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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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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- 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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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.
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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).
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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,
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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.
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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.
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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.
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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,
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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
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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."
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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
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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
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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.
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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.
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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.
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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.
Page Last Reviewed/Updated Tuesday, March 09, 2021
Page Last Reviewed/Updated Tuesday, March 09, 2021