ML20092B884
| ML20092B884 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 01/31/1992 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | |
| Shared Package | |
| ML20092B879 | List: |
| References | |
| PROC-920131, NUDOCS 9202110169 | |
| Download: ML20092B884 (18) | |
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i RELIABILITY ASSURANCE PROGRAM PLAN FOR THE l
SYSTEM 80+ NUCLEAR POWER PLANT ,
JANUARY 1992 REV. 00 ABB COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR, CONNECTICUT i.
DRAFT l
ABB ASEA BFtOWN DOVEnl l ADO 02 A PDR
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TABLE OF CONTEhfS SECTION Eggg -
TABLE OF CONTENTS-LIST OF FIGURES LIST OF-ACRONYMS
1.0 INTRODUCTION
1.1 PURPOSE- ;
1.2 SCOPE 2.0 PRA' PROGRAM ELEMENTS 2.1 PRA G0ALS 2.2- PRA METHODOLOGY.
2.3 PRA DURING DESIGN AND CONSTRUCTION 2.4 -PRA DURING PLANT OPERATION 3.0- RAMI PROGRAM ELEMENTS 3,1 RAMI ANALYSIS 3.2 PLANT RELIABILITY DATA BASE
- 3.3 CORRECTIVE ACTIONS PROGRAM 4.0 RELIABILITY CENTERED MAINTENANCE PLAN
-4.1 RCH Philosophy
-4.2 RCH PHASES
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5.0 PROCEDURES AND TECHNICAL SPECIFICATIONS 5.1 TECHNICAL SPECIFICATIONS 5.2 PLANT 0PERATING PROCEDURES 5.3 EMERGENCY OPERATING PROCEDURES 5.4 SEVERE ACCIDENT MANAGEMENT PROCEDURES C.$ SECURITY 6.0- ORGANIZATIONAL AND ADMINISTRATIVE SUPPORT
7.0 REFERENCES
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'l LIST OF FIGURES FIGURE Paae Z-1 . MAJOR PRA' TASKS ,
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t s LIST OF ACR4.rHS ALWR Advanced Light Water Reactor '
- CFR Code of Federal Regulations COL Combined Operating License E0P - Emergency Operating Procedure-EPRI Electric Power Research Institute !
FMEA Failure Modes and Effects Analysis F0AKE First Of A Kind Engineering KAG Key assumptions and Groundrules LCO Limited Condition of Operation MTBF Mean Time Between Failures l MTTR Mean Time To Repair !
PRA Probabilistic Risk Assessment PSA Probabilistic-Safety-Analysis .
RAMI Reliability, Availability,-Maintainability, and Inspectability l 4
RAP Reliability Assurance Program '
RIG Risk Based-Inspection Guide i R-Y Reactor Year '
SAMP Severe Accident Management Procedures SAR Safety Analysis Report SRP ' Standard Review Plan T/S Technical Specification
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. , l RELIABILITY ASSURANCE PLAN FOR THE SYSTEM 80+ NUCLEAR POWER PLANT 1,0 INTRODUCTION 1.1 Purpose System 80+ is g3 standard nuclear power plant design that is to be certified under 10CFR Part 52' As such, a level 111 Probabilistic Risk Assessment (PRA) has been prepared at the onset of the design and licensing processes, and will be updated during the subsequent engineeriN, procurement, and construction processes, This PRA will be maintained and updated as the design details increase, and will be delivered to the owner / operator upon completion of plant
.startup. The owner / operator will either maintain the PRA themselves or have the PRA maintained by another organization as a living document that reflects the operating plant as it evolves.
The Reliability Assurance Program (RAP) defines a program for maintaining consistency between the System 80+ PRA and the Plant configuration. The program will ensure that the Procedures and Technical Specifications and plant configuration (including maintenance) are consistent with the PRA. The program defined herein is intended to cover the entire life-cycle of a System 80+
Standard Design Nuclear Power Plant. This plan may be modified by the holder of the Combined Operating License (COL) to contain plant specific ir f The RAP is specified as part of the EPRI ALWR Ut'.!ity Requirements'3)ormation.
1.2- Scope The RAP describes the elements of the program for. maintaining the PRA. and conducting a Reliability, Availability, Maintainability, and Inspectability (RAMI) program, and a Reliability Centered Maintenance (RCM) program for the entire plant (both the NSSS and B0P) covered by the certification, it assures consistency between the PRA bases and the plant operation, maintenance and configuration. The RAP plan will be updated and expanded as appropriate as tb design moves through certification, first Of A Kind Engineering (f0AKE) and plant specific engineering. Further updates to the RAP will be included in this report as the-project progressed through procurement, construction, and operation.
The RAP describes the interface of the PRA with the plants Operating Procedures, Emergency Operating Procedures, Severe Accident Management Prccedures, Test and Maintenance Procedures, Technical Specifications and security.
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2.0 PRA PROGRAM ELEMENTS The RAP program includes the elements that are necessary to ensure that the PRA is maintained consistent with the plant configuration and operation. This requires a living PRA that reficcts the plant as it progresses from design and construction, and through the operation phase. Therefore the PRA program will be integrated with the other aspects of tie plant life cycle. The living PRA is being integrated with the design, operation procedures, maintenance procedures, emergency procedures, and general management of the plant.
2.1 PRA Goals The RAP assures that the bases of the PRA remain valid and that the plant continues to meet the ALWR reliability and safety objectives. The safety objectiveg/R-Y than 5x10' and a cancer mortality risk of less than 2x10'}/R-Y.for the ALWR is to h This implies a large release potentinl for offsite early fatalities of le s than lx10'6/R-Y.
The core damage frequency for ALWRs is to be less that lx10'g/R-Y.
2.1.2 PRA Methodology Standard methods were used by ABB-Combustion Engineering in the performance of the System 80+ PRA"). The level 1 (core damage frequency) portion of the analysis is equivalent to the base described in the PSA Procedures Guide'}ine
) probabilistic and the safety analysis methods employed (PSA) were consistent with methods outlined in the PSA Procedures Guide and methodologies described in the PRA Procedures Guide). The methods used in the PRA were also in conformance with the recommendations of the "PRA Key Assumptions and-Groundrugs" in Appendix A to Chapter 1 of the EPRI ALWR Requirements Document . The small event tree /large fault tree approach is used for the evaluation of core damage frequency.
External events are defined as those events that result in a plant perturbation or transient, but are not initiatgd g e plant systems. External events were identified by reviewin past PRAs' ' and PRA guidance documents such as the PRA Procedures guide"g the PRA Fundamentalsprepared
, gcument"6' by BNL, and the ANS guide for selecting external events' . Events with similar plant effects and consequences were grouped together. Criteria were established to determine which external events are insignificant risk contributors and thus can be excluded from detailed quantitative evaluation. The screening criteria was based on design requirements set forth in the EPRI ALWR Utility Requirements Document (7), generally acc ted regulatory practices as documented in the NRC Standard Review Plan ("RP)'g3 considerations, and gtaeric siting considerations.
Each external event identified was then evaluated against the screening criteria tc determine whether detailed quantitative analysis is needed. This evaluation also consideghginsights gained from a review of PRAs for present generation power plsnts The_ methods used for the level 2 severe accident progression, containment response and source term analyses were consistent with the methods used in NUREG-Il50ca), the methods described in the PRA Procedures Guido(6) and those methods recommended in the EPRI ALWR Requirements Document (7) . The level 3 analyses
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l (consequences). al Procedurp Guide Document,2 and9go.use the methodsmethods recommendedconsistent with Requirements in the EPRI ALWR those described in Figure 2-1 shows the major PRA tasks.
All three levels of the PRA require a bases of the plant configuration, equipment mean time between failure (MTBF), mean time- to repair - (MITR), inspection intervals, operator procedures, and other aspects of the plant operation. These bases are given in the report *Analysj)s Assumptions for the System 80+ Standard Design Probabilistic Risk Assessment " This document is an integral part of the living PRA and is updated as needed.
2.3 PRA During Design and-Construction This full-scope PRA program has been conducted using a representative site. Event tree and fault tree models were developed for the design. These models were integrated and used to estimate the feasibility of meeting the plant risk and core damage frequency goals and to provide insight into design decisions. A component reliability data base and component naming convention were established.
A baseline level 1 PRA model was developed.
In the second phase, the scope of the PRA was extended to provide detailed models of the support systems, to include a detailed containment analysis and to calculate consequences in terms of off-site doses. This phase identified the dominant core damage contributors and the dominant contributors to off-site releases. The models were used to determine the impact of design changes on core damage and on large release _ frequency and to identify the dominant contributors.
This information was be fed back to the system designers for consideration in the design.
The third phase, following Design Certification by the U.S. NRC, will involve a continuation of the interactive reliability assurance process in which, the PRA practitioners participate during the F0AKE phase. The system f ault tree models developed in earlier phases will be-modified to evaluate and reflect proposed system design enhancements and details and the engineers with system design responsibility, will be continually appraised of the reliability of their systems
! vis-a-vis achievement of the plant risk objectives.
During the procurement and construction phase, the PRA will be maintained current to reflect the "as built" and site specific design and procedures. The PRA will be delivered to the owner / operator after completion of construction and startup.
l 2.4 PRA During Operation The PRA will be installed on a computer and delivered for use at the Utility.
Appropriate training and documentation will be provided so that the Utility
, Engineering staff will be able to maintain the PRA current over the life of the l plant, and to use the PRA as -input to operations and maintenance decisions.
The Utility Staff will maintain the PRA using procedures to be developed by the Utility. The PRA will be used to evaluate potential- design changes. System
- fault trees will also be used in the RAMI Program and to support the Reliability
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Centered Maintenance Program. The PRA will be updated at regular intervals with plant specific data collected in the Plant Reliability Data Base (part of the RAMI program). In addition, The PRA will be used to support the- Significant Event Evaluation Pro fam by using the PRA to evaluate plant events as precursors to core. damage sequences. Updated system fault trees will also be used to track the Technical Specifications and LC0 conditions by tracking the dependencies of equipment and systems,
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l 3,0 RAMI PROGRAM ELEMENTS The RAMI Program will be developed by Combustion Engineering to predict and track plant availability in the same way that the PRA follows plant risk. The RAMI Program will be conducted within the context of a RAP Program. The PRA and RAMI efforts use data, methods and analyses which are similar and complementary.
After plant startup, the utility will maintain the RAMI Program and ensure that it is consistent with the plant configuration, procedures, and operating history.
3.1 RAMI Analysis in the design process, reliability engineers sub-divide the top level quantitative Capacity Factor requirements-into system level quantitative design requirements. These quantitative requirements will be addressed by each system design engineer. As the system designs evolve, the reliability engineers perform Failure Modes and Effects Analyses (FMEAs) and perform fault tree analyses for systems determined to be important to the plant's ability to meet its quantitative requirements. Standard methods will be used in the performance of the System 80+ RAMI fault trees. The RAMI modeling will be performed in manner much like the Level I (core dgage frequency) portion of the analysis described in the PSA Procedures Guide' and the methods employed are consistent with methods describe (; in the PRA Procedures Guide * . The results of '.hese analyses are provided to the -design engineers to confirm that their designs meet quantitative requirements and identify which specific design characteristics are limiting. _ As th design progresses, the system level and component level quantitative reliability requirements will be established to assure that top level requirements are met cost effectively. Through this iterative process, communication is maintained between system designers and reliability engineers.
The central features of the RAMI program are:
- 1. Reliability Analysis - Failure Modes and Effects Analysis are used to evaluate the potential impacts of component malfunctions on plant operability;
- 2. PRA Fethodology - Fault Tree Analysis methods are used tr probabilistically predict and quantify plant availability and plan; capacity factor.
- 3. Design Review - A formal design review procedure is implemented to provide a vehicle for assuring that communication is maintained between the syste designers and the engineers performing the RAMI analyses.
The compor.ent failure rate data base, for use in the fault tree analyses, will be updated during the F0AKE phase, with generic data chosen at the beginning and being replaced with design specific data as it becomes available. This revised data base will become part of the PRA data base and integrated into the Plant Reliability Data Base (see Section 3.2).
The Design Review Meetings will provide the forum for system designers, reliability engineers and project management to discuss RAMI consideration and
1 Ir tes. stems that are being specifically addressed in the i
- s> 41yses and in interactions between reliability er.gineers and syst ...lude:
1; identification of component failures, combinations of component failures, test and maintenance errors, and operator errors that can lead directly, or through the 'achnical specifications, to 'an outage or reduced-production,
- 2) Identification of Critical Components; !
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- 3) Identification of Dominant causes of outages,
- 4) Identification of each system's status with regard to meeting the RAMI l quantitative goals. '
The RAM!' Analyses will be maintained current with the design as design details become availab,a during the construction and start-up of the . plant, in addition, the final RAMI Model will be installed on a computer and delivered for use at the Utility. Appropriate training and documentation _ will be provided so that the Utility Engineering staff will be able to maintain the RAMI models current over
- the life of the plant, and to use the RAMI models as input to operations and maintenance decisions.
3.2 Plant Reliability Data Base Both the PRA, RAMI, and RCH programs need an integrated data base. The Plant Reliability Data Base started out as.the PRA dats base. The PRA data base is also used for the initial PAMI models. This data base is expanded to include plant specific data as _it is accumulated. -This insures that the Living PRA, RAMI, and the_RCM programs use consistent data, it enables an easy comparisen of generic data and ' plant specific data.
- A PRA data base was ' developed during tb detign phase.and will be expanded during the procurement and construction phases.
PRA data is needed for the quantification cf the system fault trees and the system accident sequences which result in severe core damage. The data needed for this quantification inuudrs:
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A.- Initiating event frequencies,
- 8. -Component failure rates (demand and time-dependent),
C. Component repair times and maintenance frequencies, D. Common cause failure rates, E. Human failure probabilities, F. Special event probabilities (e.g. restoration of offsite power),
- G. Error factors for the. items above.-
E Generic reliabilig data are being used appropriately per the guidance in the PSA l Procedures Guide' . The primary source of data used for the PRA in the L
'= Preliminary: Design Phase are the "PRA Key Assumptions and Groundrules" (gG) document (Appendix A to Chapter 1 of the EPRI ALWR Requirements Document ).
uther industry-accepted generic data sources will be used as needed to supplement the data in the KAG.
The Plant Reliability Data Base will contain both the generic PRA data and plant specific data on the same items. This will enable a comparison of the data, and upgrading of the PRA and RAMI analysis. Dates of maintenante requests will be stored so that Aging Analysis can be performed. The plant Reliability Data Base will-be consistent with the NPRDS, a nuclear component data base maintained by INPO.
3.3- Corrective Actions Program Part of the RAP program is the Corrective Actions Program. This program has been placed as part of the RAMI section because the most common corrective actions will deal with availability improve &nts with both the nuclear island and balance of plant. The Utility will develop a Corrective Actions Group that will review suggested plant changes to ensure that they are consistent with safety and plant availability goals. The Corrective Actions Group will also review all reactor trips, NRC SER Reports and reported events at other sites that could have significant availability or safety 5plications. The Corrective Actions program will involve senior representatives from Plant Management, Operations, Maintenance, PRA, and the RAMI groups. Its chairman will report directly to the Plant Manager. Details of the group composition and procedures will be described in this section of the RAP at a later time. Its organizational interfaces are given in Section 6.
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a 4.0 RELIABILIT) CENTERED MAINTENANCE PLAN
-The PRA incorporates the mean time between failures, mean time to repair, and inspection intervals on the equipment that support plant safety. Thor., equipment characteristics are strongly affected by the maintenance program. Reliability Centered Maintenance (RCM) is a structured, programmatic appronn to determine how to prudently and economically maintain plant equipv' t RCH embodies the attributes of reliability, availability, maintainability p t inspectability. An RCH program will be developed during the plant specific design phase of the System 80+ development with sufficient breadth and detail to support operational decisions.
4.1 RCH Phases The detailed RCH Program Guide will be developed during the Plant Specific design phase and either included here or referenced in this document. The RCH program will be integrated with the PRA program. The PRA group will supply to the maintenance planning group the HTTR, MlBF and inspection intervals used in the PRA, They will also supply to the RCH group the major sequences leading to core damage and an evaluation of the importance of each system in terms of plant risk reduction. The maintenance planning group will review the PRA bases and ensure that it is included into the RCH program. Discrepancies between the PRA and RCH programs will be eliminated in an iterative fashion. The various phases of the RCH program are outlined below.
Initiation Phase:
This phase involves developing and organizing the following information:
- 1) Description of system function for each operating mode. Description of component functions for all components within the system which can affect the system function.
- 2) Descriptior,s of how the system and its components perform their intended functions.
- 3) Descriptions of performance tasks which are required to enable the '
system and its components to continue to perform their intended functions.
- 4) Description of the PRA and RAMI models for each system including equipment data.
- 5) Develop Risk-based Inspection Guides (RIGS) based on the PRA and other available guidance.
Implementation Phase:
This phase includes developing:
- 1) Specific descriptions on how to perform each of the performance tasks,
- 2) Identification of the resources required to accomplish each of the performance tasks.
- 3) Descriptions of the scheduling process for each of the performance tasks.
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. I Testing Phase:
1 This phase in which the RCH Program is tested involves:
- 1) Identification of functional reliability factors currently being produced by the system and its components.
- 2) Descriptions of the - differences between the actual system reliabilities and intended reliabilities.
- 3) Identification of the Effectiveness Coefficients of the maintenance program for the system and its compotents.
Evaluation Phase:
This phase involves:
- 1) Identification of the causes for the differences between the actual
-system reliabilities and the intended reliabilities. .
- 2) Descriptions of.the changes in the performance of tasks which need to be made in order to reduce the differences between the a tual system reliability and the intended reliabilities.
- 3) Descriptions of any changes which need to be made to the system or the system reliability.model in. order to make the ideal reliability goals more practically achievable.
- 4) Review the age related characteristic of all components for wearout.
- 5) Review the cost-effectiveness of current maintenance procedures.
- 6) Review the maintenance requirements for evolutionary trends.
- 7) Review for consistency with the PRA and RAMI programs.
5.0 PROCEDURES AND TECHNICAL SPECIFICATIONS Imbedded in the System 80+ PRA are requirements dealing with the availability of equipment, and their inspection and maintenance frequencies. The PRA also I contains assumptions about operator actions during transients and additional recovery actions that an operator will take after system failures or during an accident sequence. The RAP Plan ensures that the bnes used in the PRA are consistent with the plant procedures and Technical Specifications.
5.1 TecHicti Specification The Plant Technical Specifications (T/S) describes the operating envelope for the plant. It specifies what equipment must be available and how long the plant can operate with a piece of safety releted equipment out of service. Surycillance requirements and frequencies are also specified. 1he operating conditions $n terms of temperatures, pressures and fluid icvels are specified in the T/S to ensure that they are bounde! by the safety analysis presented in the SAR. The validity of the PRA is also dependent on plant operating limits as specified in the T/S. The PRA group will supply to the procedures group the initial system descriptions from the PRA for review by the procedures group. All proposed or ,'
actual T/S changes will be transmitted by the procedures group to the P'lA grouc to be evaluated in the living PRA. The Plant Manager will be notified of any T/S changes that adversely tffect the plant risk with respect to the safety goals.
.If the Utility chooses to use a computerized T/S monitoring system, the PRA group it responsible to maintain the system fault trees used in the monitoring system.
This is to ensure that tha T/S models are consistent with the plant configuration and PkA models.
5.2- Plant Operating Procedures Co.. .stency between the Plant Operating Procedures and the PRA will be mair.tained. The operations group will be given system descriptions from the PRA f or their review and comment. Changes in the operating procedures will be transmitted to the PRA group to be incorporated in the living PRA. The PRA group will perform c special review of all shutdown procedures for their effect on shutdown risk.. The Plant Manager will be notified of procedural changes that adversely effect the plant risk with respect to the safety goals.
l 5.3 Emergency Operating Procedures The Emergency Operatio Procedures (E0Ps) descriN the operator actions during-transients and off r., tal events. it is important that there is consistency l between the PRA and tu E0Ps as both evolve during the plant life. The PRA L crntains assumptions about operator actions for all transients. The procedures group will be given system and transient deswriptions from the PRA for their review and comment. They will also be given the dominant (mast probable) l: sequences and equipnant failures so that they can ensure that the E0Ps reflect I the most probable accident sequences. Changes in the E0Ps will be transmitted to the PRA group to be incorporated in the living PRA. The Plant Manager will i
be notified of E0P changes that adversely effect the plant risk with respect to the safety goals.
5.4 Severe Accident Management Procedures The Severe Accident Management Procedures (SAMPs) are to provide the Plant with a framework for evaluating information on severe accidents, and ensure effective response to credible severe accidents. They will most likely be extensions of the E0Ps and will guide the operator to maintain the safety functions and safety goals within the context of the E0Ps. The SAMPs require the evaluation of phenomenological behavior of core and structural material beyond design base conditions. Both equipment and instrument performance in severe environments must be evaluated for the selection of strategies to mitigate consequences of the accident. The Severe Accident Management Procedures (SAMPs) give operators guidance during events be ' It is important that there is consistency between the PP) ond design bases.and the SAMPs as both evolve dur The PRA contains assumptior:s on operator actions for all transients and include recovery actions after s) stems fail and recovery actions during accident srtuences. The proceduras (,roup will be given system and transient descriptions anc . potential recovery actlons from the PRA for their review, comment, and inclusion in the procedures. They will also be given the a description of the most probable sequences and equipment failures predicted in the PRA so that they can ensure that the SAMPs reflect the most probable accident sequences. Changes in the SAMPs will be transmitted to the PRA group to be incorporated in the living PRA.. The Plant Manager will be notified of SAMP changes that adversely effect the plant risk with respect to the rafety goals.
5.5 Security The PRA identifies the most likely (in terms of frequency) sequences that lead to core damage and the importance of each system in preventing or mitigating core damage and large releases. This " road map" is important to plant safety. Plant
-Security will be given a summery of the importance of each system in preventing accidents anel the dominant (most likely) sequences that accidents mig 1t follow.
This information is important to the prevention of sabotage and in emergency preparedness planrIng.
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6.0 ORGAN!IATIONAL AND ADMINISTRATIVE SUPPORT This section will contain the organization charts for the Utility, plant Staff and Designers who support the RAP program when such information becomes available.
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7.0 REFERENCES
- 1. 10 CFR Part 52; Early Site Permits: Standard Design Certification and Combined Licenses for Nuclear Power Reacto.as; 54FR 15372; April 18,1989.
- 2. EPRI, Advanced Light Water Reactor Utility Requirements Document, Vol. !!!, Chapter 1: Overall Requirements, Paragraph 6.2.1.1, Reliability Program, August 31, 1990.
- 3. " Analysis Assumptions for the System 80+ Standard Design Probabilistic Risk Assessment", ABB Combustion Engineering Nuclear Power, DCTR-RS-03 Rev.0, May 1991.
4 Base Line Level 1 PtababilistlLH11LAisessment for_ the_1y11em 80-NS.51 Ds11gn; Enclosure (1)-p to LO-88-008; Combustion Engineering, Inc.;
January,1988.
- 5. NUREG/CR-2815; finhal 111 tit.11fety Anah11s_ProcedargLGuide; Papazoglou, I . A. , Bari, R. A. ; :tre okhaven National Lab.; January 1984.
- 6. NUREG/CR-2300; PRA Procedures Guidg; April 1982.
- 7. Adyinced Liaht Water Reactor Utility Re.guirements Docuneat, EPRI, Draft, April, 1987.
- 8. NUREG-ll50; igrere Accident Riihi: An Assessment for five_11._S._Nutltar Egger Plants, Second Draft for Peer Review; U. S. Nuclear Regulatory Commission; June, 1989.
- 9. Zion Probabilistic Safety Study, Commonwealth Edisun.
- 10. Oconen PRA: A Probtbilistic Risk Antisment of Oconeg_llnit 3, NSAC/60, June 1984,
- 11. Seabroo_k Station Probabilistic Safety Assgism_RD1, PLG-0365; Prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company; Pickard, Lowe and Gerrick, Inc., June, 1984.
- 12. Indian Point 2 and_3 Probabilistic Safety Study; prepared for Consolidated Edison Company of New York, Inc., and Power Authority of the State of New York; Pickard, Lowe, and Garrick, Inc.; March, 1982.
- 13. itvannah River Site _Probabilislic Risk Auessment_af_Eca_clar_0nerttlan, Draft by the Westinghouse Savannah River Company, August 31, 1989.
- 14. NUREG/CR-4840; Procedures for the External Event Core Damage Freausnc.y Analyses for NUREG-llEQ; Bohn, M. P., Lambright, J. A.; prepared for the U.S. Nuclear Regulatory Commission by Sandia National Laboratories; November, 1990.
- 15. NilREG/CR-4550, Vol. 3, Rev.1, Part 3; Analysis of Care Damage freautatfl lurry Power Station. Unit 1 External Eventi; Bohn, M. P. et. al.; prepared
for the U. S. Nuclear Regulatory Connission by Sandia National Laboratories; December, 1990
- 16. P,tchtbilistic Risk Assessment: F_undamen1111; prepared by Brookhaven National Laboratory, prepared for the U.S. Nuclear Regulatory Connission; fullwood, R. R.; Harch, 1987.
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- 17. American Nuclear Society; Guideling.s for Colnbinina Naiutal and External Man-Made Hnatds at Powar Reactor Sites; ANSl/ANS-2.12-1978.
- 18. NUREG-0800; Standard Review Plan.fgr_thglerigw of Safety Analysis _flejlatt.1 !
for Nuclgar_fgyer Plants. LWR [htign; U.S. Regulatory Connission; July, 1981.
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