ML20235W768
| ML20235W768 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1976 |
| From: | Adensam E, Carter J, Goller K NRC |
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| ML20235W743 | List: |
| References | |
| FOIA-87-620 NUDOCS 8710190018 | |
| Download: ML20235W768 (73) | |
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I TASK FORCE REPORT ON THE SYSTEMATIC EVALUATION OF OPERATING NUCLEAR POWER PLANTS i
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1 TASK FORCE K. Goller, Chairman E. Adensam J. Carter
. D. Davis B. Grimes ]
D. Mcdonald R. Purple R. Silver R. Stuart D. Ziemann i
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. TABLE OF CONTENTS Page 1
I.- INTRODUCTION'
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SUMMARY
AND RECOMMENDATIONS 3 A. Plants ' Licensed in the Future 3.
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B. . .Previously LApproved Plants 5
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C.. , Implementation 6
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, , III.. THE. EVOLUTION OF LICENSING REVIEWS AND REQUIREMENTS A.: . Background'. 8 B.- Ad_ Hoc Upgrading: 10
'C. Terminol ogy 11
~14 IV.- OBJECTIVES AND SCOPE OF THE SYSTEMATIC EVALUATION OF OPERATING UUCLEAR POWER PLANTS A. Obj ectives 14 B. Scope of the Evaluation 18
" 23 V.- ALTERNATIVE APPROACHES FOR THE SYSTEMATIC EVALUATION 0F OPERATING NUCLEAR POWER PLANTS A. Alternative Approaches 23 B. Combinations of Various Approaches 27 C. Implementation Approaches 28 1;
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i S. TAB LE : 0F . CONTENTS ' ( con t' d) .' '
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'VI. SEL' ECTION OF RECOMMENDED APPROACH 30-
~ A. ' Decision Criteria 30 B. l Approaches. Selected t. c Con.ideration 33
.c -C.. Recommended Approach 41' j% ., o VII. IMPLEMENTATION OF RECOMMENDED APPROACH 43' A. Scope and Priority of Program 43 B. Initial Phase of.the Systematic Evaluation 44
' Program-
;C. Procedures for Systematic Evaluation of 47- l Individual. Plants O.' : Resources and Schedules 51 f
E. Other . Considerations 58-
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1 APPENDICES l Appendix A - Task Force For Rereview Program for Operating Nuclear Power Plants
. Appendix B - Possible Safety Issues' for Comprehensive Topic List ,i v
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3 , I. INTRODUCTION ! On May 3,1976, the Director, Division of Operating Reactors, established a Task Force
- to develop a " program plan for (1) evaluating licensed j nuclear power plants against' current criteria, and (2) developing a: l framework from which backfitting (10 CFR 50.109) decisions can be evaluated considering all plant features relating to safety." The q
Task Force was composed of representatives having a~ broad spectrum - j
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of technical and management backgrounds. Consultation was provided I to the Task Force by representatives from other NRC Offices and NRR. .l
-Divisions. l The Task Force was directed'to examine all reasonable alternative approaches for the programs, to recommend one approach, to develop an implementation plan for that approach and to present the study and recommendations to the Office of Nuclear Reactor Regulation I
(NRR) Management, the Advisory Committee on Reactor Safeguards (ACRS), and, when appropriate, the Commissioners. The present l report has been prepared in response to that directive. i The need for a systematic evaluation of operating plants in the light of current knowledge and licensing practices has been recognized for some time and, in fact, gave impetus to the creation of the Division of Operating Reactors within the Office of Nuclear Reactor
*The full text of the Task Force directive is given in Appendix A of_this report.
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. Regulation. The staff,'in the .recent Joint Committee on Atomic Energy (JCAE) hearings into charges relating to nuclear reactor safety, acknowledged that the documentation of the' bases for some past decisions on safety issues was incomplete. The ACRS, in reporting on its independent review of the same charges, recommended that the staff improve its documentation of why backfitting to current . licensing criteria is not required *.
LThe-program plan developed by the Task Force is called "The Systematic Evaluation Program". The title is' intended to give an accurate characterization of the proposed program, i.e., a documented, systematic evaluation of the design and operating requirements of-licensed nuclear power plants with respect to current NRC-regulations and licensing requirements to determine whether safety-related improvements are warranted. I 1
*The ACRS had also recommended, in its reports of June 14, 1966, August 18, 1967, November 14, 1970 and April 16,1976, that a !
periodic review (every 10 years) of operating experience at operating reactors, be conducted. Such a periodic review of operating experience is included as one element of the Task Force's proposed program. l I I j
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SUMMARY
AND RECOMMENDATIONS _
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,The Task Force first exam'ined the need for a systematic evaluation
,of operating reactors by reviewing the evolution of licensing require-ments*, recent ACRS recommendations for more comprehensive reviews of. ;
licensed operating power plants and past ACRS recommendations for 1 periodic reviews of operating experience. The Task Force concluded ! 1 that a systematic evaluation program is needed and that its objectives f should include: an integrated and documented assessment of each licensed operating plant's safety to determine whether that plant ; i meets current licensing requirements; the development of the infonnation- J J base needed for balanced and prompt decision making regarding backfitting; l early identification of any significant safety deficiencies; and efficient utilization of NRC and industry resources. A. Plants Licensed in the Future In reviewing the evolution of the licensing requirements, the i Task Force found that no systematic program was in place to evaluate ] i l previously approved plant designs as new licensing requirements l were established. Action was on an ad hoc basis and generally limited to issues of obvious safety significance. Furthennore, Ms used in this Report " licensing requirements" include both the ; requirements set forth in NRC regulations and the recommendations i of NRC staff as set forth in Regulatory Guides and Standard Review 1 P1ans. I l l l I I _ _ _ __________ ___ i
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current staff evaluations at the ' operating license stage have: i' not adequately. documented either the deviations from current licensing requirements'or the bases for acceptability of those
. deviations.
The Task For e recommeMnds* that, a.t .the earliest practical date, U / the bases for^ali deviationo from the Standard Review Plan be. )< documented in each operating license Safety Evalnation Report. Speciai attention should .be given to documenting depauures from the Standard Review Plan Acceptance Criteria. The Task Force also'found that procedures were.not developed to ! J systematically evaluate and document the status of operating. g plants with respect to new changes in licensing requirements, m s
' The Task For recommends t) tat procedures be established in the s /
Division of OpcAating Reactors to systematicatty evaluate operating 1 plants with respect to all new licensing requirements identified by the Regulatory Requirements Review Committee iRRRC) for backfitting con 6ideAation. j l 1 While the RRRC provides an effective means of identifying the backfit potential of new licensing requirements for operating
*As discussed later in this Report, some of the Task Force reconnendations relating to in-house actions are now being implemented.
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l l l l plants, staff knowledge of the bases for not requiring backfitting l would improve the efficiency and effectiveness of any systematic
'N evaluation of operating plants.
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The Task For recommends that t.he basis for the RRRC decisions not to require backfLtting of new ticensing requiresnents on ; operating plants be documented. j 1
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By implementing the recommendations above, all plants will be evaluated and documented against all new licensing requirements 4 approved by the RRRC and plants licensed in the future will have documented the bases for acceptance of all deviations from current licensing requirements. B. Previously Approved Plants The Task Force considered what should be done to systematically evaluate and document previously approved reactors to today's licensing requirements. The selection and priority of plants to be evaluated was considered. The Task Force estimated that approximately 70 nuclear power plants with Safety Evaluation Reports (SERs) for operating licenses issued prior to January 1977 would require some degree of systematic evaluation with respect to today's licensing requirements. The Task Force concluded that the first priority for performing these evaluations should be assigned to plants with the earliest construction permit date. I
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. The evaluations would be based on systematic review according to a carefully selected . list of safety topics. Alternative approaches ;
were. considered ranging in comprehensiveness fromo' ur current.ad hoc reviews to complete operating license-type reviews. Different; 1 approaches for different age groups of, plants were also-considered. l The approach that best satisfied the Task Force's selection criteria l is a review and evaluation of each operating plant based on a list j of selected safety. issues or topics. One topic which would be i included is 'the. plant operating experience, as recommended by the ACRS'in the past.
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The Task Force recommends dat each current operating plant be reviewed and evaluated wiu respect to a selected list of safety issues or topics to confirm the safety of de plant, to document de basis for backfitting or not backfitting each plant to curred licensing requirements, and to provide an information base for I balanced decision making with respect to backfitting in essenilat , (' safety areas. C. Implementation A safety evaluation documenting the bases for acceptability and identifying any essential area where improvement is needed would be prepared for each facility reviewed. Although not part of its charter, the Task Force feels that early development of backfit i __ -__.____.__m__-__._._.-_ u _ _ - _ "._ _ _ _ _ - - . -
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lF decision. criteria or . guidance would increase the consistency and efficiency .of the evaluations.
' To implement the evaluation, experienced staff personnel. on inter >
disciplinary teams ~ should review groups of plants of'similar type an'd vintage. The Task Force estimates t.ne total effort required
' to implement -such a. program to 'be at least 170 man years over 'a six-year period.
- . ThelTask.Forc'e. posed certain ' egal questions that implementation
-of:the proposed systematic ev luation program night raise. OELD 1
- does not anticipate 'a major c bstacle to_ such a' program as it is.
. presently' envi sa ged.
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l 8_ 111. THE EVOLUTION OF LICENSING REVIEWS AND REQUIREMENTS A. Background i In the early days of the civilian nuclear power industry, the Conmission's regulatory staff review of the acceptability of ; proposed designs was based on less documented design information i than is pres'ently required of applicants. In addition, the Commission's " standards of acceptability" were established during each new licensing review. Many of these " standards" were not for-malized; rather they evolved as a precedent from each new licensing action. As the nunber of applications for construction permits increased, more uniform and consistent guidelines for acceptability I were necessary. ! 2 In 1966, the Commission issued the " Guide to the Organization and Contents of Safety Analysis Reports". From this point on, the amount of documented information submitted in support of construction permit applications increasd significantly. Applicants were reouired to address specified areas of safety concern and to present specified data and analyses that provide the staff the infomation needed for requlatory eval ua tion. Subsequent guidance was provided to applicants in the I l
" Standard Format and Content of Safety Analysis Reports for Nuclear 1
Power Plants", initially issued in February 1972 and revised in ) October 1972 and September 1975. This document has resulted in even more extensive documentation of each proposed reactor design. q l l l i
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' In'a similar way,' th'e criteria used cy-the staff to determine licensing acceptability have evolved over the years. In 1967, the Commission
. published for comment the' proposed. General Design Criteria ~(GDC)~-
. for nuclear power plants. Although these criteria, which established-the' minimum requirements for' the principal design criteria'.for water-cooled nuclear power plants were not fonnally adopted in the regulations until 1971, they have been used since 1967 by the regulatory staff in reviewing new plant applications
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, ' written in such general terms that, frequently, they w t;6 Cfq fied by.several ~ alternative design techniques.
. In a further effort to provide guidance to the industry and to increase-staff review efficiency and effectiveness, the Commission began issuing Safety Guides in 1970. These guides presented methods I
acceptable. to the regulatory staff for implementing specific parts of the regulations, including the GDC. In 1972 the Safety Guides j were superseded by the broader-based Regulatory Guides. At present more than 120 Regulatory Guides relating to power reactor safety have been issued. Regulatory Guides are not substitutes for regulations and compliance with them is not a legal requirement. Methods and solutions different from those set out in the guides are acceptable, ( l if they provide a basis for the findings requisite to the issuance i l or continuance of a permit or license by the Commission. 1 1
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M y As the knowledge of nuclear power plant design increased and operating lG
. experience began' to be accumulated, and as' the regulatory staff was
' enlarged. to review and evaluate an increasing number of applications, .
it became necessary to' document-the methods and criteria used by the -
. regulatory staff in the licensing review. Accordingly, the Commission staff developed and published the Standard Review Plans (SRP) in 1975. The'SRP provide specific guidance to NRC staff reviewers to improve the ' quality and uniformity of.the review and' evaluation of t
nuclear. power plant safety, they also provide the public and the nuclear industry with a clearer understanding of the staff review process.
- 8. Ad Hoc Upgrading Each of the plants reviewed and approved in the past was found to be in conformance with the licensing requirements in effect at the i time of licensing. As noted above, however, the requirements have i
changed and become more detailed over the years. Consequently, if 1 a licensed facility were now to be reviewed by current -licensing l
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procedures it might be found to be at variance; the older the plant 1 d ' the more at variance it would likely be. Vhether or not the variances would have significant safety implications is our primary concern. The i program to systematically evaluate operating plants must be directed toward early resolution of that concern. 1 I l
7-g g n u.' 11; - flany:important safety issues'have' aiready 'heen evaluated with r' espect to the implications for each older. reactor and, where necessary, essential safety._ improvements were backfitted. - Examples of" such issues include. ECCS performance, postulated breaks in high-energy-lines and flooding of safety equipment. Consideration of_ the safety implications of certain of these issys has even lead to the shutdown of individual facilities. Howeve'r, as.noted, these evaluations were, and continue , 1 to !)e, of an ad hoc nature. Io requiring p1 ant changes and safety improvements on an ad hoc basis, considerable- staf f and industry resources have been involved.
'(an' AIF spokesman indicated that tha annual backfitting cost per plant has been in excess of $1 million). The Task Force recognized that unless
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a systematic and comprehensive evaluation of each plant was undertaken this ad' hoc procedure would continue. The Task Force feels that
,-, such a procedure is not as cost-ef fective as an integrated evaluation of the entire plant. Moreover, innovative approaches to resolve ,
I several issues at one time are possible with an integrated evaluation. I
, C. Terminology The Task Force, in its early deliberations, discovered that discussions l related to " criteria", " requirements" and " regulations", quickly became confusing because of a 1ack of a common interpretation of the 1
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' definitions of these terms for.use in the study.
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-As used in_ this report, the term " licensing requirements"'means I all of those conditions or constraints that the NRC'-licensing .
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process (staff, ACRS, and hearing boards) decides' must be _ imposed ) l 7- - before a construction permit will be granted or an operating license j i issued. The term includes all. the explicit requirements set forth j
.. .a in the regulations (as in' Appendix J,- Appendix K, etc.). These (
will be called " regulation requirements". It includes all the !
" acceptance criteria" and " Branch Technical Positions" in the SRP and the staff positions identified in the Regulatory Guides (recognizing, of course, that alternative methods to achieve the desired result can be proposed and accepted). The term "licensina requirements" is not limited to design features but includes 4
administrative control requirements as well (e.g., physical security, quality assurance, or emergency plans). It includes industry standards and codes to the extent that these standards and codes are endorsed by NRC in the regulations and in Regulatory Guides.
" Regulation requirements" must be met, as specified, by all licensees (unless relief-is obtained by formal exemption). Similarly, all
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i requirements imposod as license conditions by hearing boards or j by the Director of Project Management exercisina his authority - .l l in contested proceedings, are obligatory. The remainder of the
" licensing requirements", (including the various acceptance criteria, 1
-l Regulatory Guides and staff positions) however, are relatively flexible i
in the sense that the need for compliance rests largely on the technical judenent of the staff. There can be many instances during the licensing review of a plant where the NRC staff reviews and accepts features (or methods) differing from those presented in the SRP or Regulatory i Guides because the staff has found that tcey provide a basis for l 1 the findings requisite to the issuance or continuance of a license. ! l Such variances have not been consistently documented in staff . safety i evaluation reports; they would be revealed in a systematic evaluation i of any currently approved plant. Moreover, the variances are not likely to reduce the level of safety, eve- though significant staff j
.1 effort may be required to reconstruct and document the basis for their i i
earlier acceptance. Because of the evolutionary nature of NRC's licensing process, there j is a growing backlog of safety-related plant design features that j were never systematically reviewed and evaluated following promul-qation of new licensing requirements. The primary objective of the q systematic evaluation program is to perforin this evaluation on operating l i pl a nts. In the following section, tne scope and purposes of the j progran are discussed. 1 1 1
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' 'IV . OBJECTIVES AND . SCOPE OF THE' SYSTEMATIC EVALUATION OF OPERATING
~RUCLEAR POWEP. PLANTS-A. Objectives The: principal . objective of the program .which ~ the Task Force was charged:with developing is to evaluate licensed nuclear power plants with. respect to current safety criteria and requirements and 'to provide clear supporting documentation.. The documentation is to include the rationale for'any departures from current criteria.-
The Task Force subdivided the objective into six parts 1which were used as decision criteria when evaluating the alternative program app'r oaches (described 'in 'Section VI). Each of these sub-objectives
.or decision criteria is discussed belcw.
- 1. Safety Assessment The systematic. evaluation program must assess the adequacy of the safety-related aspects of the design and operation of all currently licensed nuclear power reactors. An important part of this. assess-ment is documentation of the process by which the assessment is made and the bases for concluding that the safety-related aspects
- of each plant are adequate. The Task Force recognizes that it l should be possible to draw conclusions of adequacy for most facilities without a full-scale licensing review because documentation of safety adequacy already exists for many
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-g 2. Documentation with Respect .to Current Licensing Requirements q The ' program should document how each operating plant meets -I I
current licensing requirecients and the rationale' for. departures. j r , from;these requirements = Thi s -objective 'was made.' separate' from . j
'that of safety' assessment because the Task Force found that certain alternative programs had the potential for making a 1
documented case with respect to safety adequacy:without J establishing'how each plant compared ~to current criteria. For example, if a WASH-1400 risk analysis were carried out" j 1 for each facility (and assuming that an " acceptable" level- !
.'of risk could .be. defined), a well-docunented case for safety
- adequacy could result. However,' thi s documentation,; which
'would likely be in the form of fault trees and probabilistic .]
consequence evaluations, would not provide a means for comparing j 1 the plant design and operation against the acceptance criteria' j l
. currently used in licensing reviews.
- 3. Balanced Backfitting Decisions
. The program should provide for balanced decisions with respect to any backfitting. If more than one backfitting issue is being considered, it is possible that measures taken to resolve one issue may ' depend on how otner issues are resolved. Integrated j
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evaluation of each plant can provide the necessary information to assure adequate safety margins and allow licensees to address i several backfitting issues at once. 1 i 1 l
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l-l' L , . l , 1 BackfittingLdecisions should also be ccnsidered :in the context ' of' increasing the level of; protection of the' public at~each
. facility regardless of vintage or location. ' These decisions R may vary substantially from the oldest, low power facilities j in sparsely. populated areas to-recently licensed higher power J
facilities in more. densely populated areas. A rationale must ]
.be provided, however,'for differentiation in backfitting treatment. =First,, a " minimum acceptable level .of safety" may be required regardless of the impact on the facility.
This approach might be appropriate .with' respect to certain types of. ECCS protection and natural phenomena design, but ~ the " minimum acceptable level" might still vary between groups of' units because, for example, lower power reactors may present significantly less potential hazard. Second, an impact /value rationale might be used to detennine which of several additional i safety measures should be required. This approach may, of course,
, lead to different backfitting requirerrents for facilities of the
'same vintage and. power level at different sites in certain cases. j l;
- 4. Early Identification of Significant Deficiencies The program should be structured so that significant deficiencies l can be identified and pursued early in the program. The entire evaluation of a particular facility should not have to be performed
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- before.thelmore significant safety-related issues can be identified and corrective action can be taken... This pennits priority imple ;
L mentation ~of the most important modifications if backfit decisions
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.- 5. . Resource Requirewnts I'-
The program should utilize available resources efficiently and.
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l minimize requirements for commitment of additional NRC and industry resources. Although'it appears to the Task Force that I any program which fulfills the other objectives ~ set forth above. will' require significant additional resources, the objective of minimiring resource requirements is an important consideration in deciding between alternative approaches which may meet other obj ectives.
- 6. Elapsed Time to Complete Each systematic evaluation program considered by the Task Force .,
l was studied with respect to the probable time for completion. : { 1 Assuming results of comparable quality are achieved, a program , amenable to rapid completion was considered more desirable than l
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.B'.. Scope of the Evaluation 3' : As changes to licensing requirements .have been made, the. need for f.
making each new requirement applicable 'to' operating plants has not been systematically evaluated. Continuation of this practice will: result in La steadily increasing number of facilities that need an-updating review to assess the facility against today's.1icensing ); requirements. To stop1this. increase, steps have-been taken by_ NRR management to ~ assure that all deviations from . licensing
' # requirements and the bases for their acceptance are documented-in future operating license reviews. (NRR Office Letter No.' 9,
" Documentation of Departures from Standard Review Plan",' June 18, 1
-1976).. In addition to this, all new licensing ' requirements which
' are identified .by the Regulatory Requirements Review Committee (RRRC) as applicable to operating facilities will be assessed for each facility end the conclusions documented. It is anticipated that these two actions will be fully implemented by January 1,1977.
Assuming implementation of this new policy begins in January 1977, about 72 plants could be involved in the systematic evaluation program. Of these, 62 plants have an operating license today, 4 more will have both an OL and a SE't issued before January 1977, and 6 others will o f
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' i, 35 7%s have an SER only issued before January 1977. The first SER expected to be issued after January 1,1977 is Hatch-2. All plants receiving
-an SER prior to Hatch-2 may be included in the systematic evaluation l
program. Staff reviews prior to Hatch-2 have been based on many requirements and procedures' that are still acceptable. During the staff review prior to issuing the SER, decisions were made concerning the accepta-bility of approaches other than those recommended by the staff. These decisions, and especially their bases, have not always been documented in the SER. 1 An attempt was made to find groupings of nuclear power plants that could simplify the systematic evaluation process. Because of the evolving nature of both the licensing review process and the licensing requirements, useful groupings were not obvious. Table 1 lists significant licensing milestones that were considered for developing possible groupings. A plant-by-plant examination is required for most . of the milestones to establish when the staff review was actually performed. Regulations that have explicit backfit statements may However, group facilities by the date of either the CP or OL. because compliance with the regulations is mandatory and the staff has already completed these backfit reviews, a grouping according to regulation does not appear useful.
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TABLE 1
" LICENSING MILESTONES' i'
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' Guide for Preparation of SARs . June 1966 General Design. Criteria published proposal July 1967 3 1.,
- \tn NEPA 1969
.y July 1970
"\ - Quality Assurance Criteria - Appendix B y November 1970 Begin Publishing Safety Guides May 1971 General' Design' Criteria - Appendix A ECCS. - Interim' Acceptance Criteria June 1971 July 1971
~ Codes and Standards (50.55a)
Rev. O February 1972 ' Standard Format & Content of SAR Rev. 1 October 1972 Standard Fonnat & Content of SAR December 1972 Begin Regulatory Guide Program September 1973 Anticipated Transient Without Scram ( ATWS) January 1974 1 ECCS - Appendix X August 1974 WASH-1400 Draft Report Standard Technical' Specifications first issued October 1974
.(D.~C. Cook).
Rev. 2 September 1975
" o#. Standard Format & Content of SAR September 1975 Standard Review Plan October 1975 WASH-1400 Final Report I
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A milestone that does appear useful is the 1966 SAk Guide. : Prior. to 1966. the content or' scope of information to be provided in the applicant's SAR was not specified. .Thus the review perfonned for; In addition, the early licensed plants varied from current practice. between August.1962 to March 1967 no reactors were licensed ~ for operation. As a' consequence there seems to be a natural break,. in terms of staff review, between reactors licensed for operation r-b'efore and after San Onofre Unit No.1 (1967). Table 2 lists operating plants by date the operating license 'was issued. The WASH-1400 study also recognized that a change occurred in the type
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of facility being licensed in 1967. As a result the WASH-1400 study' encompassed only those facilities, i.e., San Onofre Unit No. I and after, and excluded the older facilities with an operating license (Dresden l J 1, Indian Point 1, Yankee Rowe, Big Rock, Humboldt Bay, and Lacrosse). Of these plants, Indian hint 1 is shut down pending a licensee decision whether to extensivel; modify the facility or to decommission it. Therefore, there are only f tve such "old" (pre-San Onofre) facilities in operation. I l
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'" 1 TABLE 2 LISTING OF OPERATING FACILITIES i l
1 Date of t. .ce Cate of taen nee of Full Power of Full fover License License A. Licensed to Overste Licensed to Operate A.
- 20. Tennessee valley Authority 12/20/73 FTL )
- 1. Commonwealth Edison Company 06/02/60 POL i 10/14/60 FTL Browns Ferry 1 (BWR)
Dresden 1 (SW) 08/09/73 m l 07/29/60 POL 31. Omaha Public Power District
- 2. Tankee Atomit Power Company Pt. Calhoun (PWR) 06/23/61 PTL 09/26/73 m Tankee-Rowe (rba) ., 32. Consolidated Edison Company
- 3. Coneotidated Edison Company 01/26/62 POL
- Indian Point 2 (PWR) 10/06/?) PTL ladise totat 1 (M) 08/28/62 POL 33. Duke Power Company
- 6. Pacific Cas 6 11u tric Company Oconee 2 (PVR) puebeldt Bay ( M ) 01/21/69 FTL 10/19/73 m 08/30/62 POL 34. Commonwealth Edison Company
- 5. Consur.-re Power Company Eton 1 ( M )
Sig Rock Point (9WR) 05/01/64 FTL 10/25/73 PTL j 03/27/67 POL
- 35. Philadelphia Electric company >
- 6. Southern Calitarnia Edison Company Peach letton 2 (SWR) 11/14/73 FTL )
6ea Onofre 1 ( M ) 06/30/67 POL 36. Commonwealth Edison Co.
- 2. Connecticut Tankee Atomic Power Company Eton 2 (PWR) 12/27/74 FTL 04/05/74 FTL Baddam Neck (PWR) 37. Northern States Power Co.
- 8. Dairyland Cooperative Power 07/03/67 POL
- Prairie Isisad 1 (PWR) 12/21/73 PTL }
Lacrosse (BWRL 38. Public Service Co.'of Colorado
"- -T"~7sPsfy'Te~n'TY1TPower 6 Light Company. 08/01/69 POL
- Pt 5t vrain (UTCR) 12/21/73 PTL oyster Creek 1 (8WR) 39. Wisconsin Public Service Corp
- 10. Utagara Mohawn Power Corporation 08/22/69 POL 12/26/74 FTL Kewaunee (PWR)
Mine Mile Potet 1 (SWR) 01/18/74 PTL
- 11. Rochester Cao 6 Electric Corp. 09/19/69 POL
- 40. Nebruka Public Fower District Cooper station (SWR)
R.3, Ginna ( M ) 61. Iowa Electric Light 6 Power Co. 02/22/74 P7L
- 12. Commonwealth Edison Company 12/22/69 POL
- Arnold (8.fR) 06/19/74 FTL Dresden 2 (SWR) 42, Metropolitan Edison Company
% Carolina Power 6 Light company 09/23/70 PTL Three Kile Isir.nd 1 (PWR) 05/21/74 FTL E.3. Robinson 2 (fVR) 01/19/71 POL
- 43. Arkansaa Power & Light Co.
. Northern States Power Company Arkanssa 1 ( M )
Montice11e (SWR) 64. Tennusu valley Authority 06/28/74 PTL
- 15. Wisconsin-Michisan fower company 20/05/70 FTL Browns Ferry 2 (BWR)
Fotnt Beach 1 (PVR) 45. Philadelphia Electric Co. 07/02/74 FTL
- 16. Northeast Nuclear Energy Company 10/26/70 POL' s Peach letton 3 (SWR) 07/19/14 m I M111 stone 1 (SWR) 03/02/71 PTL 46. Duke Power Company
- 17. Commonwealth Edison Company l Oconee 3 (PWM) !
Dresden 3 (8VR) 12/14/72 FTL 67, Baltimore Can & Electric Co. 07/31/74 FTL 18, Commonwealth Edison Company Calvert Clif f s 1 (PWR) 08/06/74 f L f Quad-cittee 1 (8WR) 10/16/72 POL
- 46. Georgia Power Co.
- 19. Consumers Power Company f Natch 1 (8VA) 08/16/74 F*L Palisades (M) 12/14/72 PTL 49. Sacramento Municipal Utilities District
- 20. Cosssonvulth Edison company I Rancho Seco (PVR) j Quad-Cities 2 (BWR) 05/25/72 FTL 50. Power Authority of the State of hew Tork 10/17/74 PTL
- 21. Virginia tiectric e tower Company sorry 1 ( M ) Fic: Patrick (SVR) 10/25/74 F*L f
!!, Florida Power 6 Light Comparty 07/19/72 FTL 51, Indiana & Michir,an Electric Co. j Turkey Point 3(M) c.,g g ( m )
09/15/72 FTL 10/29/74 FTL
- 23. Soston tdison Company 52. Northern States Power Co.
P11stie (BVR) Prairie Island 2 (PWR) 12/27/74 m
- 24. Vermont Yankee pucioar Power Corporation 02/26/73 FTL 53. Carolina Power & Light Co.
Vermont Tankee (BWR) Brunswick 2 (Se/R) 08/01/75 PTL ! 01/29/73 FTL 54. Northeast nucinar Energy Crepany
- 25. Virginia Electric 6 Power Company Surry 2 (M) 02/06/,73 FTL Hilletone 2 (PWR) 11/21/75 PTL !
- 26. Duka Power Cospany * *
- 55. Portland Ceneral Electric Co.
Oconee 1 ( M ) Trop a (rVR) 04/05/76 m ,
- 27. Winonsin-Michtsan Electric Company 03/08/71 ffL 56. Power Authority of the state of New York Point luch 2 (M) Indian Point 3 (PWR) 01/30/76 FTL j 28, Florida Fower 6 Light Company 04/10/73 FTL 57. Duquesne Light Company Turkey Point a (PWR) go , ,y,gg,y g gpgg) 06/29/73 FTL 58. Flordia Power 6 tight 03/01/76 P!L -
- 20. Maine Tankee Atomic Power Company Maine Tankee ( M )
gg,g,,g, g (pyg)
- 59. Tennusee Valley Authority 07/02/76 FTL ,
Browns Ferry 3 (SWR) ) 06/13/76
- 60. Baltimore cas & Electric Co. Fuel Leading l
Calvert Clif f s 2 (PWR) 08/13/76
- 61. Public Service Electric 6 Cas Co. Fuel Leading
$ ales 1 (PWR) 09/06/76
- 62. Caroltr.a Power 6 Light Compsey
*P a* applied f or a Full Ters License Brunewick 1 (SWE) fuel Loading 1
rovtalonel Operatir.g License (ull-Ters License J4l i I
p H V. ALTERNATIVE APPROACHES FOR THE SYSTEMATIC EVALVATION OF OPERATING NUCLEAR POWER PLANTS A. Alternative Approaches This section of the report describes various alternative approaches
.for a systematic evaluation progran and discusses possible combinations of these alternatives and implementation approaches.- The next section of this report describes the decision criteria utilized, compares the alternatives and their combinations, and presents a recommended approach and the basis for its selection.
- 1. Status Quo One alternative is to continue past practice in which the applicability of existing and new licensing requirements or safety issues to operating plants or plants under OL review is decided on an ad hoc basis.
I Consideration of whether specific plants must meet licensing require- 1 l ments instituted after the facility operating license was received is limited to generic safety issues and new requirements generally ] i selected on the basis of their perceived safety significance. Many d 1 I times the determination of which safety issue is pursued or dropped is made by management or staff reviewers without a documented basis. In the past there was no requirement for publishing or preparing a ] written record of such decisions and generally no such record has been kept. However, in some cases staff safety evaluations for q operating licenses or post-operating license amendments do discuss alternative resolutions for a certain licensing requirement or i
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- 24 .
y safety issues. The ad hoc approach has provided a mechanism for d raising issues which are considered by individual reviewers to have great enough safety significance to pursue on operating plants.
- 2. Selected Licensing Requirements or Safety Issues The basic difference between this approach and maintaining the status quo is the performance of a systematic, comprehensive review of current licensing requirements or existing safety issues followed t
by selection of those issues of sufficient significance to be included in the systematic evaluation program for operating nuclear power plants. In developing the list, the basis for including or deleting licensing. requirements or ~ safety issues would 'have to be documented. The development of a selected list is not within the scope of the Task Force's assignment. However, several methods were considered that could be used in defining the selected list. The methods could be used individually or in conbi.9ation and applied on a plant-by-plant or plant-grouping basis. (a) Culled SRP List of Licensing Requirements or Safety Issues l
\ j Using a culled SRP list would require a review of the SRP ;
1 (and other criteria) to identify the licensing requir6.4ents and staff acceptance criteria that would contribute signi-ficantly to overall safety when applied to operating plants. i i
)
1The' list of selected licensing requirements would then be sent to the licensees and evaluated:for each specific L ' f acility design. If essential safety issues were uncovered, the reed to modify a plant could be evaluated. a ( b) Selected Safety Issues This method would require developing a list of generic safety issues or topics based on the Commission's experience in
; regulating nuclear power plants. An example of a topical listing is the letter from R. F. Fraley to P. A. Morris dated January 13, 1971. This listing identified the topics of concern that the Advisory Comittee on Reactor Safeguards-( ACRS) had in relation to the conversion of Provisional Operating Licenses (POL) to Full Tenn Licenses.
In evaluating an operating plant, each safety issue or topic j on the list would be reviewed in the depth indicated by the relevant SRP. (c) Use of Standard Technical Specifications This method would involve updating, to the extent practical, 1 each operating plant that has customized technical specifi-- l cations to the format and content of the Standard Technical
Specifications (STS). The uniform documentation of all the operating plants technical , specifications in the STS format would provide a common basis for systematically identifying significant areas of non-conformance to current. l licensing requirements. s (d) WASH-1400 Reactor Safety Study (Limited Application)
- The WASH-1400 Reactor Safety Study provides a quantitative g risk assessment of potential accidents at nuclear power plants. The risk assessment includes consideration of the. probability and consequences of potential accidents.
The results of the study are estimates for an " average plant", i.e., an average plant that represents 100 reactors starting with San Onofre, Unit 1. The results of the study could be used as a basis for determining the extent of the ; l review required for plants after San Onofre Unit 1, or plants after the two plants included in the detailed WASH-1400 Reactor Safety Study.
- 3. WASH-1400 Reactor Safety Study This approach would involve a complete accident risk assessment j; ,
for each of the operating plants, using the methodology of the WASH-1400 Reactor Safety Study. The detailed information that was required for the two plants included in the WASH-1400 Study L___ _A ___ __ _ _ _ -_
b
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would have. to b'e provided by the licensee for each of the operating plants. This approach would provide a quantitative L 'l risk assessment for each plant reviewed, which would provide [ a basis for backfit decisions.
- 4. De Novo Review This approach would be a complete Operating License review of each operating plant utilizing the current licensing requirements L
as documented in the Standard Review Plan. The licensees would be required to provide a complete Final Safety Analysis Report in accordance to the latest version of the Standard Fonnat and Content for Safety Analysis Reports (Regulatory Guide 1.70). B. Combinations of Various Approaches Each of the basic approaches to systematic evaluations of operating plants may be used independently or in combinations. Several basic approaches best lend themselves to accomplishment of certain objectives of the evaluation program. Combinations of these approaches may provide an optimum means to accomplish several objectives. For instance, several basic approaches result in early identification of safety concerns whereas others provide a more thorough identifi-cation of shortcomings although identified later. As an example, a short list of significant licensing concerns could be given evaluation priority followed thereafter with an NRC staff or licensee assessment of the remaining licensit, conc erns.
'Certain basic evaluation approaches are most suited to certain vintages of plants. For- example, safety analysis documentation is minimal on the five oldest currently operating nuclear power facilities. For these facilities 'the approach could provide for adequate FSAR documentation by use of a complete or de-novo OL review. For newer facilities, one of the other evaluation approaches could be selected. Therefore, in selecting the alternative or combination of alternatives, the Task Force considered the relative capability of that combination to accomplish the evaluation program objectives for several vintages of plants.
i C. Implementation Approaches There are several methods of implementing the selected basic approach ; or combination thereof. In implementing the selected evaluation approach, both NRC and industry manpower and organizational arrange-ments that are needed to apply the licensing requirements must be taken into account. The licensing requirements may also be applied individually, in total, or in groups with pre-established orforities which may vary from plant to plant or among types of plants. I The implementation method used should be that which is best suited to optimize the accomplishment of the evaluation program objectives. . l In particular, the implementation technique should be cost-effective i l
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1
l and timely and avoid unnecessary impact on the industry and the NRC staff. To minimize the impact on the current NRC staff, task , groups could be fomed within the staff and consultants could j
- provide input to these groups. Likewise, the industry could fom groups to study generic topics of mutual concern.
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F i
, VI. SELECTION OF RECOMMENDED APPROACH To. identify the alternative that would provide the most effective. andL efficient approach for meeting the objectives of the systematic evalua-tion program, the alternatives discussed in Section V'were considered.
d1one and in combination against the decision criteria as discussed' immediately below. In all . cases it was recognized that should a significant safety issue not bp implemented promptly by any alternative, an ad hoc review of the safety issue may go forward when; appropriate. Descriptions of how the decision criteria were used, which- alternatives. or combinations of alternatives were considered, and how the Task Force performed the comparative evaluation are presented in this section. The results of the comparative evaluation are summarized in Figure 1 and the- ! recommended approach and its implementation are discussed in Section VII. A. Decision Criteria The evaluation of each alternative was based on the following decision cri teria:
- 1. Documentation with respect to current licensing requirements. l There are two levels of documentation to consider: a) documenta-tion of design adequacy and b) documentation of the canparison with current criteria. These were considered together in evaluating each alternative. Documentation of design adequacy is considered ;
to be the type of documentation already employed by the staff to describe resolution of current safety issues, i.e., the staff has l detennined that operation with a system, as designed, will present
7_ [ no undue. risk to public health or safety. Documentation of the f' ~ 7/' comparison of system design.against current criteria is not
! presently available in a comprehensive or consistent manner.
- 2. Resource requirements.
'For each alternative, the additional. resources over current commit-ments, both from the staff and the industry, required to meet the objectives were weighed. Staff resource requirements were, of. l course, better understood. However, an effort to consider the -
resource impact on each licensee of the alternative being considered , was'made. i
- 3. Elapsed time to complete.
This was considered as the total time required to complete the evaluation and document the conclusions.
- 4. Early identification of significant deficiencies.
This was considered to be the ability of each alternative to provide identification of significant safety issues early in the evalustion f process.
- 5. Balance of any backfitting required.
Each alternative was considered in light of how well any decisions 4 to backfit additional requirements are balanced, e.g., will the progran assure that the backfits do not conflict with each other;
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unB si e R s nog ot ns i t istt n ct e iceim epf e Dsk r B B B A A eci C A B dR a u e B q ch e nt yR ain l wa a B y nt eoe dif i t a vaS oc s rit e Pf ni i ac C D - ot cn C B A B t nie B C efi ydi c tI ni i gf l yi e il SD a b r i Aaf r Eo e t S ir C e me E it H n T e C o d p l D D A i C C C O s em so C B B R S i pC P N c P O e a A I D l o S Et E U v L i C e T N eh s A O ct e T N C r e RI R goon urs E E G I T L C R n esf ec i T A O F z RsitL F il n A B C C C C D D O K maed S i nmn N A O T n oea ir S itif I Mi uf R d qa A d et P ARS M O g C f yon octi s a st nut nn oqcee i eacm td pi e D A aAsLr B B A A t e C C nnRt nq iu eg mih ee ust rR c ei r - oDwu dI I a r D C e t ro r l l t er e ;nof ao e cro r se at ia nn f ir i v eff l o f o icmnt uos ; eo no f0 en0 n pcii ami ;o f e O oosl usl a igf t4 0 t sO0 4 v td u l e ircul eol a y dt t r ss s1 r sf n1 id n e r - e f ea rgch tdc un o iroaSH- tS onnt i nr t e f f u os iHt fS a f w S t c ii r gi cni o sms sm s Al era ser sd m f ys ea ni eW eioW rl o . i ef s euro ul or Pd oa r t e l h r o y h d h vf es cg t nn t e n f s e scff sn R ufi Sl r f s I a o itt oa rsoc . ff ct si iiss cse aa or ce f f o ;i y s ooii em f net s rP ef o y1 oisi k i p l a s pl e pl yts e pw es wee l weoei wwoi t sat s ar es . eet c eibie ttbc pi y s ierr ieit rt ss yrt A ie c v ,fo t v i r ;l i ii vvrf a fl a eft l ii ef t i ev ;i . ors esns el o1i eeo seh o'l l lh os el ve c rr r tt l t e 0ti i R uee reOe ri caya ae d e ni d e ni 0ec t - qni i a e et ot i nt viai oairf vf rr o;it v s ra el f oc t poia etf ot t eoii 4l a 1 pf r nR s gi ul ol Sl o eur ot el cm tf cl tl -m e t ti Ni i N etS e N nt eone epnei Hoh anc crc vi- t aid l coll l mol c 5cc t eaoa eiroa el rn e i el eoiea A a l A t aa S cf Dfff Df ctl Dp<a S atda S ctdf Wae 1 2 3 4 S 6 I
i that sufficient infonnation is availa:>1e to assure that only substantial, essential improvements in safety are made; and that an integrated evaluation allows more innovative and ~ cost-effective approaches.
- 6. Safety assessment.
Although the assessment of overall safety is the overriding consideration of the program, this decision criterion was was considered as the incremental change in assured safety attained by each alternative. B. Aporoaches Selected For Consideration Each alternative was evaluated on a comparative basis with the other alternatives on a scale of A through D (A = excellent, B = adequate, C = poor and D = unacceptable) for each decision criterion. The consensus rating of the Task Force is shown on Figure 1. l
- 1. Status quo.
Because this approach continues wnat has been the NRC practice with respect to operating facilities, no additional resources would be needed if it were selected. Its most significant de-ficiency is that it is not a systematic or comprehensive program and, as a consequence, the detailed documentation of the technical bases for deviation from current licensing requirements is limited to that presented in past safety evaluation reports. The approach j I 1 I l
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should not, however, be sunnarily dismissed as not being a viable option. The safety recora of operating plants has been good, significant safety improvements have been achieved, and the system does make maximum effective use of available staff resources. The approach rates low in the objective of achieving balanced backfit decisions since review of each safety topic is initiated as the need arises without particular regard to its importance relative to other safety concerns. Also, because it is not a comprehensive program and does not provide for systematic reconciliation of plant design with all current licensing requirements, it rates low with respect to elapsed time to compl ete. >
- 2. Detailed review of first five plants; rely on findings of WASH-1400 for plants f rom San Onofre 1 on.
Because of the detailed review necessary for the first five plants, more staff effort would be required for them than that of the status quo approach. However, since the total effort would only be about one-fifth of that required for Alternatives 5 and 6, and much l less than that compared to Alternatives 7, this alternative was l l given a relatively good rating for additional staff resources ] requi red. l l l l l l 1 I I j
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q For most licensed plants (i.e., those from San ~0nofre on for O
- which WASH-1400 conclusions are -declared sufficient) .this approach - l l
will not identify any significant design concerns. . Because an - f averaging philosophy is presented by WASH-1400, and it' provides .j
.the bases for plant acceptability, some backfitting might be j I
justified for some 'of the plants if a detailed review were to . . be' performed. 'In addition, some concern has been expressed : l
- that the WASH-1400 ' estimate of the probability of core melt of 5 x 10-5.per reactor year is not low enough. Accordingly, this approach is rated low for "Resulting in Improved Safety" and "Early Identification of Significant Safety Deficiencies".
The balance achieved with respect to backfitting and the assurance of increased safety would be excellent.for those plants with de novo reviews but poor for all other plants. l
- 3. Detailed review of the first five plants; review of selected I criteria or topics for San Onofre i to Surry 1; rely on findings of WASH-1400 for plants af ter Surry 1.
Similar to approach 2 except that the evaluation of plants between San Onofre 1 and Surry 1 will involve reviewing against selected topics. The selected topics are envisioned to be less than that considered for Alternative 5. Any method, or a combination of several methods described in Section V, could L 1
L be used to identify the selected topics. Those plants, after i Surry 1, would be declared acceptable based on the WASH-1400 conclusions. l In comparison with Alternative 2, then, this alternative is better in " Balance", "Resulting Improvement in Safety" and
" Documentation of Design Adequacy Against Current Requirements" but because of the lack of review of certain plants, it was not rated " excellent" in any of these categories. Since a more thorough evaluation is required (than Alternative 2), it is rated lower for additional resources required.
- 4. Detailed review of first five plants; review of selected criteria or topics from San Onofre 1 on.
Similar to approach 2 and 3 except that all plants after San Onofre will be reviewed against selected topics. ! Since all plants would receive some oegree of evaluation, this alternative rates higher than Alternatives 2 or 3 l in "Resulting Improvement in Safety" and " Documentation". However, the Task Force believes that it should not be rated as high in these categories as Alternatives 5, 6 and 8. This alternative requires greater staff resources than l l either Alternative 2 or 3. l
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_ _ _ _ _______________ - - _ w
1 I d-f I l The selected topics are envisioned to be less than that considered for Alternative 5 and so this approach does not rate as high in the area of " Balance" as alternatives 5 and 6.
- 5. Culled complete topical list. 1 j
Since this approach starts with a complete listing of all i safety concerns (topics) that the staff evaluates in licensing new facilities, it is rated excellent in supporting the objective of improving the safety of licensed reactors. The written bases for deletion of any topics will result in a comprehensive record of how each licensed plant meets current licensing requirements. The record, however, will not be as explicit as would result from a de novo OL-type ! review ( Alternative 8), nor would it be as directly relateable to current licensing requirements as would a review done I with the SRP as a guide ( Alternative 6). Nonethel ess, because it would result in a complete, documented review of significant safety issues, it was rated as excellent (with respect to documentation) by the Task Force. Probably the most difficult phase of this approach is generating the list of topics to be reviewed and culling that list to create a final list that includes only topics judged to be of sufficient safety significance as to warrant 4
eval ua tion. Opinion among the Task Force members, and among other staff members who were queried, varied considerably i as to the feasibility of being able to develop such a list of topics. Some believed that it would be an effort as difficult as that which went into the development of the high confidence that the list would be complete (particularly with respect to the SRP); and others were not sure of the usefulness of the effort since, in the end, each safety concern would be judged against the detailed criteria of the SRP. The Task Force spent a few days starting a list to confirm the feasibility of this approach. Appendix B is a tentative and and incomplete listing of possible topics. Those in favor of the approach felt that the use of specific safety topics would greatly enhance the ability to provide i for early identification of significant safety concerns and would provide more flexibility (compar+d to the " acceptance q criteria" of the SRP) in dealing with variances from current licensing requirements that may be revealed during the evaluation process. The approach would also provide a more 3 practicable framework on which to make the balanced assessments I needed to decide backfit questions. l i I t
/
This approach would require a' considerable effort, both on the part'of the staff and the licensees. The Task Force concluded that it could.not be 'done effectively in any reasonable peiiod of time with the existing NRR staff. Even with an. augmented staff to handle the increased workload, the program would take several years to complete. More; definitive estimates of manpower impacts are presented later in this report.
- 6. Culled list of SRP criteria.
. In .this alternative, each operating plant would be fully documented as to how its design conforms to current staff I'
review criteria (or why it does not have to); the alternative, therefore, rates high under " Documentation".- As a large effort-is required to evaluate all plants, this alternative would involve considerable manpower and take a significant amount of time to complete (the time depending, to soue extent, l on the available resources). The amount of effort required j l i
'is difficult at this time to estimate since it is uncertain how much " culling" can be achieved.
The Task Force believes that this alternative is not as amenable ; i
- to providing early identification of significant safety concerns l as is Alternative 5. A topical approach ( Alternative 5) is believed more likely to highlight significant problems ]
1 l l II l
0- .j
-.40 -- '!
' sooner than an approach that utilizes the detail 'of the SRP' accept'ance cri teria.
~
'7. . WASH-1400 type review.
s This alternative requires an effort greater than an OL review. I In addition, very few people are familiar.with the analysis techniques used in WASH-1400. Accordingly, it would be very _. l i
' difficult to conduct a detailed review of all operating facilities and, for this reason, the alternative was rated lowest under " Resources Required" and " Elapsed Time". I i
The results of a WASH-1400 type analysis.are in tenns of risk. If a' specific quantitative level of. risk were to be defined as " acceptable", it is conceivable that a plant 1
.i whose design is not in conformance with many of today's ]
1 licensing requirements might be found to have a lower level l of rirk. Because the WASH-1400 type analysis detennines acceptability on the basis of risk, it would not indicate how each plant conforms to today's licensing requirements. The fault-tree analysis approach inherent in a WASH-1400 I type analysis provides a basis for balanced backfit decisions. Accordingly, the alternative rates high for this decision l 1 c ri terion. I l l l t. l
a l l I i De Novo review of all plants. J 8.
' Since this approach entails a complete and detailed review identical to an OL review, it woula result in maximum identification of potential safety concerns and in the most complete documentation of findings. It woul d, however,- ]
result in substantial expense for licensees (estimates run ] between $2 to $10 million just to produce a new FSAR, which would be only the beginning of the effort) and would require j substantial additional staff resources. Because of the large amount of time that would be involved and the mass l
}
of information that
- 11d have to be developed, it is not likely that this a u rnative would provide early identification of significant safety concerns. On the other hand, because of the completeness of the infonnation that would be developed, balance in backfitting decisions should be relatively easy to achieve.
C. Recommended Approach 4 The comparative evaluation presented in Figure 1 shows that Alternative 5, Culled Complete Topical List is superior in meeting the decision criteria. Accordingly, the Task Force recommmends this approach for the systematic evaluation program. It offers the most advantages compared to other alternatives despite its larger resource requirements and the time required
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4 i I l l, l to complete the overall task. The Task Force felt this option offered the potential for properly assessing safety while still l maintaining flexibility to tailor each review to fit the specific circumstances of the individual plant (e.g., its age and location). The excellent rating of the approach with respect to providing ) a basis for balanced backfit decisions was also viewed by the Task Force as an important consideration in support of the )
)
i recommendations. 1 i I i l 1 l l
1
' VII. -IMPLEMENTATION OF RECOMMENDED APPROACH This section presents a program for initiation of the approach recommended by the Task Force to accomplish the systematic evaluation, and a preliminary estinate of the schedule and NRC manpower requirements for implementation.
A. Scope and Priority of Program 1 All facilities for which an operating license safety evaluation was issued prior to January 1977 are candidates for a systematic evaluation process. In general, the required level of staff and industry effort for each plant will be a strong function of the I length of time from past staff reviews of that plant. As indicated in NRR Office Letter No. 9, all facilities receiving an OL-SER after January 1977 will docur..ent the basis for all deviations from the acceptance criteria of the Standard Review Plan. To , preclude the need for a future systematic evaluation program involving these facilities, D0R is establishing procedures for prompt identification and application, where appropriate, of changes in regulatory requirements and licensing criteria. I In addition, these procedures assure that, as new information relevant to plant safety is developed, its implications with respect to licensed plants will be promptly evaluated. ! l In an effort to establish an order of priority for conducting I the systematic evaluation of operating plants for which SER !
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i s dated prior to January 1977, a number of factors were cm n:idered. These included: (1) age of facility, (2) date of last NRC review, (3) power level, (4) power density, (5) existence of a full term or provisional operating license, (6) present operating status, (7) special safety considerations and (8) any other special consideration (such as the population density or the recent ACRS request to re-review Zion by June of 1977). After consideration of these factors as they apply to various plants, it was concluded that the dominant factor was age of the facility; therefore, except for Indian Point 1 i which is shut down for an indefinite period, the systematic evaluation should be performed in the order that the facilities were licensed for construction. B. Initial Phase of the Systematic Evaluation Program The Task Force envisions the initial phase of the systematic evaluation, i.e., the development of a list of selected licensing requirements or safety issues, will be accomplished
- in the following sequence.
- 1. A comprehensive list of safety topics for all reactor types will be developed. This is estimated to be a 10 man-week effort to be completed by a group of D0R personnel appointed by the Director, DOR, with assistance and review by other NRR divisions. ( A 50-60% complete draft list of topics l .
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has been prepared by the Task Force as an example and is presented in Appendix C of this report).
- 2. The complete topic list applicable to all reactor types will be reduced to a list of those topics with sufficient safety significance to warrant consideration in the evaluation process. The basis for deletion of topics from the complete list will be provided. (Such a reduced topic list shall be referred to as the " systematic evaluation topic list").
Development of the basis for deletion of all the items from the complete list of topics is expected to be a three man-month effort. This task will be perfonned within DDR with assistance from other NRC personnel.
- 3. Topic lists will be prepared specifically applicable to groups of reactors, e.g., a separate list for each of the four major NSSS vendors. This will be accomplished by selecting topics from the " systematic evaluation topic list" which are applicable to a specific reactor group.
Again, the bases for deletion of topics will be documented. This is estimated to be a 3 man-week effort.
- 4. A " systematic evaluation topic list" will be prepared for a specific facility with documentation of the bases for deletion of topics, This is expected to be a ten man-week
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l' i i I I effort. This list with bases for deletion of the topics will be developed by an evaluation team including the ORpM, j OT personnel and essistance from I&E.
- 5. The " systematic evaluation topic list" concept for evaluation of operating facilities will be discussed with the Regulatory Requirements Review Committee (RRRC) and ACRS, with specific emphasis on the methods and criteria used to reduce the total list to individual plant lists. The following items should be available for presentation to the committee:
- a. Comprehensive list of safety topics applicable to all reac to rs, List of topics to be included in the systematic evaluation j b.
i with bases for deletion of topics.
- c. List of systematic evaluation topics for one reactor group )
i with bases for deletion of topics. j l I
- d. List of systematic evaluation topics for one specific facility with bases for deletion of topics.
- 6. Af ter the comments of the RRRC and ACRS have been appropriately 1
incorporated in the evaluation program, a meeting will be held with the licensee of the facility for which the plant specific l topic list was develooea to informally discuss the systematic j l
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oval'u ation program, the specific ' topics to be. reviewed for that facility, and a schedule for completing the evaluation.
- 7. 'The ' licensee will be requested, by letter,'to provide an ]
1 evaluation of the . topics identified for the specific facility. OELD will be requested to review and approve this approach. and the transmittal letter'to assure the legal bases for the l request. ;
.Following initiation of the program described above, implementation will proceed by preparing plant specific " systematic evaluation-i topic lists" for each facility to be evaluated with documentation ,
of the bases for a deletion of topics from the " systematic evaluation l topic list" for the appropriate facilities. Various methods for j
-increasing the efficiency and effectiveness of the reviews of the j 1
the plant specific lists were considered by the Task Force, including. sending review teams to the sites. 'All available documented infoma- ; tion on each plant will first be scrutinized in preparing the topic : list. A subsequent site visit by the team may be advantageous. Although a specific procedure for conducting the reviews has not i been established, it is presently envisioned that the reviews will be accomplished in the same manner as current reviews are performed. C. Procedures for Systematic Evaluation of Individual Plants After approval of the program including the topic list for a' sample facility, the actual systematic evaluation would follow. l
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i The' Task Force considered several organizational approaches to
. the evaluation process. The creation of evaluation teams seemed to offer several advantages over other organizational approaches. . In recommending this approach, the Task Force-considered it essential that (1) the team should be isolated from the day-to-day DDR responsibilities, and (2) the organizational
< structure should provide a means for assuring balanced backfit decisions. Two potential disadvantages of using evaluation teams are the potential for technical inconsistency in applying i
safety criteria (between teams themselves and between teams and other DOR reviewers) and the lack of a reserve of engineering experience from which to draw. Both of these disadvantages could be overcome by having team members administratively report to their normal 00R organization, in which case the ; management structure can assure consistency in applying criteria and each individual on the team can draw from the reservoir of { d experience of the branch to which he is assigned. The professional make-up of the teams would depend upon the specific technical topics chosen for evaluation. A team is expected to consist of ! an average of 6-7 members including the ORPM. The assistance of , i an I&E inspector would be included to the degree possible. J Af ter an evaluation team is assembled, its first step would be to prepare the topic list for the specific facility assigned to 1 1
/
it. The bases for all deletions of topics would be documented. l After the topic list is prepared, an informal meeting would be held with the licensee to discuss the various topics on the list. l l As a follow-up to the meeting, additional issues could be deleted and the bases for their deletion established. At this point, the evaluation topic list would be forwarded for NRC management review and approval . Generic items on the topic list which licensees and industry groups wished to separate and resolve through test programs may be deleted from the topic list for each individual plant provided the schedule and approach for generic resolution is accepted by the staff. After management approval is obtained, the list of topics is submitted to the licensee with an explanation of the safety concern and a request for a comparison against current criteria in the SRP and a description of any design changes the ifcensee feels are needed. Where current methods of meeting the acceptance criteria are not used, the licensee will be requested to provide justification for their use. The specific schedule for response will be aetermined af ter discussion with the licensee. Significant safety issues would be rated higher in priority and given a shorter response time. In general, the responses to all issues would be expected within
' 12 months for the older plants with shorter schedules possible for
g , P e n
',f",
L In the intervening time, while the
~
1-more recently' licensed plants. licensee-is preparing its response, the team may prepare topics _.
' lists for other facilities and meet with other licensees. - The g i
same procedure would be followed in these activities. When a j 1 licensee responds with a significant amount of the information f requested (about 80%), the evaluation to assess the adequacy- f i of safety margins relevant to each topic can begin. l
)
For those items where it is detennined that the existing safety margins are adequate, the bases for this decision should be )
)
, documented as soon as available. Items requiring potential backfit decisions or additional information should be grouped together. f (Prompt action will be taken if any item is considered to involve P
an undue risk to the public safety). All additional information H needed to reach a potential backfit decision would be obtained { prior to initiating the decisional process. This allows all the backfit judgments affecting a plant to be grouped together for proper balancing. Guidance or criteria for reaching a backfit decision would be used at this point. The evaluation team would then consider all items which potentially warrant backfitting and balance these judgments over the entire plant to assure that the final backfit decision will result in significant increase in the level of safety. The impact of the proposed backfitting versus the impact of other ways l
/ -
t.
- 51.-
a (e.g., administrative restrictions, added surveillance or lower power levels) of achieving acceptable levels of safety should be considered. Innovative approaches to resolve several issues at one time would be encouraged. The schedule for implementation should also be considered. In general, prompt action should be required if a substantial safety question is uncovered. Otherwise the implementation schedule should recognize the availability of interim measures, such as inspection, while modification proceeds. The next step in the systematic evaluation process.would be to submit any items where backfitting is recommended to management for review. The management review would consist of the usual D0R approval process for such matters. The entire evaluation process for a specific plant would be subject to licensee appeal to 00R management at any time. This would take the form of the appeal process available for CP and OL reviews. If resolution is not reached in DOR, the appeal ~ "~ ~ would continue to the Director of NRR. D. Resources and Schedules The Task Force's recommended implementation program is based on the premise stated in Section IV that about 70 nuclear power 3 plants may require some form of systematic evaluation. The 1
l 1 { I i i program is also based on the premise that the evaluation of two or more identical reactors at the same site with cps issued I within about one year of each other can be considered as one j evaluation for purposes of resource commmitment estimation. Based I on these two premises, the program would involve about 50 separate { eval ua tions. The program summarized below does not include the N.,4 followup efforts required to initiate and review any resulting l 1 plant modifications. l I i The plan is based on the use of teams of experienced professionals, I with each team performing 5 evaluations within a three-year j period. The first 4 teams, reviewing the first twenty reactors would each include 6 engineers. Thereafter, it is anticipated that the full time equivalent team size could be reduced to four men each. Smaller team sizes after the first four teams are projected because it is anticipated that the scope of evaluation in more recently licensed plants will be smal?er, and the teams will be able to take advantage of the experience gained in the earlier evaluations. ) l The schedule for each evaluation is projected to include about a ; i two-year span and is expected to follow a course similar to past full-term operating license reviews. The sequence involves identification of topics, a letter to the licensee, a response 1 l 1
-__-__ - l
i from the licensee, two rounds of staff questions and ~ issuance of a staff SER. Based on past experience, the schedule allows one year between issuance of the NRC. letter and the licensee response and.another year for staff review, questions, licensee reply to questions and preparation of a staff safety evaluation.- Past experience indicates the actual schedules from review to: review may be highly variable. The schedules assume no public hearings are conducted. The overall schedule assumes that the initial two teams will begin plant-specific reviews in January 1977 and two teams will be added every six months through June 1979. Each team is ' expected to initiate a new review every two months until five reviews are underway. With the proposed plan, the first two safety evaluations would be issued January 1979 and all safety evaluations would be issued by July 1982. The major constraint extending the overall schedule ! is expected to be time required for licensees and their consultants l to respond to topic lists and questions. The first twenty plants, which would include the oldest reactors and reactors with provisional operating licenses, would be completed in 1980. The review sequence assumes that reactors with the earliest dates of construction permit issuance would be reviewed first. It also l
- - - - = - - -- . _ - - _
i o assumes that each team would review only.BWRs or PWRs~ to increase L .eff ii c ency and effectiveness of the review team. Base'd on the program and schedule' outlined above, about 170 professional man-years would be required to perfonn the systematic evaluation program. The manpower estimate and time for-completion are considered to be minimum numbers which would be increased if unanticipated situations,.such as public hearings, occur. The number of personnel and time required would also be expected to be significantly greater if experienced personnel cannot be utilized. for the program. The proposed chronology of the review effort is shown in Figure
- 2. Table 3 includes an initial proposal for the plants to be assigned to each team..
An alternate implementation plan was considered in which complete topic lists would be issued to all licensees with plants subject to evaluation and each licensee's reply would be evaluated as it is received. The intent of such an approach would be to shorten the overall program and identify significant safety issues early. The Task Force concluded that this approach would not be as effective or efficient as the approach outlined above. The approach outlined above would determine the sequence of review
55 - REVIEW 0F CHRONOLOGY
SUMMARY
0 '
' LETTER Og . 02 i 'i 8
' ' ' DET AILED SC*EDULE LETTER SER 'SER TEAM :1
' LETTER ' SdH l l ,' SER
_ LETTER ' SdR SER LETTER' SR SER l I l
- 1/1/77 1/1/78 1/1/79 1/1/80 78 63 73' CUMULATIVE 78 REVIEWS
- OF REACTOR FIRST REVIEWS REVIEWS COMPl.6TE ON A/1/79 COMPLE TED ALL REVIEWS COMPLETE ON 7/1/82 37 19 7
1 2 l I f I 1/1/77 1/1/78 1/1/79 1/1/80 1/1/81 1/1/82 . 1/1/83
~
- OF 168 ADDITIONAL
- - ADDITIONAL PRO F E SSION AL
~
PE RSONN E L MAN YE ARS. 56 - IN EACH HALF
- 48 YEAR INTERVAL 40 44 -
24 32 24 12 16 1/1/77 1/1/78 1/1/79 1/1/80 1/1/81 1/1/82 1/1/83 _~_ n OF TE AMS iP.
- 12 -
10 10 EACH HALF 8 8 2 6 6
] , YE AR INTE RVAL 1/1/77 1/1/78 1/1/79 1/1/80 1/1/81 1/1/82 1/1/83 TWO TEAMS ADDED EVERY 6 MONTHS.
4 SIZE OF E ACH TE AM 7P 4 DECRE ASES TO 4 6P 4 AFTER 20 PLANTS 48 4 SCHEDULING ARE UNDER REVIEW. 4 AND SIZE E ACH TE AM REVIEWS SP OF TE AMS 5 RE ACTORS WITHIN 4P 4 A 3 YE AR PERIOD. 3P 4 , IDENTICAL RE ACTORS 38 4 AT SAME STATION ARE COUNTED AS 1 UNIT. 2 6 1P 6 IB 6 I l l 1/1/77 1/1/78 1/1/79 1/1/80 1/1/81 1/1/82 1/1/83
, figure 2
/ -
7* y -TABLE 3 TENTATIVE-TEAM' ASSIGNMENTS
.[Date of CP Issuance in Parentheses Unless'Otherwise Noted]~
~
BWR Teams PWR Teams
\ Team 1B - 6 men Team 1P - 6 men-
\
Dresden-1 (5/4/56) . Yankee Rowe (11/4/57) . Big Rock Point (5/31/60) San Onofre 1. (3/2/64 )* Humboldt Bay (11/9/60)' ConnecticutYankee(5/26/64)- Lacrosse (3/29/63)* Ginna (4/25/66)* Oyster Creek (12/15/64)* Indian Point 2(10/14/66) Team 2B - 6 men Team 2P - 6 men Nine Mile Point (4/12/65) Palisades (3/14/67)* _ Dresden 2/3 (1/10/66* 10/14/66) Robinson 2 (4/13/67) Millstone 1 (5/19/66)* Turkey Point 3/4 (4/27/67) Quad Cities 1/2 (2/15/67) Point Beach 1/2(7/19/67;7/25/68) Monticello(6/19/67)* Oconee 1/2/3(11/6/67) Team 3B - 4 men Team 3P - 4 men _ Browns Ferry 1/2/3 (5/10/67; 7/31/68) Diablo Canyon 1/2(4/23/68; i 12/9/70) Vermont Yankee (12/11/67) Three Mile Island 1 (5/18/68) Peach Bottom 2/3 (1/31/68) FortCalhoun(6/7/68) Cooper (6/4/68) Prairie Island 1/2(6/25/68) Pilgrim 1 (8/26/68) Surry1/2(6/25/68) Team 4B - 4 men Team 4P - 4 men Hatch 1 (9/30/69) Kewaunee (8/6/68) Brunswick 1/2(2/7/70) Crystal River 3(9/25/68) Fitzpatrick(5/2/70) Salem 1 (9/25/68) Arnold (6/22/70) RanchoSeco(10/11/68) Fermi 2 (9/27/70) Maine Yankee (10/21/68) Team SP - 4 men Arkansas 1 (12/6/68) Zion 1/2 (12/26/69) D. C. Cook 1 (3/25/69)
, Calvert Cliffs 1/2 (7/7/69)
Indian Point 3 (8/13/69)
- Full Term Operating License not yet issued.
57 - Y - Team 6P - 4 men- l Three Mile Island '2 (11/4/69) Sequoyah 1/2 (5/27/70) Beaver Valley 1 (6/26/70)-
' St. Lucie 1 (7/1/70)
Millstone 2-(12/11/70)'
. Team.7P - 4 men, Trojan 1L(2/8/71)
North Anna 1/2(2/19/71)- Davis Besse 1-(3/24/71) Farley 1/2:(8/16/72) Arkansas 2 (12/6/72) Team 8P - 4 men Fermi J (9/27/72) McGuire 2 (2/23/73) j l
-t
_ .. t t f s I
< . ;~ ~
L l
the nuclear industry should follow, would focus staff efforts on one plant at a time and would permit orderly scheduling and ; manpower additions. E. Other Considerations , In the course of its work'the Task Force, with the assistance of the Office of the Executive 1.egal Director, identified certain legal considerations involved in the implementation of the proposed program plan. None of these considerations would preclude the implementation of the plan but they could, under some circumstances, impose additional burdens. Although the program plan assumes the cooperation of ifcensees in providing the additional information required there may be instances where a licensee may refuse or may be reluctant to provide i
. that information. In such instances, we believe the statutory I
authority of the NRC coupled with the implementing NRC regulations :j in, for example,10 CFR E 5 50.54(f) and 50.71 should be adequate , to require the furnishing of necessary additional information I with enforcement to be accomplished, if necessary, under authority l of 10 CFR 92.200 g seq. Since requests for information will be handled on a licensee by licensee oasis and directed to specific aspects of individual facilities there appears to be little likelihood that the 1
% l
{b p requirements.of the Federal Reports Act will come into play.- If, however, blanket and generalized requests. for additional i information are detennined to be best suited for the purpose . of implementing the program plan, it may be necessary.to obtain necessary clearances'from the GA0. The standards of. the General Design Criteria specified in 10 CFR 550.34(a) would be applicable to the facilities subject to the program plan. Deviations from .the General Design Criteria would be required to be justified and a sound basis established for' any deviation in order to provide an equivalent degree of-safety. i It was deemed premature' to speculate as to whether the back-fitting requirements of 10 CFR 550.109 would be applicable until the re-review of a particular facility was completed. In some instances the requirements of this section of the . regulations may be reouf red. . ] An environmental impact statement is required under the National Environmental Policy Act and the NRC's implementing regulations p . for " major federal actions affecting the quality of the human environment". Since the program is only to be an updating and
- upgrading of documentation forming the basis for acceptance of l
l l 1 l I - - _ _ _ _ _ - _ _ _ _ _ _ l
' . 1 l
'..-i .
s .l q the-various systems and component's of individual facilities presently operating, it'does not appear 'that the program would j reouire the preparation l of an environmental impact statement. - T f ~ If, however, the program impacts appear to be more significant , in actual, implementation, the. preparation.of environmental ] 1 impact statements or appraisals may be required. To the extent.the program requires the issuance of regulations, statements of general NRC policy or amendments of existing licenses, the' Federal Register notice requirements specified q J
- by statute or. NRC regulations must be followed.
.i 1
I 1 l ..
HUCLEAR HEGULATORY COMM:!SION
~
;k-Qf.t}
nasmucron. v. c. 2 cuss
- APPENDIX A hh((*/
***** MAY 0 31975 l
l K. Goller
' l
- B ' Grimes ' "
(
-l R. Purple /-
D. Davis ' l J. Carter ( R. Silver i D. Stewart { D. Mcdonald TASK FORCE FOR REREVIEW PROGRAM FOR OPERATING NU l You have been designated as a member of a task force to develop a
>' j program plan for determining the need for further backfitting of 1 licensed nuclear power plants to achieve conformance withmeeting A kici-off current '
i criteria. Karl G011er will serve as chairman. will be held in a few days to provide more detailed guidance on the - task force missions. Missicn Develop a rereview progrcm plan for (1) evaluating licensed nuclear power plants against current criteria, and (2) developing a framework from which bachfitting (10 CFR 50.109) decisions can be evaluated con-sidering all plant features relating to safety. . Tasks.
- 1. Develop a staff study that examines cil reasonable alternative Recommend one.
approaches for a Rereview Program. l
- 2. For the selected approach, develop an action plan to implement it.
- 3. Present the study results and selected approach to HRR management, the ACRS and, possibly, the Co:nmissioners. .
l f 4 e-4 e
s .
\ -
Milestones . 1. Weekly status report's (oral) to Director, OR [
- 2. Study outline - May 14'
$. Draft study - May 28
- 4. Draft action plan - June 4
- 5. - Present draft plan to.NRR Management - June 15
- 6. Redraft of study and action plan - June 25 , j, ACRS Presentation - July meeting
/
7. Guidelines
- 1. The Rereview Program should be in sufficient depth (and breadth) so that backfit (10 CFR 50:109) decisions for a given plant can be based on a balancing of the ralative safety enhancements achievable. ,',
- 2. This effort is top priority; members are directed to commit the
. necessary time to complete this effort. The task force should solicit assistance as needed from within DDR. Contact points within other NRC organizations are being estab71shed.
- 3. Consider the need for and usefulness of using contract assistance (e.g., the National Labs) for portions of the Review Pr.ogram.
- 4. Consider the need for, degree of, and timing of, public and utility pariticpation in the progrcm.
- 5. Include resource requirements and schedules for each alternative
~
'. seriously considered. At least one alternative should assume no significant 8dded staff for the program. Include impact on industry
^
(during the evaluation phase). 5 . Victor Stello,I Jr. ,' Director Division of Operating Reactors Office of Nuclear Reactor Regulation cc: E. Case , NRR Directors .
. O a
(' '12
~ '
APPEND 1X B
., - e POSSIBLE SAFETY' ISSUES FOR COMPREHENSIVE TOPIC LIST
- l. ' Population Distribution .
- 2. Transportation Routes in' the Vicinity of the Plants
- 3. Aircraft Traffic in the Vicinity of the Plants .
- 4. Industrial and Military Facilitics in the Vicinity of the Plants ,
- 5. ' Industrial Security
- 6. Energency Planning
. 7. Exclusion Area Authority and Control -
- 8. Nuclear Plant Personnel Qualifications l' O. Secupation Exposure - AIARA .
- 10. Appendix I - AIARA
- 11. QA Programs for Operation
- 12. Outage Time of Safety Related Systens in Operating' Facilities to Identify Substandard Operations
.13. Severe Weather Phencrnena (Stonns, Hurricanes, Snow, etc.)
~ 14. Atmospheric Diffusion Estimates (X/Q)
- 15. Tornado Winds and Pressure Drop Protection -
- 16. Tornado Missile Protection
. 17. Turbine Missile Protection
- 18. Flood Protection ,
- 19. Geology .
- 20. Seisnology ,
- 21. Foundation .
. I I
l .
'. ,i l
_ _ _ = __ _ __ l
, u
. . . 1
. j
. j 1
q
- 22. Boron Addition Systons
- 23. Fission Product Ranoval Systan (Sprays and ESF Filters)
- 24. Control Roan Habitability. Systems -
- l
- 25. Overhead Crane Handling Systems J
- 26. Auxiliary Cooling Water Systems Including 02nponent Cooling Water .
1
- 27. Ultinnte Heat Sink and Safe Shutdown Systems . l 4
- 28. Spent Fuel-Pool Ventilation Systen .
J. .
- 29. Scisnic Qualification of Category I Instrumentation and Electrical Equignent 1
- 30. Environmental Qualification of Category I Instrumentation and I l Electrical Equipnent .
- 31. Energency Power . 1 1
- 32. Offsite Power System Stability
- 33. Safety Related Control Power
- 34. Safety Related Instrumentation Power 35.' Multi-Unit Generating Station shared Oasite Dnergency Electric Power
- 36. Transfer of Ssdoty Related Equipnent fran Nonnal to Energency Ibwer and Return
- 37. Diesel Generator Qualification and Reliability Testing
- 38. Ability of Diesel Engines to Operate at Light Ioads for Sustained Periods of Time Without Degradation or Failure ,,;
^
- 39. Diesel Engine Fuel 011: (a) quality requirements, (b) test l procedures to evaluate condition of oil, (c) surveillance test frequency and periodic cleaning or renoval of sludge fran tanks i
- 40. Station Battery Capacity Test Requiranents' l
. 1 1
s p
r 2
. .l f
1 l 41.* Scram Breaker Periodic Test Requironents i L l
' 42. Testing of Reactor Trip Systan and Engineered Safety Features .l
-l l= 43. On-Line Testability of ECCS Actuation Systen and Component -
Availability
- 44. Shared Safety and Service Systans for Multiple Unit Facilities
- 45. Thermal-Overload protection for Motors of Motor Operated Valves' 3
)
- 46. Motor-Operated Valves in ECC3 Accumulator Lines 1
~;
1
- 47. Protection Systen Autanatic Trip Point Changes for Operation With d' Reactor Coolant Pumps Out.of Service ,
]
48.. Control Elanent Assanbly Interlocks in Canbustion Engineering . Reactors .
- 49. Failure of Containment penetrations from Electrical Faults Inside Containment During LOCA
- 50. Fire Protection
- 51. Instrumentation to Follow Course of an Accident
- 52. loose Parts Monitoring .
- 53. Standard Technical Specification Requirarents .
- 54. Capability to perfonn Operation Outside Control Roan to Shut Plant Down
- 55. Separation of Counterpart Safety Related Equipnent i l
- 56. Effect of Failures in Non Safety Related Systems During Design Basis Events
- 57. Ioss of Plant Air Systans (affect on plant control and monitoring)
- 58. Reactor Coolant Pump Overspecd During IDCA
- 59. RC Pump Under Frequency Trips ,
GO. RCS Overpressure Protection e 4 9
- W 9
/ -
4
. 4. ,
61.- TUR Systan Overpressure Protection
- 62. IMR Systen Single Failure as per GDC - 19/3 4 !
i 63.' Pump and Valve Operability. .
- 64. Vulnerability of. Air Operated Relief Valves to Failure (Brotms .
'J i
Ferry Fire Example)' ' k
-j
- 65. Fuel Surveillance and Instrumental' ion In Core to Detect Failures
- 66. In-Core Vibration Failures ,
L67. Ilydrogen Explosions in Off Gas Systans , p -
- 68. Fuel Damage Due to IDCA Forces j
' ~
- 69. Presence of Superheated Steam During Blowdown -l
- 70. Switchover frcm Injection to Recirculation Mode .
- 71. Doron Precipitation b .
- 72. - Filter Technical Specification Changes that will Require hbre - j j
Frequent Inspection ~ Intervals 1'
- 73. Auxiliary and Radunste Area Ventilation Systan as Related to 10..
CFR Part 100 Doses ,
- 74. Spent Fuel pool Effective - Design Modifications for Increased Storage I
- 75. Seisnic Re-Evaluation
- a. "g" value b .' Analysis technique's and design criteria - t r
- c. Seisnic qualification of mechanical and electrical canponents
- d. Selsnic instrumentation .
I
- 76. Effect of Postulated Failure of Non-Category I Structures on the Safety of Category I Structures, Systens and Components i
i O a 9*W' P
i
- 5'- ,
)
- 1 I
77 ,Ocupanent Support Integrity V
- a. Reactor vessel 'l j
- b. Steam generator -
- c. Other ,
- 78. Snubber Reliability {
.f
- 79. MSRV'Line Restraints .
1
- 80. Piping Integrity M
- a. Austenitic Stainless Steel Piping
- b. Bimetallic welds
- 81. Appendix G Cm pnance for Older Plants
- a. Low toughress of Irradiated reactor vessel welds
- b. Thermal s'iock to reactor vessel 1 i
- 82. Chemical Decontamination
- 83. Design C#es, Design Criteria and loading Otznbinations for
- Cbncrete and Steel Structures, Systens and Canponents
- 84. Inservice Inspection of Prestressed Concrete Containments with Either Grouted or Ungrouted Tendons
- 85. l'0 CFR 50.55a(g) " Inservice Inspection Requiranents", canpliance for older Plants 1 i
- 86. Containment Isolation l
- 87. Containment leak Testing (Appeildix J) ..
. 88. Ocmbustibio Gas Control in Containment i
- 89. Asynmetric Pressure Irads in Containment Subcompartments Containment Purging - Potential release due to open purge lines, closure of purge. valves e
t i
.... ~6- . .v 90.' Generation of and Consequences of Missiles Inside Containment
- 91. Effects of Pipe Break on Structures, Systens and Canponents Inside u ' Opntainment Including Flooding
< 92. Pipe Break Effect - Outside Containment .
93.: Bypass Irakage Paths for Secondary Containments.
' 94. ECW' Recirculation. Sump Design and Test
- 95. Main Steam Line Isolation Valve Irakage .
- 96. Cbre Spray Nozzle Effectiveness P
Er7. Steam Generator Tube Failure Analysis.
- 98. Steam Generator Tube Integrity .
- a. Water chanistry control j ,
- b. Denting f
- c. Plugging criteria and repair techniques
- 99. Steam Generator Irakage Tech Specs ,
I 100. Steam Line Break Accident Analysis (BhBs and PWRs) 101. Feedwater Systan Pipe Breaks 102. Cbrrosion of Condenser Tubes t . 103. Water Hamner - Flow Instability 1,04. B1m'CRD Collet Housing Cracks 105. B1m Feedwater Nozzle Cracls ; 106. Mark I pool Dynamics 107. S ticipated 3 ansient Without Scram 108. Spurious Withdrawals of Single Control Rods in P','"s j l 1 g I
+ >
i e
' - 7.
103. Control Pod' Ass mbly Withdrawal Analysis 110. Fuel Densification and Other Fuel Perfomance Areas (Bowing) .
. i
-111. Inadvertent Dilution Accident During Refueling
<r 112. Reactor Coolant Break Accident Analysis . ;
113. loss of Coolant Accident Analysis 114. Fuel Handling Accident Analysis 115. Spent Fuel Cask Drop Accident Analysis 116. Inadvertent Loading and Operation of Fuel Assm blies in an j, Inproper Position Analysis . 117. Waste Gas Syst m Failure Analysis 3
- 118. Liquid Waste Syst s Failure Analysis I k
. 4 l' i 4 9 E
. {
e
- j 4
9 1 l I l , , _ , S P 1 j 1 l}}