ML20055C704
| ML20055C704 | |
| Person / Time | |
|---|---|
| Issue date: | 05/30/1990 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| TASK-PINV, TASK-SE GL-88-20, SECY-90-192, NUDOCS 9006050053 | |
| Download: ML20055C704 (104) | |
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POLICY ISSUE May 30, 1990 SECY-90-192
- Egn, The Commissioners Front James M. Taylor Executive Director for operations subiectr INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT VULNERABILITIES DUE TO EXTERNAL EVENTS (IPEEE)
Purcomet To request Commission approval to issue a Supplement 4 to Generic Letter 88-20 to all holders of operacing Licenses requesting that they conduct an individual plant examination for external event severe accident vulnerabilities and submit this information to the staff.
Backaroundt In the Commission policy statement on severe accidents in nuclear power plants published August 8, 1985 (50 FR 32138), the Commission concluded, based on available information, that existing plants pose no undue risk to the public health and safety and that there is no present basis for immediate action for any regulatory requirements for these plants.
However, the Commission was convincad, based on NRC and industry experience with plant-specific probabilistic risk assessments (PRAs), of the need for a systematic examination of each existing plant to identify any plant-specific vulnerabilities to severe accidents.
The staff implemented this policy, in part, with Generic Letter 88-20, issued November 23, 1988, which requested that all licensees conduct an in'dividual plant examination for severe accidt.nt vulnerabilities due to internal events, including internal flooding.
The request for an indiv!. dual plant examination of external events (IPEEE) for severe accident vulnerabilities was postpon'sd to permit the staff to (1) identify the external "hnnrs,RES NOTE:
TO BE MADE PUBLICLY AVAILABLE 492-3900 WHEN Tile FINAL SRM IS MADE AVAILABLE
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hazards that need a systematic examination, (2) identify examination methods and develop guidance and procedures, and (3) coordinate the IPEEE with other ongoing NRC programs that deal with various aspects of external event evaluations to ensure that there is no duplication of staff and industry efforts.
To accomplish these objectives, an External Events Steering Group (EESG) was established in December 1987 to make recommendations regarding the scope, methods, and coordination of the IPEEE.
The EESG recently completed its task and has recossended guidelines for the treatment of external events.
Dimeussient ~is a proposed supplement to Generic Letter 88-20, which would implement the EESG recommendations.
It describes the objectives, scope, and schedule of the IPEEE and identifies acceptable methods of examination. is a draft of a proposed document based on the EESG final report that.contains further guidance to the industry on the conduct of the IPEEE and on the structure and content of the IPEEE submittal.
If the Commission approves issuance of the proposed generic letter and the draft guidance document, a workshop will be scheduled to review these documents with the licensees and answer any questions they might have on either the generic letter or the draft guidance document.
The staff will then revise the pidance document,.if necessary, and issue it as a final document, starting the IPEEE process.
If necessary, the staff will also issue another supplement to the generic letter at that time.
After the guidance document is reissued, licensees will have 60 days to respond to the staff with their plan for conducting the IPEEE.
The information stemming from the IPEEE itself chould be submitted within 3 years of the issuance of the final guidance document.
The staff expects to have the final pidance document ready for issuance to licensees in November 1990, which would mean all licensees'
'IPEEE submittals would be due in November 1993.
However, for those licensees who choose to perform the internal events IPE and the IPEEE in series, i
requests to extend the time period to submit the IPEEE report will be considered.
I The IPEEE is to be directed toward the same two areas related to severe accidents as was the IPE:
(1) core damage prevention capability and (2) capability to mitigate the radionuclide releases I'
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associated with severe accidents.
The staff believes that, in most cases, the insights gained from the internal event IPEs-can be readily adopted in performing the IPEEE.
That is, for external events, the containment performance needs to be assessed to identify if vulnerabilities and sequences different from those obtained from internal. events are predicted.
Because the maximum benefit from the IPEEE would be realized it the licensee 8s staff becomes involved in all aspects of the examination and the knowledge gained from the examination becomes an integral part of plant operations and training, the NRC is requesting strong licensee participation in both the IPE and IPEEE examinations.
Currently, six licensees plan to submit their external event analyses along with their IPEs; therefore, it is important to provide the staff guidance on-IPEEE to the industry in a timely manner.
The staff is planning to review all IPEEE submittals and serve as a clearing house to disseminate all important IPEEE findings.
However, the staff has not yet finalized the scope and depth of the review.
The staff currently estimates that a review of the IPEEE submittal will take between 1.5 and 4 person-months per plant depending on the level of further review deemed necessary following an initial overview type of review.
Adequate resourcss for this workload have been included in the Five-Year Plan.
As in the' case of the internal event IPE, the staff intends to review the results of the IPEEE to identify. severe accident vulnerabilities generic to a class or several classes of plants.
Such generic vulnerabilities would be used to determine if deficiencies exist in the regulations.
If deficiencies are identified, the benefits of modifying the regulations would be assessed against the Commission's safety goal policy as part of determining whether modifications to the regulations are needed.
Identification of all reasonable actions (whether related to structures, equipment, maintenance, surveillaryht, operating procedures, staffing, or training) to reduce the chances lof a severe accident or to mitigate the consequences of a severe accident is an important objective of both the internal event IPE and IPEEE.- The staff expects each licensee to identify all such actions j
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and to implement them, if appropriate, in a timely manner.
Each action would have to be assessed against the criteria of 10 CFR 50.59 and, if appropriate, submitted for Commission review in accordance with 10 CFR 50.90.
If the staff disagrees with a licensee or a group of licensees and concludes that certain actions are necessary or otherwise justified, the Commission's backfit rule, 10 CFR 50.109, will be used to decide which actions need to be implemented.
During the development of the generic letter and the guidance document, the staff has interacted-frequently with the ACRS and the CRGR.
The staff also has worked with NUMARC during the development process.
NUMARC is also developing a simplified procedure for the assessment of internal fires.
NRC will review the method when available and determine its acceptability as an alternative analytical method for the IPERE.
In summary, issuance of the generic letter to request that licensees perform the IPEEE will complete implementation of a key portion of the Commission's Severe Accident Policy Statement for operating plants.
The staff intends to periodically brief the-Commission on the progress of both the internal event IPE and IPEEE.
Coordinatient oGC has reviewed the proposed generic letter and has no legal objection. The ACRS has been briefed and sent a letter on May 15, 1990 stating that the ACRS agrees with the proposed approach.
Recommendatient That the Commi';t lon (1) approve issuance of the generic letter contained in Enclosure 1 and the draft guidance document contained in Enclosure 2 to all holders of operating licenses for comment, and (2) approve the staff issuing the guidance document in final form after appropriate consideration of comments.
mes M.
ylor xecutiv Director for Operations I
- 1. Proposed Generic Letter
- 2. NUREG-1407 IPEEE Guidance, Draft for Comment I
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Commissioners' comments or consent should be provided directly to the Office of the Secretary by COB Friday, June 22, 1990.
Commission Staff Office comments, if any, should be submitted to the Commissioners NLT Monday, June 11, 1990, with an infor-mation copy to the Office of the Secretary.
If the paper is of such a nature that it requires additional time for analytical review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.
This paper is tentatively scheduled for discussion at an Open Meeting on Friday, June 15, 1990.
DISTRIBUTION:
Commissioners OGC OIG LSS GPA REGIONAL OFFICES EDO ACRS ACNW ASLBP ASLAP SECY 1
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oo en Table of Contents Generic Letter No. 88-20, Supplement 4 Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities 1.
Summary 1
2.
Examination Process 2
3.
Identification of External Hasards 2
4.
Examination Methods 3
5.
Coordination with other ongoing Programs 5
6.
Severe Accident sequence selection 7
7.
Use of IPEEE Results 7
8.
Accident Management 8
9.
Documentation of Examination Recults 9
10.
Licensee Response 9
11.
Regulatory Basis 10 12.
References 11 Figure 1 13 Appendix 1 seismic IPEEE Enhancements 14 Appendix 2 containment Performance 16 Appendix 3 Criteria for selecting Important Severe-Accident Sequences 17 Appendix 4 Documentation 20 Appendix 5 10CFR50.54(f) Analysis for'IPEEE Generic Letter 28 4
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To All Licensees Holding Operating Licenses for Nuclear Power i
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SUILTECT:
INDIVIDUAL PIANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) i j
FOR SEVERE ACCIDENT VULNERABILITIES - 10CFR 50.54 (f)
(Generic Letter No. 88 Supplement 4) l 1.
Summary In the commission policy statement on~ severe accidents in nuclear power plants issued on August 8, 1985 (Ref. 1), the Commission concluded, based on available information, that existing plants pose no undue risk to the public health and safety and that there is no present basis for immediate action on any regulatory requirements for these plants.
However, the Commission i
rec gnizes, based on NRC and industry experience with plant-specific probabilistic risk assessments (PRAs), that systematic examinations are beneficial in identifying plant-specific vulnerabilities to severe accidents that could be fixed with low-cost improvements.
As a key part of the implementation of the policy statement, the staff issued Generic Letter 88-20 (Ref. 2) on Nov. 23, 1988, requesting that each licensee conduct an individual plant examination (IPE) for internally initiated events only.
F Current risk assessments indicate that the risk from external events could be a significant contributor to core damage in some instances.
The staff, however, delayed the issuance of ths request for a systematic individual plant examination for severe accidents initiated by external events (IPEEE) to allow the staff to carry out additional work to (1) identify which external hazards need to be evaluated, (2) identify acceptable examination methods and develop procedural guidance, and (3) coordinate with other ongoing external event programs.
The staff has completed this work (Ref. 3) and is now requesting that each licensee perform an individual plant examination of external events (IPEEE) to identify vulnerabilities, if any, to severe accidents and report the results to the Commission.
The general purpose of the IPEEE is similar to that of the internal event IPE--that is, for each licensee (1) to develop an appreciation of severe accident behavior, (2) to understand the i
most likely severe accident sequences that could occur at its plant, (3) to gain a qualitative understanding of the overall i
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2 probability of core damage and radioactive material release, and (4) if necessary, to reduce the overall probabilities of core damage and radioactive material releases by modifying, where appropriate, hardware and procedures that would help prevent or mitigate severe accidents.
It must be emphasized that for the IPEEE the key outcome is the insights obtained from such an examination.
Besides the completion of the internal event IPE and IPEEE, closure of severe accident concerns involves future NRC and industry efforts in the areas of accident management.
Additional discussion is provided in SECY-88-147 (Ref. 4) on the interrelationships among these three areas and the role they play in closure of severe accident issues for operating plants.
Therefore, consistent with the Commission's Severe Accident Policy Statement and pursuant to 10 CFR 50.54 (f), you are requested to perform an Individual Plant Examination of External Events (IPEEE) for plant-specific severe accident vulnerabilities initiated by external events and submit the results to the NRC.
2.
Examination Process The examination process for an IPEE.E, in general, is similar to that for internal event IPE (Ref. 2).
Basically, the event / fault trees from the internal event IPE can be extended for external event PRAs or used to identify important equipment for other acceptable evaluation methods, for instance, the seismic margin methodology.
As in the internal event IPE:
(1)
The quality and extent of the results derived from an IPEEE will depend on the vigor with which the licensee applies the method of examination and on the licensee's commitment to the intent of the IPEEE.
(2)
The maximum benefit from the IPEEE would be realized if the licensee's staff were involved in all aspects of the examination; that involvement would facilitate integration of the knowledge gained from the examination into operating procedures and training programs.
Therefore, each licensee is requested to use its staff to the maximum extent possible in conducting the IPEEE, participating in the analysis and technical review, and validating both the process and its results.
3.
Identification of External Hazards The external events to be considered, consistent with past PRAs, are those events whose cause is external to all systems used in normal operation and emergency operation situations.
A comprehensive list of the external events can be found in NUREG/CR-2300, "PRA Procedures Guide" (Ref. 5).
Some external events listed may not pose a significant threat of a severe
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Some external events may have been considered in the i
design at some plants to have sufficient protection that the i
predicted core damage frequency is sufficiently low.
se-me events may have been or will be reviewed under ongoing programs; for instance, the important impact of Idghtning and severe cold i
weather conditions that cause loss of offsite power, which is being addressed as an internal event in the IPE.
Also, internal floods have been included in the internal event IPE request (Ref.
2).
Based on staff's evaluation of Refs. 3 and 6-8, the staff 1
recommends only five events be included in the IPEEE.
- However, licensees should confirm that no plant-unique external events known to the licenses as of the date of this letter with the I
potential-to initiate severe accidents are-excluded from the IPEEE.
For instance, volcanic activities should be assessed as part of the IPEEE process at plant sites in the vicinity of active volcanoes and lightning effects should be assessed as part of the IPEEE process at those sites where lightning' strike effects may not be limited to partial or complete loss of offsite i
power.
The five external events are listed below
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seismic Events l
2.
Internal Fires j
3.
High Winds and Tornadoes 4.
External Floods 5.
Transportation and Nearby Facility Accidents A detailed discussion regarding the evaluation of external hazards can be found in NUREG-1407 and References 6-8.
4.
Examination Methods l
The NRC has identified the following approaches as being acceptable for the examination requested by this letter:
4.1 Seismic Events.
A seismic IPEEE can be accomplished by performing a seismic probabilistic risk assessment (PRA) with enhancements or a seismic margins review with enhancements.
The seismic PRA should be at least a Level 1 plus a containment performance analysis that uses current methods i
and plant information.
Containment performance analysis guidance is summarized in Appendix 2 to this generic letter.
The containment performance analysis needs to identify if
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seismically induced vulnerabilities and sequences different from those obtained from the IPE are predicted.
The staff considers the procedures described in NUREG/CR-2300-(Ref.
1 5), NUREG/CR-2815 (Ref. 10), and NUREG/CR-4840 (Ref. 15) to be adequate for the seismic IPEEE provided the enhancements discussed in Appendix 1 of this generic letter and NUREG-1407 are also included.
The hazard curves (Refs. 11 and 12)
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i (LLNL) and the Electric Power Research Institute (EPRI), if available, should both be used in performing the PRA, since this will help to focus on delineation of dominant sequences rather than emphasize bottom line numbers.
Detailed procedural guidance is provided in NUREG-1407.
j Two seismic margins methods (SMMs) with enhancements, one j
developed by NRC and one developed by EPRI, can also be used for the seismic IPEEE.
The SMMs in their current form are not suitable for plant sites located in high. seismic areas.
l For certain sites where the seismic hazard is low, a 4
reduced-scope margins method emphasizing plant walkdowns is l
considered adequate (see NUREG-1407).
The lists of review level earthquakes (RLEs) for all U.S. sites, defined by the staff for use in SNM that may be performed in response to this letter, are presented in Appendix 3.
The RLE does not represent safety adequacy criterion or a threshold for vulnerability for the individual plant.
The RLE is intended as a reporting criterion if the plant capacity is lower than J
the specific RLE.
Detailed descriptions of the seismic margins methods can be found in NUREG/CR-4334 (Ref. 14) and EPRI NP-6041 (Ref. 13).
The requested enhancements are discussed in NUREG-1407 and summarized in Appendix 1 of this
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i 4.2 Int?rnal Fires.
A fire risk IPEEE can be accomplished by i
performing a Level 1 fire PRA as_ described in NUREG/CR-2300 or a simplified fire PRA as described in NUREG/CR-4840 (Ref.
15).
The staff considers the fire code COMPBRN adequate for a search for fire vulnerabilities in nuclear plants.
When the licensee assesses the effectiveness of manual fire fighting, it should use plant-specific data from fire brigade training to determine the response time of the fire fighters.
The effectiveness of fire barriers should be assessed and the use of separation in determining fire zones critically examined.
The walkdown procedures should be specifically tailored to assess the remaining issues identified in the Fire Risk Scoping Study'(Ref. 16): (1) seismic / fire interactions, (2) effects of_ fire suppressants on safety equipment, and (3) control system interactions for severe accident vulnerabilities.
The containment performance (Appendix 2) needs to be assessed to determine if vulnerabilities and sequences different from those obtained in the internal event analyses are predicted.
4.3 High Winds, Floods, and Transportation and Nearby Facility Accidents.
A screening type approach as shown in Figure 1 is considered adequate for evaluating high winds, external floods, and transportation and nearby facil.ity accidents.
The steps shown in Figure 1 represent a series of analyses in increasing level of detail, effort, and resolution.
The licensee should first determine if the 1975 standard review l
5 plan (SRP) criteria are met.
If plant does not meet the 1975 SRP criteria, the licensee should examine it further using the recommended optional steps.- However, the licensee may choose to bypass one or more of the optional steps so long as vulnerabilities are either identified or proved to be insignificant.
Again, the containment performance needs to be arsessed to determine if vulnerabilities and sequences different from those obtained from the, internal event analyses are predicted.
The detailed description of this screening approach is presented in NUREG-1407.
The above methods are recommended by the NRC; however, the NRC recognises that other methods capable of identifying plant specnfic vulnerabilities to severe accidents due to external events may exist.
For instance, a simplified fire risk evaluation method is being developed by NUMAAC.
The staff will review any systematic examination methods proposed, such as the simplified fire risk method being developed by NUMARC, to determine their acceptability for IPEEE..
5.
coordination with other External Event Proarama Three programs, i.e.
(1) the external event portion of the USI A-45, (2) GI-131, and (3) the Charleston Earthquake Issue, are subsumed in the IPEEE.
A brief discussion on these programs is provided below USI A-45. " Shutdown becay Heat Removal Reauirements# 1 USI A-45 had the objective of determining whether the decay haat removal function at operating plants is adequate and if cost-effective improvement can be identified.
A part of the USI A-45 activities consists of assessing the seismic adequacy of the decay heat removal system (DHR).
This aspect of DNR issue should be specifically addressed in the seismic review of the IPEEE.
The external event insights obtained from the USI A-45 study on 5 plants are presented in GL 88-20.
GI 131." Potential Seismic Interaction Involvina the Novable 2
In-core Flur Mannina System Used in Westinchause Plants #
1 Generic Issue (GI) 131 (Ref. 17) deals with the seismically induced failure of the flux mapping. transfer cart leading indirectly to the rupture of instrumentation tubes at the seal table.
This could lead to core damage if loss of coolant through the ruptured instrumentation tubes is combined with unavailability of other mitigating systems.
This scenario is applicable only to Westinghouse plants.
Affected plants should explore the potential for this scenario and achieve a resolution of this concern through the IPEEE.
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"The Charleston Earthauake Issue"t As a result of work j
carried out by LLNL and EPRI to help to resolve the Charleston Earthquake Issue, probabilistic seismic hazard estimates (Refs. 11 & 12) exist for all nuclear power plant I
sites east of the Rocky Mountains.
These estimates can be used directly by any licensee opting to satisfy the seismic i
l IPEEE by means of a seismic PRA.
The hazard estimates also played a key role in determining the review level earthquake to be used in the seismic margin option.
Therefore, the IPEEE will constitute a resolution of the Charleston t
l Earthquake Issue.
5 other external event programs listed below are either resolved or nearing completion. Their plant-specific implementation may require a plant-specific examination which should be coordinated
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with the IPEEE to minimize unnecessary duplication of examination and review efforts.
l USI A-17. " System Interactions in Nuclear Power Plants". USI A-40. " Seismic Desian Criteria. A Short-Term Proaram", and USI A-46. " Seismic Oualification of Eauinment in Onerating Plants"t The scope of USI A-46 has been expanded to contain the seismic spatial system interaction of USI A-17 and the seismic capability of safety tanks of USI A-40 (NUREG-1407).
The USI A-46 review is required on approximately 70
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operating plants, which constitute a subset of all the nuclear power plants that are expected to perform an IPEEE.
i USI A-46 should be coordinated with the IPEEE so that the objectives of both activities may be accomplished at the licensee's option with a single walkdown effort.
(Both A-46 plants and non-A-46 plants will address spatial interactions within the IPEEE program through the seismic walkdown, which is guided by the EPRI methodology.)
The most efficient way to address the seismic IPEEE for the USI A-46 plants is to conduct the USI A-46 review and walkdown to gather relevant information for the seismic IPEEE.
In order to facilitate this approach the activities of USI A-46 and the seismic IPEEE need to be coordinated and the plant walkdown needs to be well planned, t
" Eire Risk Scooina Studv a (NUREG/CR-5088) and GI 57.
" Effects of Fire Protection System Actuation on Safetv-Related Eaulement":
The licenses should address the fire issues identified in the Fire Risk Scoping Study (Ref. 16) as discussed in Section 4.2 in NUREG-1407.
However, it should be noted that in parallel with the IPEEE, additional research related to GI 57 is being performed to obtain more rigorous and realistic estimates of risk; and this research may identify other potential vulnerabilities.
The specifically tailored walkdown procedures for potential fire vulnerabilities should enable the licensee to collect
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Licensees may propose l
corrective measures that could resolve some or all GI 57 s
concerns.
A brief description of the above programs is presented in NUREG-1407.
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Resolution of Other Generic Issues If, during its IPEEE, a licensee (1) discovers a potential I
vulnerability that is topically associated with any other USI or i
GI and proposos measures to dispose of the specific safety issue or (2) concludes that no vulnerability exists at its plant that is topically associated with any USI or GI, the staff will consider the USI or GI resolved for a plant upon review and acceptance of the results from the IPEEE.
The licensee's, submittal should specifically identify which USIs or GIs it is proposing to resolve.
Further guidance is given in NUREG-1407.
6.
Severe Accident Secuence Selection l
In performing an IPEEE using a PRA, it is necessary to screen the severe accident sequences for the potentially important ones and for reporting to the NRC.
The screening criteria to determine the potentially important sequences that lead to core damage or unusually poor containment performance that should be reported to the NRC with your IPEEE results are listed in Appendix 3 of this generic letter.
If a seismic vargins method is used in the IPEEE, the licensee should report all functional sequences and success paths considered in the analysis and their associated high confidence-u low probability of failure (HCLPFs) values.
In addition, the licensee should report all HCLPFs related to containment and containment systems performance.
A HCLPF value lower than the specified review level earthquake (RLE) does not necessarily represent a plant vulnerability, but, rather, defines a reporting level.
The licenses should assess the significance of HCLPF values lower than RLE and take any actions that are deemed
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appropriate by the licensee.
NUREG-1407 describes the documentation needed for the accident sequence selection and the intended disposition of these sequences.
A summary is provided in Appendix 4, 7.
Use of IPEEE Results Licensee After each licensee conducts a systematic search for severe accident vulnerabilities in its plant (s) and determines whether potential improvements in structures, equipment, maintenance, i
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expeditiously to correct those vulnerabilities that it determines warrant correction.
Information on changes initiated by the I
licensee should be provided consistent with the requirements of l
10 CFR 50.59 and 10 CFR 50.90.
Changes should also be reported in your IPEEE submittal (including reference to any previous i
submittals under 10 CFR 50.59 or 10 CFR 50.90) that responds to this letter.
NRC The NRC will evaluate licensee IPEEE submittals to obtain reasonable assurance that the licensee has adequately analyzed the p1snt design and operations to discover instances of
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particular vulnerability to core damage or unusually poor containment performance given a core damage accident.
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the NRC will assess whether the conclusions the licensee draws i
from the IPEEE regarding changes to the plant systems or components are adequate.
The consideration will include both quantitative measures and nonquantitative judgment.
The NRC consideration may lead to one of the following assessments:
i 1.
If NRC consideration of all pertinent and relevant factors indicates that the plant design or operation must be changed t
to meet NRC regulations, appropriate functional enhancements 4
will be required to be implemented without regard to cost s
except as appropriate to select among alternatives.
2.
If NRC consideration indicates that plant design or operation could be enhanced by substantial additional protection beyond NRC regulations, appropriate enhancement will be recommended and supported with backfit analysis in accordance with 10 CFR 50.109.
3.
If NRC consideration indicates that the plant design and operation meet NRC regulations and that further safety improvements are not substantial or not cost effective, enhancements would not be suggested unless significant new safety information becomes available.
8.
Accident Manaaement Licensees need not develop an accident management plan as an l
integrated part of the IPEEE, nor must an accident management plan be submitted at the time you submit your results.
Licensees i
l should plan to incorporate the results of the IPEEE and other relevant information into their accident management plan at a l
future date.
Nevertheless, in the course of conducting your l
IPEEE, you may identify orarator or other plant personnel actions j
that can substantially reduce the risk from severe accidents at your plant and that you 'oelieve should be immediately implemented 9
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We encourage each licensee not to defer implementing such actions but rather to implement such actions immediately within the constraints of 10 CFR 50.59.
These actions can be integrated later into the plant's accident management program.
9.
Documentation of Exaninatien Results The IPEEE should be documented in a traceable manner to provide the basis for the findings.
This can be dealt with most efficiently by a two-tier approach.
The first tier consists of the results of the examination, which will be reported to the NRC.
The second tier is the documentation of the examination itself, which should be retained by the licensee for the duration of the license.
NUREG-1407 specifies the minimum information to be documented / submitted.
A summary is provided in Appendix 4 of this generic letter.
10.
Licensee Resoonse A document in draft form, NUREG-1407, that provides additional licensee guidance for the performance and the submittal of the IPEEE results is attached.
A workshop will be scheduled to discuss the IPEEE objectives and to answer questions that licensees may have on both the IPEEE Generic Letter and the draft guidance document.
The NRC will consider the comments received, make appropriate revisions, and reissue the guidance document and any necessery supplement to this generic letter, as appropriate, in response to comments received.
Licensees are requested, within 60 days of receipt of the final ~
go!4ance document, to submit theit proposed programs for completing the IPEEEs.
The proposa.$
%euld:
1.
Identify the methods and approach selected for performing the IPEEE, 2.
Describe the method to be used if it has not been previously submitted for staff review (the description may be by reference), and 3.
Identify the milestones and schedule for performing the IPEEE.
Meetings with NRC during the examinations will be scheduled as needed to discuss subjects raised by licensees and to provide necessary clarifications.
Licensees are requested to submit the IPEEE results within three years from the date of issuance of NUREG-1407 in final form.
However, for those licensees who choose to perform the internal events IPE and the IPEEE in series, requests to extend the time period to submit the IPEEE report-will be considered.
The NRC encourages those plants that have not yet undergone any
10 systematic examination-for severe accidents to promptly initiate the examination.
Those licensees that choose to-use existing external events PRAs with enhancements identified in NUREG-1407 should:
i 1.
Certify that the PRA meets the intent of the generic letter, 1
in particular with respect to the licensee's staff involvement, 2.
Certify that the' existing plant design and operation are represented accurately in the PRA, 3.
Submit the results on a schedule shorter than 3 years..
11.
Reaulatory.15A313 This lettet is issued pursuant to 10 CFR 50.54 (f). - Accordingly,-
all responses should be under oath or affirmation.
This request
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for information-is covered by the office of Management and Budget under. Clearance No. 3150-0011, which expires December 21, 1990.
i The estimated average burden would not exceed 6 person-years per licensee response (Appendix 5) over a 3-year period, including assessing the request, searching data sources, gathering and
' ?
analyzing the data, and preparing the IPEEE reports.
Comments on burden and duplication'may be directed to thc Office of-Management and' Budget, Reports Management, Room 3208, New Executive Office Building, Washington, DC 20503.
Sincerely, James G. Partlow, Associate Director fer Projects-Office of Nuclear Reactor-Regulation
Enclosures:
1 Appendices 1 through 5 1
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.i 11 12.
Referances 1
1.
" Policy Statement on~ Severe Accidents," U.S. Nuclear
]
Regulatory Commission, Federal, Register,.Vol. 50, 32138, August 8, 1985.-
2.
Generic Letter 88-20, " Individual Plant Examination for Severe Accident vulnerabilities. - -10CFR 50.54 (f)," Nov. 23, 1988.
3.
NUREG-1407, "proceddral and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities,"
1990.-(Draft for Comment) 4.
SECY 88-147, " Integration: Plans for Closure of Severe Accident Issues," May 25, 1988.
5.
NUREG/CR-2300, "PRA Procedures Guide," ANS & IEEE, January 1983.
6.
NUREG/CR-5042, " Evaluation of External Hazards to Nuclear l
Power Plants in the United States," LLNL, December.1987.
l 7.
NUREG/CR-5042 Supplement 1, " Evaluation of External Hazards to Nuclear Power Plants in the United States, Seismic Hazards" LLNL, April 1988.
8.
NUREG/CR-5042 Supplement 2, " Evaluation of External Hazards to Nuclear Power Plants in the United States, other External l
Events," LLNL, February 1989.
t I
9.
NUREG-1335, " Individual Plant Examination: Submittal l
Guidance," Final Report, August 1989.
10.
NUREG/CR-2815, "Probabilistic Safety Assessment Procedures Guide," BNL, August 1985.
l l
11.
NUREG/CR-5250, " Seismic Hazard Characterization.of 69 l
Nuclear Plant Sites East of the. Rocky Mountains," LLNL, January 1989.
12.
EPRI NP-6395-D, "Probabilistic Seismic Hazard Evaluation at.
Nuclear Plant Sites in the' Central and Eastern United
+
States: Resolution of the Charleston Issue," April 1989.
13.
EPRI NP-6041, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," EPRI, Cetober 1988.
14.
NUREG/CR-4334, "An Approach to the Quantification of Seismin
{
Margins in Nuclear Power Plants," LLNL, August 1985.
._.,7
-.... m
- l i
12-15.
NUREG/CR-4840, " Recommended Procedures for the Simplified External Event Risk Analyses for NUREG-1150," Sandia, i
I September 1989.
16.
NUREG/CR-5088, " Fire Rick Scoping Study," Sandia, January.
1989.
17.
Memo from E. Beckjord to R. Houston, dated July 31, 1989,
Subject:
Generic Issue 131, " Potential Seismic Interaction Involving: the Movable In-Core Flux P'eping System Used in Westinghouse Plants".(available in Nut Public Document Room).
Initiation of the 18.
Generic Letter 88-20, Supplement 1, individual Plant Examinatjon for Severe Accident E41nerabilities--10 CFR 50.54 (f)," August 29, 1989.
19.
Generic Letter 88-20, Supplement 3,
" Completion of Containment Performance Improvement Program and Forwarding of Insights for Use in the Individual Plant Exarination for' Severe Accident Vulnerabilities,"
June 1990.
20.
Memorandum from B. Sharon to T. Speis, dated December 1, 1988,
Subject:
Staff Evaluation in Support of 10CFR 50.54 (f)
Generic Letter 88-20 Requiring Individual Plant Examination (available in NRC Public bocument Room).
21.
NUREG/CR-4458, " Shutdown Decay Heat Removal Analysis.of a Westinghouse 2-loop Pressurized Water Reactor," March 1987.
22.
NUREG/CR-4713, " Shutdown Decay Heat Removal Analysis of a Babcock and Wilcox Pressurized Water Reactor," March 1987.
l 23.
NUREG/CR-4762, " Shutdown Decay Heat Removal Analysis of a.
Westinghouse 3-loop Pressurized Water Reactor,"-March 1987.
l 24.
NUREG/CR-4767, " Shutdown Decay Heat Removal Analysis of a General Electric BWR4/ Mark I," March 1987.
25.
NUREG/CR-4710, " Shutdown Decay Heat Removal Analysis of a Combustion Engineering Pressurized Water Reactor,". March 1987.
26.
NUREG/CR-4448, " Shutdown Decay Heat Removal Analysis of a General Electric BWR3/ Mark I," March 1987.
27 NUREG-1150, " Severe Accident Risks: an Assessment for Five U.S=. Nuclear Power Plants," Second' Draft for Peer Review, June 1989.
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4 13-Figure 1 RECOMMENDED IPEEE APPROACH FOR WINDS FLOODS. AND OTHERS (1). REVIEW PLANT SPECIFIC HAZARD
. DATA AND LICENSING BASES (FSAR) j (2). IDENTIFY SIGNIFICANT CHANGES, IF ANY, SINCE OL ISSUANCE 4
(3) DOES.Pl. ANT / FACILITIES DESIGN NO-MEETz CURRENT (1975 SRP) CRITERIA-YES (QUICK SCREENING & WALKDOWN)l (4) IS THE ' HAZARD FREQUENCY
-YES
-+
ACCEPTABLY LOW 7
~
NO, l
u OR-* (5) BOUNDING ANALYSIS' VES (RESPONSE / CONSEQUENCE)
NO OR-+ (6) PRA I
(7) DOCUMENTATION (INCL. IDENTIFIED REPORTABLE' ITEMS AND PROPOSED IMPROVEMENTS) e a
s_
s.
14 APPENDIX 1
SUMMARY
OF SEISMIC IPEEE METHODOLOGY ENHANCEMENTS The following guidelines provide some specifics that are needed in a PRA, in a supplement-to an existing PRA, or in the seismic margins method for an IPEEE submittal.
A detailed discussion of these enhancements is presented in NUREG-1407.
New PR&t Perform a plant walkdown'following the procedures described in the EPRI. seismic margin report (Ref. 13).
Perform an assessment of relay chatter effects in
)
accordance with procedure described in NUREG-1407.
Calculate the high confidence of low probability of.
failure (HCLPF) values for components, sequences, and the plant.
i Perform soil liquefaction analysis, if needed, using 1
procedures described in EPRI's report (Ref. 13).
Existing PRA: Perform a supplementary walkdown following the procedures described in EPRI's report (Ref. 13).
Perform an assessment of relay chatter effects.
[
Perform a supplementary analysis of nonseismic. failures and human actions if not considered previously.
Calculate and report HCLPF values as above.
Perform soil liquefaction analysis, if needed, using procedures described in EPRI's report'(Ref. 13).
NRC SMM8 Perform an assessment of relay chatter effects in I
accordance with procedures described in NUREG-1407.
Perform soil liquefaction analysis, if needed, using procedures described in EPRI's report (Ref. 13).
Perform an analysis of.nonseismic failures and human actions using procedures described in NUREG-1407.
Perform a walkdown and prepare its documentation in
.I accordance with EPRI's recommendations (Ref. 13).
Evaluate containment and containment system performance.
EPRI 8MM Select an alternative path so that it involves different operational sequence systems, piping runs, l
~
=
+
15 and components from the preferred-success: path;to the maximum extent possible. '
Perform an analysis of: nonseismic failures. and human actions using procedures described,in NUREG-1407.
Evaluate containment and containment systems.
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APPENDIX 2 CONTAINMENT PERFORMANCE p
The public safety of nuclear power plants has been fostered by
-applying the " defense-in-depth" principle, which relies on a set
)
of independent barriers to fission product release to the environment.
The containment'and its supporting systems comprise L
one of these barriers.
1 The evaluation of the containment performance for external events 1
should-be directed toward a systematic examination of whether l
there are sequences-that involve containment failure modes
~
distinctly different from those found in the IPE internal events evaluation or contribute significantly to the likelihood.of containment functional failure (i.e. loss of containment barrier independent of core melt).
It should recognize the role of mitigating systems, and should ultimately result in the development of accident management procedures that could-both prevent and mitigate the consequences of the severe accidents.
The most efficient way to accomplish this is to use the information developed for the IPEEE to:
1.
Identify mechanisms that could lead to containment bypass, 2.
Identify mechanisms that could cause failure of the-containment to isolate, and 3.
Determine the availability and performance of the l
containment systems under the external hazard to see if it is different from that evaluated under the internal event evaluation.
Additional guidance on the containment performance associated with external events can be found in NUREG-1407.
. {
Licensees are expected to evaluate the insights learned from CPI programs as discussed in references 18 & 19-and determine their I
applicability to external events.
l-I l'
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17 APPENDIX 3 CRITERIA FOR REPORTING IMPORTANT SEVERE ACCIDENT SEQUENCES The licensee should use the reporting criteria described in the Generic Letter 88-20 for PRA analysis.to determine which potentially important functional sequences and functional failures that might lead to core damage or unusually poor containment performance should be reported to the NRC in the IPEEE submittal.
The licensee should;use the reporting criteria i
described in'NUREG-1335-to report systemic sequences to the NRC.
These criteria do not represent a threshold for vulnerability.
If a seismic margin method is used'in the IPEEE, the licensee should report'in accordance with the:NUREG-1407 all functional sequences and success paths considered in the analysis and their HCLPFs.
The review level earthquakes, (RLEs) for all US sites are i
presented in Tables 3.1 and 3.2.
In addition, the licensee should report all HCLPFs related to containment and containment systems performance.
A HCLPF value. lower than the specified i
review level earthquake (RLE) does not necessarily represent a plant vulnerability, but, rather, defines a reporting. level.
The licensee should assess the significance of HCLPF values lower than RLE and take any necessary' actions and make other improvements that are deemed appropriate by the licer.see.
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18 TABLE 3.1 REVIEW LEVEL EARTHOUAKE - PLANT SITES EAST OF THE ROCKY MOUNTAINS
[
Reduced Scone Procram Big Rock Point *Duane Arnold South Texas Turkey Point d
Comanche Peak Grand Gulf St. Lucie Waterford Crystal River River Band i
l 0.3 a Arkansas Dresden Maine Yankee Robinson Beaver Valley Farley McGuire Salem
- Bellefonte
-Fermi.
Millstone Sequoyah' Braidwood Fitzpatrick Monticello Shoreham Browns Ferry Fort Calhoun Nine-Mile Point
- Summer Brunswick Ginna
- North Anna Surry i
Byron Haddam Neck' *Oconee Susquehanna Callaway Harris Oyster Creek Three Mile Is.
Calvert Cliffs Hatch Palisades Vermont Yankee-
- Catawba Hope Creek Peach. Bottom Vogtle Clinton Indian Point Perry Watts Bar Cook Kewaunee Point Beach Wolf Creek-Cooper LaSalle Prairie Island Yankee Rowra Davis-Besse Limerick Quad Cities Zion 0.5 as i
Pilgrim Seabrook NOTE
- Special attention to shallow soil conditions is needed at these. locations (See Section 3.2.2 of NUREG-1407).
i
- Based on the staff studies, review level earthquakes greater-l than 0.3g are needed for these two sites.
Because the-component capacity. data sets associated with the' margins methods are categorized at.'two screening levels, 0.3g and 0.5g, it is necessary that the RLE for these sites be set at 0.5g.
s-a 19 TABLE.3.2 REVIEW LEVEL'EARTHOUAKE - WESTERN UNITED STATES PLANT SITES l
0.5 g
- Trojan
- Rancho Seco
- Washington Nuclear -*Palo Verde i
Seismic Marcin Methods in current Form are not Annlicable to the Followina Sitest Diablo canyon San:Onofre.
NOTES:
i Indicates a Western United States site whose~RLE'is 0.5g unless the: licensee can demonstrate that the site hazard is similar to those' sites east of theLRocky Mountains that are found in the 0.3g bin.
changes in the review' level earthquake from 0.5g to 0.3g l
should beJapproved prior to:doing significant analysis.
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4 20 APPENDIX 4
-DOCUMENTATION TLis appendix provides Nha guidelines for detailed documentation-erd reporting format and content for the IPEEE>submittals.
The Md ar parts of this appendix are the guidelines for seismic ahalysis_(Section 4.2), internal fire analysis (Section 4.3),
other analyses (Section 4.4), specific. safety features and plant.
improvements (Section 4.1.4), and-the licensee review-team-(Section 4.1.5).
The licensees are requested to submit.their -
IPEEE reports using the standard table of contents given in Table C.1 of the NUREG-1407 or provide a. cross reference.
This will-facilitate review by the NRC and promote consistency among various submittals.
The contents of the-elements of this table are discussed in sections below.
The level of detail needed in the documentation should be-l
~
sufficient to enable NRC to understand-and determine the validity of key input data and calculation models used, to assess the sensitivity of the results to all key aspects of the analysis, and to audit any calculation.
All important assumptions should be reported.
It is not necessary to submit all the documentation needed for such an NRC review, but its existence should be cited and it should be available in easily usable form.- The guideline for adequate retained documentation is that-independent-expert L
analysts should be able to reproduce any-portion of t2ue results
- ~
- of the calculations in a straight forward,. unambiguous manner.
To the extent possible, the retained documentation should be
' organized-along the lines identified in the areas of; review.
Any information that is comparable to that provided~under the IPE for internal events can be incorporated by' reference.
4.1 General 4.1.1 Conformance with Generic Letter and Supporting Material Certification that an IPEEE has been completed and' documented as requested by Generic Letter 88-20, Supplement 4.
The certification should also identify the measures taken-to ensure the technical adequacy of the IPEEE and the validation-of results, including any uncertainty, sensitivity, and importance analyses.
4.1.2 General Methodology Provide an overview description.of the methodology employed in the-IPEEE for each external event.
4.1.3 Information Assembly LReporting guidelines include:
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i 21 Reporting guidelines include:
1.
Plant layout and containment building information not contained in.the Final Safety Analysis Report (FSAR).
2.
A concise description of plant' documentation used in the IPEEE, (e.g., the FSAR; system descriptions,-procedures, and licensee event reports); and a concise discussion of the process used to confirm that the IPEEE represents the as-built', as-operated plant.
The intent of such a confirmation is not to propose new design reverification efforts on the.
j par'..of the licensees but to account for the impact of previous plant modifications or modifications conducted J
within the IPEEE framework.
3.
A description of the coordination activities of the IPEEE teams among the external events (e.g., for seismically induced fires).-
d 4.1.4 Submittal of Vulnerability Definition and Potential-Plant Improvements The licensee should provide a discussion'on how the vulnerability is defined for each external event evaluated.
The licensee should list any potential improvements (including equipment changes as well as changes in maintenance, operating and emergency procedures, surveillance, staffing, and trajning i
L programs) that have been selected for implementation based on the IPEEE (provide a schedule for implementation) or that have 7
already been implemented.
Include a discussion of anticipated benefits as well as drawbacks.to any improvements, 4.1.5 IPEEE' Team and Internal Review The basis for requesting the involvement of the 11censee's staff in the IPEEE review'is the belief that the maximum-benefit from the performance of an IPEEE would be realized if the licensee's staff were' involved-in all aspects of the examination and that involvement would facilitate integration of the knowledge gained l
from the examination into operating procedures and training participation an,the submittal should describe licensee staff I
programs.
Thus L
d the extent to which the licensee was involved L
in all aspects of the program.
The submittal should also contain a description of the internal I
review or peer review performed, the results of the review team's
~;
evaluation, and a list of the review team members.
The maximum benefit to1the-licensee will occur if the combination of persons involved'in the original analysis and the internal review, taken as a group, provides both a cadre of licensee personnel to facilitate the continued use of the results and the expertise in the methods to ensure that the techniques ~have-been correctly
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i applied.
Furthermore, an internal ' review prov ides quality control and quality assurance to the IPEEE process.
4.2 Seismic Events Section 4.2.1 describes submittal information guidelines for licensees who choose the seismic PRA for seismic IPEEE, while section 4.2.2 describes information guidelines for licensees who choose the seismic margin method for the seismic IPEEE.
The submittal should be presented in conformance with Table col of 1
4.2.1 Seismic PRA Methodology The following information on the seismic IPEEE is the minimum that should be documented and submitted to the NRC:
I A description of the methodology and key assumptions used in 1.
i performing the seismic IPEEE.
2.
The hazard curves (or table of hazard values) used and the associated spectral shape used in the analysis.
Also, if,an_
upper bound cutoff to ground motion of less than 1.5g peak ground acceleration is assumed, the results of sensitivity studies to determine whether the cutoff affected the delineation and ranking of seismic sequences.
3.
A summary of the walkdown findings and a concise description of the walkdown team and the. procedures used.
4.
All functional / systemic seismic event trees as well as data (including origin and method of analysis).
Address to what extent the recommended enhancements have been incorporated in the IPEEE.
A description of how nonseismic failures, human actions, dependencies, relay chatter, soil liquefaction, and seismically induced floods / fires are accounted for.
Also, a list of important nonseismic failures with a rationale for the assumed failure rate given a seismic event.
5.
A descriptu,n of rominant functional / systemic sequences ~
leading to core damage along with their frequencies and percentage contribution to overall seismic core damage frequencies (for both LLNL and EPRI hazard curves if available for the site).
Sequence selection criteria are provided in GL 88-20.
If either hazard-curve causes a sequence to meet these criteria than that sequence should be included.
The description of the sequences should include a discussion of specific assumptions and human recovery action.
I 1
23 6.
~The estimated core damage frequency (for both the LLNL and' EPRI-hazard curves, if available fo the site) Land plant damage state frequencies,-the-timing of the core damage, t
including a qualitative discussion of uncertainties and how they might affect the final ~results, and contributions of different ground-motions to core damage frequencies.
7.
Any seismically induced containment failures and other containment performance insights.-
Particularly, vulnerabilities found in the following three j
systems / functions: Epenetrations, isolation, and containment heat removal / pressure; suppression.
Early containment failures that might result in high-consequence sequences or' may initiate accident sequences.
Also, computed fragilities and HCLPFs of containment components.
8.
A table of fragilities, both generic and plant-specific, used for screening.as well-as in the quantification.
The j
estimated HCLPF for the plant, dominant sequences, and components with and without nonseismic failures and human actions.
9.
Documentation with regard to other seismic issues addressed by the submittal, the basis and assumptions used to address these issues, and a discussion of the findings and conclusions.
Evaluation results and potential. improvements associated with the decay. heat removal function.and movable in-core flux mapping system (for Westinghouse plants)-
should be specifically highlighted.
- 10. When an existing PRA is used to address the seismic IPEEE, the licensee should describe sensitivity studies related to the use of the initial hazard curves,' supplemental plant walkdown results and subsequent evaluations, and relay-chatter evaluations.
The licensee should examine items 1 through 9 above to fill in those items missed in the existing seismic PRA (See NUREG-1407 3.1.2).
-4.2.2 Seismic Margins Methodology The following information on the seismic IPEEE is the minimum that should,be documented and submitted to the NRC for a' full-scope SMM review:
1.
A description of the methodology and a list of important assumptions, including their basis, used in performing the seismic IPEEE.
Addsess the extent to which the following were taken into account: nonseismic failures, human actions, dependencies, relay chatter, soil liquefaction, and seismically ' induced' floods / fires.
Also, a' list of important
.i i
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-24 nonseismic failures with a rationale for the assumed failure-rata =given a seismic event.
2.
A summary of the walkdown findings and a concise description of the walkdown team and procedures used.
3.
- All functional / systemic seismic event trees data (including origin and method of analysis) when NRC SMM is used.
4.
A description of the most important sequences and important minimal cutsets (for both seismic and nonseismic failures) leading to core damage (NRC method) or.a description.of the success paths and procedures used for their selection and of each component ir. the controlling success path (EPRI method).
5.
Any seismical3y induced containment failures or vulnerabilities and other containment performance insights.
The procedure for defining the scope of the containment' performance review, Results of isolation system, penetration, and heat removal system reviews.
1 6.
A table of fragilities, both generic and' plant-specific, used for screening as well as in the quantification.
The estimated HCLPF for the plant, dominant sequences, and components with and without nonseismic failures and human actions.
7.
Documentation with regard to other seismic issues addressed by.the submittal, the. basis and assumptions used to address these issues, and a discussion of the-findings and conclusions.
Evaluation results and potential improvements associated with the decay heat removal-function and movable in-core flux mapping system (for-Westinghouse plants) should be specifically highlighted.
The following is the minimum information that should be documented and submitted to the NRC for a reduced-scope SMM review:
1.
A description of the procedures used to identify systems and cooponents for the walkdown in performing the seismic IPEEE.
2.
A summary of the walkdown findings and a concise description of the walkdown team and procedures used.
3.
A discussion and the results of any specific component capacity evaluations performed, the methods used, and assumptions.
4.
Documentation with regard to other seismic issues addressed by the submittal, the basis and assumptions used to address
}
=y
25 these issues, and'a~ discussion ofLthe-findings and l
conclusions.. Evaluation results and potential improvements-l associated with the decay heat' removal function and movable in-core flux mapping system-(for Westinghouse plants) should I
be specifically highlighted.
l 4.3 Internal Fires The information on the internal fires IPEEE identified below is the minimum that should be documented and submitted to the NRC.
1._
A description of the methodology used in performing the fire IPEEE and a discussion of the status of' Appendix R l
modifications.
~
2.
A summary of the walkdown findings and a concise description of the walkdown team and the procedures-used.
This should include a description of the efforts to ensure that cable routing used in the analysis represents as-built information and a description of the treatment of any existing dependence between remote shutdown and control room circuitry.
3.
A discussion of the criteria used to identify critical fire areas and a list of critical areas,Eincluding (a) single area in which equipment failures represent a serious erosion of safety margin,-and-(b) same'as (a), but for double or multiple areas sharing common barriers, penetration seals, KVAC ducting, etc.
4.
A discussion of the criteria used.for fire size and duration and the treatment of cross-zone fire spread and associated major assumptions.
j 5.
A discussion of the fire initiation data base, including the-plant-specific data base used.' Provide documentation in each case where the plant-specific data used is less conservative than the data base used in the approved fire vulnerability methodologies.
Describe the data handling method, including major assumptions, the role of expert judgment, and the identification and evaluation of sources of data uncertainties.
6.
A discussion of the treatment of fire growth and spread, the spread of hot gases and smoke, and_the analysis-of detection l
and suppression and their associated assumptions, including the treatment of suppression-induced damage to equipment.
2-7.
A discussion of fire damage modeling, including the definition of fire-induced failures related to fire barriers and control systems and fire induced damage to cabinets.
A discussion of how human intervention is treated and how Il
/l
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26 fire-induced and non-fire-induced failures are combined.
Identify recovery actions and types of fire-mitigating actions taken credit for in these sequences.
8.
Event trees and associated fault-trees fortfire-initiated sequences.
Discuss the treatment of manual suppression, including fire fighting procedures, fire brigade training and adequacy of existing fire brigade. equipment', and treatment of access routes versus existing barriers.
9.
The estimated core damage frequency, the timing of the associated core damage, a' list of analytical assumptions including their bases,-and the sources of uncertainties, if applicable.
4.4 Hiah Winds. Floods, and Others i
The following information on the.high-winds, floods, and others
. portion of the IPEEE is the minimum that should be documented and submitted to the NRC I
1.
A description of the methodologies used in the examination.
2.
Information on plant-specific; hazard data and licensing bases.
3.
Identified significant changes (See NUREG-1407 5.2.2), if any, since OL issuance with respect to'high winds, floods, and other external. events.
~
4.
Results of plant / facility design-review'to determine their robustness in relation to NRC's current criteria.
5.
Results of the assessment of the hazard; frequency and the associated conditional core damage frequency if step 4 of Figure 1 is used.
6.
Results of the bounding analysis if step 5 of Figure 1 is
- used, t
7.
All functional event trees, including origin and method of analysis (PRA only).
8.
A. description of each functional sequence selected, including discussion of specific assumptions and human recovery action'(PRA only).
9.
The estimated core damage frequency,.the timing of the associated core damage, a list of analytical assumptions including their bases, and the sources of uncertainties, if applicable (PRA only).
,.~.. -. ~
.,7
=,
27 l10.
A. certification that no other plant-unique ~ external event is known that poses any significant threat of. severe accident within the context of the screening approach for "High Winds,, Floods, and others."
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28 APPENDIX 5 10CFRSO.54 (f) ANALYSIS-FOR INDIVIDUAL PIANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) 10CFR50. 54 (f) requires that "... the.NRC must prepare the reason or reasons for each infexwat!on request prior to issuance to ensure that the burden to be imposed on respondents is justified e
in view of the potential safety significance of~the issue to be addressed in the requested information."
Further, Revision 4 of
.I the Charter of the Committee to Review Generic Requirements (CRGR), dated April 1989 specifies that, at a minimum, such an evaluation shall include:
I a.
A problem statement that describes the need for.the information in terms of potential safety benefit, b.
The licensee actions required and the cost to develop a l
response to the information request, and c.
An anticipated schedule for NRC use of the information.
The staff's 10CFR50.54(f) evaluation of the information request addressing the above elements follows:
a.
A problem statement'that describes the need for tha information in terms of potential safety benefit.
1 In the Commission policy statement on severe accidents in nuclear power plants-issued August 8, 1985 (50FRL32138), the d
Commission concluded, based on available information, that l
existing plants pose no undue risk to the public' health and l
safety and that there is no present basis for immediate action on any regulatory requirements for these plants.
However, the Commission recognizes, based on-NRC.and I
industry experience with plant-specific probabilistic risk assessments.(PRAs), that systematic examinations are beneficial in identifying plant-specific vulnerabilities to severe accidents that could be fixed with' low-cost improvements.
As a key part of the implementation of the policy statement, the staff issued Generic Letter 88-20 on Nov.'23, 1988, requesting that each licensee conduct an individual plant examination-(IPE) for. internally initiated events only.
An analysis (Ref. 20) was prepared to justify the burden associated with the internal event IPE which is also generally applicable to the external event'IPE request.
This current analysis provides_ additional justification to support ~the extension of the IPE to include external events.
Current risk assessments Refs. 6-8, 14, and 21-26 indicate that the risk from external events could be a significant contributor to core damage'in some instances.
Most n
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}
i 29 recently, the NUREG-1150 (Ref.- 27) study showed that.the contribution to severe' accidents initiated by internal fires and seismic events was comparable to that' initiated by L
internal events.
Examples of the severe accident sequences j
initiated by external: events can be found in references 6-L 8,-14, and 20-26.
Typically, these sequences involved external event initiated transients and small-break loss-of-coolant accidents and were frequently related to lack of redundancy, separation,-and physical. protection in safety-trains for internal fires,-floods, and seismic events.
These results suggest likely areas for cost-effective improvements from plant-specific analyses.that focus.
properly on external events (e.g., the plant support systems c
where there is less redundancy, less separation and J
independence between trains, poorer overall general arrangement of equipment from a safety viewpoint, and much t
more system sharing as' compared to the higher level systems).
Actual 1 examples of cost-effective improvements that have been found and'made'are: modification of structural design to improve capability of the control room to withstand seismic events at Indian Point;' changes-to the turbine building, control room,. turbine building-equipment, and procedural modifications to reduce plant vulnerability to internal floods at Oconee; and enlargement of drainage divertment around the plant-to withstand ~the effects of I
external flood, and installation of a dedicated independent safe shutdown system and.constructica of a separate safe shutdown system building to improve plant. capability to withstand seismic. events, tornadoes, external floods and fires at Yankee Rowe.
In addition, deficient equipment anchorages have been identified and corrected in many plants as a result of'walkdowns like those specified for performance in the IPEEE.
The staff delayed the issuance of the request for a.
4 systematic examination of external events to allow the staff.
to carry out additional work to (1) identify which external j
hazards need an examination, (2) identify acceptable 1
examination methods and develop procedural guidance, and (3) coordinate with other ongoing external event programs.- In December 1987, NRC created the External Events Steering Group-(EESG)-to coordinate the effort to address these-issues.
The EESG established three subcommittees (Seismic; Fires;-and High Winds, Floods, and others).
The staff has completed this-work and is new lequesting that each licensee perform an individual. plant examination of external events (IPEEE) to identify plant-specific vulnerabilities, if any, to' severe accidents and report the results to the Commission.
Experience with' plant specific PRAs since the issuance of the Policy Statement has continued to confirm
.that analyses of this type often reveal plant-specific vulnerabilities that can be and typically are corrected in a
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Because severe accidents dominate nuclear power plant risks, the Commission intends to take all reasonable steps to reduce the chances of occurrence of.
a severe accident and to mitigate the consequences of such an accident should one occur, j
b.
The licensee actions required and the~ cost to develop a response to the information request.
All licensees would be requested to perform an-IPEEE on their plants for plant-specific vulnerabilities to severe j
accidents and report this information to the NRC.
The i
. licensees would also report toithe NRC proposed i
modifications, if any, to correct identified vulnerabilities and indicate how the insights and lessons learned from the examination have been incorporated into plant operation.
The licensees'may perform the IPEEE using methods described F
in the Generic Letter or using other methods that the.
licensees may-propose provided NRC review has shown that i
such proposed methods are effective and applicable, i
We estimate that the cost of these systematic examinations will vary depending upon specific site conditions, but will cost no more than $1M or a maximum of about 6 person-years for the examination.
However, we feel that for most l
licensees the scope will be.less.than the maximum *and the cost will also be less.
Also, for those licensees who have already performed external events PRAs or seismic margin analyses, the actual cost of updating,and submitting the analyses would be significantly less.
We conclude that the burden to be imposed on respondents is justified;in view of the potential safety significance of ensuring that vulnerabilities that may affect' nuclear plant-safety are properly identified and corrected.
c.
An anticipated schedule for the NRC'use of the information.
We expect that mast of the IPEEEs will be submitted by the end of 1993 and that staff review of the results to ensure that the intent of the Commission's Severe Accident Policy Statement is met will be completed by mid 1995.
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, ENCLOSURE 2' NUREG 1407 (Draft) 1 0
i PROCEDURAL AND SUBMITTAL-GUIDANCE FOR THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE)
FOR SEVERE ACCIDENT; VULNERABILITIES' P
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ABSTRACT Based on a Policy Statement on Severe Accidents, the licensee of each nuclear power plant is requested to perform an individual plant examination. The plant examination systematically looks for vulnerabilities to severe accidents and cost effective safety improvements that reduce or eliminate the important vulnerabilFies. This document presents guidance for perfosming and reporting the results of the individual plant examination of external events (IPEEE). The guidance for reporting the results of the individual plant examination of internal events (IPE) is presented in NUREG 1335.
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SUMMARY
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- INTRODUCTION 1-l 1.1-Background 1'
1.2 IPEEE Objectives' 1.3 Purpose of Document 2:
2-EVENTS EVALUATED FOR INCLUSION IN THE IPEEE.
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.L 2.1 Seismic Events.
3-2.2 Internal Fires.
4-2.3 High Winds and Tornadoes 5
2.4 External Floods 5
f 2.5 Transportation and Nearby Facility Accidents-6-
' 2.6 Ughtning -
7
=6 2.7 Severe Temperature Transients (Extreme Heat, Extreme Cold) 2.8 Severe Weather Storms 7'
2.9 External Fires (Forest Fires, Grass Fires)
-7 2.10 Extraterrestrial Activity (Meteorite Strikes, Satellite Falls).
7 2.11 Volcanic Activity 7
2.12 Summary 8,
3 ACCEPTABLE METHODOLOGIES FOR PERFORMING THE SEISMIC IPEEE i
3.1. Seismic PRA 9
l 3.1.1 New Seismic PRA Analysis 9:
I 3.1.1.1 General Considerations.
9:
3.1.1.2 Hazard Selection' 10' 3.1.1.3 Fragility Estimation..
11 3.1.1.4 Seismic PRA Methodology Enhancements 12; i
3.1.1.5 Containment Performance 12 3.1.2 Use of an Existing PRA' 12 3.2 Seismic Margins Methodologies q
3.2.1 General Considerations 14
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3.2.3 NRC Seismic Margins Methodology 15 3.2.4 EPRI Seismic Margins Methodology i7-
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'3.2.5: Reduced Scope Margins Method 18-i.
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3.2.6 ~ Containment Performance 18 3.3 Optional Methodologies 19 j
4 ACCEPTABLE METHODOLOGY FOR PERFORMING-THE INTERNAL FIRES-IPEEE l
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4.1 New Fire PRA Analysis 22; 4.1.1 Identify Critical Areas of Vulnerability
' 22 22-4.1.2 Calculate the Frequency of Fire initiation in Each Area L
4.1.3 Analyze for the Disabling of Critical Safety Functions.
23 l
4.2 Use of an Existing Fire PRA 23-l 4.3 Optional Methodologies.
23-5.
ACCEPTABLE METHODOLOGY FOR PERFORMING THE HIGH WINDS, FLOODS, >
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. AND TRANSPORTATION AND NEARBY FACILITY ACCIDENT IPEEE l
d 24-5.1 Introduction
~ 24 5.2 Analytical Procedure 5.2.1 Review Plant Specific Hazard Data and Licensing Bases
- 24 l
5.2.2-Identify Significant Changes Since OL issuarice 24 l
5.2.3' Determine if the Plant / Facilities Design Meets Current; I
Criteria.
25 I
5.2.4 Determine if the Hazard Frequency is Acceptably Low -
.25 (Optional Step) l 5.2.5 Perform a Bounding Analysis (Optional Step) _
25-(
5.2.6 Perform a Probabilist!c Risk Assessment (Optional Step)
. 25 5.3 Optional Methodologies 25 1
6 COORDINATION WITH ONGOl'NG PROGRAMS
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6.2 Description of Ongoing Programs '
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6.2.1 IPE Program Related to Internal Events
-27 l
6.2.2 Ongoing Programs Related to External Events 27 6.2.2.1 Seismic Programs 27 6.2.2.2 Intemal Fires Programs'
'28 1
6.3 Approach on Coordination with. Ongoing. Programs
- 29 6.3.1 Coordination with Internal Events Program.(IPE).
29 i
6.3.1.1 Preanalyses' Planning -
29~
6.3.1.2 Plant Modifications -
29 4
6.3.1.3 Accident Management 30 0
6.3.2 Coordination Among Extemal Events Programs 30' 6.3.3 Coordination With Seismic Programs
' 30 1
6.3.3.1-USl A 45 and GI 131 31 6.3.3.2 Charleston Earthquake issue -
31' 6.3.3.3 USl A-46 31 6.3.4 Coordination With Other issues 33 7
DOCUMENTATION AND REPORTING a
' 7.1 Information Submitted to the NRC 34 7.2 Iriformation Retained for Audit 34 vi l
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REFERENCES 35 I
'i APPENDICES -
A.
~ Review Level Earthquake 38-B.
Comparison Between a Reduced Scope and Full-Scope Seismic Margins Evaluation - -
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Detailed Documentation and Reporting Guidelines 47-0 LIST OF FIGURES
!l 5.1' Recommended IPEEE Approach for Winds, Floods, and Others 26=
1 LIST OF TABLES 3.1 Review Level Earthquake - Plant Sites East of the' Rocky Mountains 20 d
3.2 Review Level Earthquake - Westem United States Plant Sites 21 j
B.1-Reduced Scope Margins Method 46 C.1 Standard Table of Contents for IPEEE Submittal 54
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ACKNOWLEDGEMENTS This document represents the staff position on the Individual Plant Examination for severe accident vulnerabilities due to external events (IPEEE) Representatives of both the Office of Nuclear Regulatory Research and the Office of Nuclear Reactor Regulation were active contributors to the process; they are named below, in addition, significant input was received from consultants and contractors to the NRC, who are also named below. Edward Hill, of NRC, provided technical editing.
NRC Thomas E. Murley Thomas L King Demetrios L. Basdekas Lawrence C. Shao Nilesh C. Chokshi Pei Y. Chen Warren Minners David C. Jeng P.T.Kuo Glenn B. Kelly Roger M. Kenneally Wililam D. Beckner Thomas M. Cheng Jack Strosnider Kazimieras M. Campe Rex G. Wescott Conrad McCracken Adel A. El Bassioni John H. Flack David P. Notley Jocelyn A. Mitchell Thomas M. Novak James E. Richardson R. Wcyne Houston Guy A. Arlotto Brian Sheron Goutam Bagchi Andrew J. Murphy Leon Reiter T. Y. Chang John T. Chen
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Consultants R. J. Budnitz Future Resources Associates, Inc.
G. E. Cummings Lawrence Uvermore National Laboratory R. P. Kennedy t ructural Mechanics Consulting, Inc.
M. K. Ravindra ROE Engineering, Inc.
Contractors and Subcontractors P. Amico Science Application International D. L Bernreuter Lawrence Uvermore National Laboratory M. P. Bohn Sandia National Laboratory R. J. Budnitz Future Resources Associates, Inc.
D. H. Chung Lawrence Uvermore National Laboraton/
B C. Davis Lawrence Uvermore National Laboratury G. S. Hardy Lawrence Uvermore National Laboratory J. R. Mcdonald Texas Tech University R.'C Murray Lawrence Uvermore National Laboratory A. M. Nafday EOE Engineering, Inc.
P. G. Prassinos Lawrence Uvermore National Laboratc;y M. K. Ravindra EOE Engineering, Inc.
J. Savy Lawrence Uvermore National Laboratonf lx
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J EXECUTIVE
SUMMARY
BACKGROUND In the Commission policy statement on severe accidents in nuclear power plants issued i
on August 8,1985, the Commission concluded, band on available irdormation, that existing plants pose no undue risk to the public hea!!h and safety and that there is no i
present basis for immediate action on any regulatory requirements for these plants.
However, the Commission has recognized, based on NRC and industry experience with t
plant specific probabilistic risk assessments (PRAs), that systematic examinations are beneficial in identifying plant specific vulnerabilities to severe accidents that could be fixed with low cost improvements. As part of the implementation of the Severe Accident Policy, the Commission issued Generi:: Letter 88 20 on November 23,1988, requesting l
that each licensee conduct an individual plant examination (IPE) for intsmally initiated events including internal flooding.
Many PRAs indicate that the risk from externe.1 evrnts could be a significant contributor to the core damage in some instances. However, the examination for externally initiated events is proceeding on a later schedule to allow the staff to carry out additional work to (1) identify which external hazards need a systematic examination, (2) identify 4
acceptable examination methods and develop procedural and submittal guidance, and (3) coordinate the IPEEE with other ongoing external event programs. In December 1987, an External Events Steering Group (EESG) was established to make I
recommerfstions regarding the scope, methods, and coordination of the IPEEE.
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The EESG has recently completed its task.
This report, based upon the EESG recommendations, provides detailed guidance to the licensees on the conduct of the IPEEE and on the structure and content of the IPEEE submittal, it provides,
specifically, the guidelines de'ining the IPEEE objectives; identify externa! events that should be included in the IPEEE; identify, cceptable methodologies; and identify coordination between the IPEEE and the ongoing NRC programs.
.QBEC71VES OF THEREEE The ganeral objectives of the IPEEE are similar to that of the IPE-that is, for each licenset, (1) to develop an appreciation of severe accident behavior, (2) to understand the most likely severe accident sequences that could occur at its plant, (3) to gain a qualitative anderstanding of the overall probability of core damage and fission product releases, and (4) if necessary, to reduce the overall probabilities for core damage and fission produc.t releases by modifying, where appropriate, hardware and procedures that would help prevent or mitigate severe accidents.
IDENTIFICATION OF EXTERNAL EVENTS INCLUDED IN THE IPEEE In supporting the implementation of the Severe Accident Policy, a study was performed, to determine which external initiators could be a potentially important accident initiator that may pose a threat of severe core damage or of a large radioactive release to the xi p
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environment. The external events considered, consistent with past probabilistic risk assessments (PRAs), are those events whose cause is external to all systems used in normal operation and emergency operation situations. The external events evaluated include seismic events, internal fires, high winds and tornadoes, external floods, transportation and nearby facility accidents, lightning, severe temperature transients (extreme heat, extreme cold), severe winter storms, external fires (forest fires, grass fires), extraterrestrial activity (meteorite strikes, satellite falls), and volcanic activity.
Based on the results of that study, the staff has concluded that five external events need to be included specifically in the IPEEE: seismic events, internal fires, high winds, floods, and transportation and nearby facility accidents, However, licensees should confirm that no other plant un;que external events with potential severe accident vulnerability are being excluded from the IPEEE.
EXAMINATION METHODS Seismic Events A seismic PRA (Level I plus containment performance) or a seismic margins methodology (SMM) is considered a viable approaches to identify potential vulnera-bilities. Guidance is provided for licensees performing a new seismic PRA or updating an existing seismic PRA; emphass is placed on the identification and ranking of dominant plant sequences that could lead to seismically induced core damage rather than the numerical estimate of absolute frequency of occurrence.
Methodology upgrades include plant walkdowns; evaluation of relay chatter and soil liquefaction effects, and results in terms of high confidence low probability of failure (HCLPF) values.
Guidance is also provided for licensees using the NRC or EPRI sponsored seismic margins methodology. The margins methodology screens components based on their importance to safety and seismic capacity. By design, the methodology utilizes two review or screening levels geared to pek ground accelerations of 0.3g and 0.5g.
Review level earthquakes (RLE) were assigned based on the LLNL and EPRI hazard estimates, sensitivity tests, and seismological and engineering judgment. The use of the 0.3g RLE for most Central and Eastern United States plants would meet IPEEE objectives. For some sites where the seismic hazard is low, a reduced scope margins methodology emphasizing plant walkdowns is considered adequate.
For Western United States sites, except California coastal sites, 0.5g RLE should be used.
Methodology upgrades include relay chatter, liquefaction, and plant walkdown enhancements for the NRC method, guidance on alternative success paths for the EPRI method; and nonseismic failure and human actions for both methods.
Internal Fir.g3 The internal fires IPEEE can be accomplished by performing a Level I fire PRA. Those issues, identified in the Fire Risk Scoping Study should be addressed using plant-specific data and a specially tailored walkdown procedure.
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The guidance does not address a simplified fire methodology currently being developed by the Nuclear Management and Resources Council (NUMARC). This methodology will be reviewed by the staff when it is completed and submitted.
Hioh Winds. Floods. and Transoodation and Nearbv Facility Accidents The recommended overall approach consists of a progressive screening.
The screening criterion for reporting potential severe accident sequences is consistent with internal event IPEs. The steps in the progressive screening approach represent a serios of analysis in increasing level of detail, effort, and retolution. However, the licensee may choose to bypass one or more steps so long as the vulnerabilities are either identified or demonstrated to be insignificant. The screer%g approach consists of the following steps:
All Plants:
1.
Review plant specific hazard data and licensing bases 2.
Identify significant changes since operating license issuance 3.
Detemine if the plant / facilities design meets current (1975 SRP) criteria If the above are not satisfied, one or more of.he following steps should be taken to further evaluate the situation.
Optional Effort:
i 4.
Determine if hazard 'friquen~cy is acceptably low 5.
Perform bounding analysis 6.
Perform a probabilistic risk assessment (PRA}
Alternative Methods The staff recognizes that other methods capable of identifying plant specific vulnerabilities to severe accidents may be acceptable. A licensee may request a review r,f any other systematic examination method to determine its acceptability for IPEEE purposes.
4 COORDINATION Guidance is provided to coordinate the IPEEE process with ongoing programs. The first coordination level is among the major elements of the severo accident policy implementation, that is, coordination among the IPEEE, the internal events IPE, containment performance improvements, and accident management.
The second coordination level is among the major elements of the IPEEE, that is, seismic events, fires, and high winds, floods, and others. The third level of coordination is within each major element of the IPEEE.
Programs subsumed into the IPEEE include the selsmic aspect of USl A 45 (Decay Heat Removal), GI 131 (In Core Flux Mapping System), and the Charleston Earthquake issue. Programs that need to be coordinated with the IPEEE include USI A-46 (Seismic xiil 4
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Equipment Qualification which also covers the seismic spatial interaction of USl A 17 and the concern of USl A 40 for the seismic capability of large safety related above-i' ground tanks) and GI 57 (Effects of Fire Protection System Actuation on Safety Related Equipment).
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1 INTRODUCTION 1.1 Backaround On August 8,1985, the Nuclear Regulatory Commission issued a policy statement on severe accidents (NRC,1985).
The Commission concluded, based on available information, that existing plants pose no undue risk to the public health and safety and that there is no present basis for immediate action on any regulatory requirements for these plants.
However, the Commission recognizes, based on NRC and industry i
experience with plant specific probabilistic risk assessments (PRAs), that systematic
.xaminations are beneficial in identifying plant specific vulnerabilities to severe accidents i
that could be fixed with low cost improvements. As part of the implementation of the policy statement, the Commission issued Generic Letter 88 20 (NRC,1988 and 1989),
requesting that each licensee conduct an individual plant examination (IPE) for internally initiated events.
Risk assessments indicate that the risk from external events could be a significant contributor to the core damage in some instances. However, licensees were requested 3
to proceed with the examinations only for internally initiated events (including internal i
2 flooding) in Generic Letter 88 20. Examination of severe accident vulnerabilities due to l
externally initiated events (IPEEE) is proceeding separately and on a later schedule to allow the staff to carry out additional work (SECY 88147) to (1) identify which external hazards need a systematic examination, (2) identify examination rnethods and develop procedural guidance, and (3) coordinate the IPEEE with other ongoing NRC programs that deal with various aspects of external event evaluations to ensure that there is no duplication of Industry efforts.
To accomplish these objectives, the staff established the External Events Steering Group (EESG) in December 1987 to make recommendations regarding the scope, methods, and coordination of the IPEEE (Beckjord, 1987, 1988).
Specifically, the EESG is responsible for developing broad guidance for dealing with (1) external events on a generic basis bcth organizationally and technologically and (2) the implementation of the severe accidsnt policy with respect to external events. The EESG established three technical subcommittees dealing with earthquakes (seismic events), internal fires, and high winds, floods, and 'other" external events. The subcommittees were chartered to l
riefine the scope of the external events examination,-identify acceptable examination methodologies, and coordinate ongoing issues and activities (for example, Unresolved Safety Issues and Generic Issues).
1.2 IPEEE Oblectives The objectives of the IPEEE, which are similar to the objectives of the internal event IPE, are for each licensee:
1.
to develop an appreciation of severe accident behavior, l
2.
to understand the most likely severe accident sequences that could occur at the licensee's plant,
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to gain a qualitative undarstanding of the overall probability of core damage and fission product releases, and 4.
if necessary, to reduce the overall probabilities of core damage and radioactive material releases by modifying, where appropriate, hardware and procedures that would help prevent or mitigate severe accidents.
However, it was recognized at the outset _that the external event initiators could not necessarily be treated in exactly the same manner as internal event initiators in the implementation of the Severe Accident Policy.
This is because the sources and treatment of uncertainties in estimates of core damage frequencies for external and internal events can be quite different, in addition, some methods have been developed for evaluating externci hazards and identifying vulnerabilities that do not produce estimates of damage frequenm.
For example, seismic margins methods produce estimates of seismic hazard levels of high confidence low probability of failure (HCLPF) for a plant rather than estimates of core damage probability.
Therefore, the staff determined that an explicit estimate of core damage frequency was not needed to meet the intent of the Severe Accident Policy and would not be a requirement of the IPEEE. Thus, objective 3 above would be addressed only indirectly for some methods to be used in the IPEEE.
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1.3 Purnose of Document The purpose of this document is to provide guidelines for conducting the individual plant examination of external events (IPEEE) and to provide guidelines on the structure and content of the IPEEE submittal. The extemal events recommended for inclusion in the IPEEE are identified in Section 2. Acceptable methodologies for performing an IPEEE along with upgrades to reflect state-of the-art improvements are identified in Sections 3 through 5. Section 3 addresses the seismic portion; Section 4 the intemal fires portion; and Section 5 the high winds, floods, and other portion of the IPEEE.
i Coordination between the IPEEE and the internal events IPE, other external events, and l
ongoing programs within each extemal event are provided in Section 6. A summary of documentation and reporting guidelines is provided in Section 7.
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2 EVENTS EVALUATED FOR INCLUSION IN THE IPEEE The external events considered, consistent with past probabilistic risk assessments (PRAs) are those events whose cause is external to all systems used in normal operation and emergency operation situations.
Internal fire and internal flood are external to the ' system" and therefore have been considered as external events in past PRAs. Flowever, internal floods are being considered in the internal events IPE process (NRC,1988).
In supporting the implementation of the Severe Accident Policy, a study of the risk of core damage to nuclear power plants in the United States due to externally initiated events was performed. The objective was to determine which external initiators have the potential of initiating an accident that may lead to severe reacter core damage or large radioactive release to the environment. Seismically initiated events are investigated in NUREG/CR 5042, Suppl.1; internal fires, high winds / tornadoes, external floods, and transportation accidents are investigated in NUREG/CR 5042; "other external events" are investigated in NUREG/CR 5042, Suppl. 2.
The "other external events" covered are nearby industrial / military facility accidents, on-site hazardous material storage accidents, severe temperature transients, sevue wepher storms, lightning strikes, external fires, extraterrestrial activity, volcanic actMty, er.h movement, and abrasive windstorms.
Some external events may not pose a significant threat of a severe accident to all plants, some events may have been considered in the plant's design to a sufficient degree, and some events may have been or will be reviewed under ongoing programs,
at some plants. The staff's evaluation and recommendations are contained in the following sections, 2.1 Seismic Events The following are based upon an examination of current seismic design criteria, previous and ongoing seismic issues and programs, and seismic PRAs:
1.
Mean seismic cora damage frequencies calculated from past PRAs (NUREG-1150, NUREG/CRG42, Suppl.1) have been found to be in the range of 10" to 4
10 per year. Identified vulnerabilities are plant specific and include yard tanks, electrical equipment, diesel peripherals, structural failures, and equipment anchorages.
2.
New data such as the occurrences of larger than anticipated earthquakes and the development of new hypotheses indicate that the plant specific seismic l
hazard may be quite different from that envisioned at the time of licensing and
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makes it difficult to rule out seismic events on the basis of initiating event frequencies.
3.
Based primarily on their vintags, the current population of plants exhibit various levels of seismic design requirements and margin. Some of the very early plants have been backfitted under the Systematic Evaluation Program to ensure certain 3
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4.
There have been modifications to plants since their original designs, for instance, the reduction of snubbers at some plants. These changes, in part, have relied on existing conservatism or risk based arguments (e.g., LOCA + SSE combinations). The systematic examination of plants by the IPE and IPEEE will i
give an integrated picture of plants as they exist. It will also allow an integrated evaluation of the effects of individual changes made to plants over time.
5.
There are unresolved safety issues and generic issues (e.g., USl A-45, USl A-
- 46) that are in various stages of implementation. The IPE/IPEEE provides a convenient as well as meaningful framework for addressing many of these issues, i
6.
PRAs and seismic margins evaluations have resutted in cost effective plant-specific improvements.
Therefore, the seismic external hazard should be included in the IPEEE.
2.2 Internal Fires Based upon the examination of past fire PRAs, the contribution of internal fires to the probability of core damage may be significant and is very plant specific (NUREG/CR-5042)..However, the numerical results always contain large uncertainties. The fire risk scoping study (NUREG/CR 5088) further confirms the following:
l 1.
The overall fire-induced core damage frequency for the four plants studied s
(Seabrook, Oconee, Umerk:k and Indian Point) increased from the original PRA studies even though, for certain fire scenarios, there was a net decrease. For all plants reviewed, fire continues to represent a dominant risk contributor.
2.
Most initiating event frequencies were increased based on a much more complete data base available on fire occurrences in nuclear power plants. Under currently applied risk assessment rnethodologies, this increase in initiating event frequency alone results in a direct increase in overall fire-induced coro damage frequency with all other factors remaining constant.
1 1
3.
Use of an expanded data baee on historical fire suppression times for nuclear power plants resulted !. a suppression probability distribution with a lower probability vi suppression within a given time than that assumed in the original risk assessments. Under current methodologies, this again results in an increase in fire initiated core damage frequency with all other factors remaining constant.
4.
Updated information o' the ignition and damage thresholds of cable insulations in some cases resulted h lower thermal damage limits, in some cases, no change in damage limits was required. A decrease in the assumed thermal damage limits would, in genval, be expected to lead to increased estimates of fire initiated core damage frequency, 4
5.
Plant modifications made as a result of Appendix R requirements reduced core damage frequency at Indian Point and Limerick for the requantified areas by factors of ten and three, respectively. For Seabrook, the identified Appendix R plant modifications did not affect the requantified core damage f,cenarios for internal fires.
The Oconee PRA had already incorporated Appendix R modifications and no modifications subsequent to its performance were identified.
Hence no Appendix R impact could be identified for either G*abrock or Oconee.
6.
A number of issues that we:e not addressed in the past fire PRAs (effectiveness of fire brigade, effectiveness of fire barrier, seismic / fire interactions, control system interactions, and effects of fire suppressants on safety equipment) could increase vulnerability to fire.
Therefore, based on the above studies, toe internel fire hazard should be ine!uded in the IPEEE.
2.3 Hloh Winds and Tomadoes For plants designed against NRC's current criteria, these events pose no significant threat of a severe accident because the current design criteria for wind are dominated by tomadoes having a frequency of exceedance of about 10, However, older plants and some modem plants having facilities not designed against these criteria need a systematic examination to identify plant specific vulnerabilities (NUREG/CR 5042).
2.4 External Floods For plants designed against current criteria as described in Regule'.ory Guide 1.59 and applicable Standard Review Plan Sections, particularly Section 2.4, floods pose no significant threat of a severe accident because the exceedance fraquency of the design basis flood, excluding floods due to failure of upstream dams, is judged to be less than 4
10 (Chery,1985), and the conditional core damage frequency for a design basis flood 4
is judged to be less than 10. Thus core damage frequencies are estimated to be less 4
than 10 per year for the plant designed against NRC's current criteria. However, the latest probable maximum precipitation (PMP) criteria published by the National Weather Service (NWS) call for higher rainfall intensities over shorter time intervals and smaller areas than have previously been considered; this could result in higher sito flooding levels and greater roof ponding loads than have been used in previous design bases (Gl 103). Ucensees are requested to assess the effects of applying these new criteria to their plants in terms of onsite flooding and roof ponding. Also, some older plants may have higher potential risk and need systematic examinations for plant specific i
vulnerabilities, j
l 2.5 Transoortation and Nearby Facility Accidents i
These events consist of accidents related to transportation and accidents at industrial l
and military facilities.
Plants designed against NRC's current criteria (NUREG/CR-l 5
l l
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o.
5042) should have no significant vulnerability to severe accidents from these events because the initiators considered in the design should have a recurrence frequency less 4
than 10 or have been shown through a bounding analysis not to affect the plant.
However, changes may have occurred since the original design and there may be exceptions that need some systematic examination. Also, some older plants may not meet the NRC's current criteria and need systematic examinations for plant specific vulnerabilities.
2.6
.U.Qdo.ing Ughtning has eeen experienced in many nuclear power plants in the United States (NUREG/CR 504:2. Suppl. 2; AEOD,1986; ACRS,1989). The impact of lightning on plant operation and the vulnerability of plants to a severe accident due to lightning has been examined. The major conclusion is that the primary impact of lightning on nuclear power plants is the loss of offsite power. The loss of offsite power is included as part of the internal events IPE, and examination for vulnerabilities due to this aspect of lightning is therefore already included in the IPE process. The staff has concluded that, in general, other effects of lightning on nur. lear power plants are insignificant. However, for certain sites where lightning strikes are likely to affect more than just loss-of offsite power, further examination on lightning 9ffects may be warranted.
Based upon an examination of historical data on lightning, as well as knowledge of plant systems, the staff concludes:,
1.
Ughtning has typically caused partial or complete loss of offsite power, which is the main impact of lightning and which is already being examined as part of the inwrnal events IPE.
2.
Ughtning is much less likely to affect the onsite power system.
3.
Ughtning has affected safety related equipment and has caused reactor trips, but these events have not been significant in terms of impact on the plant.
4.
Safe j systems (e.g., diesel generators, electrically powered pumps) are not normally in operation. Thus, while control systems may be damaged, the safety systems are less susceptible to damage and may be manually activated.
5.
Redundancy of safety systems and the capability for recovery of systems (replacing fuses or resetting breakers) further reduce the likelihood that the low frequency of lightning damcgo events will result in a severe accident.
The staff has judged that the probability of a severe accident caused by lightriing (other than one due to loss of offsite power) is relati';Lly low and further consideration on lightning effects should be performed only for plant sites whera lightning strikes are likely to cause rpore than just loss of offsite power.
2.7 Severe Ternoerature Transients (Extreme Heat. Extreme Col _d) 6 m
Severe temperature transients may affect nuclear power plants in the United States (NUREG 1032). However, the effects are usually limited to reducing the capacity of the ultimate heat sink and loss of offsite power (NUREG/CR 5042, Suppl. 2). The capacity reduction of the ultimate heat sink would be a slow process that allows plant operators sufficient time to take proper actions such as reducing power output level or achieving safe shutdown, if necessary, and maintaining the plant in a safe shutdown condition.
The other potential impact on the plant, loss of offsite power, will be considered within the realm of the station blackout rule (NRC,1988b) and the internal event IPE.
Therefore, the temperature transients need not be addressed in the IPEEE.
2.8 Severe Weather Storms Severe weather storms (ice storm, hailstorm, snowstorm, dust storm, sandetorm) accompanied by strong winds have caused several complete and partial lesses of offsite power (NUREG/CR 5042, Suppl. 2). The potential effect of loss of offsite power and station blackout will be addressed in the internal event IPE; thus severe weather storms need not be repeated in the IPEEE.
I 2.9 External Fires (Forest Fires. Grass Fires)
These are fires occurring outside the plant site boundary. Potential effects on the plant i
could be loss of offsite power and forced isolation of the plant ventilation and possible control room evacuation. Usually, external fires are unable to spread onsite because of site clearing during the construction stage (NUREG/CR 5042, Suppl. 2). However, there has been one instance during which a nearby forest fire caused a partial loss of i
offsite power. The effect of loss of offsite power will be addressed in the Internal events IPE and need not be repeated in the IPEEE. The other effects have been evaluated during operating license (OL) review against sufficiently conservative criteria; thus they do not need to be reassessed in the IPEEE.
2.10 Extraterrestrial Activity (Meteorite Strikes. Satellite Fa!!si Extraterrestrial activity is considered to be natural satellites such as meteors or artific al satellites that enter the earth's atmosphere from space. Because the probability of a meteorite strike is very small (less than 103 (NUREG/CR 5042, Suppl. 2), it can t e dismissed on the basis of its low initiating event frequency.
2.11 Volcanic Activity Most nuclear power plant sites are too far away from active volcanos to expect any effect at the plant so most licensees need not consider the volcanic effects. However, those sites in the vicinity of active volcanoes should assess volcanic activities (NUREG/CR 5042, Suppl. 2) as part of the IPEEE process.
2.12 Summarv 7
,~
In summary, based on the above evaluation, five events need to be included by all licensees in the IPEEE: seismic events, internal fires, high winds, floods, and transportation and nearby facility accidents. Alllicensees should confirm, however, that no plant unique external events known to the licensee today with potential severe accident vulnerability are being excluded from the IPEEE.
i.
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3 ACCEPTABLE METHODOLOGIES FOR PERFORMING THE SEISMIC IPEEE For the purposes of an IPEEE, two methodologies are considered acceptable to identify potential seismic vulnerabilities at nuclear power plants.
The first is a seismic probabilistic risk assessment (NUREG/CR 2300, NUREG/CR 2815, Vol. 2), the second is one of the seismic margins methodologies (SMM) described in NUREG/CR-4334 and EPRI NP-6041 or the reduced SMM described later in this section, in meeting the objectives of the IPEEE, the examination should focus on qualitative insights from the systematic plant examination rather than only on absolute core damage frequency estimates. Guidance for performing the seismic IPEEE using a PRA or margins methodology is provided below.
3.1 Seism!c PRA This discussion deals with the use of PRA techr.lques in the selsmic IPEEE. The PRA should be at least a Level 1 plus containment performance analysis.
The basic elements are (1) hazard analysis, (2) plant system and structure response analysis, (3) evaluation of component fragilities and failure modes, (4) plant system and sequence analysis, and (5) containment and containment system analysis including source terms, to identify unique seismic sequences or vulnerabilities different from the internal event analysis. Specific guidance and enhancements are provided for licensees performing a new PRA or updating an existing seismic PRA.
L 3.1.1 New Seismic PRA Anaihsis 3.1.1.1 General Considerations Ucensees choosing to do a seismic PRA built on an internal events PRA should be aware of important considerations that, if incorporated in the planning of the internal events PRA, will minimize their resource expenditure and speed the staff reviews. For example, (1) a well-organized walkdown team and a properly planned walkdown will enable many issues to be addressed at the same time; (2) the independent peer review group should consider the need to review both internal and external event analyses; (3) fault tree analysts for internal events should be aware of spatial interactions (including Internal flooding effects) and common cause effects and the culling or pruning of trees should be done with these considerations in mind; and (4) internal event models should i
l be developed knowing that, in the seismic analysis, the fragilities of a component are sensitive to elevation. Also, a component arf its peripheral equipment may have l
different fragilities.
PRA calculations that account for all uncertainties are clearly acceptable. However, the staff believes that, for the seismic IPEEE, it is not necessary to carry out complete uncertainty quantifications defining a distribution of core damage frequencies in order to identify vulnerabilities. Mean point estimation using a single hazard curve (rather than a family of hazard curves) and a single fragility curve (rather than a family of fragility curves) for each component is sufficient to get insights into potential seismic 9
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)
vulnerabilities. Mean point estimates using hazard curves described in NUREG/CR-5250 and EPRI NP 6395D should be obtained. This will encourage the most pertinent results/ insights from the seismic portion of the IPEEE to be highlighted.
The above point estimation approach is valid only because of the IPEEE objective: to identify dominant sequences and components and where post,ible rank them. (This point estimate should not be confused with a
- Phase 1" type PRA analysis where point estimate calculations are used only to define scopes for more detailed Phase 11 and Phase ill studies). Fragilities used in this point estimate,where possible, should be plant specific and rigorous to be able to identify dominant components and rank them.
Correlations and other aspects of analysis should be treated so that, when a mean i
seismic hazard curve is used with the mean plant fragility curve, the resulting core damage estimate approximate'y represents the mean estimate that would be derived from the full uncertainty analysis.
The recommendation of performing point estimation type calculations is made primarily to highlight insights needed for the severe accident behavior perspective. This should 1
not be construed as deemphasizing or ignoring uncertainties, Analysts are encouraged to make careful study of the origins of the possible uncertainties, including those that are hardest to quantify.
Many of the insights obtained from a PRA analysis are obtained by trying to gain a better understanding of the uncertainties. Consideration of uncertainties may affect how the results of a PRA are implemented in piant changes.
3.1.1.2 Hazard Selection For the United States east of the Rocky Mountains, two highly sophisticated seismic hazard studies were conducted by Lawrence Uvermore National Laboratory (LLNL)
(NUREG/CR 5250) and the Electric Power Research Institute (EPRI) (EPRI NP-6395-D) For many sites, these studies yield significant differences at the low probability and high level ground motions. The initial PRAs carried out using these eulmates (Surry and Peach Gottom in NUREG 1150) indicate that, despite large differences in absolute l
numerical estimates, the identification, ranking, and relative contributions of the dominant seismic sequences are virtually the same for both LLNL and EPRI hazard estimates.
This equivalence is apparently due to the fact that the slopes of the seismic hazard curves are not significantly different over those ground motion levels that, in conjunction with the fragilities, control the relative distribution of seismically induced core damage frequencies. Although these results are very encouraging, there is no guarantee that this will be true for all sites in the Central and Eastern United States.
The staff position is that, while a full seismic hazard uncertainty analysis is not necessary in performing a seismic PRA for the IPEEE, mean (arithmetic) hazard estimates from both the LLNL and EPRI should be used to obtain two different point (mean) estimates. The use of both of these estimates will serve to identify differences, if any, in the delineation of dominant seismic sequences (minor variations in contributions and rankings are anticipated),
Such differences would have to be 10
l addressed by the licensee in its identification and listing of vulnerabilities. The use of both the LLNL and EPRI mean hazard curves has another advantage in that the extent of unce.1sinty will become obvious and the emphasis on the bottom line numbers is reduct d.
For plants in the Western United States, for which there are no counterparts to the LLNL and EPRI studies, a licensee should conduct 5ts own study to define the mean hazard estimate for use in the IPEEE. The licensee should also provide reasonable assurance that any significant uncertainty in those elements of hazard (for example, slope) that control the identification, ranking, and relative contribution of seismic j
contributors to core damage is addressed in sensitivity studies. As in the Central and Eastern United States, the identification and listing of vulnerabilities should take this uncertainty into account.
Most seismic PRAs use peak ground acceleration as the hazard parameter, if this is done, si, tral shapes that are consistent with current estimates of ground motion should be used. In the Central and Eastern United States, current spectral estimates can be found the LLNL and EPRI studies. Since similar spectral shape are obtained l
from LLNL and EPRI hazard studies, separate analyses using both spectral shapes are not needed.
Medium spectral shapes of 10,000 year retum period provided in NUREG/CR 5250 along with variability estimates are recommended for use in the analyses.
Other site specific spectral shape estimates may be proposed (that is, derived from a suite of appropriate recorded earthquakes). For the Western United States, site specific spectral shape should be established and used.
If an upper bound cutoff tc Ground motion at less than 1.5 g peak ground acceleration is assumed, sensitivity studies should be conducted to determine whether the use of this cutoff affects the delineation and ranking of selsmic sequences.
3.1.1.3 Fragility Estimation The following guidance on fragility estimation is included to clarify the use of fragility in the context of the ' point estimation' approach discussed above. Details and methods for fragility and high confidence-low probability of failure (HCLPF) calculations are discussed in a number of references, for example, NUREG/CR 2300, NUREG/CR-4334, EPAl NP 6041, and NUREG/CR 5270. It is recognized that large uncertainties l
exist in fragilities estimation (NUREG/CR 5270). A perspective on how this uncertainty affects the results of analysis (numerical and other insights, for example, dominant sequences and components) should be maintained. Acceptable seismic component fragilities in a HCLPF format are given in NUREG/CR 4659.
Consistent with the point estimation approach, one can use a single mean component fragility curve for each component and hence for sequence level and plant level assessmen;s. This mean curve is defined by the median capacity, E, and composite 8
2 l-uncertainty, Gc, where Oc = Br + Bu, when Br and Du are estimated separately. Gr I
and Du represent random uncertainty and modeling uncertainty, respectively, it is also acceptable to use a family of fragility curves instead of a single curve, 11 P
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When a single mean fragility curve is available, HCLPF capacity for a component (sequence, or plant) can be approximated by 2.3 Bc below the median (i.e.,1%
composite probability of failure is essentially equivalent to 95% confidence of less than 5% probability of failure). While developing sequence level and plant level fragilities, licensee should retain the ability to report HCLPFs with and without nonseismic failures and human actions.
3.1.1.4 Seismic PRA Methodology Enhancements Based on a review of past seismic PRAs, certain areas have been treated inconsistently or not at all.
The following areas should be included:
1.
Plant Walkdowns. Walkdowns are performed to find as designed, as built and as operated seismic weaknesses in plants.
Each licensee should perform a walkdown consistent with the guidelines described in the EPRI Seismic Margins Methodology (EPRI NP-6041) (team composition, documentation, etc).
2.
Relay Chatter. An acceptable procedure for addressing relay chatter issues is described in Hardy, et al.,1989. Relays, in this context, include components such as electric relays, contactors, and switches that are prone to chatter. The 4
examination of the relay chatter effects (for example, the Hatch margins evaluation) has resulted in large resource expenditures in terms of staff hours.
Therefore, careful planning and use of generic insights, if they are appilcable to the plant, are necessary.
3.
HCLPF Calculations. Ucensees should report plant level, sequence level, and component level HCLPFs. In several PRAs (for example, Millstone 3 and Diablo Canyon), HCLPF 6stimates are reported along with other PRA results. These PRAs can be used for guidance to derive HCLPFs from fragilities. HCLPFs are to be reported both with and without the effects of nonseismic failures and human actions.
4.
Uquefaction. The potential for soll liquefaction and associated effects on the plant need to be examined for some sites because of specific site conditions.
The impact on plant operation should be assessed from the point of view of both potential for and consequences of liquefaction. Procedures for assessing soll liquefaction are described in EPRI NP 6041.
3.1.1.5 Containment Performance The primary purpose of the containment performance evaluation is to identify sequences and vulnerabilities that involve containment, containment functions, and containment systems (e.g., fans, sprays, igniters, and suppression pools) seismic failure modes or timing that are significantly different from those found in the IPE internal events evaluation. Additional guidance is presented in Section 3.2.6.
12
j 3.1.2 Use of an Existing PRA The use of an existing seismic PRA to address the seismic IPEEE is acceptable provided the PRA reflects the current as built and as operated condition of the plant and 4
some of the deficiencies of past PRAs discussed below are adequately addressed.
1 1.
Hazard Selection. For PRAs at sites east of the Rocky Mountains that did not 4
use the LLNL and EPRI hazard estimates, sensitivity studies should be conducted to determine if the use of these results would affect the delineation or ranking of seismic sequences. For PRAs in the Western United States, the sensitivity studies should be carried out to determine the effect of uncertainty in hazard on the delineation and ranking of seismic sequences.
1 L
l 2.
Walkdowns. Since a walkdown is considered to be one of the most important ingredients of the seismic IPEEE, a supplementary walkdown in conformance with the procedures described in the EPRI margin methodology (EPRI NP 6041) (team composition, documentation, etc.) should be performed. it may be necessary to amplify the earlier analysis based on the walkdown outcome. These results should be reported.
3.
Relay Chatter. Relay chatter effects either have not been considered or were assumed fully recoverable in past PRAs.
Relays, in this context, include components such as electric relays, contactors, and switches that are prone to chatter. Ucensees should analyze the effect of relay chatter or determine thct -
the type of relays used in the safety systems are not subject to relay chatter.
Results of this effort that lead to a PRA revision or plant fixes should be reported.
Additional guidance is provided in Hardy et al.,1989.
4.
Nonseismic Failures and Human Actions, in several selsmic PRAs, nonselsmic failures ( e.g., failures of th0 auxiliary feedwater system and failure of feed and bleed mode of core cool,ng, battery depletion, power-operated relief valve failures) and human actions (e.g., delays or failures in performing specified actions, or operator misdiagnoses a situation and tskes an improper action that is not be related to the actual, current plant situation) have been important contributors to seismically inc'uced core damage frequencies or risk indices.
Unless nonseismic failures are considered, improper deelslons may be made regarding plant modifications or procedural changes.
The licensee has the option to exoand its PRA or demonstrate that the exclusion of nonseismic failures will not sign?icantly alter the PRA results or insights. The scope of nonseismic failures and human interactions that might affect seismic sequences should be defined by the lunsee based on the internal events analyses.
5.
Uquefaction. The potential for soil liquefactico and associated effects on the plant need to be examined for some sites becausa ci specific site conditions.
The impact on plant operation should be assessed from the point of view of 13
~.-
l l
both potential for and consequences of liquefaction. Procedures for assessing soil liquefaction are described in EPRI NP-6041.
6.
HCLPF Calculation. Licensees should extract and report plant level, dominant-sequence. level, and dominant component level HCLPFs both with and without the effects of nonselsmic failure and human actions.
i j
7.
Containment Performance. Ucensees should ensure that the performance of l
containment and containment systems are addressed. Section 3.2.6 contains guidance.
3.2 Seismic Maroins Methodolooies I
This discussion deals with the use of the seismic margins methodology in the seismic iPEEE. Specifically, guidance and enhancements are provided for a licensee using either the NRC or EPRI margins methodology.
3.2.1 General Considerations lhe seismic margin methodology is considered acceptable for addressing seismic concerns in the severe accident policy implementation.
Two methodologies are currently available: one developed under NRC sponsorship and the other developed under EPRI sponsorship. The staff has determined that both methods (with the noted enhancements) will adequately address IPEEE objectives.
The two methods use different system aneJysis philosophies. The NRC method is based on an event / fault tree approach to delineate accident sequences. For example, for PWRs, two safety functions are considered to be most important to plant seismic safety: reactor subcriticality and early emergency core cooling. If these functions are ensured for a given earthquake, there is high confidence that core damage would not occur at that level The EPRI methodology is based on a systems ' success path" approach. This approach defines and evaluates the capacity c" hose components required to bring the plant to a stable condition (either hot or colo.
2n) and maintain that condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Several possible success pat,
- my exist.
Each licensee should examine its plant critically to ensure that the generic insights used in margins methodology development to identify critical functions, systems, and success path logic are applicable to its plant. This is particularly vital for older plants where systems and functions may differ greatly from the plants considered in the development of the margins methodologies (NUREG/CR-4334, NUREG/CR 5076, and EPRI NP.
l 6041).
~
Both NRC and EPRI methods should include an independent peer review to ensure proper implementation of the margins methodology. A peer review group meeting prior to a plant walkdown will provide insights into the appropriateness of the proposed plant walkdowns. Peer review group endorsement of the final results adds to the credibility of the HCLPF numbers. Review group members should have combined experience in 14 l
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i the areas of systems engineering, seismic capacity engineerinb, seismic PRAs, and
]
seismic margins methodologies, j
s 3.2.2 Review Level Earthquake and Associated Spectral Shapes i
The seismic margins methodology was designed to demonstrate sufficient margin over SSE to assure plant safety and to find any ' weak links' which might limit the plant shutdown capability to safely withstand a seismic event bigger than SSE. The seismic
]
margins method utilizes two review or screening levels geared to peak ground accelerations of 0.3g and 0.5g. It is the staff's judgement that the use of a 0.3g review 4
level earthquake (RLE) for most of the nuclear power plant sites in the Central and Eastern United States (east of the Rocky Mountains) would serve to meet the objectives of the IPEEE. However, all sites east of the Rocky Mountains are not subject to the sama level of earthquaka hazard. For some sites where the seismic hazard is low, a reduced scope margin approach centered on walkdowns is acceptable.
For two Eastern US sitt 5 whMs the staff studies indicated that the seismic hazard is relatively high and a 0.31 RL'd it, Judged not adequate. Because the component capacity data sets associated with the margins methods are categorized at two screening levels,0.3g and 0.5g, a 0.5g RLE should be used. For westem sites other than California coastal sites, a 0.5g RLE should be usef for the margin approach. The RLEs defined for U.S.
sites, as well as sites that can perform a reduced scope SMM, are presented in Tables 3.1 and 3.2.
The seismic margins evaluations should utillze the NUREG/CR-0098 median rock or soil spectrum anchored at 0.3g or 0.5g depending on the g level and primary condition at the site. Further discussion on the review level earthquake is presented in Appendix A.
The ground motion should be considered at the sudace in the free field. If secondary E
conditions such as shallow soil conditions are being considered, appropriate procedures should be used to determine the free field motion in the vicinity of those affected structures and components, and the capacity evaluation of structures and components should take into account soll structure interaction effects.
Because recent ground motion estimates such as those included in the LLNL and EPRI hazard studies indicate relatively higher ground motion at frequencies greater than 10 Hz than that shown in the NUREG/CR 0098 spectrum, the margins evaluation of only l
nonductile components (if required), for instance, relays, that are sensitive to high frequencies should be performed as discussed in Section 3.2.3(1). No plant specific response analysis is anticipated to address high-frequency ground motion concerns.
1 However, if a licensee decides to evaluate plant response for high-frequency ground motion, the response spectral shapes derived from the appropriate site specific median uniform hazard response spectra (10,000 year return period) shown in NUREG/CR-5250 anchored at 0.3 or 0.5g should be used, j
3.2.3 NRC Seismic Margins Methodology l
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1 1
Several enhancements to the NRC methodology are needed before it can be used for IPEEE implementation. These enhancements, along with the means to accomplish them, are described below:
1.
Relay Chatter. Relays, in this context, include components such as electric l
relays, contactors, and switches that are prone to chatter. The NRC method as originally developed did not address the relay chatter issue because of the lack l
of information on the subject. Hardy et al.,1989, has summarized results of
]
several efforts in this area and has provided guidance to address this issue in j
i an IPEEE context. Relay chatter analysis could be very resource intensive, and l
careful planning and use of generic insights if they are applicable to the plant are i
necessary.
l Attempts to address high frequency ground motion concerns by analysis is very j
likely to entail extensive efforts including the development of new and much more complex building models that transmit and ampilty high frequency input and l
generate accurate and meaningful floor spectra at high frequencies. Estimates of high frequency amplification in cabinets containing relays will also have to be developed. Rather than using analysis, the following alternative approach is acceptable:
)
a.
Prepare a list of relays that are known to have high frequency sensitivity, b.
Screen relays that are kreuwn to have very high r1CLPFs (that is, they can be eliminated from further consideration without performing - specific response calculations).
i c.
Assume that the remaining relays will chatter at the review level earthquake.
I d.
Screen the remaining relays by either showing that the electrical circulty is insensitive to high frequency chatter or that they can be recovered from changes of state and associated false alarms, e.
Finally, replace the remaining relays with ones that are not sensitive to high frequency (an alternative approach is to show that the remaining relays are rugged by conducting tests).
Although stated in the context of high frequency ground motion, the above approach can be used to address relay chatter issue.
2.
Uquefaction. An evaluation of the potential for liquefaction should DJ conducted, however, no guidance was provided in NRC SMM EPRI NP 6041 can be used for guidance to carry out a liquefaction analysis.
3.
Nonseismic Failures and Human Actions. These activities should be included; guidance on including nonseismic failures and human actions is provided in I
10 i..
NUREG/CR-4826 (Maine Yankee evaluation) and Budnitz,1989a (extension of the Maine Yankee, Catawba, and Hatch analyses).
4.
Plant Walkdown.
The walkdown should be performed and documented in accordance with the recommendations contained in EPRI NP-6041, 5.
HCLPF Calculations. Two approaches, fragility analysis (FA) and conservative deterministic failure margin (CDFM), for computing component and plant HCLPFs are acceptable. By developing complete fragilities for components that remain in the plant level Boolean equations if the CDFM method is chosen to calculate plant HCLPF, it is possible to make plant HCLPF statements with and without the inclusion of nonseismic failures and human actions.
As noted in EPRI NP-6041, use of the Generic Equipment Ruggedness Spectrum (GERS) to estimate HCLPFs should take into account the latest results for ongoing work on the reconciliation of GERS and HCLPF.
{
3.2.4 EPRI Seismic Margins Methodology Several enhancements to the EPRI methodology are needed before it can be used for IPEEE implementation. These enhancements along with the means to accomplish them are described below:
1.
Selection of Altemative Success Paths. The EPRI methodology as currently constituted requires evaluation of a preferred path and an alternative path. The NRC panel that reviewed the EPRI methodology recommended:
" that a reasonably complete set of potential success paths be set down initially, rather than a very small number, since limiting the number of success paths too quickly can prevent the identification of some plant-level HCLPF insights, and can mask plant differences regarding defense-in depth. The Panel believes that preliminary analysis to narrow the number of paths to the required two or three should begin with the fuller set, and it recommends that th!s narrowing be documented in detail."
For IPLdE purposes, it is desirable that, to the maximum extent possible, the 4
alternative path involve operational sequences, systems, piping runs, and l
components different from those used in the preferrt path. The procedure used in the trial application of the EPRI methodology (EPRI NP 6359) provides an acceptable approach for use in selecting success paths (pr3ferred and
. l alternative).
2.
Nonseismic Failures and Human Action. Success paths are chosen based on a screening criterion applied to nonselsmic failures and needed human actions.
It is important that the failure modes and human actions are clearly identified and i
have low enough failure probabilities to not affect the seismic margins evaluation.
The screening criteria used in the Maine Yankee margins evaluation (NUREG/CR-17 4
4
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4826) addressing both single train and multitrain systems is an acceptable approach. The redundancies along a given success path should be specifically analyzed and documented when they exist. (in a complementary sense, where a single component is truly *alone* in performing a vital function along a success path, this should be highlighted too.) This information will serve to indicate the i
extent to which a single failure would or would not invalidate the plant's ability to respond safely to a given earthquake level.
3.
Use of GERS. Although not an enhancement per se, the use of the Generic Equipment Ruggedness Spectrum (GERS) to estimate HCLPFs should take into account the latest results for ongoing work on the reconciliation of GERS and HCLPF.
4.
Relay Chatter for High Frequency Ground Motion. Guidance for addressing the high frequency ground motion issue is discussed in Section 3.2.3(1).
3.2.5 Reduced Scope Margins Method For sites where the seismic hazard is low, a reduced scope seismic margins method emphasizing the walkdown is adequate. Thorough walkdowns have demonstrated to be the most important tool for identifying seismic weak links whose correction is highly cost beneficial. Applicable sites are identified in Table 3.1.
1he initial steps of the full scope margins methodology up to and including the initial plant walkdown are performed regardless of method selected (NRC or EPRI). Basically, pertinent activities up to and including the initial plant walkdown need to be performed.
These activities include gathering system information, classifying front line systems and identifying front line components, classifying support systems and identifying support system components, and identifying plant unique features.
Further guidance on the differences between the reduced scope and full scope margins methods, that is, elements preserved and elements eliminated are provided in Appendix B.
The evaluation should be documented in a walkdown team report and subjected to a peer review.
3.2.6 Containment Performance The primary purpose of the evaluation for a seismic event is to identify sequences and vulnerabilities that involve containment, containmwat functions, and containment systems (e.g., fans, sprays, igniters, suppression pools, ice buckets) and seismic failure modes that are significantly different from those found in the IPE internal events evaluation.
Each licensee should develop a plan to address containment performance during a seismic event. Some general guidance is provided based on past PRA experience and some generic capacity estimates of typical components involved in containment 18 l
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systems.
From a survey of past PRAs (Amico,1989), it appears that high-consequence sequences involve gross structural failure of the containment itself or failure of major equipment or structures within the containment at very high accelerations (HCLPF values greater than 0.5g) and isolation failure due to seismically induced relay chatter, i
Generally, containment penetrations are seismically rugged; a rigorous fragility analysis is needed only at review levels greater than 0.3g, but a walkdown to evaluate for unusual conditions is recommended. An evaluation of the backup air system of the equipment hatch and personnel lock that employ inflatable seals should be performed at all review levels. Also, some penetrations need cooling, and the possibility and consequences of a cooling loss caused by an earthquake should be considered.
Valves involved in the containment isolation system are expected to be seismically I
rugged (NUREG/CR-4734). A walkdown to ensure that they are similar to test data and have known high capacities and that there are no spatial interaction issues will suffice.
1 Selsmic failures of actuation and control systems are more likely to cause isolation system failures and should be included in the examination. For valves relying on a backup air system,' the air system should also be included in the seismic examination.
Components of the containment heat removal / pressure suppression functional system that are not included elsewhere and are not known to have high capacities should be
)
examined. An example of such a component might be a fan cooler unit supported on l
Isolator shims. The walkdown should include examination of such components and their anchorages. Similarly, support systems and other system interaction effects (e.g.,
I relay chatter) should be examined as applicable, j
For Mark I and ice condenser containments, it is not feasible to screen out components (e.g., torus, ice bucket support) on a generic capacity basis. The potential for accident i
sequences initiated by a containment functional failure should be examined.
3.3 Ootional Methodologies A licensee may request a review of any other systematic examination method to determine its acceptability for IPEEE purposes.
19 t
4 a...,,,.,
TABLE 3.1 REVIEW LEVEL EARTMOUAKE - PLANT SITES EAST OF THE ROCKY MOUNTAINS Reduced Scooe Procram Big Rock Point
- Duane Arnold South Texas Turkey Point Comanche Peak Grand Gulf St. Lucie Waterford Crystal River River Bend
.01.9 Arkansas Dresden Maine Yankee Robinson Beaver Vetley Farley McGuire Salem Beliefonte Forml Millstone Sequoyah Braidwood Fitzpatrick Monticello Shoreham Browns Ferry I' ort Calhoun Nine Mile Point
- Summer Brunswick Ginna
- North Anna Surry Byron Haddam Neck *Oconee Susquehanna Callaway Harris Oyster Creek Three Mile Island Calvert Cliffs Hatch Palisades Vermont Yankee
- Catawba Hope Creek Peach Bottom Vogtle Clinton Indian Point Perry Watts Bar Cook Kewaunee Point Beach Wolf Creek Cooper LaSalle Prairie Island Yankee Rowe Davis Besse Limerick Quad Cities Zion 0.5 a#
Pilgrim Seabrook NOTE:
- Special attention to shallow soll condluns is appropriate for these locations (See Section 3.2.2).
- Based on the staff studies, review level earthquakes greater than 0.3g are appropriate for these two sites. Because the component capacity data sets associated with the margin methods are categorized at two screening levels,0.3g and 0.5g, the RLE for these sites is set at 0.5g.
20 yy
)
i I
TAB!,E 3.2 i
REVIEW LEVEL EARTHOUAKE - WESTERN UNITED STATES PLANT SITES i
M i
- Trojan
- Rancho Seco
- Washington Nuclear
- Palo Verde i
Seismic Marcin Methods Do Not Aoolv To Followino Sites:
r Diablo Canyon San Onofre 1
i NOTES:
Indicates a WesterrilJnited States site whose default bin is 0.5g unless the licensee can demonstrate that the site hazard is similar to those sites east of the Rocky Mountains that are found in the 0.3g bin.
Changes in the review level earthquake from 0.5g to 0.3g should be approved
[
prior to doing significant analysis.
- i 1
21
o
?
4 ACCEPTABLE METHODOLOGY FOR PERFORMING THE INTERNAL FIRES IPEEE I
For purposes of an IPEEE, a Level 1 probabilistic risk assessment (PRA) is considered acceptable to identify potentialinternal fire vulnerabilities at nuclear power plants. Some l
fire issues identified in the Fire Risk Scoping Study, (1) seismic / fire interactions, (2) effects of fire suppressants on safety equipment, and (3) control system interactions, should be addressed in the IPEEE. The walkdown procedures of the IPEEE should
(
address the above issues - and be specifically tailored to assess the potential w!nerabilities related to these issues. The licensee should use a plant specific c:sta base on fire brigade training _in the IPEEE to assess the effectiveness of manual fire fighting to determine the response time for the manual fire fighters. The licensee should also show the effectiveness of fire barriers in the IPEEE. The current fire PRA method has its limitations (NUREG/CR 5088,1989) and that significant " engineering judgment" must be brought to bear once the PRA has been accomplished to allow for sensible application of the results. The staff believes that the type of " engineering judgment" needed to interpret the results of a PHA is fully within the competence of most fire-safer / experts, including experts within the regulatory staff. Further, despite current limitations in the methodology, a fire " vulnerability search" in the spirit of the Severe Accident Policy Statement and the IPE exercise is feasible and such a vulnerability search need not wait for the completion of further methodrwv development. Finally, in meeting the objectives of the IPEEE, it is desirable to focus on relative insights rather than on absolute core damage frequency.
4.1 New Fire PRA Analvsis There are several different approaches for the analysis of fires (NUREG/CR 2300,1983, NUREG/CR 2815,1985, NUREG/CR-4840,1990, and NUREG/CR 5259,1990). _ A logic approach is described in NUREG/CR 5259. Although not all fire PRAs delineate their l
analysis steps in exactly the same way, the following steps, in one form or another, l
must be part of any analysis.
4.1.1 Identify Critical Areas of Vulnerability The crit.,rion is whethor c. f% could compromise important safety equipment. Emphasis should be placed on areas whswe n:ultiple equipment could be compromised, in pa&iar, several trains of redundant equioment to perform the same safety function.
AWtion should be given to the potential for cross zone spread of fire and the li.-U. hood that transient fuels might supplement fuels already present in a zone.
4.1.2 Calculate the Frequency of Fire Initiation in Each Area This calculation is sensitive to location within a larger area, particularly if fuel loading '
conditions, cross zone spreading potential, or other idiosyncrasies are considered.
Also, the data base on fires in various areas should be coupled with location specific information obtained from the plant walkdown and other experience to account for uncertainties.
l 22 w
m:
p
'i 4.1.3 Analyze for the Disabling' of Critica! Safety Functions Determine the likelihood of equipment being disabled by a fire. The areas to be
'l addressed include:
1.
Fire growth and spread, including the trectment of hot gases and smoke.
2.
Detection / suppression effectiveness and reliability.
l 3.'
Component fragility to fire and combustion products.
l 4.
Probability estimates (distributions) for fault tree quantification.
I 4.1.4 Identify Fire-Induced initicting Events / Systems Analysis j
Perform in a fashion similar to an internal-initiator PRA.
l 4.2 Msg.of an Existing Fire PRA The use of e." existing fire PRA to address the internal fire IPEEE is acceptable provided the PRA rome the current as built.and as-operated status of the plant-and the deficiencies of past PRAs, identified in the fire risk scoping study (NUREG/CR 5088),
are adequately addressed.
4.3 Ootional Methodologies A licensee may request a review of any other systematic examination method to determine its acceptability for IPEEE purposes.
s i
i 23 I
tx
C' a 8 i
5 ACCEPTABLE METHODOLOGY FOR PERFORMING THE HIGH WINDS, FLOODS, AND TRANSPORTATION AND NEARBY FACILITY ACCIDENT iPEEE For the purposes of an IPEEE, the staff is recommending a progressive screening apprcach to identify potential high winds, floods, and transportation and nearby facility accident vulnerabilities at nuclear power plants.
The owners of the Trojan and Washington Nuclear Plant 2, who are requested to evaluate the effects of volcanic 1
activities in assessing severe accident vulnerabilities, should determine if the recommended screening approach is applicable to their unique situation.
J 5.1 Introduction It is assumed that the IPE for internal events will be in progress or completed when the high winds, floods and transportation and nearby facility accident portion of the IPEEE is being performed. Some external events will be addressed in the internal events IPE i
analyses (e.g., the primary effect of lightning is loss of offsite power, which is included
)
in the internal events analyses); other external events will have been screened from further consideration by the staff. For those external events not in either of these categories, further consideration using the progressive screening approach shown in Figure 5.1 is recommended.
5.2 Analvtical Procedure The steps shown in Figure 5.1 represent a series of analyses in increasing level of detail, effort, and resolution. However, the licensee may choose to bypass one or more of the optional steps so long as the 1975 Standard Review Plan criteria are met or the potential vulnerabilities are either identified or demonstrated to be insignificant, in general, the containment structure, equipment hatch, personnel air lock, and other penetrations are designed and constructed to have high capacities in resisting the effects of high winds, floods, and overpressure induced by transportation or nearby facility accidents.
Therefore, no additional containment performance assessment (beyond that discussed for the seismic portion of the IPEEE in Sections 3.1.1.5,3.1.2.6, and 3.2.6) is needed unless a licensee predicts or identifies plant-unique accident sequences different from those determined by the internal events IPE, 5.2.1 Review Plant Specific Hazard Data and Ucensing Rases All licensees should review the plant design hazard information and the licensing bases, including the resolution of each event.
5.2.2 Identify Significant Changes Since OL issuance All licensees should review the site for any significant changes since the issuance of the operating license with respect to (1) military / industrial facilities within 5 miles, (2) onsite storage or other activities involving hazardous materials, (3) transportation, or (4) l developments that could affect the original design conditions.
24
--,y
c.
5.2.3 Determine if the Plant / Facilities Design Meets Current Criteria All licensees should compare the information obtained from 5.2.1 and 5.2.2 for conformance to current criteria and perform a confirmatory walkdown of the plant. The walkdown would concentrate on outdoor facilities that could be affected by high winds, onsite storage of hazardous materiais, and offsite devedpments, if the compwison indicates that the plant conforms to the current criteria (1975 NRC SRP criteria) and the walkdown reveals no potential vulnerabilities not included in the original design basis analysis, it is judged that the contribution from that hazard to core damage frequency is less than 10 per year and the IPEEE screening criterion is met.
Otherwise, one or more of steps 5.2.4, 5.2.5, and 5.2.6 should be taken to further evaluate the situation.
I 5.2.4 Determine if the Hazard Frequency is Acceptably Low (Optional Step)
If the original design basis does not meet current regulatory requirements, the licensees may choose to demonstrate that the original design basis is sufficiently low (i.e., less 4
than 10 per year) and the conditional core damage frequency is judged to be less 4
than 10,
if the original design basis hazard combined with the conditional core damage frequency is not sufficiently low, i.e., lower than the screening criterion of 10* per year, additional analysis may be needed, l
5.2.5 Perform a Bounding Analysis (Optional Step)
This analysis is intended to provide a conservative calculation showing that either the hazard would not result in core damage or the core damage frequency is below the reporting criterion. The level of detail is that 18 vel needed to demonstrate the point; I
judgment is needed for determining the proper level of detail and needed effort.
5.2.6 Perform a Probabilistic Risk Assessment (Optional Step)
A probabilistic risk assessmsnt (PRA) consists of the following key elements: hazard analysis, fragility evaluation, plant systems and accident analysis (event / fault trees), and radioactive material release analysis.
The detailed procedure is described in NUREG/CR 2300, NUREG/CR-2815, and NUREG/CR-5259. A core damage frequency 4
less than 10 per year would screen the event from further consideration. The level of detail is that level needed to conclude that the core damage frequency is low or to find vulnerabilities.
5.3 Ootional Methodoloales A licensee may request a review of any other systematic examination method to determine its acceptability for IPEEE purposes.
25 f
.-,.-m-,~.-
-m-
o.-
lgsp %
- I ';'
a t
Figure 5.1 =
ItECOMMENDED IPEEE APPROACH FOR WINDS. FLOODS. AND,0,THERS -
(1) REVIEW PLANT SPECIFIC liAZARD.
DATA AND LICENSING IIASES (FSAR)-
4
' i (2) IDENTIFY SIGNIFICANT CHANGES, I
IF ANY, SINCE.OL.ISSUANCEi (3) DOES PLANT / FACILITIES DESIGN.
~ NO-MEET CURRENT (1975 SRP) CRITERIA 4'ES (QUICK SCREENING &.WALKDOWN)-
I (4) IS THE HAZARD FREQUENCY
-YES
~
ACCEPTABLY LOW 7 NO
-.=.
I b
OR+ (5) BOUNDING ANALYSIS
-Y S (RESPONSE / CONSEQUENCE) i
(
1 NO 1
OR- (6) PRA i
h (7) DOCUMENTATION l
(TNCL, IDENTIFIED REPORTAllLE ITEMS i
AND PROPOSED IMPROVEMENTS) 1 26
(
e t
b t i~
9 e ?
t - Sg. s M*
e-
+
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i 6
COORDINATION WITH ONGOING PROGRAMS -
6.1 introduction if unnecessary duplication of effort is to be avolded, coordination with ongoing programs is necessary. The first coordination level consists of the three major elements of the severe accident policy implementation, that is, coordination of the IPEEE with the internal events IPE and accident management. The second coordination level consists of the three major elements of the IPEEE, that is, seismic events, internal fires, and high winds, floods, and others. The third coordination level consists of each major element of the IPEEE, for example, seismic events, and the ongoing programs related to that element.
6.2.Queriotion of Onacing Pr% rams 6.2.1 IPE Program Related to Internal Events in accordance with Generic Letter 88-20 (NRC,1988) each existin m..
equested to perform a systematic examination to identity any plant spec,ific voinerabilitie so severe accidents and report the results to the staff. The process was defir
's an individuai plant examination (IPE). Licensees were requested to proceed w.: the examinations for internally initiated events only (including internal flooding). Examination of externally initiated events would proceed separately and on a later schedule. However, while
_ performing the IPE for internally initiated events, licensees were advised to document i
and retain plant specific data relevant to external events so that they can be readily j
retrieved in a convenient form when needed for later external event analyses.
6.2.2 Programs Related to External Events 6.2.2.1 Seismic Programs The following is a brief description of the programs related to seismic events:
1.
USl -A-17, " System Interactions in-Nuclear Power Plants," addresses NRC's concerns regarding the interaction of various systems with regard to whether actions or consequences could adverseiy affect the redundancy and independence of safety systems. The evaluation of system interactions related to internal events and internal floods is included in the IPE (GL 88 20). The evaluation-of spatial system interaction under seismic conditions (the SSE) is included in USI A-46.
2.
USI A-40, " Seismic Design Criteria," investigates selected areas of the seismic design process. The staff identified alternative approaches to certain design procedures and modifications to the NRC criteria in the Standard Review Plan to reflect the current state of the art and industry practice. The concern for the seismic capacity of safety related above ground tanks (9t the SSE) is included in USl A 46, 27 l
l l
.+
- y.. ;
w l}.
L 3.
USl A-45, ' Shutdown Decay Heat Removal Requirements," has the' objective of determining whether the decay heat removal function at' operating plants-is adequate ~and if cost beneficial improvement could be identified. USI A-45 was 4
subsumed in the iPE (GL 88 20); therefore, the seismic adequacy of the decay.
L heat removal system should be included in the seismic events review of the IPEEE.
4.
USI A-46, ' Seismic Qualification of Equipment in Operating Plants," has developed.
f
~
an alternative method and acceptance criteria (to current licensing requirements) to verify the seismic adequacy of equipment in operating plants with construction permit applications docketed before abot' 1972. All these ints will be reviewed to the existing safe shutdown earthquake (SSE). The scopes of USl A 46 has been expanded to cover the seismic spatial system interaction of USl A 17 and the coricern of USI A-40'for the seismic capability of large safety related above.
~ 1 ground tanks.
5.
GI-131, " Potential Seismic Interaction involving the Movable in Core Flux Mapping System Used in Westinghouse. Plants,* was identified because portions of the in-co;a flux mapping system, that have not been seismically analyzed are located directly above the seal table. Failure of this equipment during a selsmic event could cause multiple failures ~at the seal table and could produce an equivalent small break LOCA.
6.
The " Charleston Earthquake issue
- came about as a result of e U.S. Geological Survey latter in 1982 that pointed out the possibility that large damaging earthquakes have some - likelihood of. occurring. at locations not formerly considered in past licensing decisions. The staff initiated the Seismic Hazard Characterization Project (LLNL), which provided probabilistic seismic hazard estimates for all nuclear power plant sites east of the Rocky Mountains. A similar -
project was carried out by EPRI for the electric utility industry. The staff's purpose in evaluating the probabilistic studies has been to ide_ntify plants in the Central and Eastem United States where past licensing decisions may have resulted in their
(
being outliers with respect to seismic hazard, that is, the likelihood of exceeding 3
their design bases.
6.2.2.2 Internal Fires Programs L
The following is a brief description of programs related to internal fires:
1.
NUREG/CR 5088, ' Fire Risk Scoping Study," identifies some fire issues that haC not previously been addressed in the fire PRAs: fire growth codef seismic / fire
-interaction, fire barrier effectiveness, manual fire fighting effectiveness, effects of fire suppressants on safety equipment, and control system interactions. A plant.
specific analysis (including a specifically tailored walkdown) should be performed to assess the actual risk impact of these issues at a plant, p
+
l L
[
28 L
m.
- - ~ - - - - -
2.
GI 57, " Effects of Fire Protection System Actuation on Safety Related Equipment,"
l.
assesses the impact of inadvertent actuation of fire protection systems on safety.
i systems. This is one of the issues identified in the Fire Risk Scoping Study.
l The' industry, through EPRI, has a program collecting ~ data on the effects of suppressants on the safety equipment. '
L l.
6.3: Acoroach on Coordination with Onacina Proarams y
L If dupHeation of effort by the staff and licensees is to be avoided, it is important that the
(~
above programs be coordinated.
6.3.1 Coordination With Internal Events Program (IPE)
The coordination between the internal' events IPE and the IPEEE can be categorized i
into three phases: preanalyses planning, plant modifications, and accident management.
6.3.1.1 Preanalyses Planning f
Considerations that would enhance' efficiency between internal event and seismically related extemal event analyses were discussed in Section 3.1.1.1. These considerations include (1) definition of elements and their boundaries, (2) walkdown procedures and spatial interactions, and (3) composition of the peer review group.- it is likely that the IPE will precede the IPEEE.
Careful. planning taking into account the above l
considerations will avoid a duplication of effort by the hcensee.
(
6.3.1.2 ' Plant Modifications Since the IPE and IPEEE are likely to be performed separately, it is imperative to examine the impact of modifications identified during the IPE on external events and vice versa. The staff examined several PRAs that included both laternal events and.
external events (Bohn,1989) to identify possible interactions. Highlights of the major i
findings in the seismic area are-q 1..
In general, the modifications proposed as a result of the intemal events analysis would not adversely affect the seismic risk provided they do not become weak.
s links.
2.
In. general, the modifications made could potentially contribute to an increase in-3 l
risk at the plant in the foUowing ways:
a.
Many of the modifications proposed may involve adding valves.or suction i
lines to existing systems. Thus the possibility of a violation of the pressure:
boundary and potentiaj diversion exists if the modification were to fail during an earthquake. Also, modifications may involve routing different trains of electried eswer or power from adjacent units. The possibility exists that the circuitry could be designed in such a way that failure of non safety related l
29
~,
. ~.
h y,
1 electrical components could actually defeat the circuitry that was desired to provide redundancy, and b.
The possibility that inadequate anchorage ; could defeat the planned redundancy.
3, The potential adverse effects of the modifications ihclude:
a.
Poor accessibility for maintenance,-
b.
_ Stiffening of systems leading to. higher stress due to thermal cycles during d
normal plant operation.
ti The cited study (Bohn,1989) provides specific examples of fixes and their impact on
-other initiating events.
6.3.1.3 Accident Management 1
Guidance on the integration of findings from the IPEEE and accident management is being developed (SECY-89-308, Oct.1989).-
6.3.2 - Coordination Among External Events Programs The issue of integration between seismic and other external events primarily involves interactions'between seismic events and fires and seismic events and floods, itis necessary to address seismically induced fires and floods'as part of the IPEEE. - The effects of seismically induced fires and the: Impact of inadvertent actuation of fire z
protsetion systems on safety systems should be addressed. The effects of seismically induced external flooding'and internal flooding on. plant safety should be included. The coordination between the seismic and the fire or flood analysts should be based on the following 1.
The seismic analysts should generally search for. tmd identify the initiating events (certain specific seismically initiated failures of.equiprnent or structures) that can cause fires or floods, and-2.
The seismic and fire or flood analysts should also discuss other. concurrent -
seismically: induced failures or possible effects on human: actions and then, proceed to complete the rest of the IPEEE analysis.
The coordination should include a meeting, prior to seismic walkdown,'in which the fire and flood analysts.should discuss the key issues, how the analysis will be done, and what to look for. The fire or flood analyst may need to participate in parts of the -
seismic walkdown or revisit the areas identified during the seismic walkdown to grasp the issues from the seismic-capacity point of view.-
6.3.3 Coordination With Seismic Programs 30 l
l
,.c-v
A number of programs related to seismic events requiring licensee action have been identified. Many of these programs have arisen as a result of the changing perception of hazards and revisions in the des ~gn and qualification criteria.
There are two categories of seismic programs as they relate to the seismic IPEEE. The first category involves programs, USl A-45, ' Shutdown Decay Heat Removal Requirements,' GI-131,
' Potential Seismic Interaction involving the Movable in Core Flux Mapping System Used in Westinghouse Plants,' and the ' Charleston Earthquake issue' that have been subsumed into the IPE/IPEEE and should therefore be specifically addressed as part of the seismic IPEEE The second category involves programs, e.g., USI A-46, " Seismic Qualification of Equipment in Operating Plants," that can be coordinated with the seismic IPEEE. The coordination of these programs with the selsmic IPEEE is most beneficial in reducing the resources spent by the licensea and staff.
6.3.3.1 - USl A-45 and GI 131 The methodology used in the seismic IPEEE can also be used to address USl A-45 and
- GI131, The systems and components for addressiry USI A-45 will have been determined by the internal events IPE, and the purpwe of the seismic IPEEE is to identify any significant and unique seismic vulnerabilities in the decay heat removal s
function. In addition, the seismic IPEEE will evaluate the potential seismic interaction of the movable in-core flux mapping system used in Westinghouse plants.
Capacities of decay heat removal components can be established vsing either the fragility analysis (FA) or conservat've deterministic failure margin (CDFM) approaches depending upon the methodology chosen to Nplement the seismic IPEEE. Thus resolution of these issues can be easily -J.complished during the seismic IPEEE evaluation.
The potential interaction between the seal table and non-Category I seismic systems associated with the movable in-core flux mapping system can be identified during the seismic walkdown of the IPEEE. If needed, the component capacities or consequences of component failure can be evaluated using the same procedures that are used in the seismic IPEEE.
a 6.3.3.2 Charleston Earthquake issue As a result of work carried out to resolve the Charleston Earthquake issue, probabilistic seismic hazard estimates exist for all nuclear power plants east of the Rocky. Mountains.
These should be used directly by any licensee in that region opting to satisfy the seismic IPEEE with a seismic PRA. The hazard estimates also played a key role in determining the review level earthquake used in the seismic margins methodology option. Therefore, the iPEEE will constitute resolution of the Charleston Earthquake issue.
6.3.3.3 USI A 46 31
= = <
=
__-m__m-_---_m
s' w'
implementation of the USl A-46~ program involves plants with construction permit applications docketed before about 1972. The USl A-46 plants thus form a subset of all the nuclear power plants in the U.S. that are requested to perform the seismic IPEEE.
The most efficient way to address the ongoing seismic programs for USl A-46 plants-is to conduct the A-46 review and.walkdown to gatherLrelevant information for the 4
seismic IPEEE. In order.to facilitate this' approach, the activities of USI A-46 and the i
seismic IPEEE need to be coordinated, and the plant walkdown needs-_to be well i
planned. Severalinherent differences between the A-46 program and the seismic IPEEE should be noted at the outset before attempting _to coordinate the two programs.
' First, the objectives are quite different. The'US! A-46 program has licensing implications l
on plant operation; this program will assess and _ ensure the seismic ruggedness of safety related equipment in a plant to withstand the SSE. The seismic IPEEE, on the other hanc', generally tries to identify plant vulnerabilities when subjected to earthquake-levels higher than the SSE design basis.
~
Second, the scope of the reviews are different, USl A-46 is concerned with only one success path (with some requirement on equipment redundancy) of equipment needed to bring the piant to safe shutdown in the event of an earthquake and ma!ntain it there for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The ' scenario considers an earthquake of the SSE level with a possible loss of offsite power because of this earthquake. The probabilities of a
,,,_, seismically induced LOCA (small or _large)-and a high-energy line break (HELB)-
occurring are judged to be low enough that their consideration at this earthquake level is not warranted. Piping, tubing, and structures will be examined during a walkdown only if they have the potential to cause seismic interaction with the equipment reviewed and cause damage to this equipment. The review of above-ground tanks (as part of USI A 40) is an exception.
The seismic IPEEE is concerned with the vulnerabilities of the whole plant, not just the-equipment. Also, evaluations are generally made at levels above the design basis. At this level of earthquake, seismica:iy induced LOCAs are considered, and mitigating systems and equipment to adtkess this initiator are reviewed. Therefore, even if the EPRI seismic margins methooology.is utilized to' implement the seismic IPEEE (since it is quite similar to the USI A 46 evaluation), it would need additional equipment to be reviewed over that required for implementing USI A-46.
Third, the levels of review and walkdown are different. The Seismic Qualification Utility Group (SQUG) and EPRI have developed a detailed Generic Implementation Procedure (GIP) for the USl A-46 review and walkdown that was reviewed by the NRC staff, and a ' Safety Evaluation Report (SER) was issued.
The GlP should be followed in i
performing the USl A 46 review and walkdown. The guidelines associated with the seismic PRA or seismic margins methodology are not as specific as those in the GIP.
To illustrate this poirit,.in the walkdown review of expansion anchor bolts, GIP calls for the use of a wrench test for the. bolt tightness check, whereas the margins walkdown ensures only that the anchor bolt; are. adequate to hold down the equipment as 32
designed with no specific testing requirements to confirm anchor capacity.
The completion of the seismic IPEEE does not automatically mean that the USl A 46 review is satisfactorily completec'.
There may be overlaps or differences in the equipment scope for USl A-46 and the seismic IPEEE. For equipment that is within the scope of USI A 46 or the seismic.
IPEEE enly, it is clear that either GIP or IPEEE guidelines, respectively, should apply.
For the overlapping equipment, the efficient approach is to use the GIP for both waixdowns; however, the IPEEE should use the review level earthquake.
In surnmary, it is recommended that licensees coordinate the information collection for the USl A-46 and seismic IPEEE review and walkdown in order to minimize or avoid duplication.of effort by the licensees and staff.
Care should be exercised.in the coordination to ensure that the requirements and objectives of both programs are fulfilled. Ctordination of the two programs has been shown to be feasible in the trial evaluation of the Hatch plant using the EPRI seismic margins methodology.
6.3.4 Coordination With Other lasues In addition to the specific USIs and '31s discussed above, if, during its IPEEE, a licensee.
(1) discovers a notable vulnerability
- hat is topically associated with any other USl or GI and proposes measures to dispoce of the specific safety issue or (2) concludes
+ hat no vulnerability exists at its plant that is topically associated with any USl or Gl,- the aff will consider the USl or GI resolved for a plant upon review and acceptcnce of the a uits of the IPEEE. The following should be discussed:
a.
The ability of the methodology to identify vulnerabilities associated with the USl or GI being addressed.-
b.
The contribution of each USI or GI to core damage frequency or unusually poor containment performance, including sources of uncertainty when PRA is used.
c.
The technical basis for resolving the issue.
l l
33 1
7.
DOCUMENTATION AND REPORTING The IPEEE should be documented in a traceable manner to provide the basis for the findings. This can be dealt with most efficiently by a two-tier approach. The first tier consists of the results of the examination, which will be reported to the NRC for review.
The second tier is the documentation of the examination itself, which should be retained by the licensee br the duration of the license unless superseded.
The information aubmitted to the NRC should be organized and presented.In accordance with Aapendix C. The submittal may enable many issues to be dealt with in the IPEEE revie'v. Pertinent issues are discussed in Section 6.
For some issues, for example, USl A 46, a detailed documentation requirement exists and that should be followed in the broad framework of IPEEE submittals. Specific information relevant to particular issues e.g., USis and Gis, should be identified.
Informat! n Submitted to the NRC 7.1 Q
A detDd list of information to be submitted to the NRC is provided in Appendix C.
7.2 Information Retained for Audit Retsined documentation should include applicable event trees and fault trees, current versions of the system notebooks (d applicable), walkdown reports, and the results of the examination. In general.. all. documents essential for a practitioner in the field to understand what was done in the IPEEE should be retained, in addition, the manner in which the validity of these documents has been ensured should be docurrented. If credit is allowed in the IPEEE for any actions taken by the operators, the licensee should have established plant procedures to be used by the plant staff responsible for managing a severe accident should one occur. Procedures should provide assurance that the operators can and will take the proper action, l
l 1
l 34 1
i
y c>-
8 REFERENCES-ACRS,1989, Memorandum from C. Daily to-R. Savio and M. Stella,-
Subject:
Assessment of Issue'Concerning Operating Reactors: Ughtning Induced Reactor-Events, dated August 2,1989.
-l AEOD,1986, AEOD Engineering Evaluation Report, "Ughtning Events at Nuclear Power Plants," April 1986.
Amico,1989, " Containment Considerations for the Use of the Seismic Margins Methods for Risk Screening," Science Applications International Corporation, Letter Report,
)
June 30,1989, i
Appen' dix R, Title 10 Code of Federal Regulations, Energy (10 CFR Parts 0 to 199),
Revised as of January 1,1989,10 CFR 50, Appendix R, " Fire Protection Program for Nuclear Power Facilities Operating Prior _ to January 1,1979," Office of the i
Federal Register, National Archives and Records Administration, Washington, D.C.
Bohn,1989, " Status Report on Issues Related to Internal /Extemal Event interaction and Decay Heat Removal Requirements for IPEs," Draft, May 1989, s
Beckjord,1987, Memorandum from Eric S. Beckjord to L.C. Shao,
Subject:
External Events Steering Group, Dated December 21,1987.:
Beckjord,1988, Memorandum from Eric S. Beckjord to Lawrence C. Shao,'
Subject:
Extemal Events Steering Group, Dated May 31,1988.
Budnitz,1989, Letter from Robert J. Budnitz to Conrad E. McCracken,
Subject:
Modification to my Memorandum of 12 July 1988, dated 3 January 1989.
Chery,1985, Memorandum from D. Chery to D. Moeller and D. Okrent,
Subject:
Hydrologic Engineering Presentation to Combined - Meeting of the ACRS-Subcommittees on Site Evaluation and Extreme Events Phenomena, dated October 9,1985.
Hardy et al.,1989, " Guidance on Relay Chatter Effects," EOE, Draft, June 1989.
Ravindra,1989, " Integration of Various Seismic issues," EOE, Draft, June 1989.
NRC,1985, " Policy Statement on Severe Reactor Accidents." Federal Register, Vol. 50,
- p. 32138, August 8,1985.
NRC,1988, " Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54(f)," Generic Letter No. 88-20, November 23,1988.
I 35
..,.s_,
i NRC,1989, " Initiation of the Individual Plant Examination for Severe Accident Vulnerabilities 10 CFR 50.54(f), Gemric Letter No. 88 20, Supplement _No.1.
August 29,1989
- SECY 88147, " Integration Plans for Closure of Severe Accident issues, May 25,1988.
EPRI NP 6041, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin,"
October 1988.
EPRI NP 6359, " Seismic Margin Assessment of the Catawba Nuclear Station," Vols 1
& 2, April 1989.
EPRI NP 6395 D, "Probabilistic Seismic Hazard Evaluation at Nuclear Plant Sites-in the Central and Eastern United States:. Resolution of the Charleston issue," April 1989.
NUREG 1032, " Evaluation of Station Blackout Accidents at Nuclear Power Plants," June 1988 NUREG-1150, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," Vols.1 & 2, June 1989..
1 NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuclear Power Plants," May 1978.
NUREG/CR 2300, "PRA Procedures Guide," Uanuary 1983.
NUREG/CR-4334, "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants," August 1985.
NUREG/CR-4482, " Recommendations to the Nuclear Regulatory Commission on Trial
' Guidelines for Seismic Margin Reviews of Nuclear Power Plants,", March 1986.
NUREG/CR 4659, " Seismic Fragility of Nulcear Power Plant Components, Phase I," Vol.
j 1, June 1986.
f NUREG/CR-4659, " Seismic Fragility of Nulcear Power Plant Components, Phase 11, l
Motor Control Center, Switchboard, Panelboard and Power Supply," Vol. 2, December 1987.
NUREG/CR-4659, " Seismic Fragility of Nutcear Power Plant Components, Phase 11, 3
Switchgear, I&C Panels (NSSS) and Relays," Vol. 3, February 1990.
NUREG/CR-4659, " Seismic Fragility of Nulcear Power Plant Components, Phase ll, j
3 Summary," Vol. 4, To be published.
a i
NUREG/CR-4734, " Seismic Testing of Typical Containment Piping Penetration Systems,"
December 1986.
36 l
i
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. ~,..
w.-
<.,_y
,.p.,
-,._.y
o.
., 1, i
! NUREG/CR-4826, " Seismic Margin Review of the Maine Yankee Atomic Power Station,"
Vols.1 - 3, March 1987.
NUREG/CR 5042, " Evaluation of External Hazards to Nuclear Power Plants in the United States," December 1987.
H NUREG/CR 5042, Suppl.1, " Evaluation of External Hazards to Nuclear Power Plants in the United States Seismic Hazard," April 1988.
NUREG/CR-5042, Suppl. 2, " Evaluation of External Hazards to Nuclear Power Plants in the United States - Other External Events," February 1989.
NUREG/CR 5076, "An Approach to the Quantification of Seismic Margins in Nuclear-Power Plants: The importance of BWR Plant Systems and Functions to' Seismic Margins," May 1988.
NUREG/CR-5088, " Fire Risk Scoping Study," January 1989.
NUREG/CR 5250, " Seismic Hazard Characterization of 69 Nuclear Power Plant Sites East of the Rocky Mountains," Vols.1 - 8, January 1989.
NUREG/CR 5259, " Individual Plant Examination for External Events: Guidance and Procedures," Draft.
NUREG/CR-5270, " Assessment of Seismic Margin Calculation Methods," March 1989.
I
-37 y
1 l
Appendix A REVIEW LFuEL EARTHOUAKE The seismic margins methodology was designed to demonstrate sufficient margin over SSE to assure plant safety and to find any " weak links" which might limit the plant shutdown capability to safely withstand a seismic event teigger than SSE.
The methodology involves the screening of components based on their importance to safety and seismic capacity. The seismic margins method utilizes two review or screening levels geared to peak ground accelerations of 0.3g and 0.5g. In areas' of low to moderate seismic hazard, most plants that have been evaluated using PRAs or margins studies have been shown to have HCLPFs at or below o.3g. Past experience indicates that, at the 0.3g screening level, a small number of " weak links" are likely to be identified, efficiently defining the cominant contributors to seismically induced core damage, it is the staff's judgment that the use of a 0.3g review level earthquake for most of the nuclear power p'. ant sites in the Central and Eastern United States (east of the Rocky Mountains) would serve to meet the objectives of the IPEEE, All sites east of the Rocky Mountains, however, are not subject to the same level of earthquake hazard. The recent studies by LLNL (NUREG/CR<5250) and EPRI (EPRI NP 6395-D) show significant differences depending on location and specific site conditions. Because the two studies do not necessarily agree with each other, it was deemed necessary to use both studies in determining which review level earthquake should be assigned to each site. Hazard comparisons were made using th.e median, 85th percentile, and mean from the site-specific results provided by the LLNL and EPRI studies. Based on the sensitivity tests and engineering and seismological judgment, the L
staff has defined the review level earthquake for each site (0.3g, 0.5g, or redus sa scope) in Table 3.1.
The sites in the Western United States (west of the Rocky Mountain Front) are treated
[
differently. Those sites in coastal California where the seismic hazard is much higher and the resulting design bases are grsater than 0.5g cannot make use of the margins methodology.
The other plant sites in the West should use a 0.5g review level earthquake unless it can be demonstrated that the seismic hazard level at a particular plant site is consistent with the seismic hazard at the 0.3g bin plant sites east of the Rocky Mountains. The results of the binning for the plants in the Western United States are presented in Table 3.2.
The rationale for the selection of the review level earthquakes (RLEs) and the grouping of the plants east of the Rocky Mountains is discussed below.
1 A.1 Introduction The specification of a review level earthquake (RLE) for use in carrying out an individual plant examination for external events (IPEEE) has been a complex problem involving the search for consistency, it would be preferable if the RLEs were completely consistent with the individual plant examination (IPE) for internal events, the IPEEE using a l
38 e
s-weiwr-9 m.
+ - -
+
=-m-w eww er v
seismic PRA, and the inherent strengths of the seismic margins methodologies, but it is very difficult to satisfy all of these elements in any rigorous quantitative sense. Thus, for example, attempting to equate the review level earthquake to the reporting criteria in the IPE (mean sequence frequency leading to core damage of 10 per year) is fraught w:'h difficulties because of the large uncertainties in numerical estimates of seismict.dy induced core damage, the inappropriateness of comparison between numer; cal estimates of seismically and internally induced core damage (the source and trea'. ment of uncertainty can be quite different), and the inherent difficulties in relating the output of a seismic margins study (HCLPF) to estimates of core damage frequency.
For some of the same reasons, it was recognized that external initiators, including earthquakes, need not necessarily be treated in the same manner as internal initiators in implementing the Severe Accident Policy, it should be noted that the RLE defines a reporting level. A HCLPF value lower than RLE does not necessarily represent a plant vulnerability. However, the licensee should assess the significance of HCLPF i
values lower than RLE and take any necessary actions and make other improvements that are deemed appropriate by the licensee, A.2 General Evaluation Procedure A.2.1 Data Evaluated The staff has recommended three review level earthquakes to be used when applying the seismic margins methodology to nuclear power plants east of the Rocky Mountains
'a for the IPEEE. The review levels or " bins" are 0.5g, 0.3g, and a reduced scope level.
The basic information used was the Lawrence Livermore National Laboratory (LLNL) hazard study (NUREG/CR 5250) and the Electric Power Research Institute (EPRI) hazard study (EPRI NP 8395 D). These studies represent state of-the-art estimates of seismic hazard. Because the two studies do not necessarily agree with each other, it was deemed necessary to use them both in determining which bin a particular site belonged in, in the LLNL study (NUREG/CR 5250), it was noted that, for some sites, the mean estimates of seismic hazard were dominated by the input of one ground motion expert (No. 5). This dominance was caused by the low attenuation, high uncertainty, and relatively high motion on rock found in this expert's input. This input has received a great deal of attention, and some have argued that it is inconsistent with the data. The staff requested LLNL (as a sensitivity study) to calculate the hazard at nuclear power plant sites east of the Rocky Mountains leaving out the input of this expert.
Data from the Saguenay Event in Quebec, Canada (November 1988), the largest earthquake in eastern North America in 50 years, appears to be quite different from previous data sets and has not helped to resolve the controversy. At this time, in order to avoid relying exclusively on the LLNL results that include the input of expert No. 5, the staff is treating the LLNL hazard estimates based on the other four ground motion experts as a separate study when binning nuclear power plant sites for IPEEE.
A.2.2 Comparison Procedure 39
,o l
l Hazard comparisons were made using the mean, median, and 85th percentile from the I
site specific results provided by' the LLNL and EPRI studies. Each of these pieces of information represents a different way of characterizing the distribution of seismic hazard i
estimates at each site as determined by a particular study.
Mean: The mean is a commonly used statistic that can be assigned actuarial significance. However, because of the skewed nature of the distribution, it is also a highly unstable (with respect to methodology and input assumptions) view of hazard. The mean is highly sensitive to the characterization of the extremes of the distribution.
Median: The median is more stable than the mean and shows the greatest agreement between the LLNL and EPRI studies. However, it is only the 50th percentile of the hazard and is insensitive to the extent of uncertainty, 85th Percentile: An alternative candidate to the mean is the 85th percentile. It reflects uncertainty in that it indicates the breadth of the distribution, but it is less sensitive to extreme outliers.
A.2.3 Weighting Criteria in the past, great emphasis has been placed on the likelihood of exceeding peak-ground acceleration (PGA), in this evaluation, site hazard comparisons were made using response spectra and 'PGA. The likelihoods of exceeding spectral response accelerations in the 2.5 to 10 Hz range were examined because these frequencies are more related to the types of motion that could cause damage at nuclear power plants.
Unit weights (2/7th each) were assigned to the likelihoods of exceeding spectral response ordinates at 2.5,' 5, and 10 Hz. One-half unit weight (1/7th) was assigned to -
the likelihood of exceeding the PGA.
A.2.4 Ranking Criteria Emphasis was placed on the relative ranking of sites with respect to other sites using the same seismic hazard study, statistic, and ground motion measures. Extensive use was made of a clustering methodology developed by LLNL for the NRC (Bernreuter et al.,1989a,1989b). For a given hazard study, statistic, ground motion measure and reference level, this methodology divides the ensemble of sites into groups so that the sites in any one group are 'close" to each other with respect to seismic hazard. For example, the sites may be divided into groups based on mean estimates of exceeding 0.5g PGA from the EPRI study or median estimates of exceeding the 2.5 Hz spectral response (associated with the NUREG/CR-0098 response spectrum anchored at 0.3g).
from the LLNL five expert study. Although there were a fixed r. umber of groups, no minimum number of sites were required in a group, and indeed some groups contained only one site.
A.2.5 Spectral Shape 40
t The spectra' shape associated with the 0.3g screening level was assumed to be the median NUREG/CR 0098 spectrum anchored at 0.3g. There has been some discussion that the scieening level should actually be associated with a somewhat higher ground motion (the Seismic Qualification Utility Group (SOUG) bounding spectrum) but in this relative comparison, the use of this alternative spectrum would make little or no l
difference.
A.3 Soecific Binnino Procedure A.3.1 Initial Binning Evaluation As the first step, sites that consistently fell into the group that had the highest likelihood of exceeding the 0.3g NUREG/CR 0098 5% damped median spectrum were conditionally assigned to the 0.5g bin. Sites that fell into the group that had the lowest likelihood of exceeding the 0.3g NUREG/CR 0098 5% damped spectrum were assigned to the reduced scope bin.
The ground motion measure compared was the weighted combination of 2.5 Hz,5 Hz, 10 Hz and PGA. The individual consistency criteria used were:
1.
Agreement among the LLNL five-expert, LLNL four expert, and EPRI studies, and 2.
Agreement between the median.and either mean.QI 85th percentile statistics,-
This resulted in a comparison of nine separate hazard gr$upings (three pieces of information for each of the three studies).
For example, if a particular site fell in the top group (0.5g bin) for all of the criteria-except the EPRI median, it remained in the 0.3g bin.' ' The. conclusions must be supported by all the hazard studies. ' On the other hand, if a particular site fell in the L
bottom group for all of the criteria except for the LLNL four-and LLNL five expert mean estimates, it was included in the reduced-scope bin. Only one measure of uncertainty, mean or 85th percentile, needs to be satisfied.
The candidates for the 0.5g and reduced scope bins were then subjected to additional evaluation by the staff.
A.3.2 Subsequent Binning Evaluations The candidates for the 0.5g bin were first examined to provide some assurance that although the hazard was relatively high, it was high enough to warrant inclusion-in this bin.
As a test, it was considered appropriate that a site belonged in the 0.5g bin if a hypothetical nuclear power plant at that site was assumed to have a HCLPF of 0.3g and the mean annual core damage frequency associated with that hypothetical plant was 10 or higher. The work cited in Ravindra 1989b showed that the mean annual core damage frequency was roughly an order of magnitude lower than the mean annual 41
.y
y
,e likelihood of exceeding the plant HCLPF and very roughly equal to the median annual likelihood of exceeding the plant HCLPF.
l Based on these estimates, the staff assumed that inclusion in the 0.5g' bin would be supported if:
1.
The mean or 85th percentile annual likelihood of exceeding the 0.3g spectrum d
from all three studies was 10 or greater, and 2.
The median annual likelihood of exceeding the 0.3g spectrum from all three 4
studies was 10 or greater.
This evaluation should be viewed as a ' sanity check"; it should not be viewed as a plant specific statement on core damage frequencies. The reasons are:
1 1.
The uncertainty and generic nature associated with the correlation in Ravindra,
- 1989b, 2.
The use of spectral estimates rather than peak ground acce!eration, 3.
The inclusion of the 85th percentile estimates, and 4.
Ah the previously mentioned problems associated with bottom line numbers.
Finally, the staff examined the candidates for the 0.5g-and reduced scope bins to assure itself that the classification made good seismological sense and there was no need to include additional sites in these bins. In conjunction with this examination, limited sensitivity tests were also carried out to determine the impact of slight relaxations in the consistency criteria.
A 4 Binnina of Sites - Results A.4.1 Reduced Scope Margins Methodology Bin The consistency criteria outlined in Section A.3.1 were slightly modified to identify sites L
for the reduced-scope bin. The two bottom median groups were included rather than the' bottom group alone. When this was done, five sites (South Texas, Comanche Peak, Waterford, River Bend, and Crystal River) were identified as belonging to the reduced scope bin.
Also added to this bin were several sites for which no EPRI calculations were available but were in the bottom groups in both the LLNL four-and five expert studies. They are Duane Arnold, Big Rock Point, Grand Gulf, St. Lucie, and Turkey Point. The-ten candidate sites in the reduced-scope bin lie in areas of low seismic hazard along or near the Gulf and Florida coasts and in the upper Midwest.
A.4.2 0.5g Bin 42
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As a result of the evaluations cited above,-'two sites (Pilgrim and Seabrook)_ were identified as belonging in toe 0.5g bin.
A.4.3 0.3g Bin All sites not identified'as belonging in the 0.5g or reduced-scope bins were assigned to the 0.3g bin.
l A.4.4 Other Considerations The grouping was.made assuming that each location was ass,ociated with one site
[
condition (rock or varying depths of soil). Some twelve plant sites east of the Rocky i
Mountains whose main Category I structures are located on rock also have some Category I structures or components located on shallow or intermediate depths of soll.
i Since shallow soil, less than about 80 feet thick, can significantly amplify ground motion, these sites should perform soil amplification studies to determine the effect.
ni particular, for four of the sites included in the 0.3g b n (on the basis of their primary
?
site conditions), the. hazard for structures or components on the secondary site conditions is equal to or higher than the. hazard associated with those plants in the 0.5g bin. Ucensees should, if site-specific analysis. indicate,' use the 0.5g screening i
tables for elements affected by soil amplification. Similarly, for one site in the reduced scope bin, site-specific analysis should be carried out to determine the effects on those.
1 elements affected by soil an5htication.
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Appendix B COMPARISON EETWEEN A REDUCED-SCOPE AND FULL SCOPE SEISMIC MARGINS EVALUATION-There are differences between the reduced scope and full-scope margins evaluation both in the extent of the systems analysis and in the amount of quantification of HCLPF values for equipment identified in the walkdown. The comparison is presented in Table B.1.
The emphasis on walkdown and not on ' quantification also applies to the performance of containment and containment systems, USl A-45 (Decay Heat Removal Requirements), and GI-131_ (In Core Flux Mapping System).
B.1 Elements Preserved -
The following elements of the seismic margins methodology must be preserved, that is, they must be identical in the reduced-scope and full scope evaluation:
l 1.
For either the NRC or EPRI methodology,.the systems engineers must perform significant pre-walkdown work that should be preserved in a reduced scope evaluation.
In the NRC methodology, this involves defining initiating events, defining event trees and the safety functions involved, and identifying systems and components necessary to carry out these functions. In the EPRI methodology, this involves defining success paths (primary and alternative) and the systems and components involved in these paths. For both methodologies, the thrust of this work is to narrow the scope and focus the effort of the key element of the review, the walkdown.
2.
For either the NRC or the EPRI methodology, the seismic capability evaluation engineers must perform significant pre-walkdown work that should be preserved in the reduced scope evaluation. In each methodology, this involves developing an understanding of the seismic input to the plant, the seismic dcsign basis, and t
realistic ground and floor response spectra.
It also involves pre-walkdown i
screening of the key systems and components identified by the systems engineers so as to make the walkdown itself most efficient. The thrust of the screening is to identify items thought to have very high HCLPF values, items suspected of l
having low HCLPF values, and_ therefore lists of items to be examined at various levels of detail during the walkdown.
3.
The reduced-scope evaluation should be identical in quality and effort to that required in the full scope margins methodology. One crucial feature-is that it should involve interactions among seismic capability evaluation engineers, systems -
engineers, and the licensee's plant operations personnel. The walkdown team should visually inspect pertinent structures, equipment, and anchorages. consistent 4
with the. full sgope NRC or EPRI methodology.
If-potentially. vulnerable components are found during the walkdown, a capacity check may be necessary using the applicable NUREG/CR-0098 ground response spectra. These results l
should be documented. Data sheets similar to those found in Appendix I of EPRI i
i-44 l
l l
l l
s NP-6041 should be used to document the walkdown. A review of construction drawings-for structural details that can not be seen. in the field should be:
performed.
4.
While the post walkdown assessment effort for a reduced scope evaluation should -
be identical in quality to that in the full scope margins methodology its thrust and '
level of effort are different because sequence-level-(NRC) or success path level-(EPRI) HCLPFs will not be computed, instead, its emphasis-should be on -
identifying possible weak-link items that may need strengthening.
B.2 ~ Reductions 1
The following, although needed in the full-scope margins methodology, are not needed in a reduced scope margins evaluation-B.2.1 NRC Seismic Margins Methodology 1.
The ystems engineers need not prepare or quantify fault trees and Boolean
- expressions representing accident sequences. Also, since fault trees will not be developed, these engineers need not combine nonseism!c failure basic events with seismically initiated failures in any rigorous fashion, although the existence of those non seismic failures, if identified, should.be noted and their importance assessed.
I in the course of the margin evaluation.
" '27 The seismic capability evaluation engineers need not develop HCLPF capacity values for all of the key equipment items that would be represented on-the i
sequence level Booleans (which will not be developed), it follows that they can not develop a plant level HCLPF capacity value.
B.2.2 EPRI Seismic Margins Methodology The seismic capab!!ity evaluation engineers need not develop HCLPF capacity.
' values for all of the key equipment items found on the success paths (primary and <
alternative) being studied, it follows that they can not develop any success path-level HCLPF capacity values that would be taken as representing the plant level HCLPF capacity.
1' 1
l 45
.~
s-1 TABif B.1 REDUCED SCOPE MARGINS METHOD BASED ON NRC SEISMIC MARGINS METHODOLOGY - (NUREG/CR 4482, Chapter 4)
STEP NO.
=,DfSCRIPTION l_N REDUCED PROGRAM?
1 Selection of Earthquake Not applicable, NRC desig-Review Level nates sites that quellfy 2
Initial Systems Review Yes, in entirety
' nitial Component HCLPF Yes, in entirety i
3 Categorization 4
First Plent Walkdown Yes, in entirety.
5 Systems Modeling Finalize Event Trees:
Yes Fault Tree Development:
No 6
Second Plant Walkdown Only as needed 7
Systems Model Development No 8
Margin Evaluation of No...,.
Components, Plant BASED ON EPRI SEISMIC MARGINS METHODOLOGY -' (EPRI NP-C041, Chapter 2)
STEP NO.
DESCRIPTION IN REDUCED PROGRAM 7 1
Selection of Seismic Margins Not applicable, NRC desig-Earthquake nates sites that qualify 2
Selection of Assessment Team
- Yes, in entirety 3
Pre Walkdown Preparation Work Yes, in entirety 4
Systems and Element Selection Yes, in entirety Walkdown l
S Seismic Capacity Walkdown Yes, in entirety l -
6 Subsequent Walkdowns Only as needed 7
Seismic Margin Assessment Work No 46 4
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Appendix C DETAILED DOCUMENTATION AND REPORTING GUIDELINES This appendix provides the guidelines for detailed documentation and reporting format and content for the IPEEE submittals.
The major perts of this appendix are the guidclines for seismic analysis (Section C.2), internal fire analysis (Section C.3), other analyses (Section C,4), specific safety features and plant improvements (Section C.1.4),
and the licensee review team (Section C.1.5). The licensees are requested to submit their IPEEE reports using the standard table of contents given in Table C.1 or provide a cross reference. This will facilitate review by the NRC and promote consistency among various submittals. The contents of the elements of this table are discussed in sections below.
The level of detail needed in the documentation should be sufficient to enable NRC to l
understand and determine the validity of all input data and calculation models used, to assess the sensitivity of the results to all key aspects of the analysis, and to audit any calculation, it is.not necessary to submit all the documentation needed for such an NRC review, but its existence should be cited and it should be available in easily usable j
form. The guideline for adequate retained documentation is that Independent expert i
analysts should be able to reproduce any portion of.the results of the calculations in a straight forward, unambiguous manner.
To the extent possible, the retained documentation should be organized along the lines identified in the areas of review.
Any information that is comparable to that provided under the IPE for :nternal events can be incorporated by reference.
O.1 General C.1.1 Conformance with Generic Letter and Supporting Material 4
Certification that an IPEEE has been completed 'and documented
.3 requested by Generic Letter 88 20, Supplement 3. - The certification should also identify the measures taken to ensure the technical adequacy of the IPEEE and the validation of results, including any uncertainty, sensitivity, and importance analyses.
3 C.1.2 General Methodology.
Provide an overview description of the methodology employed in the IPEEE for each external event.
C.1.3 Information Assembly a
Reporting guidelines include:
1.
Plant layout and containment building information not contained in the Final Safety Analysis Report (FSAR).
)
47
6 o
2.
A list of PRA or SMM studies on similar plants that the IPEEE team has reviewed along with a list of impormnt insights derived from these reviews.
3.
A conc se description of plant documentation used in the IPEEE, (e.g., the FSAR; syste.n descriptions, procedures, and licensee event reports); and a concise discussion of the process used to confirm that the IPEEE represents the as built, as-operated plant. The intent of such a confirmation is not to propose new design reverification efforts on the part of the licensees but to account for the impact of previous plant modifications or modifications conducted within the IPEEE framework.
4.
A description of the coordination activities of the IPEEE teams among the external events (e.g., for seismically induced fires).
C.1.4 Submittal of Specific Safety Features and Potential Plant improvements The licensee should provide a discussion of the criteria used to define vulnerabilities for each external event evaluated. The licensee should list any potential imorovements (including equipment changes as well as changes in maintenance, operating and emergency procedures, surveillance, staffing, and training programs) that have been selected for implementation based on the IPEEE (provide a schedule for implementation) or that have already been implementeo.
Include a discussion of anticipated benefits as well as drawbacks to any improvements.
C.1.5 IPEEE Team and Internal Review The basis for requiring the involvement of the licensee's staff in the IPEEE review is the belief that the maximum oenefit from the performance of an IPEEE would be realized if the licensee's staff were involved in all aspects of the examination and that involvement would facilitate integration of the knowledge gained from the examination into operating procedures and training programs. Thus the submittal should describe licensee siaff participation and the extent to which the licensee was involved in all aspects of the program.
The submittal should also contain a description of the internal review or peer review performed, the results of the review team's evaluation, and a list of the review team members. The maximum benefit to the licensee will occur if the combination of persons involved in the original analysis and the internal review, taken as a group, provides both a cadre of licensee personnel to facilitate the continued use of the results and the expertise in the methods to ensure that the techniques have been correctly applied.
Furthermore, an internal review provides quality control and quality assurance to the IPEEE process.
C.2 Seismic Events Section C.2.1 describes submittal information guidelines for licensees who choose the seismic PRA for seismic IPEEE, while section C.2.2 describes information guidelines for 48
.,w ji licensees who choose the seismic margin method for the seismic IPEEE. The submittal should be presented in conformance with Table C.1.
C.2.1 ' Seismic PRA Methodology I
The following information on the seismic IPEEE is the millmum that should be documented and submitted to the NRC:
1.
A description of the methodology and key assumptions used in performing the l
seismic IPEEE.
i 2.
The hazard curves (or table of hazard values) used; also, the associated spectral shape used in the analysis. Also, if an upper bound cutoff to ground motion of less than 1.5g peak. ground acceleration is assumed, the results of sensitivity studies to determine whether the cutoff affected the delineation and ranking of j
seismic sequences.
3.
A summary of the walkdown findings and a concise description-of the walkdown 3
team and the procedures used.
4.
All functional / systemic seismic event trees as well as data (including _ origin and method of analysis). Address to what extent the recommended enhancements l
have been incorporated in the IPEEE. A description of how.nonseismic failures, human actions, dependencies, relay chatter, soil liquefaction, and seismically induced floods / fires are accounted for.
Also, a list of important nonseismic' failures with a rationale for the assumed failure rate given a seismic event, i
5.
A description of dominant functional / systemic sequences leading to core damage 4
along with their frequencies and percentage contribution to overall seismic core i
damage frequencies (for both LLNL and FPRI hazard curves if available for the site). Sequence selection criteria are
,ced in GL 88 20 and NUREG-1335.
The description of-the sequences
?,ould include a ^ discussion of specific assumptions and human recovery acuon.
i 6.
The estimated core damage frequen.y (for both the LLNL and EPRI' hazard curves,-if available for the site) and plant damage state frequencies, the timing of the core damage, including a qualitative discussion of uncertainties and how they might affect the final results, and contributions of different ground motions to core damage frequencies.
7.
Any seismically induced containment failures and other containment performance insights. Particularly, vulnerabilities found in the following three systems / functions:
penetrations, isolation, and containment heat removal / pressure suppression. Early containment failures that-might result in high-consequence sequences or may.
initiate accident sequences. Also, computed fragilities and HCLPFs of containment components.
49 j..
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8.
A table of fra@ities, both generic and plant specific, used for screening as wel' as in the quannfication. The estir eted HCLPF for the plant, dominant sequences, and components with and without nonseismic fa. lures and human actions.
9.
Documentation with regard to other seismic issues (Section 6) addressed by the submittal, the basis and assumptions used to address these issues, and a discussion of the findings and conclusions.
Evaluation results and potential improvements associated with the decay haat removal function and movable in-core flux mapping system (for Westinghouse plants) should be specifically highlighted.
- 10. When an existing PRA is used to address the seismic IPEEE, the licensee should describe sensitivity studies related to the use of *.he initial hazard curves, supplemental plant walkdown results and subsequent evaluations, and relay.
i chatter evaluations. The licensee should examine the above list to fill in those items missed in the existing seismic PRA (See Section 3.1.2).
O.2.2 Seismic Margins Methodology l
The following Infctmation on the seismic IPEEE is the minimum that should be documented and submitted to the NRC for a full scope SMM review:
1.
A description of the methodology and a list of important assumptions, including their basis, used in performing the seismic IPEEE. Address the extent to which the following were taken into account: nonseismic failures, human actions, dependencies, relay chatter, soll liquefaction, and seismically induced floods / fires.
Also, a list of important nonselsmic failures with a rationale for the assumed failure rate given a seismic event.
2.
A summary of the walkdown findings and a concise description of the walkdown team and proc?oures used.
3.
All functbnal/ systemic seismic event trees data (including origin and method of ent ysis) when NRC SMM is used.
l 4.
A description of the most ir6portant sequences and important minimal cutsets (for both seismic a,1d nonseismic failures) leading to core damage (NRC method) or a description of the success paths and procedures used for their selection and of each component in the controlling success path (EPRI method).
5.
Any seismically induced containment feltures or vulnerabilities and other containment pedormance insights. The procedure for defining the scope of the containment pedormance review. Results of isolation system, penetration, and heat removal system reviews.
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A table of fragilities, both generic and plant specific, used for screening as well as in the quantification. The estimated HCLPF for the plant, dominant sequences, and components with and without nonseismic failures and human actions.
7.
Documentation with regard to other seismic issues (Section 6) addressed by the submittal, the basis and assumptions used to address these issues, and a discussion of the findings and conclusions.
Evaluation results and potential improvements associated with the decay heat removal function and movable in-core flux mapping system (for Westinghouse plants) shou'd be specifically q
highlighted.
The following is the minimum information that should be documented and submitted to the NRC for a reduced scope SMM review:
t 1.
A description of the procedures used to identify sys'. ems and components for the walkdown in performing the seismic IPEEE.
2.
A summary of the walkdown f;ndings and a concise description of the walkdown team and procedures used.
3.
A discussion and the results of any specific component capacity evaluations performed, the methods used, and assumptions.
i 4.
Documentation with regard to other seismic issues (Section 6) addressed by the submlital, the basis and assumptions used to address these issues, and a-discussion of the findings and conclusions.
Evaluation results and potential improvements associated with the decay heat removal function and movable in-core flux mapping system (for Westinghouse plants) should be specifically highlighted.
C.3 lo'sgal Fires The in'onnotion on the intemal fires IPEEE Identified below is the minimum that should be docu nented and submitted to the NRC.
1.
A description of the methodology used in performing the fire IPEEE and a discussion of the status of Appendix R modifications.
2.
A summary of the walkdown findings and a concise description of the walkdown team and the procedures usM. This should include a description of the efforts i
to ensure that cable routing Jsed in the analysis represents as built information and the treatment of any existing dependence between remote shutdown and control room circuitry.
i 1
i 3.
A discussion of the criteria used to identify critical fire areas and a list of critical j
areas, including (a) single area in which equipment failures repiesent a serious 51 w
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sharing common barriers, penetration seals, HVAC ducting, etc.
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4.
A discussion of the criteria used for fire size and duration and the treatmJnt of cross zone fire spread and associated major assumptions, j
5.
A discussion of the fire initiation data base, including the plant specific data base used. Provide documentation in each case where the plant specN data used is less conservative than the data base used in the approved iire vulnerability methodologies. Describe data handling method, including major assumptions, the i,
role of expert judgment, and the identification and evaluation of sources of data i
uncertainties.
l 6.
A discussion of the treatment of fire growth and spread, the spread of hot gases uid smoke, and the analysis of detection and suppression and their associated escumptior =, including the treatment of suppressioninduced damage to equipment.
7.
A discussion of fire damage modeling, including the definition of fireinduced failures related to fire barriers and control systems.and fire induced damage to cabinets. A discussion of how human intervention is treated and how fire-induced and non fire-induced failures are combined. Identify rt.covery actions and types of firo mitigating actions taken credit for in these sequences.
8.
Event trees and associated fault trees for fire-initiated sequences. Discuss the treatment of manual suppression, including fire fighting procedures, fire brigade l
training and adequacy of existing fire brigade equipment, and treatment of access routes versus existing barriers.
9.
The estimated core damage frequency, the timing of the associated core damage, a list of analytical assumptions including their bases, and the sources of uncertainties, if applicable.
C 4 High Winds. Floods. and Others The following information on the high winds, floods, and others portion of the IPEEE is the minimum that should be documented and submitted to the NRC:
1.
A description of the methodologies used h the examination.
2.
Information on plant specific hazard data and licensing bases.
3.
Identified s!gnificant changes (See Section 5.2 2), if any, since OL issuance with respect to high winds, floods, and other extemal events.
l 4.
Results of plant / facility design review to determine their robustness in ielation to NRC's 1975 SRP criteria.
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hrsuh of the assessment of the hazard frequency and the associated conditional core damage frequency if step 4 of Figure 5.1 is used.
6.
Results of the bounding analysis if step 5 of Figure 5.1 is used.
7.
A1 functional event trees, including origin and method of analysis (PRA only).
6.
A description of each functional sequence selected, including discussion of specific assumptions and human recovery action (PRA only).
g.
The estimated core damage frequency, the timing of the associated core damage, a list of analytical assumptbns including their bases, and the sources of j
uncertainties, if applicable (PRA only),
i
- 10. A certification that no other plant unique external event is known that poses any j
significant threat of severe accident within the context of the screening approach for 'High Winds, Floods, and Others,"
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I Table C.1 STANDARD TABLE OF CONTENT FOR IPEEE SUBMITTAL i
1.
Executive Summary 1.1 Background and Objectives 1.2 Plant Familiarization 1.3 Overall Methodology 1.4 Summary of Major Findings l
2.
Examination Description i
i 2.1 Introduction 2.2 Conformance with Generic Letter and Supporting Material 2.3 General Methodology 2.4 Information Assembly.
3.
Selsmic Analysis 1
3.0 Methodology Selection (PRA or SMM) 3.1e Seismic PRA 3.1.1 Hazard Analysis I
~4.1.2 Fieview of Plant Information and Walkdown l
3.1.3 Analysis of Plant System and Structure Response l
i 3.1.4 Evaluation of Component Fragilities and Fallure Modes 3.1.5 Analysis of Plant Systems and Sequences 3.1.6 Analysis of Containment Performance 3.1b Seismic Margins Method (SMM) (NRC, EPRI, or Reduced SMM) 3.1.1 Review of Plant Information, Screening, and Walkdown 3.1.2 System Analysis l
3.1.3 Analysis of Structure Response l
3.1.4 Evaluation of Selsmic Capacities of Components and Plant 3.1.5 Analysis of Containment Performance 1
3.2 USl A-45, GI-131, and Other Seismic Safety issues 4.
Intemal Fires Analysis i
4.1 Fire Hazard Analysis 4.2 Fire Growth and Propagation 4.3 Fire Detection and Suppression 4.4 Analysis of Plant Response 5.
High Winds, Floods, and Others 5.1 High Winds 54 4
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5.3 Transportation and Nearby Facility Accidents 5.4 Others 6.
Ucensee Participation and Internal Review Team 6.1 IPEEE Program Organization l
6.2 Composition of Independent Review Team V
6.3 Areas of Review and Major Comments j
6.4 Resolution of Comments i
7.
Plant improvements and Unique Safety Features 8.
Summary and Conclusions (including proposed resolution of USIs and Gis) r
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