ML19309B574
| ML19309B574 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 04/02/1980 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML19309B573 | List: |
| References | |
| NUDOCS 8004040311 | |
| Download: ML19309B574 (10) | |
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O TECllNICAL APPROACl!
TO TiiE PROEALILISTIC RIGK ASSESCMENT OF Tile 1
DIG ROCK POIf!T PLANT 1
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April 2, 1980 8004040 38/
TABLE OF CONTENTS 1.0 Introduction 1.1 Purpose 1.2 Scope 2.0 Elements of the Study 2.1 Development of Plant Models 2.2 Plant Data Evaluation 2.3 Accident and Consequence Analyain 2.h Special Studies 3.0 Form and Uce of Results
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1.0 INTRODUCTION
1.1 Purpose P
The Big Rock Point Nuclear Plant (BRP) has operated successfully without l'
a serious accident for nearly eighteen years.
During this time, as a
- result of evolving regulatory requirements and an industry-wide re-exam-4
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ination of safety issues following the' accident at the Three Mile Island
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Nuclear Plant, numerous safety concerne have been expressed. These con-
-cerns have led to an extensive array of changes which the Nuclear Regula-tory Commission has requested be implemented. The obvious question which 4~
we must ask is: Do serious safety deficiencies exist?
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In'an effort to evaluate the safety of BRP, Consumers Power Company (CPCo)
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has chosen to perform a Probabilistic Risk Assessment (PRA) of the plant.
The reasons for selection of PRA as the basis for re-evaluation are:
1.
PRA methods allow a systematic and comprehensive analysis of of the spectrum of potential accidents at a plant; 2.
PRA provides a thorovgh description of both the probability of occurrence and the consequences of potential accidents; 3.
Site characteristics such as population distribution and meteorology can be treated explicitly by the methods of PRA; h.
Experience with the operation of the plant can be incorporat-ed to include both component and system failure rates and pro-cedures which depict the involvement of operating and mainten-ance staff in plant safety; j
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5 Analyses performed to support the conclusions of PRA have his-l torically represented best estimates of accident phenomena and plant response; 6.
.The insights and models developed in a PRA vill be useful in an evaluation of plant safety issues which will continue throughout the life of the plant.
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1.2 Scope j
i Because the BRP plant is unique in many_ design features including its small power rating.ni low power density, it has been concluded that the PRA must re-exas.iae each of the major assumptions included in the Reactor l
Safety Study (WASH-lh00). This re-examination vill include the develop-l ment of a list of initiating events, preparation of plant specific reli-ability models of all plant functions important to safety, review of operating' experience to define important failure rates for key safety l
related equipment, and a reanalysis of the potential radionuclide release I
categories which will include both definition _of containment failure modes and calculation of radionuclide releases associated with each l
release category.
In addition to these modeling activities, the consequences of potential reactor accidents can be evaluated in terms of effects on human health.
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Finally, using the models developed to describe potential accident sequences at the plant, we vill conduct a series of special studies which' vill address the following issues:
- seismic risk
- risk from fires
- operator error potential The exact form which these analyses will take is currently being defined.
i Subsequent to the characterization of the BRP plant risk contributors, design or procedural alternatives will be considered as means of reduc-I ing the. probability of any risk outliers which are discovered. These
. alternatives will then be analyzed by modifying the baseline plant models to depict changes in risk associated with the modifications.
The final task of the current study will be to define several programs intended to employ the PRA models in a continuing program of safety reassessment. These programs vill be designed to maintain the PRA model as a living description of the current understanding of important safety issues. Possible programs may include:
- reliability growth models
- accident sequence precursor analysis
- revisions to emergency procedures
- root cause analysis of important failure of engineered safety features.
The remainder of this paper will present some additional details of the-PRA program. Several elements of this effort such as the special studies and the list of continuing programs are continuing to be defined in great-er detail. An important element of this overall program is our intention to sponsor NRC participation.in the review group which will meet at approx-l imately monthly intervals to assist in directing overall efforts. The j
opportunity for implementing mid-course corrective measure vill exist at
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these meetings.
2.0 ELEMENTS OF THE STUDY i.
The key elements of the PRA proposed for BRP are discussed in this section.
In advance of embarking on the PRA, the conditions under which the plant will l
be analyzed must be defined, and the potential hazards to be investigated must be tabulated. At the present stage in the program, the assessment is
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expected to be confined to analysis of BRP under full power operation with 3
an equilibrium fuel load'in. place.
In addition to the reactor core, the p
spent fuel pool vill be evaluated as a source of radioactive material with l
- potential to be released during an accident.
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2.1 Development of Plant Models The process of developing models to depict and to allow quantification of the probability of important accident sequences will be an iterative The methods used in this modeling activity will be those develop-one.
ed for application to reactors in the Reactor Safety Study (WASH-lh00).
During the first stage of the modeling activity, important initiating events will be tabulated and grouped by similarity of required plant For each of the groups of initiating events, event trees vill response.
be constructed to investigate and to depict the plant functional response required after the initiator. A compilation of plant functions and the systems required to satisfy those functions vill be developed from the event trees, and each function vill define the need for a fault tree to investigate the failures which are sufficient to disable that function.
Event trees will be constructed for both accident prevention systems and for systems designed to mitigate consequences of accidents (containment systems).
In addition to the construction of event trees, the first stage of the
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PRA vill include the development of dependency diagrams in which the key equipment dependencies for each plant safety function will be depict-ed.
These dependency diagrams will provide the first step in the program to define potentially important sources of common cause failures.
The output of the first stage of the PRA vill be the following:
- a list of important initiating events
- preliminary event trees
- key system deper.dency diagrams
- preliminary definition of the systems for which fault trees are required
- an extensive list of questions on plant system performance under a variety of conditions.
Formulation of answers to the list of questions will support refinement of the event trees and definition of the failure criteria for the systems for which fault trees are required.
It is expected that some transient and LOCA analysis of BRP vill be required to answer these questions.
The second stage of the modeling effort vill include refinement of the event tree models developed in stage 1 and development of fault trees for systems depicted in the event trees. The results of this analysis together with the processed data and the accident and consequence anal-ysis discussed later will provide the basis for the evaluation of risk for the current design of BRP.
The third stage of the modeling effort will be a sensitivity analysis of the BRP plant as it currently exists and an evaluation using the event tree / fault tree techniques of potential design or procedural modi-
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fications to the plant.
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2.2 Plant Data Evaluation Because of' the long operating history of the BRP plant, it 'is important to developJdata on initiator frequency and equipment failure rate which is specific to.the plant. The difficulty expected in trying to charac-terize the operating history of all the safety equipment in the plant is sufficient that only the failure history of equipment expected or demonstrated to be of key safety significance-vill be completely charac-terized. Types of information which will be developed will include:
initiator frequency failure rate or time to failure distributionc
- time to repair distributions.
These data vill te developed to support quantification of accident sequence probabilities using the models described in Section 2.1.
Subsequent to compilation and analysis of important plant initiator fre-quencies,-equipment failure rates, and repair time distributions, com-parisons will be made with industry-wide data being developed by INEL-
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and SANDIA, and significant differences will be investigated.
2.3-Accident and Consequences Analsyis This subsection provides a brief discussion of the specific items which will be included in the consequences analysis of the ERP risk assessment:
- the release category definitions
- the consequence calculations, that is the effect of the radionuclide release on the surrounding population.
2.3.1 Radioactive Release Categories Because of the significant differences between the BRP plant and the Peach Bottom plant analyzed in the Reactor Safety Study, the radionuclide
. release categories (l) are expected _to be considerably different. Examina-tion of both the containment ESP failures and the resulting physical pro-cesses involved in the various event tree sequences for WASH-lh00 reveals patterns of similarity that could be used to characterize the spectrum of releases of radionuclides from the plant. Recognition of these patterns suggests that the selection of sets of representative, or key, accident sequences to define the spectrum of releases is possible. Hence, it is possible to restrict the calculation of radioactive releases only to the key sequences. This representative group of sequences spans the follow-ing differences: 11 ) accident time histories, 2) system involvement in accident' sequences, and 3) magnitude of radioactive releases. The release category definition allows the decoupling of the accident sequence calcu-lations from the. consequence calculations. ' The use of this scheme, how-ever, requires that the radioactive release categories be defined to
'(1) Release ~ Category - These are significantly different categories.which were
-selected by the screening of key accident sequences to identify sequences that are significantly different.from one another in terms of radioactivity
. released from the plant to the environment.
5 represent the releases associated with the dominant contributors to risk for the specific plant, containment, and safety system arrange-ment. Therefore, the plant specific assessment of the BRP plant will be performed as discussed later.
The release categories are important in determining the following for use in the consequence evaluation.
1.
The time of radioactive release; 2.
The warning time for evacuation; 3.
The duration of the release; h.
The elevation of the release; 5
Energy release from containment; 6.
Fraction of core inventory in release.
The calculation of the radioactivity released from the containment barrier for any key accident sequence requires input data obtained from varicus sources. These input data consist of 1) the physical description of the containment; and 2) of the physical phenomena occurring, as well as, 3) the amounts of radioactive materials releas-ed to the containment.
There are a number of physical phenomena which must be modeled to ade-quately represent the potential for radioactive release following a postulated core melt.
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The physical processes of core melt; 2.
The interaction of the molten material with the containment base material and/or water; 3.
The dispersion of radicactivity into the containment spaces; h.
Potential releases of the radioactivity into the atmosphere and its subsequent distribution on the landscape.
The pattern for this type of analysis is adopted from WASH-lh00 which simplified the process of calculating offsite consequences for a wide variety of events by grouping the events into five categories of radio-active release.
The release categories for BRP will be chosen to represent various types of events which can e postulated to occur.
WASH-1400 notes that the release fractions end to be overpredictions since they are based on experiments with relatively small samples with vastly different surface to volume ratios. The rationale for this was purely one of relying upon available data.
In order to determine a unique set of release fractions for BRP, the plant specific items influencing these factors must be re-evaluated for the BRP plant. These factors include:
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. Filter removal; 2.
Plateout; 3.
Resuspension; h.
Other deposition processes; 5
Radioactive decay.
These items are to some degree influenced by the ' plant physical arrange-ment, containment type and-size, primary system characteristics, and miti-gating safety systems. These factors are then coupled with the actual 3 -
reactor operating history, and the factors listed in the table above -
time of release, elevation of release, etc to perform the consequence calculation (ie, the effects of the radioactive release on the environment).
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In summary, plant specific radioactivity release categories.will be defined which reflect the unique features of the plant such as power level, type of nuclear steam supply system, containment type and size, and the particular combination of emergency safety features that are i
provided. Some of the release categories defined in WASH-1400 may be used vith little or no change, but others should be modified with respect to timing specifications and/or radioactivity release fractions.
It is possible that'either additional or fever release categories than used in WASH-lh00 would be appropriate, depending on the existence of unique accident sequences or modified containment failure mode specifi-cations.
L The work plan would involve a careful review and comparison of the major structures and systems in the plant under study against those of the corresponding reference plant used in WASH-lh00. Differences which would
'l be expected to lead to significant changes in particular WASH-lh00 release categories vould be identified, and then analyses performed to determine the required changes. The analyses vould involve recalculation of impor-tant accident sequences with respect to plant physical response and radio-i activity behavior. The calculations would utilize the RACAP. computer code l
group, which is a set of coupled plant response codes'(CONTEMPT, BOIL, PVMELT, INTER) combined with the CORRAL radioactivity behavior code, developed by SAI for analysis of ~ time dependent consequences of severe LWR accidents. Basic core radioactivity release fraction schedules vould l
be modified as appropriate using procedures similar to those described in WASH-lh00. Close coordination would be maintained between this task and
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the work on plant specific accident sequence definition and probabilities to assure that RACAP calculations are performed for all important and l
The results of the analyses would form the.
i unique accident sequences.
basis for devising the set of plant specific release categcries which l
would provide the necessary input for subsequent site specific consequence assessment.
2.3.2 Consequence Evaluation Radiation transport and health effects from a containment failure and a l
given radioactive release category vill be calculated with the aid of an l
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SAI version of the Calculations of Reactor Accidents Code 'CRAC) which van originally developed for use in the Reactor Safety Study. SAI has taken advantage of recent work by its own offices as well as that done at Sandia Laboratories for NRC to develop a site specific model suit-able for utility response to regulatory developments. These develop-ments include a chronic health effects model which conserves radioactive inventory (feature not present in most CRAC versions), an adoption of the straight line plume trajectory model for use with important plume meander and topological effects, an improved evacuation model which better corresponds to EPA evacuation data, and improved output routines allowing for better analysis of results, including sensitivity studies where applicable.
It should also be noted that SAI's version of the CRAC code is an improved version of the NRC preferred model for site specific consequence calculations. It offers the advantages of verifi-ability for use in regulatory response, requires only the venther data that is normally available at most reactor sites, costs significantly less in computer time while maintaining higher statistical confidence, and offers simplicity in modeling.
2.4 Special Studies By employing the event trees and fault treen discusand in Section 2.1, we vill perform a number of special studies. These studies will supple-ment our understanding of the potential for common cause failures at BRP. Typical topics to be addressed in these studies include:
- seismic risk risk from fires
- operator error potential.
The exact form which these studies vill take is currently being defined and status reports on methodology will be presented at approximately one month intervals during the course of the PRA study.
3.0 FORM AND USE OF RESULTS Results from the PRA are expected to be developed in several forms including:
- core dama6e probability and uncertainties
- definition of dominant sequences by release category
- definition of failures contributing to dominant sequences
- distributions of radionuclide releases
- distributions of typical health effects.
These results will be employed as shown in Figure 1.
The principal uses will include definition of risk outliers and employment in the definition of poten-tially useful corrective actions (design or procedural changes).
In addition, comparison with any "non-acceptance criterla" supplied by the NRC will be possi-ble.
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8 FIGWiE 1 USE QE RESULTS 3
DOMINANT SEQUENCES DEFINE RISK OUTLIERS y
DOMINANT FAILURES '
SUPPORT CORRECTIVE ACTION j
DEFINITION CORE DAMAGE PROBABILITY'COMPARE TO "NON-ACCEPTANCE" RELEASE DISTRIBUTIONS F
- CRITERIA SITE CHARACTERISTICS 3
PROBABILISTIC MODELS DEFINE FUTURE DATA NEEDS EVALUATED DATA EVALUATE ACCIDENT SEQUENCE
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PRECURSORS DEPICTION OF PLANT AGING PROCESS
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