United States Nuclear Regulatory Commission - Protecting People and the Environment

Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54(f) (Generic Letter No. 88-20)



                             UNITED STATES
                     NUCLEAR REGULATORY COMMISSION
                       WASHINGTON, D.C.  20555 

                           November 23, 1988

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

SUBJECT:      INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT 
              VULNERABILITIES - 10 CFR 50.54(f)
              (Generic Letter No. 88-20) 

1.   SUMMARY 

In the Commission policy statement on severe accidents in nuclear power 
plants issued on August 8, 1985 (50 FR 32138), the Commission concluded, 
based on available information, that existing plants pose no undue risk to 
the public health and safety and that there is no present basis for 
immediate action on generic rulemaking or other 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.  Therefore, each existing plant should perform a 
systematic examination to identify any plant-Specific vulnerabilities to 
severe accidents and report the results to the Commission. 

The general purpose of this examination, defined as an Individual Plant 
Examination (IPE), is for each utility (1) to develop an appreciation of 
severe accident behavior, (2) to understand the most likely severe accident 
sequences that could occur at its plant, (3) to gain a more quantitative 
understanding of the overall probabilities of core damage and fission 
product releases, and (4) if necessary, to reduce the overall probabilities 
of core damage and fission product releases by modifying, where 
appropriate, hardware and procedures that would help prevent or mitigate 
severe accidents.  It is expected that the achievement of these goals will 
help verify that at U.S. nuclear power plants severe core damage and large 
radioactive release probabilities are consistent with the Commission's 
Safety Goal Policy Statement.  Besides the Individual Plant Examinations, 
closure of severe accident concerns will involve future NRC and industry 
efforts in the areas of accident management and generic containment 
performance improvements. Additional discussion is provided in SECY-88-147 
on the interrelationships among these three areas and the role they play in 
closure of severe accident issues for operating plants.  The portion of 
that document relevant to closure is provided as Attachment 1.  Attachment 
2 contains a list of references of the IDCOR program technical reports and 
also some related NRC and NRC contractor reports.   

Therefore, consistent with the stated position of the Commission and 
pursuant to 10 CFR 50.54(f), you are requested to perform an Individual 
Plant Examination of your plant(s) for severe accident vulnerabilities and 
submit the results to the NRC. 



                                    2                  November 23, 1988 

2.   Examination Process  

The quality and comprehensiveness of the results derived from an IPE will 
depend on the vigor with which the utility applies the method of 
examination and on the utility's commitment to the intent of the IPE.  
Furthermore, the maximum benefit from the IPE would be realized if the 
licensee's staff were involved in all aspects of the examination to the 
degree that the knowledge gained from the examination becomes an integral 
part of plant procedures and training programs. Therefore, we request each 
licensee to use its staff to the maximum extent possible in conducting the 
IPE by: 

     1.   Having utility engineers, who are familiar with the details of 
          the design, controls, procedures, and system configurations, 
          involved in the analysis as well as in the technical review, and 
          
     2.   Formally including an independent in-house review to ensure the 
          accuracy of the documentation packages and to validate both the 
          IPE process and its results. 

The NRC expects the utility's staff participating in the IPE to: 

     (1) Examine and understand the plant emergency procedures, design, 
     operations, maintenance, and surveillance to identify potential severe 
     accident sequences for the plant; (2) understand the quantification of 
     the expected sequence frequencies; (3) determine the leading 
     contributors to core damage and unusually poor containment 
     performance, and determine and develop an understanding for their 
     underlying causes; (4) identify any proposed plant improvements for 
     the prevention and mitigation of severe accidents; (5) examine each of 
     the proposed improvements, including design changes as well as changes 
     in maintenance, operating and emergency procedures, surveillance, 
     staffing, and training programs; and (6) identify which proposed 
     improvements will be implemented and their schedule. 

3.   External Events (Treated Separately) 

Licensees are requested to proceed with the examinations only for 
internally initiated events (including internal flooding) at the present 
time.  Examination of externally initiated events (i. e., internal fires, 
high winds/tornadoes, transportation accidents, external floods, and 
earthquakes) will proceed separately and on a later schedule from that of 
internal events (1) to permit the identification of which external hazards 
need a systematic examination, (2) to permit development of simplified 
examination procedures, and (3) to integrate other ongoing Commission 
programs that deal with various aspects of external event evaluations, such 
as the Seismic Design Margins Program (SDMP), with the IPE(s) to ensure 
that there is no duplication of industry efforts.  Utilities would be 
expected to examine and identify any plant-specific vulnerabilities to 
severe accidents due to externally initiated events.  Therefore, while 
performing your IPE for internally initiated events, you should document 
and retain plant-specific data relevant to external events (e.g., data from 
plant walkdowns) such that they can be readily retrieved in a convenient 
form when needed for later external event analyses that may be required.  
If a licensee chooses to submit an external event examination at this time, 
the staff would review it on a case-by-case basis. 
.

                                    3                  November 23, 1988 

While current staff efforts are focused on identifying acceptable methods 
for examining external events, the staff encourages the industry to propose 
a methodology for examining external events that meets the intent of the 
severe accident policy; namely, that it is capable of identifying 
vulnerabilities to external hazards.   We will work with NUMARC in 
developing acceptable methodologies for external hazard examinations. 

4.   Methods of Examination 

The NRC has identified three approaches that satisfy the examination 
requested by this letter.   The methods are: 

1.   A PRA, provided it is at least a Level I* and uses current methods and 
     information, plus a containment performance analysis that follows the 
     general guidance given in Appendix 1 to the is generic letter.  The 
     staff will consider those  PRA s that follow the PRA procedures 
     described in NUREG/CR-2300, NUREG/CR-2815, or NUREG/CR-4550 to be 
     adequate for performing the IPE, provided the assessment considers the 
     most current severe accident phenomenological issues (as discussed in 
     Appendix 1) and the licensee certifies that the PRA is based on the 
     most current design. 

2.   The IDCOR system analysis method (front-end only), provided the 
     enhancements identified in the NRC staff evaluation of the IDCOR 
     method (to be issued shortly) are applied.  Guidance for the back-end 
     analysis is provided in Appendix 1 and  additional guidance will be 
     issued as described in Section 11 of this generic letter. 

3.   Other systematic examination methods, provided the method is described 
     in the licensee response and is accepted by the NRC staff.  For those 
     methods with which the  staff is not familiar, a staff review might be 
     necessary to ensure that the methods are generally acceptable. 

For the phase of the evaluation associated with core melting, release of 
molten core to the containment, and containment performance, the staff 
recognizes that for a few of the phenomena, notably associated with areas 
that affect containment performance, there is a wide range of views about 
their relative probability as well as their consequences.  For these 
issues, additional research and evaluation will be needed to help reduce 
the wide range of uncertainties.  Because of the concern over the ability 
of containments to perform well during some severe accidents, the staff is 
conducting a Containment Performance Improvements Program.  This program 
complements the IPE program and is intended to focus on resolving generic 
containment challenges.  License are expected to correct vulnerabilities 
that may be identified by their IPE results but, because of the generic 
Containment Performance Improvements Program that complements the IPE, the 

____________________ *The PRA levels are defined as follows: Level I - 
determination of core-damage frequencies based on system and human-factor 
evaluations; Level II -determination of the physical and chemical phenomena 
that affect the performance of the containment and other mitigating 
features and the behavior and release of the fission products to the 
environment; and Level III - determination of the offsite transport, 
deposition, and health effects of fission product releases. 
.

                                    4                  November 23, 1988 

staff does not require industry to make any major modifications to their 
containments or other systems that can affect containment performance until 
the information associated with the containment performance generic issues 
has been developed by the staff.  Hence, industry will not be placed in a 
position of having to implement improvements before all containment 
performance decisions have been made. 

Appendix 1 provides the utility with guidance to proceed with the 
evaluation of containment performance to identify plant-specific factors 
important to containment performance.  Following the Appendix 1 guidance 
will also enable utilities to understand and develop strategies to minimize 
the challenges and the consequences such severe accident phenomena may pose 
to the containment integrity and to recognize the role of mitigation 
systems while awaiting their generic resolution. 

5.   Resolution of Unresolved Safely/Generic Safety Issues (Relationship to   
     USI A-45) 

Because the resolution of several USI(s) and GSI(s) may require an 
examination of the individual plant, it is reasonable to use the current 
IPE process for that examination.  For example, Unresolved Safety Issue 
(USI) A-45 entitled "Shutdown Decay Heat Removal Requirements" had as its 
objective the determination of whether the decay heat removal function at 
operating plants is adequate and if cost-beneficial improvements could be 
identified.  We concluded that a generic resolution to the issue (e.g., a 
dedicated decay heat removal system for all plants) is not cost effective 
and that resolution could only be achieved on a plant-specific basis.  To 
implement a plant-specific resolution would require each plant to do an 
examination of its decay heat removal system to identify vulnerabilities.  
In the IPE, each plant will do an examination of both its decay heat 
removal system and those systems used for the other safety functions for 
the purpose of identifying severe accident vulnerabilities. Therefore, we 
have concluded that the most efficient way to resolve A-45 is to subsume it 
in the IPE. 

You should ensure that your IPE particularly identifies decay heat removal 
vulnerabilities.  To achieve this assurance we have extracted insights 
gained from the six case studies performed for the USI A-45 program.  These 
insights are discussed in Appendix 5 to this letter and should be 
considered as you con-duct your IPE.  In addition, if a utility (1) 
discovers a notable vulnerability during its IPE that is topically 
associated with any other USI or GSI 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 GSI, the staff will 
consider the USI or GSI resolved for a plant upon review and acceptance of 
the results of the IPE.  Your IPE submittal should specifically identify 
which USIs or GSIs it is resolving. 

6.   PRA Benefits 

The NRC recognizes that many licensees now possess plant-specific PRAs or 
similar analyses.  Use of existing PRA analyses is encouraged in achieving 
the objectives of the IPE.  In some cases, the licensee may have to confirm 
that the existing PRA analyses reflect the current state of the art 
regarding severe accidents. 
.

                                    5                  November 23, 1988 

In addition to being an acceptable method for conducting an IPE, there are 
a number of potential benefits in performing PRAs on those plants without 
one. Some examples of potential additional benefits are as follows: 

     Support for Licensing Actions - PRAs have been used to support 
     arguments to justify technical specification changes, both routine and 
     emergency.  PRAs would also be useful in supporting other regulatory 
     actions (e.g., design modifications). 

     License Renewals - PRAs could be a basis for utilities to establish a  
     program to ensure that risk-significant components and systems are  
     identified and maintained at an acceptable level of reliability during 
     the license renewal period. 

     Risk Management - A PRA could be used to develop a risk management 
     program that systematically uses the available information about risk 
     at a nuclear power plant and identifies alternative combinations of 
     design and operational modifications, ranks these alternatives 
     according to the relative benefits of each, and selects an optimum 
     from the alternatives. 

     Integrated Safety Assessment - The staff believes that by performing a 
     PRA a licensee would have the benefit of having developed the 
     technical basis for an integrated assessment.  An integrated safety 
     assessment would (1) provide integrated schedules for licensing, 
     regulatory, and safety issues on a predictable basis, (2) evaluate 
     licensing and generic issues on a plant-specific basis such that they 
     are weighted against all other pending actions, (3) provide a licensee 
     with the opportunity to demonstrate with its PRA that various issues 
     that might be applied to other plants are not justified at that 
     facility, (4) help improve outage planning, and (5) rank issue 
     importance such that the most important are dealt with first.  This 
     prioritization of actions benefits the licensees and the NRC by 
     providing a rational schedule for implementation of actions and 
     provides a basis for the possible elimination of actions determined to 
     have low safety  significance for the individual plant. 

7.   Severe Accident Sequence Selection 

In performing an IPE, it is necessary to screen the severe accident 
sequences for the potentially important ones and for reporting to the NRC.  
The screening criteria to determine the potentially important functional 
sequences* that lead to core damage or unusually poor containment 
performance and should be reported to the NRC with your IPE results are 
listed in Appendix 2.  Appendix 4 describes 

____________________ 
*"Sequence" is used here to mean a set of faults, usually chronological, 
that result in the plant consequence of interest, i.e., either a damaged 
core or unusually poor containment performance.  A functional sequence is a 
set of faulted functions that summarizes by function a set of systems 
faults which would result in the consequence of interest.  Functional 
sequences are to be contrasted with systemic sequences.  A systemic 
sequence is a set of faulted systems that summarizes by systems a set of 
component failures resulting in a damaged core or unusually poor 
containment performance. 
.

                                    6                  November 23, 1988 

the documentation needed for the accident sequence selection and the 
intended disposition of these sequences. 

It is expected that during the course of the examination, the utility would
carefully examine the results to determine if there are worthwhile 
prevention or mitigation measures that could be taken to reduce the core 
damage frequency or poor containment performance with the attendant 
radioactive release.  The determination of potential benefits is plant 
specific and will depend on the frequency and consequence of the accident 
sequence leading to core damage and containment failure. 

8.   Use of IPE Results 

     a.  Licensee 

After each licensee conducts a systematic search for severe accident 
vulnerabilities in its plant(s) and determines whether potential 
improvements, both design and procedural, warrant implementation, it is 
expected that the licensee will move expeditiously to correct any 
identified vulnerabilities that it determines warrant correction.  
Information on changes initiated by the licensee should be provided 
consistent with the requirements of 10 CFR 50.59 and10 CFR 50.90.  Changes 
should also be reported in your IPE submittal (by reference to previous 
submittals under 10 CFR 50.59 or 10 CFR 50.90) that responds to this letter 
(see Appendix 4). 

     b.  NRC 

The NRC will evaluate licensee IPE submittals to obtain reasonable 
assurance that the licensee has adequately analyzed the plant design and 
operations to discover instances of particular vulnerability to core melt 
or unusually poor containment performance given a core melt accident.  
Further, the NRC will assess whether the conclusions the licensee draws 
from the IPE regarding changes to the plant systems, components, or 
accident management procedures 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 must be changed to meet NRC 
     regulations, then appropriate functional enhancements will be required 
     and expected to be implemented without regard to cost except as 
     appropriate to select among alternatives. 

2.   If NRC consideration indicates that plant design or operation could be
     enhanced by substantial additional protection beyond NRC regulations, 
     then appropriate functional enhancements will be recommended and 
     supported with analysis demonstrating that the benefit of such 
     enhancement is substantial and worth the cost to implement and 
     maintain that enhancement, in accordance with 10 CFR 50.109. 

3.   If NRC consideration indicates that the plant design and operation 
     meet NRC regulations, and that further safety improvements are not 
     substantial or not cost effective, enhancements would not be suggested 
     unless significant new safety information becomes available. 
.

                                    7                  November 23, 1988 

9.   Accident Management 

An important aspect of severe accident prevention and mitigation is the 
total organizational involvement.  Operations personnel have key roles in 
the early recognition of conditions or events that might lead to core 
damage.  The availability of procedures specifying corrective actions and 
the training of operators and emergency teams can have a major influence on 
the course of events in case of a severe accident. 

Because the conclusions you will draw from the IPE for severe accident 
vulnerabilities (1) depend on the credit taken for survivability of 
equipment in a severe accident environment, and (2) will either depend on 
operators taking beneficial actions during or prior to the onset of severe 
core damage or depend on the operators not taking specific actions that 
would have adverse effects, the results of your IPE will be an essential 
ingredient in developing a severe accident management program for your 
plant. 

At this time you are not required to develop an accident management plan as 
an integrated part of your IPE.  We are currently developing more specific 
guidance on this matter and are working closely with NUMARC to (1) define 
the scope and content of acceptable accident management programs, and (2) 
identify a plan of action that will ultimately result in incorporating any 
plant-specific actions deemed necessary, as a result of your IPE, into an 
overall severe accident management program.  Nevertheless, in the course of 
conducting your IPE you may identify operator or other plant personnel 
actions that can substantially reduce the risk from severe accidents at 
your plant and that you believe should be immediately implemented in the 
form of emergency operating procedures or similar formal guidance.  We 
encourage 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 immediately within 
the constraints of 10 CFR 50.59. 

10.  Documentation of Examination Results 

The IPE should be documented in a traceable manner to provide the basis for 
the findings.  This can be dealt with most efficiently by a two-tier 
approach.  The first tier consists of the results of the examination, which 
will be reported to the NRC for review.  The second tier is the 
documentation of the examination itself, which should be retained by the 
licensee for the duration of the license unless superseded.  Appendix 4 
contains the minimum information necessary for reporting and documentation. 

11.  Licensee Response 

A document that provides additional licensee guidance for the performance 
of the IPE (both core damage and containment system performance) and 
describes the review and evaluation process that the NRC staff will use for 
assessing the submittals will be issued in draft form within the next few 
months. 

.

                                    8                  November 23, 1988 

Following the issuance of the draft document, workshops with utility 
representatives will be scheduled to discuss the IPE objectives and to 
answer questions that utilities might have on both the IPE generic letter 
and the guidance document. 

Following the completion of the workshops, the NRC, as appropriate, will 
revise its guidance contained in the guidance documents to take into 
consideration comments received and will reissue them.  Within 60 days of 
receipt of the final guidance documents, licensees are requested to submit 
their proposed programs for completing the IPEs.  The proposal should: 

1.   Identify the method and approach selected for performing the IPE, 

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 schedules for performing the IPE and 
     submitting the results to the NRC.

Meetings at NRC Headquarters during the examinations will be scheduled as 
needed to discuss subjects raised by licensees and to provide necessary 
clarifications.  

Licensees are expected to submit the IPE results within 3 years.  The 
Commission encourages those plants that have not yet undergone any 
systematic examination for severe accidents to promptly initiate the 
examination. 

Those utilities that choose to use an existing PRA or similar analysis on 
their plant should (1) certify that the PRA meets the intent of the generic 
letter, in particular with respect to utility staff involvement, (2) 
certify that it reflects the current plant design and operation, and (3) 
submit the results as soon as the analysis is completed but on a shorter 
schedule than 3 years. Utilities with plants that used the initial IDCOR 
system analysis in the IDCOR test applications are encouraged to submit 
their results on a shorter schedule than 3 years.  This will  ensure review 
and resolution of any items while the utility's examination team is easily 
accessible.  In this regard, the staff also encourages licensees whose 
plants have been extensively analyzed under the NUREG-1150 program to 
submit their IPEs on an expedited basis.  This will enable the staff to 
exercise its review and decision process for determining acceptability of 
the IPE, the adequacy of the licensee identification of plant-specific 
vulnerabilities, and the associated modifications using insights and 
experience from NUREG-1150.  Finally, those licensees planning to perform a 
new Level II or Level III PRA may need more time.  The NRC staff will 
consider requests for additional time for such an examination. 

12.  Regulatory Basis 

This letter is issued pursuant to 10 CFR 50.54(f), a copy of the 10 CFR 
50.54(f) evaluation which justifies issuance of this letter is in the 
Public Document Room.  Accordingly, all responses should be under oath or 
affirmation. This request for information is covered by the Office of 
Management and Budget under 

.

                                    9                  November 23, 1988 

Clearance No. 3150-0011, which expires December 31, 1989.  The estimated 
average burden hours is 8100 person-hours per licensee response, over a 
3-year period including assessment of the new requirements, searching data 
sources, gathering and analyzing the data, and preparing the required 
reports.  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.  


                            Sincerely, 



                            Dennis Crutchfield, Acting Associate 
                              Director for Projects 
                            Office of Nuclear Reactor Regulation 


Enclosures: 
Appendices 1 through 5 
     w/ attachments 1 and 2 

.

                               APPENDIX 1
      GUIDANCE ON THE EXAMINATION OF CONTAINMENT SYSTEM PERFORMANCE
                           (BACK-END ANALYSIS)


1.   Background 

The role of the containment as a vital barrier to the release of fission 
products to the environment has been widely recognized.  The public safe%y 
record 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.  The containment and its supporting systems are 
one of these barriers.  Containment design criteria are based on a set of 
deterministically derived challenges.  Pressure and temperature challenges 
are usually based on the design basis loss-of-coolant accident; 
radionuclide challenges are based on the source term of 10 CFR Part 100.  
Also, criteria based on external events such as earthquakes, floods, and 
tornadoes are considered.  The margins of safety provided by such practices 
have been the subject of considerable research and evaluation, and these 
studies have shown the ability of many containment systems to survive 
pressure challenges of two to three times design levels. Because of these 
margins, the various containment types presently used in the United States 
have the capability to withstand, to varying degrees, many of the 
challenges presented by severe accidents.  For each type of containment, 
however, there remain failure mechanisms that could lead to either early or 
late containment failure, depending on both the accident scenarios involved 
and the containment types.  

This appendix discusses the key phenomena and/or processes that can take 
place during the evolution of a severe accident and that can have an 
important effect on the containment behavior.  In addition, general 
guidance on the evaluation of containment system performance given the 
present state of the art of analysis of these phenomena is provided.  The 
evaluation should be a pragmatic exploitation of the present containment 
capability.  It should give an understanding and appreciation of severe 
accident behavior, should recognize the role of mitigating systems, and 
should ultimately result in the development of accident management 
procedures that could both prevent and ameliorate the consequences of some 
of the more probable severe accident sequences involved.  The users of this 
appendix are referred to Chapter 7 of Volume 1 of NUREG/CR-2300, "PRA 
Procedures Guide," for a more detailed description of procedures and 
guidance on containment performance analysis.  The additional information 
provided here summarizes some more recent developments in core melt 
phenomenology relevant to containment performance, identifies areas of 
uncertainty, and suggests ways of proceeding with the evaluation of 
containment performance despite uncertainties,and potential ways of 
improving containment performance for severe accident challenges.  In this 
reloads, the Severe Accident Prevention and Mitigation Features report 
(NUREG/CR-4920) summarizes insights gained from industry sponsored PRAs, 
NUREG-1150, and IDCOR reference plant analyses.  The report identifies 
plant features and operator actions that have been found to be important to 
either the prevention or the mitigation of severe accidents for a specific 
plant containment type.  The report indicates what may be important to risk 
and suggests potential improvements in various areas of plant design and 
operation.  These insights and suggestions may be helpful when conducting 
the IPE and when making decisions on plant improvements. 

                                         1-1 
.

The systems analysis portion of the IPE identifies accident sequences that 
occur as a result of an initiating event followed by failure of various 
systems or failure of plant personnel to respond correctly to the accident.  
Although the number of possible core melt accident sequences is very large, 
the number of containment system performance analyses does not have to be 
as large.  The number of sequences can be reduced by grouping those 
accident sequences that have a similar effect on the plant features that 
determine the release and transport of fission products. 

A containment event tree (CET) could provide a structured way for the 
systematic analysis of containment phenomena provided: 

1.   The CET is quantified, i.e., branch point split fractions are 
     propagated for each sequence based on the most recent data base 
     regarding important severe accident phenomena including considerations 
     of uncertainties (e.g., letters from T. Speis, NRC, to A. Buhl, ITC, 
     "Position Papers for the NRC/IDCOR Technical Issues," dated September 
     22, 1986; November 26, 1986; and March 11, 1987). 

2.   The system analysis is integrated with the containment analysis so 
     that initiating events and system failures (resulting in core damage) 
     that also impair containment systems are not overlooked.  

3.   The duration and sequencing of the interacting events are specified, 
     e.g., the times at which core damage and containment failure occur, 
     the time of inventory depletion (in particular, as related to recovery 
     from an accident), the success or failure of equipment or operator 
     responses, and the failure or degradation of support systems that were 
     originally available at the onset of the accident. 

2.   Status of Containment Systems Prior to Vessel Failure 

The role of interfaces between the system analysis (front-end) and the 
containment performance analysis (back-end) is particularly important from 
two perspectives.  First, the likelihood of core damage can be Influenced 
by the status of particular containment systems.  Second, containment 
performance can be influenced by the status of core cooling systems.  Thus, 
because the influences can flow, in both directions between the system 
analysis (front-end) and the containment performance analysis (back-end), 
particular attention must be given to these interfaces. 

To ensure consistency within entire sequences, the analysis should include 
a cross-checking sheet of the following by sequence: (1) the sequence 
frequency, (2) whether the containment is bypassed, (3) whether the 
containment is isolated, (4) the containment system and reactor system 
availability, and (5) the approximate source term.  This cross-checking 
sheet would be reviewed by both the systems analyst and the source term 
analyst to provide added assurance that the status of key systems is 
treated consistently in the front-end and back-end analyses.  Other options 
to ensure adequate interfaces can be used instead of the cross-checking 
list identified above. 

In order to examine the containment performance, the status of the 
containment systems and related equipment prior to core melt should be 
determined.  The first CET nodal decision point is to determine the 
likelihood of whether the 

                                    1-2 
.

containment is isolated, bypassed, intact, or failed (i.e., a branch point 
split fraction).  This requires analyses of (1) the pathways that could 
significantly contribute to containment-isolation failure, (2) the signals 
required to automatically isolate the penetration, (3) the potential for 
generating the signals for all initiating events, (4) the examination of 
the testing and maintenance procedures, and (5) the quantification of each 
containment-isolation failure mode (including common mode failures).  

In the early phase of an accident, steam and combustible gases are the main
contributors to containment pressurization.  The objective of the 
containment decay heat removal systems such as sprays, fan coolers, and the 
suppression systems is to control the evolution of accidents that would 
otherwise lead to containment failure and the release of fission products 
to the environs.  The effectiveness of the several containment decay heat 
removal systems for accomplishing the intended mitigating function should 
be examined to determine the probability of successful performance under 
accident conditions.  This includes potential intersystem dependencies as 
well as the identification of all the specific functions being performed 
and the determination of the mission time considering potential failure due 
to inventory depletion (coolant, control air, and control power) or 
environmental conditions.  If, as a result of the accident sequence, the 
front-line containment decay heat removal systems fail to function, if 
their effectiveness is degraded, or if the operator fails to respond in a 
timely manner to the accident symptoms, the containment pressure would 
continue to increase.  In this case, some systems that were not intended to 
perform a safety function might be called upon to perform that role during 
an accident, If the use of such systems is considered during the 
examination, their effectiveness and probability of success for fulfilling 
the needed safety function should also be examined.  Part of the 
examination should be to determine if adequate procedures exist to ensure 
the effective implementation of the appropriate operator actions. 


3.   Phenomena After Vessel Failure 

If adequate heat removal capability does not exist in a particular accident
sequence, the core will degrade and the containment could potentially over-
pressurize and eventually fail.  Efforts to stabilize the core before 
reactor vessel failure or to extend the time available for vessel reflood 
should be investigated.  For certain accident groups that proceed past 
vessel failure, the containment pressurization rate could exceed the 
capability of the mitigating systems to reject the energy associated with 
the severe accident phenomena encountered with vessel failure.  For each 
such accident sequence, the molten core debris will relocate, melting 
through and mixing with materials in its path.  Depending on the particular 
containment geometry and the accident sequence groups, a variety of 
important phenomena influence the challenges to containment integrity. 

The guidance provided below deals with this subject at three levels.  The 
first provides some rather general considerations regarding the nature of 
these phenomena as they impact containment (Section 3.1).  The second level 
considers the manifestation of these phenomena in more detail within the 
generic high and low pressure scenarios (Sections 3.1.1 and 3.1.2).  
Finally, the third level provides some specific guidance particularly 
regarding the treatment of certain important areas of uncertainty (Section 
4). 

                                    1-3 
.

3.1  General Description of the Phenomena Associated with Severe Accident 
     Considerations 

The contact of molten corium with water, referred to as fuel-coolant 
interaction, can occur both in-vessel and ex-vessel.  If the interaction is
energetic inside the reactor vessel, it may generate missiles and a rapid 
pressurization (steam explosion) of the primary system.  Early containment 
failure associated with in-vessel steam explosions is generally considered 
to be of low enough likelihood to not warrant additional consideration 
(NUREG-1116). However, smaller, less energetic in-vessel steam explosions 
are not unlikely and their influence on fission product release and 
hydrogen generation are still under investigation.  If the fuel-coolant 
interaction occurs ex-vessel, as might happen if molten fuel fell into a 
water-filled cavity upon vessel meltthrough, it may disperse the corium and 
lead to rapid pressurization (steam spike) of the containment.  In any 
case, at one extreme, abundant presence of water would favor quenching of 
the corium mass and the continued dissipation of the decay heat by steaming 
would lead to containment pressurization.  Clearly in the absence of 
external cooling, the containment will eventually overpressurize and fail, 
although the presence of extensive, passive heat sinks (structures) 
within the containment volume would delay the occurrence of such an event.  
Fuel-coolant interactions can also yield a chemical reaction between steam 
and the metallic component of the melt, producing hydrogen and the 
consequent potential for burns and/or explosions. 

At the other extreme, when water is not available, the principal 
interaction of the molten corium is with the concrete floor of the 
containment.  This interaction produces three challenge to containment 
integrity.  First, the concrete decomposition gives off noncondensible 
gases (CO2, CO) (of certain composition) that contribute to pressurizing 
the containment atmosphere. Second, concrete of certain compositions 
decomposes and releases CO2 and steam, which can interact with the metallic 
components in the melt to yield highly flammable CO and H2, with potential 
consequences ranging from benign burns at relatively low hydrogen 
concentrations to rapid deflagrations at high hydrogen concentrations.  
Third, continued penetration of the floor can directly breach the 
containment boundary.  Also, thermal attack by the molten corium of 
retaining sidewalls could produce structural failure within the containment
causing damage to vital systems and perhaps to failure of containment 
boundary. 

Another type of fuel interaction is with the containment atmosphere.  
Scenarios can be postulated (e.g., station blackout) in which the reactor 
vessel and primary system remain at high pressure as the core is melting 
and relocating to the bottom of the vessel.  Continued attack of the molten 
corium on the vessel lower head could eventually cause the lower head to 
fail.  Because of a potentially high (approximately 2500 psi) driving 
pressure, the molten corium could be energetically ejected from the vessel.  
Uncertainties remain related to the effect of the following on direct 
containment heating: (1) vessel failure area, (2) the amount of molten 
corium in the lower head at the time of failure, (3) the degree to which it 
fragments upon ejection, (4) the degree and extent to which a path from the 
lower cavity to the upper containment atmosphere is obstructed, (5) the 
fragmented molten corium that could enter and interact with the upper 
containment atmosphere, and (6) cavity gas temperature.  Since the 
containment atmosphere has small heat capacity, the energy in the 
fragmented corium could rapidly transfer to the containment atmosphere, 
causing a 
                                    1-4 
.

rapid pressurization.  The severity of such an event could be further 
exacerbated by any hydrogen that may be simultaneously dispersed and direct 
oxidation (exothermic) of any metallic components.  Depending upon this and 
the other factors previously mentioned, this pressurization could challenge 
containment integrity early in the event.  

The BWR Mark I and Mark II containments are normally inerted.  Therefore, 
non-condensible gases such as hydrogen and oxygen released following a 
severe accident would pressurize the containment, but would not burn or 
rapidly deflagrate.  If the containment is deinerted, additional 
pressurization events or dynamic loads obtained from global hydrogen burn 
or detonations must be considered.  Local burns are also potentially 
important as they may degrade the seals around the various penetrations or 
produce a thermal environment that challenges the operability of important 
equipment. 

Even with the above limited perspective, it should be clear that given a 
core melt accident, a great deal of the phenomenological progression hinges 
upon water availability and the outcome of the fuel-coolant interactions; 
specifically whether a full quench has been achieved and whether the 
resulting particulates will remain coolable.  In general, the presence of 
fine particulates to any significant degree would imply the occurrence of 
energetic steam explosions and hence the presence of significant forces 
that would be expected to disperse the particulates to coolable 
configurations outside the reactor cavity.  Otherwise, the coolability of 
deep corium beds of coarse particulates is the major concern.  A summary of 
how these mechanisms interface and interact as they integrate into an 
accident sequence is given below. 

3.1.1    Accident Sequences - High-Pressure Scenario 

The core melt sequence at high primary system pressure is often due to a 
station blackout sequence.  The high-pressure scenario also represents one 
of the most significant contributors to risk.  The initial stages of core 
degradation involve coolant boiloff and core heatup in a steam environment.  
At such high pressures, the volumetric heat capacity of steam is a 
significant fraction of that of water (about one-third), and one should 
expect significant core (decay) energy redistribution due to natural 
circulation loops set up between the core and the remaining cooler 
components of the primary system.  Consensus appears to be developing that 
as a result of this energy redistribution, the primary system pressure 
boundary could fail prior to the occurrence of large-scale core melt. The 
location and the size of failure, however, remain uncertain.  For example, 
concerns have been raised about the possibility of steam generator tube 
failures and associated containment bypass.  If the vessel lower head 
fails, violent melt ejection could produce large-scale dispersal and the 
direct containment heating phenomenon mentioned previously.  A significant 
amount of research in the past has not, yet produced definitive results on 
this issue. 

Concerns may also be raised about the potentially energetic role of 
hydrogen within the blowdown process.  The presence of hydrogen arises from 
two complementary mechanisms: (1) the metal-water reaction occurring at an 
accelerated pace throughout the in-vessel core heatup/meltdown/slump 
portion of the transient, and (2) the reaction between any remaining 
metallic components in the melt and the high-speed steam flow that partly 
overlaps and follows the melt ejection from the reactor vessel.  The 
combined result is the release of rather large quantities of hydrogen into 
the containment volume within a short time 
                                    1-5 
.

period (a few tens of seconds).  The implication is that the consideration 
of containment atmosphere compositions and associated burning, explosion, 
or detonation potential becomes complicated by a whole range of highly 
transient regimes and large spatial gradients.  

A recent independent review of uncertainties in estimates of source terms 
from severe accidents by an NRC-sponsored panel of experts (NUREG/CR-4883) 
provided an additional perspective on these issues and made recommendations 
for their resolution.  In particular, "if direct containment heating or 
containment bypass through steam generator tube failure contribute 
importantly to risk, this may indicate a need for a hardware modification 
or a procedural measure to ensure depressurization before primary system 
failure.  An early study of relative merits of the possibilities available 
would be valuable."  The staff is in favor of adopting the panel 
recommendation and has initiated a research program to study the effect of 
depressurization on the core melt progression and the potential benefit in 
preventing direct containment heating.  

3.1.2    Accident Sequence - Low-Pressure Scenario 

At low system pressure, decay heat redistribution due to natural 
circulation flow (in steam) is negligible and core degradation occurs at 
nearly adiabatic conditions.  Steam boiloff, together with any hydrogen 
generation, is continuously released to the containment atmosphere, where 
mixing is driven by natural convection currents coupled with condensation 
processes.  The upper internals of the reactor vessel remain relatively 
cold, offering the possibility of trapping fission product vapor and 
aerosols before they are released to the containment atmosphere.  
Throughout this core heatup and meltdown process, the potential to 
significantly load the containment is small.  The first possibility for 
significant energetic loads on the containment occurs when the molten core 
debris penetrates the lower core support structure and slumps into the 
lower plenum.  The outcome of this interaction cannot be predicted 
precisely.  Thus, a whole range of behavior must be considered in order to 
cover subsequent events. At the one extreme the interaction is benign, 
yielding no more than some steam (and hydrogen) production while the melt 
quickly reagglomerates on the lower reactor vessel head.  At the other 
extreme an energetic steam explosion occurs. It may be possible to 
distinguish intermediate outcomes by the degree to which the vessel 
integrity is degraded.  In analyzing this phase of the accident scenario, 
the important tasks are to determine the likelihood of containment failure 
and to define an envelope of corium relocation paths into the containment.  
The latter is needed to ensure the assessment of the potential for such a 
phenomenon as liner meltthrough.  

Consideration should also be given to ex-vessel coolability as the corium 
can potentially interact with the concrete.  The non-energetic release 
(vessel lower head meltthrough) and spreading upon the accessible portions 
of the containment floor below the vessel needs to be examined.  There is a 
great deal of variability in accessible floor area among the various 
designs for some PWR cavity designs.  The area over which the core debris 
could spread is rather small given whole-core melts and the resultant pool 
being in excess of 50 cm deep.  In the absence of water, all these 
configurations would yield concrete attack and decomposition of variable 
intensity.  In the presence of water (i.e., containment sprays), even deep 
pools may be considered quenchable and coolable. However, the possibility 
exists for insulating crusts or vapor barriers at the corium-water 
interface. 
                                    1-6 
.

Both of these two extremes should be considered.  The task is to estimate 
the range of containment internal pressures, temperatures, and gas 
compositions as well as the extent of concrete floor penetration and 
structural attack until the situation has been stabilized.  In general, 
pressurization from continuing core-concrete interactions (dry case) would 
be considerably slower than from coolable debris configurations (wet case) 
because of the absence of steam pressurization. As a final and crucial part 
of this scenario, one must address the combustible gas effect.  This must 
include evaluation of the quantities and composition of combustible gases 
released to the containment, local inerting and deinerting by steam and 
CO2, as well as hydrogen mixing and transport.  Also included should be 
consideration of gaseous pathways between the cavity and upper containment 
volume to confirm the adequacy of communication to support natural 
circulation, and recombination of combustible gases in the reactor cavity. 

4.   General Guidance on Containment Performance 

In the approach outlined in this appendix, emphasis is placed on those 
areas that would ensure that the IPE process considers the full range of 
severe accidents.  The IPE process should be directed toward developing a 
plant-specific accident management scheme to deal with the probable causes 
of poor containment performance at each plant.  To achieve these goals, it 
is of vital importance to understand how reliable each of the CET estimates 
are, and what the driving factors are.  Decisions on potential improvements 
should be made only after, appropriately considering the sources of 
uncertainties.  Of course, preventing failure altogether is predicated upon 
recovering some containment heat removal capability.  Given that in either 
case pressurization develops on the time scale of many hours, feasible 
recovery actions could be planned as part of accident management.  

It is the staff's view that the bulk of phenomenological uncertainties 
affecting containment response is associated with the high-pressure 
scenarios.  Unless the licensee can demonstrate that the primary system can 
be reliably depressurized, a low probability of early containment failure 
should not be automatically assumed.  Similarly, for BWRs it should not be 
assumed that the availability of the automatic depressurization system 
(ADS) in an event will ensure that reactor vessel failure will always occur 
at low pressure, since the operability of the ADS, in some plants, depends 
on maintaining a requisite differential pressure between containment and 
the reactor coolant systems.  

Low-pressure sequences, by comparison, present few remaining areas of 
controversy.  For BWRs, phenomenological uncertainties are associated with 
the behavior of combustibles and the spreading of the corium on the drywell 
floor. For PWRs, these areas include the coolability behavior of deep 
molten corium pools and the behavior of hydrogen (and other combustibles) 
in the containment atmosphere.  The staff's views and guidance concerning 
each one of these areas is briefly summarized below. 

The concerns about deep corium pools arose from experiments with 
top-flooded melts that exhibited crust formation and long-term isolation of 
the melt from the water coolant.  Such noncoolable configurations would 
yield continuing concrete attack and a containment loading behavior 
significantly different from coolable ones.  On the other hand, it has been 
pointed out that small-scale 

                                    1-7 
.

experiments would unrealistically not favor coolability.  The staff views 
this as an area of uncertainty and recommends that assessments be based on 
available cavity (spread) area and an assumed maximum coolable depth of 25 
cm.  For depths in excess of 25 cm, both the coolable and noncoolable 
outcomes should be considered.  Along these lines the IPE should document 
the geometric details of cavity configuration and flow paths out of the 
cavity, including any water drain areas into it as appropriate.  

With respect to hydrogen, the staff concerns are related to completeness of 
the current understanding of hydrogen mixing and transport.  In general, 
combustibles accumulate very slowly and only if continuing concrete attack 
is postulated.  For the larger dry containments, because of the large 
containment volume and slow release rates, compositions in the detonable 
range may not develop unless significant spatial concentrations exist or 
significant steam condensation occurs.  In general, the containment 
atmosphere under such conditions would exhibit strong natural circulation 
currents that would tend to counteract any tendency to stratify.  However, 
condensation-driven circulation patterns and other potential stratification 
mechanisms could limit the extent of the containment volume participating 
in the mixing process.  For those plants with igniters (ice-condenser and 
Mark III plants), the buildup of combustibles from continuing 
corium-concrete interactions could be limited by local ignition and 
burning.  However, oxygen availability as determined from natural 
circulation flows could limit the effectiveness of this mechanism.  
Finally, in all cases inerting/deinerting thresholds and ignition aspects 
need additional attention.  The staff recommends that, as part of the IPE, 
all geometric details impacting the above phenomena (i.e., heat sink 
distribution, circulation paths, ignition sources, water availability, and 
gravity drain paths) should be documented in a readily comprehensible form, 
together with representative combustible source transients. 

For normally inerted BWRs, the concerns with combustibles relate to 
potential burns and/or explosion events in deinerted Mark I or Mark II 
containments or in the secondary containment building following containment 
failure.  The staff recommends that, unless deinerting can be 
satisfactorily ruled out by probability, its occurrence and consequences 
should be included in the event trees.  Regarding the secondary 
containment, the staff believes that consideration of combustibles in it is 
essential with respect to the reactor building effectiveness in limiting 
the source term.  

Finally, uncertainties arise for all plants because of lack of knowledge on 
how the corium will spread following discharge from the reactor vessel.  
For Mark I containments, such uncertainties impact the configuration of the 
corium-concrete interaction process and also the potential for drywell 
liner meltthrough.  It is recommended that an assessment of the debris 
coolability, based on available water sources, should be performed to 
determine the possibility for liner meltthrough.  For Mark II containments, 
uncertainties are associated with the retention of corium on the drywell 
floor (and associated corium-concrete interactions) and the extent of 
fuel-coolant interactions in the suppression pool.  For PWR containments, 
the reactor cavity configuration will influence the potential for direct 
attack of the liner by dispersed debris, as well as the potential for 
basemat failure or structural failure due to thermal attack.  The staff 
recommends that the IPE document describe the detailed geometry (including 
curbs, standoffs) of the drywell floor. 

                                    1-8 
.

As discussed earlier, a CET provides a,structured way for a systematic 
analysis of containment phenomena.  Separate CETs representing the 
high-pressure and low-pressure sequences deal with uncertainties discussed 
earlier. 

In general terms, and consistent with the overall IPE objectives, the staff
guidance on the approach to the back-end analysis can be summarized as 
follows: 

1.   The approach should focus on containment failure mechanisms and 
     timing. Releases should be based on corresponding release categories 
     and associated detailed quantifications from reference plant analyses 
     and applied to the plant being examined.

2.   All severe accident sequences that meet the criteria of Appendix 2 
     should be considered and reported.

3.   System/human response should be realistically integrated with 
     phenomenological aspects into simplified, but realistic, containment 
     event trees for the plant being examined.  Allowance should be made 
     for the probability of recovery or other accident management 
     procedures (particularly for long-term responses).

4.   The quantification of the containment event trees should both (a) 
     clearly take into account the expected progression of the accident and 
     (b) aim to envelop phenomenological behavior (i.e., account for 
     uncertainties).  This implies: 
     
     a.   Identification of the most probable list of potential containment 
          failure mechanisms applicable to the plant under consideration 
          (e.g., see Table 7-1, NUREG/CR-2300).

     b.   Use of existing structural analyses to determine the ultimate 
          pressure capability of the containment, i.e., the quasi-static 
          internal pressure resulting in containment failure.  These should 
          be modified as necessary to take into account any unique aspects 
          that could substantially modify the range of possible failure 
          pressures. 

     c.   Use of available separate-effects analyses for the other 
          potential containment failure mechanisms to determine other 
          failure modes to which the plant might be vulnerable.  As stated 
          earlier, there are some severe accident phenomenological issues 
          (e.g., direct containment heating and containment shell 
          meltthrough) where research has not produced conclusive results 
          on the challenges that these phenomena could pose to containment 
          integrity.  Consideration must be given to strategies to deal 
          with those severe accident issues.  For example, although there 
          appears to be no consensus on whether water availability will 
          fully quench the debris and keep it coolable and hence prevent 
          Mark I containment shell meltthrough, there is a broad agreement 
          that the presence of water will scrub the fission products and 
          could substantially reduce the radionuclide released even if 
          containment shell meltthrough were to occur.  Utilities should be 
          aware of these insights and experience when conducting the IPE 
          and should develop appropriate strategies to deal with those 
          phenomenological issues while awaiting their generic resolution 
          as discussed in Section 4 of the IPE generic letter. 

                                    1-9 
.

     d.   Development of a plant-specific probability distribution function 
          of failure likelihood for the range of failure pressures.

     e.   Any claim of decontamination factors for the secondary 
          containment in the analyses should consider the possibility of no 
          natural circulation, resulting in less time for aerosol 
          deposition, as well as localized hydrogen burns causing reactor 
          building failure and forcing the reactor building atmosphere out 
          into the environment. 

5.   Documentation should be presented concerning how any calculation was 
     performed, what assumptions have been made, and how these phenomena 
     couple to other aspects of the analysis.  Any use of codes within the 
     IPE to calculate accident progression up to and including the source 
     term calculation should be described along with the circumstances 
     under which the code was used, the version of the code used, any code 
     revisions used, the key modeling and input assumptions, and the 
     calculated results.  

6.   The insights gained from the containment performance analysis should 
     be factored into the utility's accident management program. 

                                    1-10 
.

                                  APPENDIX 2 

          CRITERIA FOR SELECTING IMPORTANT SEVERE ACCIDENT SEQUENCES 

Sequence Selection Criteria 

The following screening criteria should be used to determine which 
potentially important functional sequences* and functional failures (based 
on the procedure established in NUREG/CR-2300) that might lead to core 
damage or unusually poor containment performance should be reported to the 
NRC in the IPE submittal. They do not represent a threshold for 
vulnerability.  All numerical values given in this appendix are 
"expected"** values. 

1.   Any functional sequence that contributes 1E-6*** or more per reactor 
     year to core damage,

2.   Any functional sequence that contributes 5% or more to the total core 
     damage frequency,

3.   Any functional sequence that has a core damage frequency greater than 
     or equal to 1E-6 per reactor year and that leads to containment 
     failure which can result in a radioactive release magnitude greater 
     than or equal to the BWR-3 or PWR-4 release categories of WASH-1400,

4.   Functional sequences that contribute to a containment bypass frequency 
     in excess of 1E-7 per reactor year, or

5.   Any functional sequences that the utility determines from previous 
     applicable PRAs or by utility engineering judgment to be important 
     contributors to core damage frequency or poor containment performance.


____________________
*" Sequence" is used here to mean a set of faults, usually chronological, 
that result in the plant consequence of interest, i.e., either a damaged 
core or unusually poor containment performance.  A systemic sequence is a 
set of faulted systems that summarizes by systems a set of component 
failures resulting in a damaged core or unusually poor containment 
performance.  A functional sequence is a set of faulted functions that 
summarizes by function a set of systems faults which would result in the 
consequence of interest. 

**For those cases where only point estimates are generated, the licensee 
shall propose a suitable factor that adjusts the overall value to the 
"expected" level. 

***lE-6 denotes abbreviated scientific notation for I x 10-6. 

                                    2-1
.

                                APPENDIX 3

                            ACCIDENT MANAGEMENT

There already is an international consensus that the cause and consequences 
of a severe core damage accident can be greatly influenced by the 
operator's actions. In addition, the ability of essential equipment to 
survive the environment resulting from severe accidents is an important 
consideration in mitigating a severe core damage accident and managing its 
progression.  The failure of essential equipment can (1) incapacitate or 
remove systems needed to respond to severe accidents or (2) misinform the 
operator. 

The NRC has initiated a research program to examine the efficacy of generic 
accident management strategies.  We intend to periodically meet with 
industry (NUMARC) to compare the results of our respective programs.  
However, the staff has done some preliminary work in defining the key 
elements of a severe accident management program. 

Since your IPE results will ultimately play a significant role in the 
development of such a program for your plant, we are providing you with the
results of our work at this time.  The main elements of an accident 
management program should address: (1) the organizational responsibilities 
and structure needed to direct the responses to a severe accident, (2) the 
instrumentation, procedures, and alarms needed to diagnose severe 
accidents, and the procedures and equipment needed to accomplish the 
functions necessary to prevent and to mitigate leading accidents, and (3) 
the procedures and training needed for operators to be skilled in possible 
remedial actions. 

Suggested Elements of an Accident Management Program 

1.   Organization 

The first element of any severe accident management program is to assign 
responsibilities for dealing with these accidents and to identify the 
necessary organizational structure. 

The utility should decide which operators are to be trained to manage 
severe accidents or if a separate evaluation team is to be established to 
direct the operators.  Clear lines of decision making authority should be 
established.  For example, if containment venting is an option that could 
conceivably be considered during the course of an accident to prevent 
overpressure failure, then the person responsible for making that decision 
should be clearly identified to all involved personnel.  Analyses of 
ultimate containment strength, the venting pressure, and the advantages, 
disadvantages, and potential consequences should also have been evaluated 
beforehand, and the decision makers should be properly trained from the 
evaluation results to make an informed decision. 

2.   Instrumentation and Equipment 

Practically every aspect of plant operation is likely to be involved in 
accident management.  Coordination among the various organizational units 
is vital for communicating the status and the control of needed equipment.  
It should be clear (1) what information is needed to make decisions, (2) 
who is responsible 

                                    3-1
.

for obtaining the information, (3) what instruments plant personnel can 
rely on to determine the status of the plant, and (4) what essential 
equipment is needed to mitigate severe accidents and the time interval for 
which it is needed. Survivability of specific equipment needs to be 
evaluated by establishing whether the qualification of equipment for design 
basis events is sufficient to support the assumed performance of this 
equipment during severe accidents. 

For sequences with a significant potential to progress beyond core melt, 
means of maintaining containment integrity is the main goal.  Heat removal 
from the containment and retention of fission products are the most 
important functions. Equipment needed to accomplish these functions should 
have been identified and appropriate preparations made.  All reasonable 
preparations to enable operators to recognize approaching containment 
failure, to assess possible remedial actions, and to accomplish the 
necessary functions should be provided. Potentially adverse action should 
be identified and evaluated.  For example, recovery and initiation of 
containment sprays after the containment has a substantial quantity of 
steam and hydrogen can condense the steam and may leave a detonable mixture 
of hydrogen.  Similarly, spraying into a containment that has been vented 
could result in a vacuum and possible implosion.  

If special equipment might be needed to both prevent and mitigate severe 
accidents, provisions might be made to ensure its timely availability.  The
responsibility to take such action should be assigned, and the individuals 
responsible should know where to procure the needed equipment. 

3.   Procedures and Training 

The accident management plan should be developed to accomplish these 
functions for each set of the leading accident sequences despite the 
degraded state of the plant.  There should be consistency and smooth 
transition between the emergency operating procedures and the accident 
management plan.  The plan should be checked against the existing 
organizational structure to ensure that responsibilities for managing each 
accident are clearly defined and the responsible personnel are adequately 
trained. 

                                    3-2
.

                                APPENDIX 4

                               DOCUMENTATION

At a minimum, the following information on the IPE should be documented and
submitted to the NRC: 

1.   Certification that an IPE has been completed and documented as 
     requested by the provisions contained in this generic letter.  The 
     certification should also identify the measures taken to ensure the 
     technical adequacy of the IPE and the validation of the results, 
     including any uncertainty, sensitivity, and importance analysis. 

2.   A list of all initiating events, the containment phenomena, and the 
     damage states examined.  

3.   All function event trees and containment event trees (including 
     quantification) as well as all data (including origin and method of 
     analysis).  The fault trees (or equivalent system failure models) for 
     the systems identified, using the criteria of Appendix 2, as main 
     contributors to core damage or unusually poor containment performance 
     should also be provided.  

4.   The support state models for the IDCOR IPEMs, including descriptions 
     of all applicable findings from the visual inspections.  

5.   A description of each functional sequence selected by the criteria of 
     Appendix 2, including discussion of accident sequence progression, 
     specific assumptions, and human recovery action.  

6.   The estimated core damage frequency and the likelihood or conditional 
     probability of a large release.  The timing of significant large 
     releases for each of the leading functional sequences.  A list of 
     analysis assumptions with their basis should be provided along with 
     the source of uncertainties.  

7.   Identification of the USI(s) and GSI(s), if applicable, that have been
     assessed to estimate their contribution to the core damage frequency 
     or to unusually poor containment performance.  

8.   A description of the technical basis for resolving any USI or GSI when
     applicable.  

9.   A list of the potential improvements, if any (including equipment 
     changes as well as changes in maintenance, operating and emergency 
     procedures, surveillance, staffing, and training programs) that have 
     been selected for implementation and a schedule for their 
     implementation or that are already implemented.  Include a discussion 
     of the anticipated benefit as well as any drawbacks.  

10.  A description of the review performed by a utility party not directly 
     involved in producing the IPE to evaluate or oversee the IPE review.  

11.  Documentation on the level of licensee staff involvement in the IPE. 

                                    4-1
.

Retained Information 

The documentation pertaining to the examination that must be retained by 
the utility for the duration of the license or until superseded includes 
applicable event trees and fault trees, current versions of the system 
notebooks if applicable, walk-through reports, and the results of the 
examination.  In general, all documents essential to an audit of the 
examination should be retained.  In addition, the manner in which the 
validity of these documents has been ensured must be documented.  For any 
actions taken by the operators for which credit is allowed in the IPE, the 
licensee should establish a plant procedure, to be used by those plant 
staff responsible for managing a severe accident should one occur, that 
provides assurance that the operators can and will take the required 
action.  Plant owner groups are encouraged to develop generic guidelines 
from which utilities can develop plant-specific accident management 
programs and/or procedures. 

                                    4-2
.

                                APPENDIX 5

                 DECAY HEAT REMOVAL VULNERABILITY INSIGHTS

As part of the Unresolved Safety Issue (USI) program, six limited scope 
PRAs were performed under the USI A-45 project, "Shutdown Decay Heat 
Removal Requirements," to assess the decay heat removal (DHR) function in 
existing plants.*   The results showed that DHR-related core damage risk is 
in a range, on some plants, where attention may be warranted regarding 
whether or not such risks can be lowered in a cost-effective manner.  The 
results also showed that the sources of DHR-related core damage risk are 
highly plant specific. 

The following insights have been gained as a result of those six PRAs.  The
insights are summarized here in order to assist licensees in the conduct of
their IPEs as they relate to their search for potential core damage risk 
associated with DHR-related severe accident sequences.  Although licensees 
are requested in the generic letter to proceed with the examination only 
for internally initiated events at the present time, insights from both 
internal and external events are provided in this appendix to indicate what 
may be important to decay heat removal function vulnerabilities when 
performing the IPE for externally initiated events. 

Areas where such cost-effective improvements might be possible were 
identified for severe accident sequences initiated by 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, sabotage, and seismic events. 

Such areas for possible improvement were particularly apparent in plant 
support systems.  At the support system level, there is often 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.  These situations 
suggest the possible need to investigate corrective actions that could 
reduce the probability that single events such as a fire, flood, or insider 
sabotage could disable multiple trains (or single trains with a multiple 
purpose) thereby creating an inability to cool the plant. 




_____________________ *  See the following NUREG/CR reports:

4448,     "Shutdown Decay Heat Removal Analysis of a General Electric BWR3/ 
          Mark I," March 1987.  
4458,     "Shutdown Decay Heat Removal Analysis of a Westinghouse 2-Loop 
          Pressurized Water Reactor," March 1987.  
4713,     "Shutdown Decay Heat Removal Analysis of a Babcock and Wilcox 
          Pressurized Water Reactor," March 1987.  
4762,     "Shutdown Decay Heat Removal Analysis of a Westinghouse 3-Loop 
          Pressurized Water Reactor," March 1987.  
4767,     "Shutdown Decay Heat Removal Analysis of a General Electric 
          BWR4/Mark I," July 1987.  
4710,     "Shutdown Decay Heat Removal Analysis of a Combustion Engineering 
          Pressurized Water Reactor," July 1987. 

                                    5-1
.

Human errors were found to be of special significance.  The six studies 
modeled errors of omission (e.g., delays or failures in performing 
specified actions), and it was found that in many cases the resulting risk 
was very sensitive to the assumptions made and to the way such errors were 
modeled.  

Consequently, great care is warranted in the development of human error 
models. In addition, it is likely that errors of commission are also 
important (i.e., where the operator misdiagnoses a situation and takes an 
improper action that is not be related to the actual, current plant 
situation).  Although such "cognitive" errors are much more difficult to 
model, efforts to take them into account will result in a more complete 
picture of DHR-related risk.  

Of equal importance to human errors is the credit that is allowed for 
recovery actions, which can have a very significant effect upon the 
resulting risk.  Some of the more important recovery actions are recovering 
offsite power, fixing local faults of batteries or diesel generators, 
actuating safety systems manually, realigning auxiliary feedwater steam and 
feedwater flowpaths, and manually opening locally failed motor-operated 
valves.  Considering the importance of such human recovery actions, 
considerable effort is justified in the development of the methods and 
assumptions used in these areas. 

Transient events that are initiated or influenced by a loss of offsite 
power were found to contribute significantly to risk.  A new rule, 10 CFR 
50.63, has been issued June 21, 1988 (53 FR 23203) as a resolution to USI 
A-44, "Station Blackout." Implementation of this rule will reduce the risk 
from such events. 

For PWRs, the ability to cool the plant through "feed and bleed" operations
could have a significant effect upon the DHR-related core damage risk.  
However, care must be taken that feed and bleed operations would actually 
be undertaken in a real emergency situation in sufficient time to prevent 
core uncovery and subsequent damage.  In view of the potential benefits, 
significant effort might be justifiable in ensuring that procedures and 
training are actually in place sufficient to warrant credit for feed and 
bleed cooling.  

Just as the origins of DHR-related risk are plant specific, the effects of 
corrective actions are also quite plant specific and must be evaluated on a
plant-by-plant basis.  In choosing which potential corrective actions to 
investigate in more detail, a general principle is that the modifications 
having the highest potential for reducing the risk, for the lowest cost, 
will be those that increase the redundancy or availability of systems 
shared between units. 

In summary, both the DHR-related risk and the effects of various corrective
actions are highly plant specific.  The dominant risks are divided between 
internal and external causes, and the areas of support systems and human 
response are of particular significance.  Studies show that various cost-
effective corrective actions may be possible to reduce DHR-related core 
damage risk after its source has been identified. 

                                    5-2
.

                               ATTACHMENT-1

         CLOSURE OF SEVERE ACCIDENT ISSUES FOR OPERATING REACTORS

                       (Excerpted from SECY 88-147)

The Commission has ongoing a number of programs related to severe accident 
behavior in operating light water reactors.  Each program addresses a 
specific aspect of severe accident behavior and may in fact result in a 
proposed specific action on the part of the staff or Commission towards the 
regulated industry. However, neither the staff nor Commission has yet 
defined for the industry which programs are critical to resolving the 
severe accident issues for their plants and what specific steps must be 
taken by each licensee to achieve this resolution. 

Completion of this resolution process is termed "closure" of severe 
accident issues.  Actions resulting from two tracks; namely, generic issues 
and plant-specific issues, must be taken for severe accident closure.  
Closure for generic severe accident issues will be obtained when the 
Commission takes action in the form of rulemaking, or states whatever its 
required approach is.  Closure for plant-specific severe accident issues 
will be obtained when each licensee has completed certain evaluations and 
implemented certain programs such that events which comprise the dominant 
contributions to risk for each plant are identified and that practical 
enhancements to the design, procedures, and operation are made such that 
further improvements can no longer be justified by backfit analysis 
pursuant to 10 CFR 50.109.  However, specific plant and operational 
improvements may be identified which do not meet the backfit rule, but if 
implemented, would significantly alter the risk profile of the plant, 
improve the balance of reliance on both prevention and mitigation, or 
substantively reduce uncertainties in our understanding.  Any such 
improvements identified will be brought forward to the Commission with 
recommended action on a case-by-case basis.  Closure of a single issue or 
combination of issues is achieved when the above is satisfied for that 
issue or those issues addressed.  

It should be noted that "closure" does not imply that all severe accident 
activities will cease.  Certain activities, such as research in the areas 
of severe accident phenomena and human performance will continue beyond 
"closure." These activities are designed to provide confirmation of 
previous judgments.  It is expected that as a result of continuing 
research, experience, and other activities, additional issues or questions 
regarding judgments related to severe accidents may arise.  These will be 
considered and disposed of on a case-by-case basis, and are not expected to 
bring into question the previous conclusions regarding closure.  

The following sections describe in detail the steps that each licensee is 
expected to complete in order to achieve severe accident closure for each 
of its operating reactors. 

                                    A1-1
.

1.   Completing Individual Plant Examinations (IPEs)

The IPE program is intended to be "an integrated systematic approach to an 
examination of each nuclear power plant now operating or under construction 
for possible significant risk contributors (sometimes called "outliers") 
that might be plant specific and might be missed absent a systematic 
search."  

Each licensee is expected to perform an IPE using a method acceptable to 
the staff.  As will be described in the staff generic letter implementing 
the IPE, the staff expects that in many cases utilities, in the performance 
of their IPEs, may find and will voluntarily remedy uncovered 
vulnerabilities by making the necessary safety improvements (conforming to 
the requirements of 10 CFR 50.59).  However, through the review of IPE 
submittals, the staff may find it necessary to employ established 
plant-specific backfit criteria to assure that justifiable corrections are 
made. 

For the phase of the evaluation associated with identification of dominant 
core melt sequences (commonly referred to as the "front end" analysis of a 
PRA), there is little controversy regarding methods, and we expect the 
industry decision process with respect to potential modifications to be 
straightforward. For the phase of the evaluation associated with core 
melting, release of molten core to the containment, and containment 
performance, the staff recognizes that for a few of the phenomena, notably 
in areas which affect containment performance, there is a wide range of 
views about their relative probability as well as their consequences.  For 
these issues additional research and evaluations will be needed to help 
reduce the wide range of uncertainties. Because of concern over the ability 
of containments to perform well during some severe accidents, the staff is 
conducting a Containment Performance Improvements Program (for more details 
see Item 3 below).  This program complements the IPE program and is 
intended to focus on resolving generic containment challenges, including 
issues associated with the phenomena mentioned above. 

The NRC and industry currently have ongoing research programs to address 
these few issues.  However, until a sufficient understanding of these 
phenomena is developed, each licensee will be faced with the need to be 
able to understand the potential range of probabilities and consequences 
associated with these issues. 

Accordingly, we would expect each licensee to implement a Severe Accident 
Management Program which provides training and guidance to their 
operational and technical staff on understanding and recognizing the 
potential consequences of these phenomena. 

We do not plan to require a licensee to consider external events in its IPE 
at this time.  The staff is currently studying methods it would find 
acceptable for examining plants for severe accident vulnerabilities from 
external events, and will be meeting with NUMARC regarding these methods as 
well as the scope of an external event examination.  We expect completion 
of the methods development within 12 to 18 months.  Closure with respect to 
external events will be achieved upon completion of an examination of each 
plant, as needed, for external event vulnerabilities consistent with the 
conclusions of the staff studies described above. 


                                    A1-2
.

2.   Accident Management. 

The staff has concluded that significant risk reductions can be achieved 
through effective severe accident management.  We also believe that the IPE 
conclusions reached by licensees for their plants will explicitly rely on 
certain operator actions, or on operators not taking actions which could 
adversely affect both the probability and consequences of a severe 
accident. 

Hence, a key element to severe accident closure for each plant will be the 
implementation of a Severe Accident Management Program.  Since information 
on severe accident phenomena and effective accident management strategies 
will continue to be developed by both NRC and industry over the next 
several,years, closure is not predicated on having a "complete" accident 
management program in place.  Rather, closure is based on each licensee 
having an Accident Management Program framework in place, that can be 
expanded, modified, etc. to accommodate new information as it is developed.  

3.   Containment Performance Improvements 

As a result of concerns related to the ability of containments to withstand 
some generic challenges associated with severe accidents, the staff has 
undertaken a program to determine what, if any, actions should be taken to 
reduce the vulnerability of containments to severe accident challenges, and 
to reduce the magnitude of releases that might result from such challenges. 

Staff efforts have first focused on the BWR MARK I containment.  The staff 
studies are primarily focused on the potential generic vulnerabilities of 
these containments, and not plant unique vulnerabilities, which is the 
primary focus of the IPEs.  The staff schedule calls for an interim report 
on BWR MARK Is to be submitted to the Commission in June of this year, with 
final recommendations due in the fall of this year.  The other types of 
containments are to be assessed by the fall of 1989. 

The IPE generic letter is now expected to be issued by July of this year, 
and licensees will have approximately four months to respond identifying 
their plan for conducting the IPEs.  Following the four-month period, it is 
expected they will commence with their IPEs.  It is further expected that 
any modifications to Mark I containments that the staff may recommend will 
be available to the industry before they start their IPEs.  For the other 
containment types, the fact that any staff recommendations will not be 
available until after they have commenced with their IPEs is a concern.  
However, the IPE generic letter will state that the staff does not expect 
the industry to make any major modifications to their containments until 
the information associated with the generic issues which affect containment 
performance has been developed by the staff.  Hence, the industry will not 
be placed in a position of having to implement improvements before all 
containment performance decisions have been made. 

4.   Use of Safety Goal in the Closure Process 

The staff expects to use safety goal policy and objectives, including the 
10(-6)/reactor-year "large release" guideline, to assist in the resolution 
and 10 closure of severe accident issues.  Resolution and closure of issues 
are expected to be of two different types, either plant unique or generic.  
Safety 

                                    A1-3
.

goals and objectives are to be used only for the resolution of generic 
issues, i.e., severe accident issues common to a defined generic class of 
plants. Resolution of plant unique issues is to be accomplished on a case 
by case basis,using the information developed by Individual Plant 
Examinations (IPE) as is described in Section 1. 

The staff is preparing a Safety Goal Policy Implementation Plan (Revised) 
that incorporates the following, as directed by the Commission (Staff 
Requirements Memorandum dated November 6, 1987): 

(1)  Information on how the staff proposes to implement OGC guidance on the 
     use of averted on-site costs in backfit analyses.

(2)  Whether averted off-site property damage costs should be included in a 
     more explicit manner in backfit analyses.

(3)  Whether $1,000/person-rem remains an appropriate cost/benefit 
     criterion.

(4)  A discussion of options for defining a "large release." 

(5)  A discussion of options for specifying appropriate plant performance 
     objectives.

(6)  Responses to Commissioner Bernthal's questions regarding population 
     density considerations, and whether it would be acceptable for a plant 
     to have no containment if it met the large release criterion by 
     prevention of core melt (core damage) alone.

This plan will also reflect the consideration given by the staff to ACRS 
recommendations and the results of several meetings with the ACRS on this 
subject. 

Resolution of severe accident generic issues using safety goal objectives 
is expected to proceed as follows.  PRA information from a variety of 
sources, including both staff generated PRAs, (e.g., NUREG-1150) and 
utility generated PRAs (IPE) will be used to make comparisons with 
applicable safety goal objectives in accordance with the implementation 
plan.  The staff will identify the reasons why particular plants appear to 
meet or not meet these objectives and assess these reasons in relation to 
current regulatory requirements.  This assessment will constitute a testing 
of the effectiveness of these requirements or their implementation and is 
expected to result in the identification of potential changes to regulatory 
requirements that, for some plants, would be expected to result in safety 
enhancements.  These, in turn, will be subject to appropriate regulatory 
analysis as provided in the Commission's backfit rule 10 CFR 50.109.  Those 
that can be shown to provide substantial safety benefit and are 
cost-effective will be proposed to the Commission for backfit, possibly in 
the form of rulemaking.  The staff expects that this process would have no 
impact on classes of plants for which there is reasonable assurance that 
safety goal objectives are met.  This expectation is based upon the intent 
to identify those features of design and/or performance that are already in 
place at plants meeting safety goal objectives and to structure any new 
requirements such that they do not require changes or additions at these 
plants. 

                                    A1-4
.

The staff's revised Safety Goal Implementation Plan is scheduled to reach 
the Commission in August, 1988.  The first application is expected to be 
reflected in the staff's recommendations to the Commission in the Fall of 
1988 on potential improvements to BWR MARK I severe accident containment 
performance. 

5.   Summary of Closure Process 

In summary, the steps which each licensee is expected to take to achieve 
closure on severe accidents for its plants are as follows: 

o    Complete the IPEs; identify potential improvements, evaluate and fix 
     as appropriate. 

o    Develop and implement a framework for an Accident Management Program 
     that can accommodate new information as it is developed. 

o    Implement any Commission-approved generic requirements resulting from 
     the staff Containment Performance Improvement Program; this should 
     constitute closure of containment performance generic issues.   

While programs for improved plant operations and research in the area of 
severe accidents will continue, completion of the above by a licensee is 
considered to constitute "closure" of the severe accident issue for the 
plant in question. Specific issues that may arise in the future as a result 
of ongoing research will be treated on a case-by-case basis and will not 
affect the closure process. 

                                    A1-5
.

                               ATTACHMENT 2

    LIST OF REFERENCES OF THE IDCOR PROGRAM REPORTS AND KEY NRC REPORTS

                               IDCOR Reports

Tech. Report No.                                 Title

1.1           Safety Goal/Evaluation Implications for IDCOR 
2.1           Ground Rules for Industry Degraded Rule Making Program 
3.1           Define Initial Likely Sequences 
3.2           Assess Dominant Sequences 
3.3           Selection of Dominant Sequences 
4.1           Containment Event Trees 
5.1           Human Error Effects on Dominant Sequences 
6.1           Risk Significant Profile for ESF and Other Equipment 
7.1           Baseline Risk Profile for Current Generation Plants 
9.1           Preventive Methods to Arrest Sequences of Events 
              Prior to Core Damage w/Revision 1 
10.1          Containment Structural Capability of LWRs 
11.1/11.5     Estimation of Fission Product and Core Material 
              Characteristics 
11.2          Identifying Pathways of Fission Product Transport
11.3          Fission Product Transport in Degraded Core Accidents 
11.6          Resuspension of Deposited Aerosols 
11.7          FAI Aerosol Correlation 
12.1          Hydrogen Generation During Severe Core Damage Sequences 
12.2          Hydrogen Distribution in Reactor Containment Buildings 
12.3          Hydrogen Combustion in Reactor Containment Buildings 
13.2-3        Evaluation of Means to Prevent, Suppress or Control 
              Hydrogen Burning in Reactor Containments 
14.1A         Key Phenomenological Models for Assessing Explosive 
              Steam Generation Rates 
14.1B         Key Phenomenological Models for Assessing Non-Explosive 
              Steam Generation Rates 
15.1          Analysis of In-Vessel Core Melt Progression 
15.1A         In-Vessel Core Melt Progression Phenomena 
15.1B         In Vessel Core Melt Progression Phenomena 
15.2A         Effect of Core Melt Accidents on PWRs with Top Entry 
              Instruments 
15.2B         Final Report on Debris Coolability, Vessel Penetration, 
              and Debris Dispersal 
15.3          Core-Concrete Interactions 
16.1          Assess Available Codes, Define Use and Follow and 
              Support Ongoing Activities 
16.1A         Review of MAAP PWR and BWR Codes  
16.2-3        MAAP Modular Accident Analysis Program User's Manual, 
              Vols. I & II
16.4          Analysis to Support MAAP Phenomenological Models 
17            Equipment Survivability 

                                    A2-1

.

                         ATTACHMENT 2 (Continued)

17.5          Draft Final Report: An Investigation of 
              High-Temperature Accident Conditions for Mark-1 
              Containment Vessels 
18.1          Evaluation of Atmospheric and Liquid Pathway Dose 
18.2          Completion of Conditional Complementary Cumulative 
              Distribution Functions 
19.1          Alternate Containment Concepts 
20.1          Core Retention Devices 
21.1          Risk Reduction Potential 
22.1          Safe Stable States 
23.1          Uncertainty Studies for PB, GG, Zion, Sequoyah 
23.1B         Peach Bottom - Integrated Containment Analysis 
23.1Z         Zion - Integrated Containment Analysis 
23.1S         Sequoyah - Integrated Containment Analysis 
23.1GG        Grand Gulf - Integrated Containment Analysis 
23.4          MAAP Uncertainty Analysis 
23.5          Containment Bypass Analysis 
24.4          Operator Response to Severe Accidents 
85.1          IDCOR 85 Program Plan 
85.2          Technical Support for Issue Resolution 
85.3          IPEM A1 Thru B2 
              IPE Applications PB, Susquehanna, Zion, Oconee, 
              BWR User's Guide  
85.4          Reassessment of Emergency Planning Requirements 
              With Present Source Terms 
85.5A         Revised Source Terms 
85.5B         Source Terms and Emergency Planning 
86.20C        Verification of IPE for Oconee 
86.3A2        IPE Source Term Methodology for PWRs 
86.3B2        IPE Source term Methodology for BWRs 
86.20G        Verification of IPE for Grand Gulf 
86.25H        Verification of IPE for Shoreham

                                    A2-2
.

                      NRC and NRC Contractor Reports

Tech. Report No.             Title 

NUREG-0956                   Reassessment of the Technical Bases for 
                             Estimating Source Term 
NUREG-1032                   Evaluation of ion Blackout Accidents at 
                             Nuclear Power Plants 
NUREG-1037                   Containment Performance Working Group Report 
NUREG-1079                   Estimates of Early Containment Loads from Core 
                             Melt Accidents 
NUREG-1116                   A Review of the Current Understanding of the
                             Potential for Containment Failure from 
                             In-Vessel Steam Explosions 
NUREG-1150 Volumes 1-3       Reactor Risk Reference Document 
NUREG-1265                   Uncertainty Papers on Severe Accident Source 
                             Terms 
NUREG/CR-2300                PRA Proceed Guide 
NUREG/CR-2815                Probabilistic Safety Assessment Procedures 
                             Guide 
NUREG/CR-4177 Volumes 1-2    Management of Severe Accidents 
NUREG/CR-4458                Shutdown Decay Heat Removal Analysis of a 
                             Westinghouse 2-Loop PWR 
NUREG/CR-4550 Volumes 1-4    Analysis of Core Damage Frequency from 
                             Internal Events 
NUREG/CR-4551 Volumes 1-4    Evaluation of Severe Accident Risks and the
                             Potential for Risk Reduction 
NUREG/CR-4696                Containment Venting Analysis for the Peach 
                             Bottom Atomic Power Station 
NUREG/CR-4700 Volumes 1-4    Containment Event Analysis for Postulated 
                             Severe Accidents 
NUREG/CR-4767                Shutdown Decay Heat Removal Analysis of a GE 
                             BWR4/Mark I 
NUREG/CR-4881                Fission Product Release Characteristics into 
                             Containment Under Design Basis and Severe 
                             Accident Conditions 
NUREG/CR-4883                Review of Research on Uncertainties in 
                             Estimates of Source Terms from Severe 
                             Accidents in Nuclear Power Plants 
NUREG/CR-4920 Volumes 1-5    Assessment of Severe Accident Prevention and 
                             Mitigation Features 
NUREG/CR-5132                Severe Accident Insights Report

                                    A2-3
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