United States Nuclear Regulatory Commission - Protecting People and the Environment

Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10CFR 50.54(f) (Generic Letter No. 88-20, Supplement 4)


                                   June 28, 1991


To All Licensees Holding Operating Licenses and Construction Permits 
for Nuclear Power Reactor Facilities

SUBJECT:  INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS 
          (IPEEE) FOR SEVERE ACCIDENT VULNERABILITIES - 10CFR 50.54(f) 
          (Generic Letter No. 88-20, Supplement 4)


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 recognizes, 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 which 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.
  
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 the 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, (3) coordinate with other ongoing external event programs, 
and (4) conduct a workshop to explain the IPEEE process and to obtain 
comments and questions on the draft generic letter supplement and 
associated guidance document.  The staff has completed this work and 
has revised this supplement and the guidance document (Ref. 3) and is 
now requesting that each licensee perform an individual plant 
examination of external events to identify vulnerabilities, if any, to 
severe accidents and report the results together with any 
licensee-determined improvements and corrective actions 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 


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appreciation of severe accident behavior, (2) to understand the most 
likely severe accident sequences that could occur at its plant under 
full power operating conditions, (3) to gain a qualitative 
understanding of the overall likelihood of core damage and radioactive 
material release, and (4) if necessary, to reduce the overall 
likelihood of core damage and radioactive material releases by 
modifying hardware and procedures that would help prevent or mitigate 
severe accidents.  It must be emphasized that for the IPEEE the key 
outcome is the knowledge and appropriate improvements resulting from 
such an examination which can be conducted using any of the approaches 
discussed below or an alternate approach, if acceptable to the NRC.  
Besides the completion of the IPEEE, closure of severe accident 
concerns involves the completion of the internal event IPE, including 
applicable items resulting from the Containment Performance Improvement 
(CPI) Program, and 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), licensees are requested to 
perform an IPEEE for plant-specific severe accident vulnerabilities 
initiated by external events and to submit the results to the NRC.  
NUREG-1407, which is enclosed, provides additional guidance for the 
performance and submittal of the IPEEE.  (It is not the intent of 
NUREG-1407 to go beyond the information request contained in this 
generic letter supplement.)
 
2.    Examination Process

The examination process for the IPEEE, in general, is similar to that 
for the 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, training 
     programs, and appropriate hardware changes. 

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                                   3

Therefore, each licensee is requested to use its staff to the maximum 
extent possible in conducting the IPEEE, by participating in the 
analysis and technical review, and by validating both the process and 
its results through a peer review by individuals who are not associated 
with the initial evaluation.

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 associated with 
normal and emergency operation situations.  A comprehensive list of 
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 accident.  Some external events may have been 
considered at the design stage and have sufficiently low contribution 
to core damage frequency or plant risk.  Some events may have been or 
will be reviewed under ongoing programs; for instance under IPE, the 
significance of lightning and severe cold weather conditions that could 
cause loss of offsite power will be assessed.  Also, internal floods 
have been included in the internal event IPE request (Ref. 2).  Based 
on staff's evaluation of References 6 through 8, the staff recommends 
that only five events be included in the IPEEE.  However, licensees 
should confirm that no plant-unique external events known to the 
licensee with the potential to initiate severe accidents are excluded 
from the IPEEE.  For example, 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, based on past operating experience, 
lightning strikes may fail equipment in addition to causing partial or 
complete loss of offsite power, (i.e., affecting safety-related 
instrumentation and control systems).  The five external events 
requested to be assessed include:
  
1.   Seismic Events 
2.   Internal Fires 
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 through 8.

4.   Examination Methods

The NRC has identified the following approaches (details are provided 
in NUREG-1407) as being acceptable for the examination requested by 
this letter.  However, the NRC recognizes that other methods capable of 
identifying plant-specific vulnerabilities to severe accidents due to 
external events may exist.  The staff will review any systematic 
examination methods proposed to 
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determine their acceptability for IPEEE.  A brief discussion of the 
staff identified approaches is provided below:

4.1  Seismic Events.  A seismic IPEEE can be accomplished by performing 
     a seismic probabilistic risk assessment (PRA) with enhancements or 
     by using one of two seismic margin methods with enhancements.  

     The seismic PRA should be at least a Level 1 plus a containment 
     performance analysis that uses current methods and plant 
     information.  Containment performance analysis guidance is 
     provided in Appendix 2.   The containment performance analysis 
     should concentrate on identifying seismically induced 
     vulnerabilities and sequences different from those obtained from 
     the IPE.  The staff considers the procedures described in 
     NUREG/CR-2300 (Ref. 5), NUREG/CR-2815 (Ref. 9), and NUREG/CR-4840 
     (Ref. 10) to be adequate for the seismic IPEEE, provided the 
     enhancements discussed in Appendix 1 of this generic letter are 
     also included.  The staff prefers that licensees use both mean 
     (arithmetic) hazard curves (Refs. 11 and 12) developed by the 
     Lawrence Livermore National Laboratory (LLNL) and the Electric 
     Power Research Institute (EPRI), if available, in performing the 
     PRA, since this will help to focus on the delineation of dominant 
     sequences rather than on the bottom line numbers.  If a licensee 
     chooses to perform only one analysis, then the higher of the two 
     mean (arithmetic) hazard estimates should be used.

     Two seismic margins methods (SMMs) with enhancements, one 
     developed by NRC and the other developed by EPRI, can also be used 
     for the seismic IPEEE.  However, the SMMs in their current form 
     are not suitable for plant sites located in areas of high 
     seismicity.  For the remaining sites, a graded review approach 
     (full scope, focused scope, and reduced scope) is defined (see 
     NUREG-1407).  The lists of review level earthquakes (RLEs) and 
     review scope defined by the staff for all U.S. sites, and for use 
     in SMMs, are presented in Appendix 3.  The RLE does not represent 
     a safety adequacy criterion or a threshold of vulnerability for 
     the individual plant.  The RLE is intended as a reporting 
     criterion if the plant capacity is lower than the specific RLE.  
     Detailed descriptions of the seismic margins methods can be found 
     in NUREG/CR-4334 (Ref. 13), NUREG/CR-4482 (Ref. 14), NUREG/CR-5076 
     (Ref. 15), and EPRI NP-6041 (Ref. 16).  The requested enhancements 
     are discussed in NUREG-1407 and summarized in Appendix 1 to this 
     generic letter.

4.2  Internal Fires.  Fire initiated events can be treated by 
     performing a Level 1 fire PRA as described in NUREG/CR-2300 or a 
     simplified fire PRA as described in NUREG/CR-4840 (Ref. 10).  The 
     COMPBRN code can be used to model fire 
     .     

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     propagation, provided that the shortcomings identified in Ref. 17 
     are addressed.  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. 17): (1) seismic/fire interactions, (2) effects of 
     fire suppressants on safety equipment, and (3) control system 
     interactions for severe accident vulnerabilities.  Containment 
     performance (Appendix 2) should be assessed to determine if 
     vulnerabilities stemming from sequences that involve containment 
     failure modes distinctly different from those obtained in the 
     internal event analyses are predicted.  
     
     An alternative fire vulnerability evaluation (FIVE) method is 
     under review by the staff at this time, and may become a viable 
     option for the treatment of fire in the IPEEE.

4.3  High Winds, Floods, and Transportation and Nearby Facility 
     Accidents.  A screening type approach as shown in Figure 1 can be 
     used to evaluate the impact of high winds, external floods, and 
     transportation and nearby facility 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 Plan (SRP) criteria (Ref. 
     18) are met.  If the 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, provided that vulnerabilities are 
     either identified or proved to be insignificant.  Again, the 
     containment performance should be assessed to determine if 
     vulnerabilities and sequences different from those obtained from 
     the internal event analyses are predicted.  

The application of the above approaches involves considerable judgment 
with regards to the requested scope and depth of the study, level of 
analytical sophistication, and level of effort to be expended.  This 
judgment depends on how important the external initiators are likely to 
be compared with internal initiators, and a perceived need for 
accurately characterizing plant capacity or core damage frequency.  The 
detailed guidelines presented in NUREG-1407 do not preclude use of this 
type of judgment.  Consistent with engineering practice, expert 
opinions, simplified scoping studies, and bounding analyses (which 
should be documented), are expected to be used, as appropriate, in 
forming these judgments.  At sites that have multiple units, some 
utilities may wish to reduce their review scope after completing the 
initial IPEEE plant evaluation.  The licensee should discuss 
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any proposed reduction in the scope of the IPEEE with the NRC on a 
case-by-case basis.

5.   Coordination with Other External Event Programs

Three programs, i.e., (1) the external event portion of USI A-45, (2) 
GI-131, and (3) the Eastern U.S. Seismicity Issue (formerly called the 
Charleston Earthquake Issue), are subsumed in the IPEEE.  A brief 
discussion of these programs is provided below:

-    USI A-45, "Shutdown Decay Heat Removal Requirements":  USI A-45 
     had the objective of determining whether the decay heat 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 adequacy of the decay heat removal 
     system (DHR) to deal with external events initiators.  This aspect 
     of the DHR issue should be specifically addressed in the review of 
     the IPEEE.  The external event insights obtained from the USI A-45 
     study on five plants are presented in GL 88-20 (Ref. 2).

-    GI 131, "Potential Seismic Interaction Involving the Movable 
     In-Core Flux Mapping System Used in Westinghouse Plants": GI 131 
     (Ref. 19) deals with the seismically induced failure of the flux 
     mapping transfer cart that would lead 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.

-    The Eastern U.S. Seismicity (The Charleston Earthquake) Issue:  As 
     a result of work carried out by the NRC, LLNL, and EPRI to resolve 
     the Charleston Earthquake Issue, probabilistic seismic hazard 
     estimates (Refs. 11 & 12) exist for all nuclear power plant sites 
     east of the Rocky Mountains.  These estimates can be used directly 
     by any licensee opting to satisfy the seismic IPEEE by means of a 
     seismic PRA.  The NRC/LLNL and EPRI work in this area also played 
     a key role in determining the review level earthquakes to be used 
     in the seismic margin option.  The IPEEE will provide a resolution 
     of the Eastern U.S. Seismicity issue without the need for 
     utilities to perform any additional work.

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 with the IPEEE 
to minimize unnecessary duplication of examination and review efforts.    
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                                   7

-    USI A-17, "System Interactions in Nuclear Power Plants," USI A-40, 
     "Seismic Design Criteria, A Short-Term Program," and USI A-46 
     "Verification of Seismic Adequacy of Equipment in Operating 
     Plants,":  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 operating plants, which 
     constitute a subset of all the nuclear power plants that are 
     expected to perform an IPEEE.  USI A-46 should be coordinated with 
     the IPEEE so that the objectives of both activities may be 
     accomplished 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.)

-    NUREG/CR-5088, "Fire Risk Scoping Study" and GI 57, "Effects of 
     Fire Protection System Actuation on Safety-Related Equipment":  
     The licensee should address the fire issues identified in the Fire 
     Risk Scoping Study (Ref. 17) as discussed in Section 4.2 in 
     NUREG-1407.  However, it should be noted that additional research 
     related to GI 57 is being performed in parallel with the IPEEE to 
     obtain more rigorous and realistic estimates of risk; this 
     research may identify other potential vulnerabilities.  A 
     specifically tailored walkdown for potential fire vulnerabilities 
     should enable the licensee to collect information related to GI 
     57.  Licensees may propose corrective measures that could resolve 
     some or all of the GI 57 concerns.

If, during its IPEEE, a licensee (1) discovers a potential 
vulnerability that is topically associated with any other USI or GI and 
proposes 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 IPEEE submittal should specifically identify 
which USIs or GIs it is proposing to resolve.

6.    Severe Accident Sequence Selection

In performing an IPEEE using a PRA, it is essential to screen for 
potentially important severe accident sequences.  The screening 
criteria that should be used to determine which of the potentially 
important sequences that lead to core damage or unusually poor 
containment performance, should be reported to the NRC with your IPEEE 
results, are listed in Appendix 3 of this generic letter.  

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                                   8

If a seismic margin 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-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.  The licensee should 
assess the significance of HCLPF values lower than the RLE and take any 
actions that are deemed appropriate.

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

It is expected that the licensee will move expeditiously to correct any 
vulnerabilities that it determines warrant correction.  Information on 
changes initiated by the licensee should be documented in accordance 
with the requirements of 10 CFR 50.59 and 10 CFR 50.90.  Changes should 
also be reported in the IPEEE submittal (including reference to any 
previous submittal under 10 CFR 50.59 or 10 CFR 50.90) in response to 
this letter.

NRC

The NRC will evaluate licensee IPEEE submittals and will serve as a 
clearing house to disseminate all important IPEEE findings.  These 
evaluations are intended to obtain reasonable assurance that the 
licensee has adequately analyzed the plant design and operations to 
discover instances of particular vulnerability to core damage or 
unusually poor containment performance given a core damage accident.  
Further, the NRC will assess whether the conclusions the licensee draws 
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:

1.   If NRC consideration of all pertinent and relevant factors 
     indicates that the plant design or operation does not meet the 
     facility's current licensing basis, then appropriate actions will 
     be required consistent with the Commission's rules and 
     regulations.

2.   If NRC consideration indicates that plant design or operation 
     could be enhanced by substantial additional protection beyond NRC 
     regulations, appropriate enhancement 
     .     

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     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 are not cost effective, enhancements would not be 
     required.

8.   Accident Management

Licensees need not develop an accident management plan as an integrated 
part of the IPEEE.  Licensees should plan to incorporate the results of 
the IPEEE and other relevant information into their accident management 
plans at a future date.  Nevertheless, the IPEEE process may identify 
operator or other plant personnel actions that can substantially reduce 
the risk from severe accidents at the plant and that the licensee 
believes should be immediately implemented in the form of emergency 
operating procedures or similar formal guidance.  The staff encourages 
each licensee to not defer implementing such actions until a more 
structured and comprehensive accident management program is developed 
on a longer schedule, but rather to implement such actions within the 
constraints of 10 CFR 50.59.  These actions can be integrated later 
into the plant's accident management program.

9.   Documentation of Examination 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.  A summary of the 
documentation format and content is provided in Appendix 4 of this 
generic letter.

10.  Licensee Response

Licensees are requested to submit within 180 days from the issuance 
date of this generic letter a response which describe their proposed 
programs for completing the IPEEEs.  The proposal should:

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, and 
     submitting the results to the NRC.

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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 issuance date of this generic letter (Supplement 4 to Generic 
Letter 88-20).  The NRC encourages those plants that have not yet 
undergone any systematic examination for severe accidents to promptly 
initiate the examination.

11.  Regulatory Basis

This letter is issued pursuant to Section 182a of the Atomic Energy Act 
and 10 CFR 50.54(f).  A 10 CFR 50.54(f) analysis is provided in the 
Appendix 5.  Accordingly, all responses should be under oath or 
affirmation.  This request for information is covered by the Office of 
Management and Budget under an Interim Clearance No. 3150-0011, which 
expires on June 30, 1991.  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.  A 
value/impact analysis for the implementation of the IPEEE is provided 
in the attachment to Appendix D of NUREG-1407.  Comments on burden and 
duplication may be directed to the Office of Management and Budget, 
Reports Management, Room 3208, New Executive Office Building, 
Washington, DC 20503.




                    James G. Partlow, Associate
                      Director for Projects
                    Office of Nuclear Reactor Regulation


Enclosures:
1.  Appendices 1 through 6
2.  NUREG-1407
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                               Figure 1
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                              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 PRA:       Perform a plant walkdown following the procedures 
               described in the EPRI seismic margin report (Ref. 16). 

               Perform an assessment of relay chatter effects in 
               accordance with scope and procedure described in 
               NUREG-1407.

               Perform soil analysis, if needed, using procedures 
               described in NUREG-1407.

               Calculate the high confidence of low probability of 
               failure (HCLPF) values for components, sequences, and 
               the plant (optional).

Existing PRA:  Include the enhancements noted above for new PRA 
               and add the following if not considered previously:  

               Perform sensitivity studies to determine if the use of 
               LLNL or EPRI mean hazard estimates would affect the 
               delineation and ranking of sequences.

               Perform a supplementary analysis of nonseismic failures 
               and human actions.

               Perform containment performance assessment.   
          
NRC SMM:       Perform an assessment of relay chatter effects in 
               accordance with scope and procedures described in 
               NUREG-1407.

               Perform soil analysis, if needed, using procedures 
               described in NUREG-1407.

               Perform an analysis of nonseismic failures and human 
               actions using procedures described in NUREG-1407.

               Perform a walkdown and prepare its documentation in 
               accordance with EPRI's recommendations (Ref. 16).

.

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               Evaluate containment and containment system performance.

EPRI SMM:      Select an alternative path so that it involves to the 
               maximum extent possible systems, piping runs, and 
               components that are different from the preferred success 
               path.

               Perform an analysis of nonseismic failures and human 
               actions using procedures described in NUREG-1407.

               Evaluate containment and containment systems 
               performance.

               Perform an assessment of relay chatter effects in 
               accordance with the scope and procedures described in 
               NUREG-1407.
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                              APPENDIX 2
                        CONTAINMENT PERFORMANCE

The protection of public safety from any hazard 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 
one of these barriers.  

The evaluation of the containment performance for external events 
should be directed toward a systematic examination of whether 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 functional failure of the 
containment (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 containment 
     systems under the external hazard to see if they are different 
     from those 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 20 & 21 and determine their 
applicability to external events.   
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                              APPENDIX 3
      CRITERIA FOR REPORTING IMPORTANT SEVERE ACCIDENT SEQUENCES

The licensee should use the reporting criteria described in 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 described in NUREG-1335 (Ref. 22) 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 NUREG-1407 all functional sequences and 
success paths considered in the analysis and their HCLPFs.  The review 
level earthquakes (RLEs) for all applicable U.S. sites are 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 review level earthquake (RLE) does not 
necessarily represent a plant vulnerability.  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 licensee. 
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                               TABLE 3.1
                  
   REVIEW LEVEL EARTHQUAKE - PLANT SITES EAST OF THE ROCKY MOUNTAINS
Reduced Scope  
     
     Big Rock Point Duane Arnold*  South Texas    Turkey Pt.
     Comanche Peak  Grand Gulf     St. Lucie      Waterford
     Crystal River  River Bend
               
0.3g Focused Scope   

     Arkansas #2         Dresden        Limerick       Quad Cities
     Beaver Valley       Farley         McGuire        Salem
     Bellefonte          Fermi          Millstone      Shoreham
     Braidwood           Fitzpatrick    Monticello     Summer* 
     Browns Ferry        Fort Calhoun   Nine Mile Pt.  Surry
     Brunswick           Ginna          North Anna*    Susquehanna
     Byron               Haddam Neck    Oyster Creek   Three Mile Is. 
     Callaway            Harris         Palisades      Vermont Yankee
     Calvert Cliffs      Hatch          Peach Bottom   Vogtle
     Catawba*            Hope Creek     Perry          Watts Bar
     Clinton             Kewaunee       Point Beach    Wolf Creek
     Cook                LaSalle        Prairie Island Zion
     Cooper
     Davis-Besse    

0.3g Full Scope

     Arkansas #1         Maine Yankee   Robinson       Yankee Rowe
     Indian Point        Oconee*        Sequoyah
 
     
Committed to Perform a Seismic PRA**

     Pilgrim             Seabrook


NOTES: 

*    Special attention to shallow soil conditions is appropriate for 
     these locations (see NUREG-1407, Section 3.2.2 and Appendix A).

**   Relay chatter evaluation should be similar to a full-scope review.
.

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                               TABLE 3.2
                  
      REVIEW LEVEL EARTHQUAKE - WESTERN UNITED STATES PLANT SITES



0.5g*
   
     Trojan                   Rancho Seco
     Washington Nuclear       Palo Verde  


Seismic Margin Methods Do Not Apply To the Following Sites:

     Diablo Canyon            San Onofre



NOTES:

*    Indicates a Western United 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.
.

                                   18

                              APPENDIX 4 
                             DOCUMENTATION

This appendix provides the guidelines for documentation and reporting 
format and content for the IPEEE submittal.  The major parts of this 
appendix are the guidelines for seismic analysis (Section 4.2), 
internal fire analysis (Section 4.3), other analyses (Section 4.4).  
Licensees are requested to submit their IPEEE reports using the 
standard table of contents given in Table C.1 of NUREG-1407 or provide 
a cross reference.  This will facilitate review by the NRC and promote 
consistency among various submittal.  The contents of the elements of 
this table are discussed further below.

The level of detail needed in the documentation should be sufficient to 
enable the 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.  Relevant documentation should be cited in the IPEEE submittal, 
and be available in easily retrievable form.  The guideline for judging 
the adequacy of retained documentation is that independent expert 
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 internal events can be 
incorporated by reference.

4.1  General 

4.1.1          Conformance with Generic Letter and Supporting Material 
          
Certification should be provided that an IPEEE has been completed and 
documented as requested.  The certification should also identify the 
measures taken to ensure the technical adequacy of the IPEEE and the 
validation of results.

4.1.2          General Methodology

An overview description of the methodology employed in the IPEEE for 
each external event examined should be provided.

4.1.3          Information Assembly

Reporting guidelines include:

1.   Plant layout and containment building information not contained in 
     the Final Safety Analysis Report (FSAR).

.

                                   19

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 part of the licensees but to account 
     for the impact of previous plant modifications or modifications 
     conducted 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).

4.1.4          Submittal of Vulnerability Definition and Potential 
               Plant Improvements

The licensee should provide a discussion on how a vulnerability is 
defined for each external event evaluated.  The licensee should list 
any improvements (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 (a schedule for implementation should 
be provided) or that have already been implemented.  A discussion of 
anticipated benefits, in terms of averted potential risk or increased 
plant seismic capacity, as well as drawbacks to any improvements should 
be provided.  Those improvements that have been taken credit for in the 
analysis and have not yet been implemented at the plant, should be 
specifically highlighted in the submittal.

4.1.5          IPEEE Team and Peer Review

The basis for requesting the involvement of the licensee'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 from the examination 
into operating procedures and training programs.  Thus, the submittal 
should describe licensee staff 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 peer review 
performed, the same type of review as requested for the internal event 
IPE, the results of the review team's evaluation, and a list of the 
review team members.   

4.2  Seismic Events

Section 4.2.1 describes guidelines for submittal of information by 
licensees who choose the seismic PRA for the seismic IPEEE, 
.

                                   20

whereas 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 the table of contents 
provided in Table C.1 of NUREG-1407.

4.2.1          Seismic PRA Methodology

The following information on the seismic IPEEE should be documented and 
submitted to the NRC:

1.   A description of the methodology and key assumptions used in 
     performing the seismic IPEEE.  
 
2.   The hazard curve(s) (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 overall results and 
     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 description of dominant 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 used).  Sequence selection criteria 
     are provided in GL 88-20 and NUREG-1335.  If either hazard curve 
     causes a sequence to meet these criteria, that sequence should be 
     included.  The description of the sequences should include a 
     discussion of specific assumptions and human recovery actions.

6.   The estimated core damage frequency (for both the LLNL and EPRI 
     hazard curves, if used) and plant damage state, 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. 

.

                                   21

7.   Any seismically induced containment failures and other containment 
     performance insights.  Particularly, vulnerabilities found in the 
     systems/functions which will lead to early containment failure 
     that might result in high consequences.  This includes: isolation, 
     bypass, containment integrity and systems (e.g., igniters) 
     required to prevent early failure.  The computed fragilities of 
     containment components, systems, and functions as applicable 
     should be provided.  The licensee may submit computed HCLPFs 
     associated with containment (Optional). 

8.   A table of fragilities, both generic and plant-specific, used for 
     screening as well as in the quantification.  The estimated 
     fragilities for the plant, dominant sequences, and dominant 
     components should be reported.  (Optional: The estimated HCLPF for 
     the plant, dominant sequences, and components with and without 
     nonseismic failures and human actions may be submitted by the 
     licensee.)

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.  A discussion of nonseismic failures and human actions that are 
     significant contributors, or have impacts on results.
  
11.  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 10 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 should be documented and 
submitted to the NRC for a full-scope and a focused-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, soil liquefaction, and seismically induced floods/fires.  
     Also, a list of important nonseismic failures with a rationale for 
     the assumed failure rate given a seismic event.
.

                                   22

2.   A summary of the walkdown results 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 in 
     the controlling success path (EPRI method).

5.   Any seismically induced containment failures and other containment 
     performance insights.  Particularly, vulnerabilities found in the 
     systems/functions which will lead to early containment failure and 
     high consequences.  This includes: isolation, bypass, containment 
     integrity and systems (e.g., igniters) required to prevent early 
     failure.  Also, computed fragilities (if used) and HCLPFs of 
     containment components, systems, and functions as applicable. 

6.   A table of fragilities (if used) and HCLPFs, both generic and 
     plant-specific, used for screening as well as in the 
     quantification.  The estimated fragilities (if used) and HCLPFs 
     for the plant, dominant sequences, and dominant components should 
     be reported.  

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.

8.   For NRC method provide a discussion of nonseismic failures and 
     human actions that are significant contributors, or have impacts 
     on results.  

The following information 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 
     components 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.

.

                                   23

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 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.

4.3  Internal Fires

The following information on the internal fires IPEEE should be 
documented and submitted to the NRC:

1.   A description of the methodology and key assumptions 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 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, including (a) single areas 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, HVAC 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.

5.   A discussion of the fire initiation data base, including the 
     plant-specific data base used.  Describe the data handling method, 
     including major assumptions, the role of expert judgment, and the 
     identification and evaluation of sources of data uncertainties.  A 
     discussion of each case where the plant-specific data used is less 
     conservative than the data base used in the approved fire 
     vulnerability methodologies.

6.   A discussion of the treatment of fire growth and spread, the 
     spread of hot gases and smoke, and the analysis of detection and 
     suppression and their associated assumptions, 
     .     

                                   24

     including the treatment of suppression-induced damage to 
     equipment.  

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 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.   Discuss the treatment of detection and suppression, including fire 
     fighting procedures, fire brigade training and adequacy of 
     existing fire brigade equipment, and treatment of access routes 
     versus existing barriers.  

9.   All functional/systemic event trees associated with fire initiated 
     sequences.

10.  A description of dominant functional/systemic sequences leading to 
     core damage along with their frequencies and percentage 
     contribution to overall fire core damage frequencies.  Sequence 
     selection criteria are provided in GL 88-20 and NUREG-1335.  The 
     description of the sequences should include a discussion of 
     specific assumptions and human recovery action.  

11.  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.  

12.  Any fire induced containment failures identified as being 
     different than those identified in the internal events analysis 
     and other containment performance insights.  

13.  Documentation with regard to fire risk scoping study 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 should be 
     specifically highlighted.  

14.  When an existing PRA is used to address the fire IPEEE, the 
     licensee should describe sensitivity studies related to the use of 
     the initial hazard supplemental plant walkdown results and 
     subsequent evaluations.  The licensee should examine the above 
     list to fill in those items missed in the existing fire PRA.  

4.4  High Winds, Floods, and Others

.

                                   25

The following information on the high winds, floods, and others portion 
of the IPEEE should be documented and submitted to the NRC:

1.   A description of the methodologies used in the examination.

2.   Information on plant-specific hazard data and licensing bases.

3.   Identified significant changes not reported per 10CFR 50.71(e) 
     (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.

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).

10.  A certification that the licensee knows of no other plant-unique 
     external event that poses any significant threat of severe 
     accident within the context of the screening approach for "High 
     Winds, Floods, and Others." 
.

                                   26

                              APPENDIX 5
                        10CFR50.54(f) ANALYSIS 
      FOR INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE)

10CFR50.54(f) requires that "... the NRC must prepare the reason or 
reasons for each information request prior to issuance to ensure that 
the burden to be imposed on respondents is justified in view of the 
potential safety significance of the issue to be addressed in the 
requested information."  Further, Revision 4 of the Charter of the 
Committee to Review Generic Requirements (CRGR), dated April 1989 
specifies that, at a minimum, such an evaluation shall include:

     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 
          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 the information in 
     terms of potential safety benefit.
     
     In the Commission policy statement on severe accidents in nuclear 
     power plants issued August 8, 1985 (50FR 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 on any regulatory 
     requirements for these plants.  However, the Commission 
     recognizes, 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 on Nov. 23, 1988, requesting that each licensee 
     conduct an individual plant examination (IPE) for internally 
     initiated events only.  An analysis prepared to justify the burden 
     associated with the internal event IPE (Ref. 23) 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, 13, and 24-29 indicate that 
     the risk from external events could be a significant 
     .     

                                   27

     contributor to core damage in some instances.  Most recently, the 
     NUREG-1150 (Ref. 30) study showed that the contribution to severe 
     accidents initiated by internal fires and seismic events was 
     comparable to or greater than that initiated by internal events.  
     Examples of the severe accident sequences initiated by external 
     events can be found in References 6-8, 13, and 23-29.  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 where there is less 
     redundancy, less separation and independence between trains, 
     poorer overall general arrangement of equipment from a safety 
     viewpoint, and much more system sharing as compared to the higher 
     level systems).  Actual examples of cost-effective improvements 
     that have been found and made are modification of structural 
     design to improve the 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 external flood and installation of a 
     dedicated independent safe shutdown system and construction 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 systematic 
     examination of external events to allow the staff to carry out 
     additional work to (1) identify which external hazards need an 
     examination, (2) identify acceptable 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 now requesting 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 
     .     

                                   28

     Statement has continued to confirm that analyses of this type 
     often reveal plant-specific vulnerabilities that can be and 
     typically are corrected in a cost effective manner see the 
     value/impact analysis presented in the Attachment to Appendix D of 
     NUREG-1407.  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.

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 accidents and 
     report this information to the NRC.  The licensees would also 
     report to the NRC proposed modifications, if any, 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 in the Generic Letter or using other 
     methods that the licensees may propose provided NRC review has 
     shown that such proposed methods are effective and applicable.

     We estimate that the cost of these systematic examinations will 
     vary depending on specific site conditions, but, on the average, 
     will cost no more than $1M or a maximum of about 6 person-years 
     for the examination.  However, we feel that, for most licensees, 
     the scope will be less than that and the cost will also be less 
     (see cost estimates presented in the Appendix D to NUREG-1407).  
     Also, for those licensees who have already performed external 
     event 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 most of the IPEEEs will be submitted in mid 1994 
     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.
.

                                   29

                              APPENDIX 6
                              REFERENCES 

1.   U.S. Nuclear Regulatory Commission (USNRC), "Policy Statement on 
     Severe Accidents," Federal Register, Vol. 50, 32138, August 8, 
     1985.

2.   USNRC Generic Letter 88-20, "Individual Plant Examination for 
     Severe Accident Vulnerabilities--10CFR 50.54(f)," November 23, 
     1988.

3.   USNRC NUREG-1407, "Procedural and Submittal Guidance for the 
     Individual Plant Examination of External Events (IPEEE) for Severe 
     Accident Vulnerabilities," May 1991.

4.   USNRC SECY 88-147, "Integration Plans for Closure of Severe 
     Accident Issues," May 25, 1988.

5.   USNRC NUREG/CR-2300, "PRA Procedures Guide," Amarican Nuclear 
     Society and Institute of Electrical and Electronic Engineers, 
     January 1983.

6.   USNRC NUREG/CR-5042, "Evaluation of External Hazards to Nuclear 
     Power Plants in the United States," Lawrence Livermore National 
     Laboratory, December 1987.

7.   USNRC NUREG/CR-5042, Supplement 1, "Evaluation of External Hazards 
     to Nuclear Power Plants in the United States, Seismic Hazards" 
     Lawrence Livermore National Laboratory, April 1988.

8.   USNRC NUREG/CR-5042, Supplement 2, "Evaluation of External Hazards 
     to Nuclear Power Plants in the United States, Other External 
     Events," Lawrence Livermore National Laboratory, February 1989.

9.   USNRC NUREG/CR-2815, "Probabilistic Safety Assessment Procedures 
     Guide," Brookhaven National Laboratory, August 1985. 

10.  USNRC NUREG/CR-4840, "Recommended Procedures for the Simplified 
     External Event Risk Analyses for NUREG-1150," Sandia National 
     Laboratory, September 1989.

11.  USNRC NUREG/CR-5250, "Seismic Hazard Characterization of 69 
     Nuclear Plant Sites East of the Rocky Mountains," Lawrence 
     Livermore National Laboratory, January 1989.

12.  Electric Power Research Institute, 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.
.

                                   30

13.  USNRC NUREG/CR-4334, "An Approach to the Quantification of Seismic 
     Margins in Nuclear Power Plants," Lawrence Livermore National 
     Laboratory, August 1985.

14.  USNRC NUREG/CR-4482, "Recommendations to the Nuclear Regulatory 
     Commission on Trial Guidelines for Seismic Margins Reviews of 
     Nuclear Power Plants," Lawrence Livermore National Laboratory, 
     March 1986.

15.  USNRC 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," Lawrence Livermore 
     National Laboratory, May 1988.

16.  Electric Power Research Institute, NP-6041, "A Methodology for 
     Assessment of Nuclear Power Plant Seismic Margin," October 1988.

17.  USNRC NUREG/CR-5088, "Fire Risk Scoping Study," Sandia National 
     Laboratory, January. 1989.

18.  USNRC NUREG-75/087, "Standard Review Plan for the Review of Safety 
     Analysis Reports for Nuclear Power Plants--LWR Edition," December 
     1975.
 
19.  USNRC Memorandum from E. Beckjord to R. Houston, dated July 31, 
     1989, Subject: Generic Issue 131, "Potential Seismic Interaction 
     Involving the Movable In-Core Flux Mapping System Used in 
     Westinghouse Plants" (available in NRC Public Document Room). 

20.  USNRC Generic Letter 88-20, Supplement 1, "Initiation of the 
     Individual Plant Examination for Severe Accident 
     Vulnerabilities--10 CFR 50.54(f)," August 29, 1989.

21.  USNRC Generic Letter 88-20, Supplement 3, "Completion of 
     Containment Performance Improvement Program and Forwarding of 
     Insights for Use in the Individual Plant Examination for Severe 
     Accident Vulnerabilities,"  June 1990.

22.  USNRC NUREG-1335, "Individual Plant Examination: Submittal 
     Guidance," Final Report, August 1989.

23.  USNRC Memorandum from B. Sheron 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 Document Room). 

24.  USNRC NUREG/CR-4458, "Shutdown Decay Heat Removal Analysis of a 
     Westinghouse 2-loop Pressurized Water Reactor," Sandia National 
     Laboratory, March 1987.
.

                                   31

25.  USNRC NUREG/CR-4713, "Shutdown Decay Heat Removal Analysis of a 
     Babcock and Wilcox Pressurized Water Reactor," Sandia National 
     Laboratory, March 1987.

26.  USNRC NUREG/CR-4762, "Shutdown Decay Heat Removal Analysis of a 
     Westinghouse 3-loop Pressurized Water Reactor," Sandia National 
     Laboratory, March 1987.

27.  USNRC NUREG/CR-4767, "Shutdown Decay Heat Removal Analysis of a 
     General Electric BWR4/Mark I," Sandia National Laboratory, March 
     1987.

28.  USNRC NUREG/CR-4710, "Shutdown Decay Heat Removal Analysis of a 
     Combustion Engineering Pressurized Water Reactor," Sandia National 
     Laboratory, March 1987.

29.  USNRC NUREG/CR-4448, "Shutdown Decay Heat Removal Analysis of a 
     General Electric BWR3/ Mark I," Sandia National Laboratory, March 
     1987.

30.  USNRC NUREG-1150, "Severe Accident Risks: An Assessment for Five 
     U.S. Nuclear Power Plants," Sandia National Laboratory, December 
     1990.
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