottom, Unit 2
Main Report
C. N. Amos, A. S. Benjamin, G. J. Boyd, J. M. Griesmeyer, F. E. Haskin,
J. C. Helton, D. M. Kunsman, S. R. Lewis, L. N. Smith
Prepared by
Sandia National Laboratories
AlbLiquerque, New Mexico 87185 and Livermore, Calitornia 94550
for the United States Department of Energy
under Contract DE-AC04-76DPOO789
Prepared for
U. S. NUCLEAR REGULATORY COMMISSION
NOTICE
This report was prepared as an account of work sponsored by an
agency of the United States Government. Neither the United
States Government nor any agency thereof, or any of their employ-
ees, makes any warranty, expressed or implied, or assumes any
legal liability or responsibility for any third party's use, or the
results of such use, of any information, apparatus product or
Process disclosed in this report, or represents that its use by such
third party would not infringe privately owned rights.
Available from
Superintendent of Documents
U.S. Government Printing Office
Post Office Box 37082
Washington, D.C. 20013-7082
and
National Technical Information Service
Springfield, VA 22161
NUREG/CR-4551, VOL. 3 DRAFT REPORT FOR COMMENT (FEBRUARY, 1987)
NUREG/CR-4551
SAND86-1309
Volume 3
EVALUATION OF SEVERE ACCIDENT RISKS AND
THE POTENTIAL FOR RISK REDUCTION:
PEACH BOTTOM, U1T 2
C. N. Amos*
A. S. Benjarnin
G. J. Boyd**
J. M. Griesmeyer
F. E. Haskin
J. C. Helton***
D. M. Kunsman
S. R. Lewis'
L. N. Smith****
Printed February 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 RegulatoryResearch
U. S. Nuclear Regulatory Commission
Washington, DC 20555
Under Memorandum of Understanding DOE 40-550-75
NRC FIN No. A1322
*Technadyne Engineering Consultants, Inc.., Albuquerque, NM
**Safety and Reliability Optimization Services, Inc., Knoxville, TN
'Arizona State University, Ternpe AZ
****Science Applications International Corporation, Albuquerque, NM
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 reviews
will be completed and corrections made, if necessary, prior to
final publication.
This report contains the main body of the report only. The
appendices will be published separately.
PEOUEST FOR COMMENT
This report, NUREG/CR-4551, "Evaluation of Severe Accident Risks and
Potential for Risk Reduction," was prepared for the U.S. Nuclear
Regulatory 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 road public interest such as
probabilistic risk analyses, emergency response planning, siting, NRC
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 Murphy, Division of Reactor
System Safety. These comments will be most useful to the staff if they
are received by June 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 tudy 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.
D. F. Ross, eputy Director
Office of Nuclear Regulatory Research
iii
NUREG/CR4551, VOL. 3 RAFT REPORT FOR COMMENT (FEBRUARY, 1987)
ACKNOWLEDGMENTS
A number of individuals have contributed to this program. In
particular, the authors would like to thank the members of the review
groups listed in Section 2 who provided valuable comments on the
work as it progressed and who participated directly in the collection of
input for the uncertainty analysis. Ile following Sandia personnel also
provided analysis and input for use in this report: R. L. Iman, M. J.
Shortencarier, and D. C. Williams.
ABSTRACT
The Severe Accident Risk Reduction Program (SARRP) has completed a
rebaselining of the risks to the public from a boiling water reactor with a Mark I
containment (Peach Bottom, Unit 2. Emphasis was placed on determining the
magnitude and character of the uncertainties, rather than focusing on a point
estimate. The risk-reduction potential of a set of proposed safety option backfits
was also studied, and their costs and benefits were also evaluated. It was found
that the risks from internal events are generally low relative to previous studies;
for example, most of the uncertainty range is lower than the point estimate of risk
for the Peach Bottom plant in the Reactor Safety Study (RSS). However, certain
unresolved issues cause the top of the uncertainty band to appear at a level that is
comparable with the RSS point estimate. These issues include the modeling of
the common-mode failures for the dc power system, the likelihood of offsite
power recovery versus time during a station blackout, the probability of drywell
failure resulting from meltthrough of the drywell shell, the magnitude of the
fission product releases during core-concrete interactions, and the
decontamination effectiveness of the reactor enclosure building. Most of the
postulated safety options do not appear to be cost effective, although some based
on changes to procedures or inexpensive hardware additions may be marginally
cost effective. This draft for comment of the SARRP report for Peach Bottom
does not include detailed technical appendices, which are still in preparation. The
appendices will be issued under separate cover when completed. This work
supports the Nuclear Regulatory Commission's assessment of severe accidents in
NUREG-1150.
iv
NUREG/CR-4551, VOL. 3 RAFT REPORT FOR COMMENT (FEBRUARY, 1987)
CONTENTS
Section
1. INTRODUCTION ................................................................................ 1-1
1. 1 Background and Objectives ............................................................... 1-1
1.2 Scope of Analysis .......................................................................... 1-2
1. 3 Overview of the Peach Bottom Plant ..................................................... 1-6
1.4 Organization of the Report. ............................................................... 1-7
References for Section ....................................................................... 1-10
2. METHODOLOGY FOR REBASELR*flNG OF RISK ...................................... 2-1
2.1 Overview of Risk Integration and Review Activities .................................. 2-1
2. 1. 1 Integration of Project Activities ................................................. 2-1
2.1.2 Review and Quality Assurance ................................................... 2-9
2.2 Development of the Central Estimate of Risk. ........................................ 2-11
2.3 Characterization of Uncertainties ....................................................... 2-12
2.3.1 Overview of Uncertainty Treatment .......................................... 2-14
2.3.2 Description of the LLH Approach ............................................. 2-17
2.4 Reassessment of Dominant Core-Damage Sequences ............................... 2-29
2.5 Evaluation of Contaim-nent Response .................................................. 2-31
2.5. 1 Development and Quantification of the Containment Event Tree .......... 2-31
2.5.2 Plant Features Important to Containment Response ......................... 2-38
2.5.3 Definition of Containment-Release Modes .................................... 2-40
2.6 Assessment of the Radiological Source Term ....................................... 2-41
2.6.1 Integration With Other Risk-Assessi-nent Tasks ............................. 2-42
2.6.2 Source Term Code Package Calculations .................................... 2-45
2.6.3 Overview of RELTRAC ........................................................ 2-47
2.6.4 Development of Source Terms for the Central Estimate .................... 2-50
2.6.5 Derivation of Source Terms for the I=LH Uncertainty Analysis ........... 2-53
2.7 Offsite Consequence Analysis .......................................................... 2-54
References for Section 2 ....................................................................... 2-59
3 METHODOLOGY FOR EVALUATION OF RISK-REDUCTION OPTIONS .......... 3-1
3.1 Identification of Safety Options .......................................................... 3-1
3.1.1 GenericRisk-ReductionOptions ............................................... 3-2
3.1.2 Risk-Reduction Measures for Peach Bottom .................................. 3-2
3.2 Evaluation of Costs and Other Impacts ................................................. 3-9
3.3 Evaluation of Effects on Risk .......................................................... 3-10
3.4 Value/Impact Assessment .............................................................. 3-11
3.5 Treatment of Uncertainties .............................................................. 3-13
References for Section 3 ....................................................................... 3-16
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NUREG/CR4551, VOL, 3 RAFT REPORT FOR COMMENT (FEBRUARY, 1987)
CONTENTS (CONTUSTUED)
Section
4. EVALUATION OF KEY UNCERTAINTY ISSUES ....................................... 4-1
4.1 Sequence Frequency Issues .............................................................. 4-2
4.1.1 FailuretoActuatetheStandbyLiquidControlSystem ...................... 4-2
4.1.2 Frequency of Dc Power System Common-Mode Failure .................... 4-2
4.1.3 Probability of Failure to Vent During an ATWS .............................. 4-3
4.1.4 Power Recovery Uncertainties .................................................. 4-3
4.2 Containment Loading and Performance Issues ......................................... 4-4
4.2.1 Probability of Stuck-Open Vacuum Breaker .................................... 4-4
4.2.2 Use of the High Pressure Service Water System Spray as a Recovery .... 4-6
4.2.3 Probability that the Operations Staff is Unable to Vent During
Station Blackout .................................................................. 4-6
4.2.4 Probability that the Operations Staff is Unable to Vent After Ac Power
R ecovery .......................................................................... 4-7
4.2.5 Level of Suppression Pool Bypass Through a Stuck Open Safety/
Relief Valve Vacuum Breaker .................................................. 4-7
4.2.6 Containment Failure Pressure and Location ................................... 4-8
4.2.7 Containment Failure Size ....................................................... 4-11
4.2.8 Vessel Failure Mode ............................................................ 4-12
4.2.9 Containment Pressure Prior to Vessel Breach for Station Blackout
Scenarios ......................................................................... 4-14
4.2 10 Containment Pressure Rise at Vessel Breach ................................ 4-16
4.2. 1 1 Probability of Drywell Shell Meltthrough .................................... 4-18
4.2.12 Probability of Hydrogen Bums in Reactor Building Sufficient to
Cause Bypass ................................................................... 4-19
4.3 Radiological Source Term Issues ....................................................... 4-21
4.3.1 In-Vessel Release from the Fuel ............................................... 4-21
4.3.2 Amount of CsI Decomposition ................................................ 4-23
4.3.3 Retention in the Reactor Pressure Vessel ..................................... 4-23
4.3.4 Suppression Pool Decontamination Factor for Aerosols ................... 4-25
4.3.5 Suppression Pool Scrubbing of Volatile Iodine Species .................... 4-27
4.3.6 Revolatilization Following Vessel Breach RVOL) ........................ 4-27
4.3.7 Release from the Melt During Core-Concrete Interactions ................. 4-29
4.3.8 Retention in the Reactor Building and the Refueling Bay .................. 4-31
4.3.9 Late Releases of Iodine ......................................................... 4-33
References for Section 4 ........................................................................ 4-34
5 RESULTS OF RISK REBASELINING ...................................................... 5-1
5.1 Core-Damage Frequency Results ........................................................ 5-1
5. 1. 1 Accident Frequencies ............................................................. 5-1
5.1.2 Uncertainty Representation ...................................................... 5-3
5.1.3 Observations Concerning the Core-Damage Frequency ...................... 5-8
5.1.4 Comparison to the Reactor Safety Study ....................................... 5-9
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NUREG/CR4551, VOL. 3 RAFT REPORT FOR COMMENT (FEBRUARY, 1987)
CONTENTS CONTINUED)
Section
5.2 Containment Analysis Results .......................................................... 5-10
5.2. 1 Central Estimate of Containment Response .................................. 5-10
5.2.2 LLH Containment Analysis Results .......................................... 3-14
5.2.3 Comparison to Other Studies ........ .......................................... 5-17
5.3 Results of the Radiological Source Term Analysis .................................. 5-20
5.3. 1 Source-Term Results for the Central Estimate ............................... 5-20
5.3.2 LLH Source Term Results ...................................................... 5-20
5.4 Offsite Consequence Results ............................................................ 5-29
5.5 R isk R esults ............................................................................... 5-37
5.5. 1 Results of Risk for Latent Cancer Fatality and Early Fatality .............. 5-37
5.5.2 Results for Other Risk Measures .............................................. 5-61
5.5.3 Observations Concerning the Risk Results ................................... 5-64
5.6 Limi tations ................................................................................ 5-68
References for Section ........................................................................ 5-70
6 RESULTS OF RISK REDUCTION ANALYSIS ............................................ 6-1
6.1 Effects of Prevention Options ............................................................ 6-2
6.2 Effects of Mitigation Options on Containment Response ............................. 6-3
6.3 Effects of Safety Options on Risk ........................................................ 6-5
6.4 Costs of Risk-Reduction Options ....................................................... 6-8
6.5 Comparison of Costs and Benefits ....................................................... 6-9
7. INSIGHTS AND CONCLUSIONS ........................................................... 7-1
7.1 Insights and Conclusions from the Rebaselining of Risk ............................. 7-1
7.2 Risk-Reduction Insights and Conclusions .............................................. 7-4
7.3 Lim itations .................................................................................. 7-5
vii
NUREG/CR-4551, VOL. 3 RAFT REPORT FOR COMMENT (FEBRUARY, 1987)
CONTENTS (CONTRTUED)
Section
APPENDICES
A. CONTAINMENT ANALYSIS UNCERTAINTY ISSUES FOR TBE LIN(lITED
LATIN HYPERCUBE STUDY OF GRAND GULF ...........................................
B . SUPPORTING ANALYSIS FOR THE GRAND GULF SOURCE TERM
UNCERTAINTY ASSESSMENT .................................................................
C. DETAIT LISTINGS OF RISK RESULTS ....................................................
D. ASSESSMENT OF RISK REDUCTION MEASURES .........................................
E. SUMMARY OF REVIEW TEAM CONMENTS CONCERNING TBE LIM[ITED
LATIN HYPERCUBE ANALYSIS AS RAPLEMENTED IN SARRP .......................
F. DATA BASE FOR ESTIM[ATION OF COST AND PERSONNEL DOSE FOR
PROPOSED M ODIFICATIONS ...................................................................
viii