ML20236N048

Requests Commission Approval of NRR Program to Evaluate Operating Power Reactors Re Current Licensing Criteria & Document Results of Evaluation

ML20236N048
Person / Time
Issue date:	11/12/1976
From:	Rusche B Office of Nuclear Reactor Regulation
To:
Shared Package
ML20236N045	List: IA-87-620, Requests Commission Approval of NRR Program to Evaluate Operating Power Reactors Re Current Licensing Criteria & Document Results of Evaluation ML20236N042 ML20236N048
References
FOIA-87-620, TASK-PS, TASK-SE SECY-76-545, NUDOCS 8711160002
Download: ML20236N048 (58)

Text

{{#Wiki_filter:' i 'r q ' OFFHCUL USE ONLY uwtr o sTbts' SECY 545 November 12, 1976 NUCt.E AR REGtlLATORY COMMISSION POLICY SESSION ITEM 1 For: The Commissioners From: Ben C. Rusche, Director, Of fice of Nuclear Reactor Regulation Thru: Executive Director for Operations

Subject:

TdE SYSTEMATIC FNALUATION OF OPERATING NUCLEAR POWER PLANTS ObtainCommissionApprovalofaMajorPolicyIssue]_

Purpose:

Issue: NRR plans to initiate a program that will evaluate. operating power reactors with respect to current licensing criteria'and document the results of this .i evaluation, including th'e need for' any necessary plant changes. Discussion: Introduction The NRR staf f has recognized for some time that it needs a more thoroughly documented evaluation of the disparity between current technical-positions on safety issues and those that ex'isted when a particula'r: plant was licensed. In addition, the staff has recognized-that improvements. in documentation on operating power reactors will facilitate balanced decisions regarding backfit requirements.. These improvements are especially needed for the oldest docketed cases.. An increasing amount of staf f time is being spent documenting ljusci-fication for'the safety of existing reactors and for judging the need for additional analysis and equipment. The continuation of evaluations on an ad hoc basis' will not t'esult in the most efficient use-of professional ~ time and will not lead to the most-ef fective regulatory decisions. The problems encountered by the staf f and the increasing amount of staf f time.~ being devoted' to these ad hoc responses is. reflected 'in similar problems' and increasing workloads for the licensees and vendors.s

Contact:

V. Stello, NRR Tel: 492-7672 8711160002 871112 PDR FOIA HAMLIN87-620 PDR

gffy yg j 'n - n -s s {a s g: e ) i /1 3 1 g y o l 4 2: l to ef f ectively and 'convini aglyOM/.Um the level of-l andi o maintain. se ety providiad in. operating plantv t 8 3 the. documented acceptability. of.thpA 1.avel of saf ety,- it.is proposed to.begh a, Systedatir Evaluation' Pro- 't ~ gram.*. It is' expect (i that' this' program will success- ~ g ] i itsly ' address the c.etcerns in; the pre. ceding paragraphs-

with thk. leas t posLible.- impac t to the industry and maxi-

-) mure certainty to the public.that the; safety reviev of' all planta by NRO-is being acomplishsd by a continuing,, .j well planned and thoroughly documented' program.- l ec 9 -Althoughthe'staylbelievesthataclear'needexis'tsl-

j

~ for a more systea.atic and comprehensive' documentation'

j of the staff safety;teview o'f operating reactors, O.is

,l need does not necessarily imply;the presence ofa any. Lj critical' safety defect in currently 1 operating plants. N In the, past the scaf f has examicad.311, operating 1 nuclear L ' facilities to' assure their adequacy.-with respect to.cer-i ~ tain new saf ety is sues, e.g., ECCS, and postulated pipe 7, breaks outside of containment. Otherl safety issues, judged-to be of lesser significance,lhave not.been examihed -j explicitly. The systematic evaluation of operating.re-- actors -is expected to confirm and document. the staf f! judg-ment that present safety margins are adequate. iloweve r,~ i as a result of their re-review,'the staff may conclude-that some plants should provide increased. safety margins.

Evaluations in accordance with 10 CFR 50.109 will-: be per-4 formed to support.such conclusions.

The next paragraphs-q J describe the systematic evaluation program-proposed'to achieve these objectives. Program Objective Because of the recognized deficiencies in documentation-of the acceptability of licensed plants and the. increasing amount of staff. time ~being devoted'to developing such docu ~ mentation on an'ad hoc basis, a. Task? Force was established within NRR to develop a plan and program-for performing.at -systematic review of-operating nuclearipower: facilities.- I 5 'i'

• Attached is a discussion paper, "The Systematic -

a -Evaluatio*a of-Operating Nuclear Power Plants."- i 1 l, ( _m- ___w_

yrgy ,, grg q'; Qp t, 3, 7 ]4., y-q -v

.w, s.

r cn4 [ f., x s .[ h '[f a h y, s s 'V

o. 3' l-1

't ~ {he 1 t Y, .j. i ' The foll iing program objectives have beed established t; yM by,theTasbJothpV! ,Pti, . ](I Dassess the j J . safety adequacy of ' the~ design ) cf. c'peration~ of cur /.. 1 -l l'. The Systeinic Evaluatlon. Prog ' d'tus t rently. licensed nuclear' power uts. py1 r.i 4 A s ~ 2. The 3 program.should establi$h ]! hentation.Ich O s. Y. reviewe'd com'are show1you'eachoperating~plantwith cvdrent: criteria on signi 'i i l . and's(suld, ozavidela ratiotale%r acceptab1'ef de-l f 'parttires frttmihese criteria, j qi i. y', The.prograd,Whouldprovidethe;} capability'tomake ' ~j f, 6,. < J 3H 1. 3. I- -intqtated.and balanced - decisions with respect.tof . ~ S)/ ,H anyL, t eguir'ed L backfit ting. 1 l v >j .f 4. The' progra.u ~should be /stk[; i pl for "early,identif' .. catide 'ahd = resolutions of =p 4 aificant deficiencies.,' 4 I 'l -gs e d l ' TM pr'ogram'should. ef fi ter.ly. ise avail $ble re. f ; i 5. Escurce's and. minimize rey.,h'fremer ts for ~ additiona15 reseprces',ty Ntg] or LgJig ry. y; i fJ5qq g 1 The,Ptogram s. s } } = .g < The. approach receprenced y the T &h to review-j ~ all operating relators.agair.st a sr'sc N of' safety' , ' M,,M is sues ' hased o'or el2rrent licensing r/ uirement identified i s I in the SRP. l ./ 1 -f The.firstphase.obthe-f'o)yr'ainwillbe'ogener/ ate A e - ( 1. t r a comp 7ehensivefirv of ia1\\j)j.gnificant safety _ y issuep A croup og senior 4e eone'el.will devg'lovthe l_ list. 62 copIcs ~ to be revievd c (such 'as 2smic(pgaY. [; i 7 adequacy, flood prMeetio'n, and 'shiitdd .pl1inglystems)..f The initial compth /i.ve '.ist will-be re n};ed to those-1 copics, vich,su[J'.cih rdade y digni.ficAnc 'to warrant-I Thebasf,-(hr.11l topics-evalum:ionr initial listfsill be:Lc/dQ ceL The as'esk.d[.y or the eleM fr 'nclude -j,, d WAsil.400ctype results a' 'vell.~as o r i s 'r 'being[,ncepiled. i a t The Licepees will\\nol:be involv ' inQ ( 5-4 this' pras s, tt m, g 2. The dNa h5ase. of _ the1 pre. gram will be revb/ M ~ of a limited o2mberlof reabedes usin listfof kh l l safety.issu'ef determined ts be p(ige /fica 3 ch' E L of these~ reactors.1 The reahtoR i (dially revieued j l

• _j p* a' p' " f L

s> e 7 3 -~& t-c i (.i 7 g h a dl, ~3 hsm;[~. la

j

{ jff ly s-

w yf.

x xus ,x +dra e, m

<g6 if ' gg, O; f L - o y x

c d

41 4 -) .i.I

p

.\\s

\\>

/ J YQ - h h h'., h ' ('4 .V Ia will be' thq eight -plants eceiving operating licenses N..- ( ) 7 . prior. to' 1969. These reactors are considered toi 7 t'[ 7 -havelehe least:adeqdatepdohumentation'andfto' involve-the greatest disparity betseen curre;ntLeechnical '4 r positions 1 and those which existed'when the' reactors l (' were licensed.-

)

j ,y '4 ,1

y...

For each.f acilitygaviewed during the second' phase; I the issues :to be t evaluated :will be established by ga [ l i ' senior NRC' review team ' consisting of an'NRR project I manager and various technical specialists. LThe'Lidensee.

e

(' will fitst become. involved in the program at.this'.. point? > \\ (see i discussion paper for a more!p.113 of the accompany ngcomplete explanation).' TheS u with the Licens~ee,- will' establish la plant-specific d list of? issues-totwhich the Licensee will respond'with d o. [' - 9 > his evaluation. The team will then review the responses l i (the depth pf the' review will' vary' depending.onfthe- '4 plant /issuelsafety significance). The detailed evaluation.: 1 'J S procedures for these teams, including. procedural guidelines; Ei, for arriving at the: disposition-of.each issue' considered,. U 2 i \\, will be ' developed early. in the program (in' parallel sich the screening of. issues). The detailstof this procedur'el l h [r ~ Wave not 'yet been worked out, : however, it - is ' expected 'M ): hat the_ relative significance' of some issues map be 'de - l cided on. probability bases already known or under investi"-

l

~ <. p J (1 'I gation. -Another acceptable approach ~would obviously be lg to show. compliance with current-requirements. It will? ..1 T Ep al'so be the responsibility.y of this team to assure that N all backfi"t decision's on-individua1L reactors have' been 4 given appropriate NRR'.and management review &nd thatT l these decisions' are consistent with the backfit 'decisionsL of other teams. q 3. During the second phase,.a management evaluation.will be ' condqtted to determine the' appropriate course of' action ~ forfthe remainder of the operating reactors. -It is.esti-f l . mated that this evaluation: can:be made about.30 month's af ter ,,yf program initiation;- it will be a : substantial' decision and" l 0' will be based on the? record'of achievement 11n. meeting-( program ' goals to that date, and' on the prospects for. 4 / p' ' f real additions;to operating saf etyL with' positive-impact /, j value consideration for the, remainder 'of the' plants.: j 0 L '\\) Consultation with the-original Systematic; Evaluation. O Task Force will be maintained'in the decision process; V H l i I g, l b

Q j

J ,I o .JP c ~ <1 h H ll l,f. 't' 7; Ohlhru_ ___r_______

,1,, A .a ,) .\\ 5 c The management evaluation.is expected to determine j vhether the' most ef ficient means of proceeding with the remaining reactors 'is through continuation' of, the initial' team approach,'a modified case-by-case approach or an evaluation'of classes of reactors, or whether there R is suf ficient bases ~ to terminate th.e; reevaluation. program.. 4. The -planned systematic evaluation would confirm the.ade-quacy of.all current operating power reactors lwith respect to saf ety and provide clear written documentation of ' the basis for this -' conclusion. The' review ef fort, when.completen. will include all presently l licensed. reactors either-by a precess of case-by-case reviews or by:a combination of case-by-case reviews and reviews of classes of reactsrs. .For future reactors, the, operating license review will document deviations. from alli licensir.g require-ments-then current and the basis l fer acceptance. - In - 1 addition to this, each new licensing requirement which is identified' by the Regulatory Requirements Review l Committee as applicable to operating f acilities will i . be assessed ' for each. f acility and the. conclusions. docu-mented, thus ' keeping the. evaluations ' current:in the - ~ 'l future. It Lis anticipated that 'these two. procedures will be fully implemented by. January 1, 1977.- Coupled j with' the completion of ' the Systematic Evaluation Program,.. they will' insure that every operating plant will have a record of, the' results of 'staf f review for.all safety

I

encerns and that the record will be continuously updated

-{ as. new issues are' identified by the st'af f. Resources Required 1 About 70.f acilities will be. involved in the Systematic Evaluation Program to' varying degrees depending largely on-the vintage of the facility.. It is estimated that-reviews for 'the eight' initial cases will take about ~ - three years ' to complet.e. j q l It.is planned to accomplish 'the initial reviews with two teams. Each team will be l assigned. an additional 1fcci? 8 ty for review every two to three' months until the teams are working simultaneous' sly onl all eight:of.tihe initial reactors

m I l 1 6 l 1 to be reviewed. 'After a canagenent evaluation of the j progress of the reviews on the first eight reactors, the j f program will be redirected as necessary. If a case-by-case team review approach is' still censidered desirable, teams will be added until a peak of about 40 professionals are assigned full-time to review teams. The effort of these-1 people plus suppo'rt from other technical staff are esti-l i mated to total about 170 professional manyears over a seven year period. It is expected that the.last review would j ' be completed by' about the end of 1983 if the reviews are j done entirely through case-by-case evaluation-and sooner { if done by a combination of case-by-case and reactor class ' reviews. Upon completion of this effort the personnel.would be f needed to perform' continuing evaluations of the increased j 1 number of operating plants anticipated in the late 1970's and early 1980's. l l The staf f. ef fort of 63 manyears for the evaluation program 4 involving the oldest 8 f acilities is based on 23 manycars l to develop guidelines and cull the list of topics and 5 l manyears/ facility for the specific reviews. This total j staff effort is equivalent to 3 manyears per plant or $700,000 for each of the oldest 8 f acilities.. The cor-responding average cost of staff effort for each of the i ~ remaining 62 facilities is about $200,000.. No contractual l 'l support is envisioned. -) The workload forecast for the program identifies a need for. i reprogramming within NRR 21 additional professional position

l in FY1977.

A substantial effort is required in TY1977 if .l the Systematic Evaluation Program is to complement the con-l siderable ongoing review of generic problems'such as fire j protection. An early and vigorous start of the Systematic j Evaluation Program will also enable NRR to integrate ongoing 'j i generic. reviews with systematiac; evaluations without signifi-l j cantly delaying the resolution of the ongoing review efforts. The total workload schedule represents a balance between the goal of a timely completion of an integrated review of generic and plant specific topics and an orderly development of NRR capability to perform these reviews without unduly impacting l other functions. 1 d i l 1 L 1

i 7 The cost of the program to the licensees of.the eight oldest f acilities is estimated to average $2-4 million - per facility over a 3 year period excluding possible equipment modifit.tions or~ additions. The estimate ;is very approximate since the actual number of topics to be- ~ -reviewed and.the complexity of the required evaluation will not 'be known until-the individual reviews are in-l progress. By comparison, in 1974, the cost to. prepare a combined PSAR and~FSAR was. estimated to be about'$4 million. It should.be noted the the program cost is not a completely incremental additional cost to the Licensees, because each~ Licensee could reasonably expect to incur the cost of'several ad hoc or. generic reviews-during the period of the program. The Atomic Industrial j Focum has estimated.the ongoing cost per facility for analysis, procedural changes and facility modifications q to be about $1 million/ year. The program cost for each of the remaining licensees is estimated to be about half' that for the first 8, i.e., $1-2 million over a 2 year period. A single impact /value assessment for this complete' program cannot be made because of the uncertainties associated with'possible required plant modifications versus the'ex-pected advantages of treating several issues at once. There will, however, be an impact /value consideration-involved in any backfit. decision for each plant. ' ^ ' Recommendation: That the Commission: l. approve the program concept as outlined in the paper 2. note: the JCAE will be informed via either a paper or presentation Coordination: The Of fices of Inspection and Enforcement, Research, and Standards Development concur in-the program outlined in this paper. 'The Office of the Executive Legal Director has no legal objection. 0GC has no legal objections but supports OPE comments responded to at enclosure. OPA prepared the Press, Release,

1. d Scheduling:

At an early Policy Sessioken C. Rusche, Director % Office of Nuclear Reactor Regulation 1 DISTRIBUTION

Enclosures:

1) Discussion Paper

2) Tentative press Release Commissioners

3) Response to Comments from OGC/0PE Comission Staff Offices Executive Director for Operations ACRS Secretariat l

/

DISCUSSION PAPER THE SYSTEMATIC EVALUATION OF OPERATING NUCLE?.R_ POWER P NRR STAFF OCTOBER 5, 1976 Introduction The NRR staff has recognized for some time that it needs a better documented technical basis to substantiate our opinion that currently In ) operating licensed nuclear power plants are acceptably safe. particular, staff safety evaluations have not been conducted concerning l the acceptability of many nf the numerous disparities between current technical positions on safety issues and those that existed when In addition, there particular plants were licensed for operation. are major deficiencies in the documentation needed by the staff to reach balanced decisions regarding the need for backfitting of operating plants, l in that information needed to determine, pursuant to Section 50.109 of the Commission's Regulations, whether certain backfit measures would provide substantial additional protection required for the public health and safety or the common defense and security is not currently available; 1 The need for such improved documentation concerning operating plants was further noted during the recent JCAE hearings into allegations relating to I I

s. nuclear reactor safety. The staff, in one of its reports [1] made available to the JCAE, acknowledged the lack of documentation providing l 1 the basis for some of its past decisions on safety issues. l The ACRS, at the conclusion of its independent review of the allegations, i recormlended that the staff improve its documentation of the bases for decisions that backfitting to current licensing criteria is not required [2]. l l i A related concern, identified by the ACRS over the past few years and more recently in several letters, is the need for a systematic and comprehensive review of operating reactor experience at periodic intervals (e.g.[3]). l An increasing amount of staff time is being spent as individual issues l arise in assessing the safety significance of such issues for existing l reactors and in judging the need for additional analysis and equipment.as f a result of operating experience and in response to show cause_ requests l The submitteci pursuant to Section 2.206 of the Cormiission regulations. / continuation of such evaluations on an ad hoc basis does not result in l.. efficient use of professional time, nor does it lead to effective regulatory decisions. The problems encountered by the staff and the increasing amount of staff time being devoted to these ad hoc responses discussed above are also reflected in similar problems and increasing workloads for the Licensees and vendors. They report problems with the establishment of-work priorities The f and in the efficient use of their professional manpower resources. 1 L -

q 'l - l 1 present approach also results in.'an. atmosphere.of uncertainty for the industry. Plant modifications.made in response to an ad hoc review may later have to :be modified as 'a result of subsequent reviews..There l is, at present, no suitable means of_ establishing confidence that. a ( ) major investment in an analysis or a. modification believed required by; the staff will provide assurance that plant operations will be stabilized for an extended period of time, nor is there means for a licensee to. determine the total-investment needed to bring the plant to a level of' j acceptability for extended future' operation. To effectively and convincingly' confirm the level-of safety provided in operating plants, and to maintain documentation of the acceptability-

• of' that level of safety, it is proposed to begin a Systematic Evalua~-

tion Program. It is expected that.this program will successfully address the concerns in the preceding paragraphs with minimization of l the necessary impact to the industry and maximum certainty to the NRC l and the public that the safety review of all plants carried out by the l NRC is a continuous, well-planned and thoroughly documented program. j j The staff believes that the clear need for a more systematic and compre-hensive design review of operating reactors does not necessarily imply ] the presence of any critical safety defect in currently operating plants.- I In the past the staff has examined all operating nuclear facilities to assure their safety with regard to certain important new safety issues, l ~ i e.g., ECCS,and postulated pipe breaks ' outside' of. containment. In some q

l 4. 1 i cases, changes in NRC Regulations (e.g.,10 CFR 50.46) clearly required i these re-examinations, while in other cases (e.g., pipe breaks outside 1 containment) safety significance of.the issue dictated re-examination i on all plants. The result of these reviews of selected new safety issues as they arose combined with the inherent large safety margins l that have been required historically, form the bases for our judgment -1 that no critical safety defects exist at this time, j However, it must be acknowledged that other safety issues that have been implicitly judged to be of somewhat lesser significance have not l been routinely examined in all operating plants. In fact, a major complementary element to the proposed Systematic Evaluation Program is to explicitly treat each new licensing requirement identified by the RRRC in such a way as to always have a documented decision as to its 1 resolution on all plants. Of course, for this complementary element 1 to be effective, the proposed program leading _ to initial documented i i evaluation of facilities to current safety criteria must first be completed. The complete program is as follows: I i (1) New procedures in NRR will provide detailed baseline documentation in staff Safety Evaluation Reports for plants licensed to operate after January 1977; (2) A systematic evaluation program for operating reactors j will provide.this evaluation and documentation for those l plants reviewed prior to this time, and (3) Future RRRC approved licensing requirements will be systematically reviewed for each-operating plant and the 1 evaluation documented. The reevaluation of operating reactors is expected to confirm and document .I \\ the staff judgment that present safety margins are adequate for continued j .i A

, q l 1 l< s operation of all plants. However,-as a result of their reevaluation,; { the staff may conclude that some plants Tshould provide increased safety j margins. The next paragraphs address _the methodology.of any required ] 1 backfitting needed to achieve this confidence in long-term operation. ] An important element in the reevaluation process ~ will' be the decision-as to whether backfitting of particular current licensing requirements is-needed. Paragrapn 50.109 of the regulations states that the NRC may-j l require backfitting on finding that " substantial, additional protection l which is required for the public health and safety or the corrmon defense-f I and ' security" is' involved. Paragraph 50.100'also states in-part that ] "a license...may be... modified...because of conditions revealed by...any report, record, inspection or other means,'which would warrant the Com-mission to refuse to grant a license on an original application...." 1 A finding of substantial additional protection would be made by.the NRC~ staff for each item of significant cost impact required of a Licensee. l a l In dealing with the question of " substantial, additional protection" for a particular safety issue, all aspects of the facility design which bear on the issue will be considered. For example, consideration will be-given to capability to accomplish the safety objective with existing plant-equipment, other than that nonna11y dedicated for this purpose,' existing-design' margi_ns, calculational conservatism, the probability of. serious - b l l A

events during the remaining facility lifetime as well as the potential-and probable consequences of these events, and prior operating experience where this is applicable. It is not believed practical with current: technology and data base to quantify the " safety worth" of all of these items, each of which may tend to tip the balance for or against ' requiring a change when considering the standard of substantial additional protection. Nor do we believe that adopting specific quantified acceptable values of risk will be appropriate in the foreseeable future for all safety issues, or for all facilities. Each backfit decision will be a judgment reached by the staff and approved by NRR management considering the specific safety issues and all relevant plant-related and site-related factors, a value/ impact appraisal and whatever quantitative risk information i may be available and relevant. The decision process which is the end i result of the Systematic Evaluation Program is being carefully structured to assure that all of the above factors are considered. It is important to the success and credibility of the staff effort that this decision process be carried out in a well-structured manner in which both ~ technical reviewers and cognizant management personnel participate. As will be discussed in more detail later in this paper, a series of review I teams will be established for evaluating the reactor facilities. Each of the review teams will consist of a project Manager and various specialists, i depending on the group of facilities being evaluated. It will be the responsi-bility of each team to assure that all backfit decisions on individual reactor-l l have been given appropriate technical and management review and that these l l

.n

..y. 1 The: decisions are: consistent with the backfit decisions' of other teams. i evaluation procedures for these teams,. in.cluding procedural.. guidelines - 1 for arriving at' the dispos; tion of each issue considered, will be= developed - early. in the program (in parallel with the-sr.reening of issues). While the details of this procedure have not yet been wor'ked out. aL11st of i iving at a'backfit' _ factors to be-addressed by the review team n arr In-decision (such as the factors mentioned above) will be developed. addition, the Regulatory Requirements Review Cemittee will be kept apprised of the review status and. consulted'early in the decisional processiesp'ecially-when an issue that may affect a class of plants "is' identified. t c l As discussed in the " Program"'section of.this document, a key element in. the plan is the early identification-of the list of. safety concerns to-be reviewed. Any new safety issue identified as a result of. operating. l experience o'r new information during the course of the Systematic'Evalu-Issues of; ation Program will be handled on a case-by-case' basis. imediate 'and significant safety concern will be resolved as is presently. done on an ad hoc basis. Resolution of generic'iissues of a longer term nature will be accomplished by developi.ng appropriate interim measures - while the necessary investigations'are proceeding. When these investi-a gations are completed, the long-term resolutions will be integrated ~ into the Systematic Evaluation process 1 i [ t_____-

(

b B -

l. y l

Background

the: Coamissi n's In the early days; of the civilian nuclear power program,! scope of review of proposed pcwer reactor designs was. not wellL def .) ~ l The requirements for acceptability evolved as new f5cilities' or documented. In 1967, the Commission published for coment and: interim ' 3 1 were reviewed. j use proposed General Design Criteria for Nuclear. Power Plants 1(G The: j established minimum requirements for the principal design criteria. GDC were fonnally adopted, though somewhat modified in 1971,. and h' been used as guidance in reviewing new plant applications.since.then; l These' guides ,l In 1972 the development of Regulatory Guides' was initiated. describe methods acceptable to the. staff for'. implementing specific portio 1 techniques of the regulations, including the GDC, and formalize stafft In 1975 the Standard Review Plan (SRP)' for performing a facility review. was published to provide.further guidance for improving the' quality and 1 unifonnity of staff-reviews,. enhance communications.and understanding. O the review process by interested menters of the public.and the nuclear. For the most power industry, and stabilize the licensing process. ~ part, the detailed acceptance criteria prescribed in the SRP do not' i represent new criteria; rather, they are current methods of review that,3 in many cases, were not previously' published in anytregulatory document. i i j l

i Because of the evolutionary nature of the licensing requirements i discussed above and the developments in technology over the years, operating nuclear power plants inchde a broad spectrum of design features and requirements depending upon when the plant was constructed and licensed for operation. Listings of plants by vintage is included in the attached Tables 1 and 2, along with manpcwer estimates for their The amount of documentation that defines these safety reevaluation. design characteristics also decreases with the age of the plant - the older the plant the less documentation as well as the more variance - from current licensing requ rements. Although the earlier safety i evaluations of operating facilities did not address many of the topics discussed in current safety evaluations, it should be recalled that all operating facilities have rare recently been reviewed against a sub-stantial number of major :)fety issues which have evolved since OL Therefore, as discussed in the Introduction, conclusions of issuance. overall adequacy with respect to these major issues (e.g., ECCS, fuel design, pressure vessel design, quality assurance) are a matter of record. On the other hand, a number of issues of lesser safety significance (e.g., seismic considerations, tornado and turbine missiles, flood protection, equipment qualification, pipe break effects inside containment, piping whip) have not been reviewed on many operating plants and documentation does not exist that explains why the issue lacks the significance to warrant a review.

l t 1 1 procram Objective Because of the 61iciencies in documentation Of the acceptability of currently licensed plants and the increasing amount of staff time being devoted to developing such documentati:n on an ad hoc basis, a 1 Task Force was estabijhjed within NRR to develop a plan and program 1 for performing a systematic review of all operating nuclear power f acilities. The Task Force, under the direction of th'e Division of Operating Reactors, solicited input from other Offices of NRC and from the ACRS. It also considered the suggestions of the AIF, at its l i request. The following five objectives of the program were established l i by the Task Force: 1. The review program must assess the adequacy of the design l and operation of all currently licensed nuclear pcwer plants. 2. The program should establish documentation which shows how each operating plant compares with current criteria on significant safety issues, and provide a rationale for acceptable departures from these criteria. 3. The program should provide for the capability to make integrated and balanced decisions with respect to any required backfitti ng. 4. The program should be structured for early identification l and resolution of significant deficiencies. l

1 The program should efficiently utilize available resources 5. and minimize requirements for additional resources by NRC or i indus try. The planned systematic evaluation would establish the adequacy of all operating power reactors with respect to safety and provide clear written documentation bases for tnis conclusion. To limit the potential number of facilities that need an updating review to assess the facility against today's requirements, steps have been taken by NRR management at the recommendation of the Task For:e to assure that deviations from all licensing requirements and their basis for acceptance is documented in future operating license reviews. In addition to this, all new licensing requirements which are identified by the Regulatory Requirements Review Committee, (RRRC) to be applicable to operating facilities will be assessed for each of these facilities and the conclusions documented as the new requirements art adopted. It is anticipated that these two actions will be fully implemented by January 1, 1977. It should be cmphasized that these two actions, coupled with the completion of the Systematic Evaluation Program, will insure that in the future every operating plant will have a record of the results of the staff review of all safety concerns and,ne the record will be continuously updated as new issues are identifit/ " staff. Thus, the ACRS concern for " periodic review" of operating, plaats will be encompassed by such a continuous program when combined with the continuing feedback of operating information into the licensing process by ongoing formal mechanisms recently established by the Division of Operating Reactors.

} 12 - The Program The approach recoratended by the Task Force is to review all operating-1 1 reactors against a' selected list of safety issues which are saiected I fram current licensing requirements identified in the SRP. Table 3 f shows how the list will be developed. i A l The Task Force recommendation will be implemented in two steps. program that reviews the oldest 8 facilities will be started irrmediately, i l After this program has progressed to a sufficient extent management will evaluate the worth of the review and determine the appropriate l action for the remaining operating facilities. I 1. The first phase of implementing the program will be to A ] generate the list of all significant safety issues. i group of NRR personnel will develop the list of topics (such as seismic design adequacy, flood protection'and j shutdown cooling systems) to be reviewed. This initial-l i l comprehensive list will be reduced to those topics with i sufficient safety significance to warrant evaluation. l The basis for all topics deleted from the initial list l l will be documented. 1 .1 The topic list will be further reduced for each facility 'l ~ l to be reviewed, in accordance with those that have been l l previously evaluated for that facility and the evaluations l l 1 1 ) i i i J t

f m

1 li ~ i' t documented by'the" staff either for a groupior'for a' + '~ 4 q parti: alar, type of facility. o.r 2. The second phase of-the program wiltbe thelrev.iew 'of ' J a lir.ited number. of reactors ;using -lists of safety issues- } 5 deternined to b significant:for each of these' reactors. The reactors' initially reviewed will be the eight flants a-receiving operating;1icenses prior'to 1969.:'These reactors. are considered to have the' least adequate docum' ntation an'd ' e to involve the greatest disparity between: current technical pc 'tions and those which existed whsn'the reactors were. 4 4 L l icer. sed. Each Licensee then will.be required to provide a comparison; of design against current criteria for 'the' remai.ning topics-q l identified by the staff, and wil.1 be required:to provide. a.] s 1 a description of any changes believed'by the Licensee to bei .1 o' This plant-specific topic list will then be reviewed necessary. by a team of about'5 technical specialists, the. assigned. Projecti Manager and the IAE inspector (as available).. This team nill meet with the Licensee to explain the safety concern of each . topic; 'The t6am will assess the response' from the Licensee: } and ultimately prepare a safety evaluation. :It is anticipated that it will take;each Licensee about 12 ' months to respond to - ~ all the topics applicable-to his facilit/. -When Lit:is deter.- mired that adequate safety. margins are. present for anyfissile s )' l: N i i y 7, N 2 --. J ^ ' ~ " L

7 m i 4' fl 'j .n .s14 -' \\> even though the present requirements for new plant. design-are not explicitly met,. the bases for:the edecision will ' be' documented by the staff. Those' topics potentially requiring; a backfit will be held for subsequent-resolutf on'. This will3 ~ allow all backfit judgments affecting' a plant lto be con- ~ sidered:at the same time.for_ proper balancing. Guidance for ~ .I making backfit decisions will be applied at.this' time. LIf-'

)

imediate safety concerns.are identified, prompt corrective ~ J action would be taken. l

3. - After the second; phase.has. progressed to alsufficients extent to validate its worth and rectify procadural-problems, a management evaluation-will b~e conducted

i to dete.cmine the appropriate course of action for,

the remainder of the operating reactors. Consultation - with the original Systematic Evaluation Task Force I will be maintained in the decision process.- The ma'nagement evaluation is expected 'to determine whether t the most efficient means'of. proceeding with the re-maining reactors is continuation of'the. initial.' team approach, a modified case-by-case approach or an i evaluation of classes of reactors, orLwhether there is sufficient bases to terminate the reevaluation programL During' the entire systematic evaluati_on~ process the.us'ual NRC manage. ment review and approval processes will be-used. Licensees will have: 'a the opportunity to informally appeal. any adverse decisions to NRC management. t 'l

4 s. It is anticipated that at the conclusion of the review program, both the NRC and the Licensee will have the means for assessing the modifi-cations and investment required by the staff to assure an acceptable level of safety for an extended period of time. This process could involve both hearings and appeals. Resources The initial review involving the oldest 8 facilities will require about 63 manyears of effort expended over a 3 year period. A management decision to proceed with the review of the remaining facilities is expected to be made during the third year. If the decision is to proceed, about 70 facilities will be involved in the l Systematic Evaluation program to varying degrees depending largely l on the vintage of the facility. The detailed scheduling of the systamatic evaluation program is based on the premise that identical units on the same site may be considered in one review; therefore, about 50 separate reviews are anticipated. It is estimated that each review will take about two years to complete. 1 It is planned to accomplish the remainder of the review with about twelve Each team will be assi ned an additional facility for review 5 teams. every two months until each team is simultaneously working on up to five reviews. The total program will involve a peak of about 40 professionals l assigned full-time to the review teams pl.us support from other technical l staff and is estimated to require about 170 professional manyears of effort over a six year period. It is expected that the last review could l be completed by about the end of 1983. These estimates will be re-evaluated after topical lists are prepared. Upon completion.of this effort the personnel /

would be needed to perfom continuing evaluations of the increased number of operating plants anticipated'in the late 1970's and early 1980's. The allocation of these resources over the next seven years is presented in the attached Table 4. Figure 1 shows how the manpower required for the evaluation program will be utilized to cope with new plants and the added task of keeping up to date the continuous review The total work-of all plants against new standards as.they evolve. load schedule represents a balance between the goal of a. timely com-pletion of an integrated review of generic and plant-specific topics and orderly development of DOR capability to perfom these "eviews without un' duly impacting other NRR functions. Program Costs The cost of the program to the Licensees of the eight oldest facilities is estimated to average $2-4 million/ facility over a 3 year l 1 period. This cost does not include any possible equipment modifications or. l The estimate is very approximate since the actual number of 1 additions. 1 d topics to be reviewed and the complexity of the required evaluation will not be known until the individual reviews are in progress. By comparison, j in 1974 the cost to prepare a combined PSAR and FSAR was estimated to be about $4 million. However, the analyses on one major topic, j It should such as seismology, could cost as much as $1 million. be noted that the program cost is not a completely incremental additional i cost to the Licensee, because each Licensee could reasonably expect i to incur the cost of several ad hoc or generic reviews during the period l J The Atomic Industrial Forum has estimated the ongoing of the' program. a J

f ~ i. q'

g

- l7 - l i L o cost per facility'for an'alysis, p'rocedural changes and facility modifi. cations to 'be about $1 million/ year. The' program cost for each of. the-remaining Licensees is estimated to be about. half that for.the'first 8,. 1.e., $1-2 million over a 2 year period. The-staff effort of 63 manyears for; the evaluation program involving: the. oldest 8 facilities'is based lon>23 manyears to develop guideline's: and cull the list of topics and 5 manyears/ facility for '.the' specific-reviews. This total' staff effort is equivalent to 8 manyears per plant or $700,000 for each of the ' oldest 8 facilities- .It is expected: i that the Licensee will require from 2 to 5 times the manpower expended by the staff or 16 to 40 manyears of effort. per plant. The corresponding. t l average cost of staff' effort for each of the remaining 62 facilities is about $200,000. Since the culling effort must' be done' first, Lit was : a considered part of the effort for :the first 8 facilities' in the' above estimate, however, the culling would' also be applicable to subsequent' reviews. l l l a 1 l' L j

.g f f9 t 1 4 l 4. 3 i , J +Qj i Table 1' h TENTATIVE SCHEDULE TO

l REREVIEW OPERATING PLANTS-NUMBER OF-OPERATING PLANTS BY GROUPS PHY/ Case [

PHY.- l Groups OL's Issued

1. Plants A

Pre 1969 8. 5.0-40 B FY 1970 4 ~ 3. 0 - 12 C FY 1971/ 9 2.5: 22.5. 1972 0 FY 1973 8-2.0 16.0' E FY 1974 -15 1.5 -22.5 F Post 1974 26 1.0 26 Devel op -Reeval uation Guidelines 30 170 TOTALS Current PHY per custom OL review is 6 PHY of which 4 PMY Lis associated 1/ with the safety review effort. LNRR is currently reviewing some Technical aspects on all operating reactors,(ECCS, ' Appendix I,. and It.is assumed that the'o1 der plants issued operating ~ Appendix X). licenses will require more effort than the.new facilitie.s' issued ;0L's. i, i

p 33

r \\' \\ c TABLE.2 PLAllT N'AMES BY VINTAGE GROUP- 'Groun B Group A Oresden 1 Oyster Creek;l_. 'Nine Mile Point l? Yankee Rowe . Indian Point'1 'Ginna Dresden 2. Humboldt Bay. Big Rock Point ._Grouo-D' l San Onofre 1 Haddam Neck Turkey._ Point 3 q' Lacrosse' Pilgrim-Vennont: Yankee Group C Surry. 2 Robinson 2 Oconee'l Point Beach 2 m Monticello Point Beach 1 Turkey-Point 4 l Millstone 1 -Maine Yankee Dresden 3 Groun F-Quad Cities 1 ~~ Palisades Quad Cities 2 . Peach: Bottom '3 Oconee 3-l

l Surry 1 Calvert Cliffs 1 Ha tch,1 l

. Rancho:Seco l Group E , FitzPat' ick d. r -Cook'l; l Prairie Island 2 Browns Ferry 1 Fort Calhoun Brunswick 2 Indian Point 2 ' Millstone 2-Oconee 2 Troj an - Zion 1 Indian Point 3 Peach Bottom 2 Beaver, Valley 1 Zion 2 . St. Lucie l' Prairie Island 1 . Browns Ferry 3-Fort St. Vrain Salem 1 Calvert Cliffs 2-Xewaunee Brunswicki-1 Cooper Station Arnold Crystal River 3. ~ l Three Mile Island 1 Davis Bessel' Arkansas 1 ' Farl ey L 1 Browns Ferry 2 North Anna l' .Diablo Canyon?1' Ofablo' Canyon:2~ North Anna 2 Three MileLIsland 2i l

l .,m. - TOPIC LIST DEVELOPMENT 1 ] Elapsed' Staf' Resources Time Schematic Representation' wks) (Wks) of Topic List 1 List'of 118 topics ' developed by Task l Force to illustrate safety. issues to. ! be considered. y' 10 37 Comprehensive list' of topics prepared by a group of 00R staff personnel wit; assistance from others-in-NRR, IE a.t l GENERIC CULLING - DELETION AND DOCUMENTATION BY STAFF l Topics of lesser safety significance, j f /g7 i.e., those that contribute to defen a in depth but by themself are not wort re-rev iew, e.g., Loose parts monitoring i 12 4 Turbine missiles Tepics that have. already 'been generic 4 resolved. e.g., ECCS criteria ( Pipe break outside contair. ) 3 >.} Review of Topic List by R C. - Deletic q 2 1 of any other identified topics. l lA Topics that are not applicable to a ~ 3 1 particular. reactor type 'or design { e.g., BWR or pWR i Type of containment ] Total for development of general list l T/ 9 ptANT SPECIFIC CULLING - DELETION AND DOCUMENTATION BY STAFF j Topics that hav'e already been ] Mh u-2 .36 6 Nd $Y resolved for the particular facility 1 e.g., Primary System Leek Detec-l l l Inservice Inspection Proc j i hI Topics that as a result of infomal 12 2 h m ' d.. discussions with licensees and/or. l licensees' identification of eyalua ! l-tions 'not previously 1ocated by .l staff or other explanation, were l i concluded.as acceptable without detailed re-review. l f List of Topics sent fomally to-1 1 If censee for re-review. Ti ~f Total per plant after general list;is" developed - not ccmpletely. repetitive because 'of similarities between plants. / 1

Y 1 F ~ ~

g Y

470_%2l 2 i 8 2 O0 3 2 3 l F 1 2_ 8 01 4000 7 2 Y 11 1 3 5 E ~ F 0 R 8 7 9 A 2 08225 2 3 1 E Y Y F LA m C a D S r A) I 9 g ROY F 7 3' 36 00 5 5 o 2 3 r OLM 3 TK Y 1 CRl F ~ P AOa n 4 EWn o R o E .Ni d L GOs n B NI sTe 8 e A I T TAf 7 = p AUo 7 800000 5 5 e 2 3 d RL r Y 1 EAP F L l PV l OE n i Ei w R( g n 7 i 7 2 9000O0 ~ f f Y 1 F a t se t 7' a 7D 9 n 1 n o Yo Fi i y ' t t b a l a u s S ai l w N ut O ti a e I cn v i ee v ABCDEF T AI en e Y I Ri R M S l / O ~ pe t p P P od cu li uo L 'L eu d r A A vG nG T T e o O O LT D C T. l U" m

Enclosure'2i =,. Tentative Press Release '.= NRC STAFF PLANS PILOT PROGRAM IN CONTINUING. ' REVIEW OF OPERATING NUCLEAR ~ PLANTS The Nuclear Regulatory Commission's Office of-Nuclear Reactor. Regulation is implementing.a pilot program te assist in its continuing evaluation of licensed nuclear power plants in light of present-day safety requirements. The goal of the progran is to assure that'the NRC staff-has. l adequate documentation of data to compare. safety requirements in effect-1 at the time a plant was licensed for operation and those.in -effect a today. These data may be useful to help balance the costs and benefits of requiring improvements..in older operating plants to increase margins of safety. l { The pilot program has these objectives: (1). assure the adequacy of the safety aspects of the design and operationsof currently licensed l j nuclear power plants; (2) develop:information showing how each of these. .l a plants compares with current safety criteria and assess' the acceptability. ] of. areas where existing requirements are not being literally mett (3) provide a capability for balancing the costs and benefits;of. q mar.ing improvements to meet the intent of, current' safety requirements; if and(4) identify and resolve any significant' safety deficiencies 1that j might be found. The first phase of the program will involve development of a comprehensive list of all.significant safety issues. Next, there will be a review of a limited number of operating reactors-using' lists LEnclosure 21, u r ] "-~ ---..--.-m_'__._ _,_,E

1. REFERENCES '4 3 " Nuclear' Regulatory Staff / Report Concerning ' A11egatio'ns by I. Robert Pollard." February'28',1976 ' " Report on Review of Statements by Messrs. Bridenbaugh, Hubbird, 1 2. Minor and Pollard,'? ACRS Report by _ Working Group' No. 3, May-'19,. L' 1976. ACRS Lettar dated February -13,1974'on Generic Iteme, (See. item 3. IIA-10). q p ..j l i .q.- ') 4 r I f () e

. -( 7 mr - ' l, A { 4 1 ' d u_,[ -40 1 J. 37 -:L 29 . l '{. . 3, 231 l 25 3, 20 / /g ,;\\ > l, / SYSTEMATIC B;/ALUATION. t

1. 1 O t'

) VQ . 19 ll ffJl=>. O' e, v-( r.- \\ d f!(, .s' t /[.3g 4' 40 - ,f-'k g. 7 -- - 30 - 20t. ,i . KEEPING RE EVALUATED F (T? 20 4'. 10 I I . CURRENT.- a. ,. t. 1 0-f, n 4 's {i f t (? i 1 / ,l I w <4[ -f I g' 4 t l < 200 g g .'*y y [ 2 + 1 E t t .( g 0 180 - (, 1- / .) o. I '161 157 155 l ~ I i 160 152_j ) /, NEW FACILITIES - 1 2 2 140 TOTAL i l R5 EVALUATION I> i I 3 EFFORT 115 l 120-120 114 'h N l 108-A 100 .g 3 1 ' i' 95 ( - gh 81 ('(y j 1 80 g e~.~ yg 61 a -l 60 I NEW FACILITIES .]j 43 52 o 40 I } l 1 35 l. f 20 / 22

• [4

'i ,(I f \\ Y /f i i t I f f l' d l' 77 78 79 80 81' '82 83' 84- 'j g h l FISCAL YE AR '7a b j A, s3,i 1 .] d 4

y 7 a , e 3. f> c n -; 0 .lg > s,. 9: n - ]..} 'N,! 2 f] f[sl.,, } of safety issues /detennined.to be significant for-each ~ individual. <l Qi{ d F t J 7, facili ty. Decisions on plant _. improvements, if any.will be consistent 1 4 ' -j.with requirements imposed on!cther reactors included in the: review g l v.,. s

s, ' (.

f rogram. - Q Upon completion of the first two phases of the program,. a, j sc 3 de'tpnnination will be niade'on'what' actions shculd be.taken with respect-c q s' f to reviey of the remaining operati6g reactors.. g, f ffr fut'#Weactors,-deviatioryt frem safety requirements, andLthk j ng' kvia$iens, will be documented fduringLthe - justification foi p ro h operating license review. They, each new safety requirement identified: a 1 1 'y e \\,4 %' the NRC's Regulatory Requirements Review Committee (a senior staff, i e 1 t' group) as applicable to oper aNng facilities wil'1 be asses' sed on a c \\ \\ b s (\\ b;l-c;(se' basi, and +he' conclusions documented. It is anticipated that - l i 4 s pf,gyIm,1will byglmpkemented ful.ly.by January 1, s, 31 04 thic phase of 6 [y I 8s / c s, /. \\p~ ' c s \\ y d .,,M i ?',\\, g ' (j-I e i i A \\l, ja .A r h /; ', ' e s 5 i t, { s. r c y + t s q, s 6' g k (s l ,4 ? f e <, o I' U! \\ 'M 'f~ N- ,v. , f t i on,\\ l '} N j ,(, p e i \\ ,(' f<_ l ? a l 'p l y%', l l\\gq 'i s ,\\ f D 'h. tf ' i 3 4 , }d d.y{s e j -,.l. .y a

H

} i n\\. /j, 6 l(lt, ld, .[ [g((s. y [.g d i g, ? \\} '1 - -.c (,- .1

4 a q l ENCLOSURE 3; l RESPONSE TO COMMENTS-FROM OGC/0PE ON PREVIOUS' DRAFT 0F THIS COMMISSION PAPER We have reviewed the memorandum from Ben Huberman dated / October 13, h i -1976, which provided' comments on the Commission Paper on t e re-rev ew of operating reactors ~. Many of these comnents apparently were caused 1 ~ by 0GC/0PE not being aware of.the staff effort which culminated.in'the Commission Paper. A Task Force was established within' DOR,which, with input from the other Divisions in NRR, OI&E and OSD; considered the-alternatives and recomended the program for'the' systematic evaluation l of operating power plants. presented in the Commission Paper.. A Task Force l l Report on the Task Force's efforts addresses many of the OGC/0PE comments. Unfortunately this report was still in draft form when OGC/0PE reviewed ' the Commission Paper and therefore did not accompany-the Paper..The l Report is now being finalized and will be published shortly. The Task Force Report discusses NRc urrent practice and the-need for a systematic evaluation of operating reacMet. In the report, various ways are considered for evaluating operating plants and how to develco the needed documentation. The documentation should' include - the rati m ic for any departures from current crit ria;and should-provide the basit tse concluding that the safety margins in' operating. plants' are adequate. l 4 i

! l 1 l In many cases, the conclusion on the adequacy of safety margins will have to be based on infomed judgements of experienced and knowledgeable It is the staff's considered opinion that it is not possible people. to establish meaningful criteria in advance for making these judgements. Inherent in this are the questions of what-is an " acceptable level of I 1 risk" and "how safe is safe enough". However, although it may not be possible to provide a quantitative answer to these questions in general, I it is possible to arrive at. logical conclusions for specific cases by balancing the costs and benefits. It is the documentation of these judgements, including perhaps some facility modifications,'that will be f the result of the proposed program. We expect that this systematic l evaluation will resolve any doubts about the safety of individual plants. The Huberman memo suggests consideration of a stepwise approach in the systematic evaluation program involving staff evaluation and Commission approval at each critical stage. An ongoing evaluation by NRR management of the program as it is developed and implemented was ) and is still intended. The program will be carefully controlled and directed by NRR management. Numerous specific management review and However, the proposed program approval steps are built into the program. and the Commission paper have been revised to provide for a specific, formal management review of the program after it has been applied to the t first eight plants and before it is continued to any others. i l

.3'- The Huberman' memo asks the-question,'"How will the public perceive the significance of the program?" The major objective of.the systematic ] evaluation of' operating plants is to. positively determine and document. the safety adequacy of operating plants..Therefore it definitely.should help to resolve any doubts about the safety'of these plants. ~ The Huberman memo suggests that, "there is a need for an impact / value assessment". The Comission Paper does include an impact /value assessment to-the extent that it is practical, and we think necessary, to generate such an assessment at this time..The value of the proposed l program is the clarification and documentation of the safety of operating - plants that it will provide. The estimated impact or costs of the proposed program are presented in the' paper. The balance, in our opin' ion, is overwhelming on the side of the safety benefits that will be derived, especially for the first'eight plants. It may be possible to develop a more definitive impact /value assessment for continuing the. program as part of the management review of the' program after.it has-been. applied I to the first eight plants. Attached are ~ responses to each of. the specific.0GC/0PE coments, In some cases, the response to coments may include quotations from ~ the Comission Paper or Discussion Paper that are not precisely the same as those in the final versions-'of these documents. This is due to last-minute, minor editing of-these documents. In no' case:should this change the basic ' response to the coment. l l

b RESPONSE TO OPE /0GC CCMMENTS ON CRAFT PAPER "THE SYSTEMATIC EVALUATION OF OPERATING ELEAR POWEP, PLANTS" Pace 1 Aporoval of an " issue"? What does statement of purpose mean?Or is it approval of first phase Q. Isn't it approval of a program? \\\\ of a propram? The present statement is unclear. The purpose of the paper is to acquire Comission approval of j A. the Svstematic Evaluation pr'ogram as described. The program-1 is considered a major policy issue. It is not intended to obtain piece meal aporoval of the program. y The stated " issue" does not indicate any options, l Q. l A discussion of alternate approaches for accomplishing theA synopsis A. systematic evaluation was studied by'a Task Force.o l ] l In the discussion there are references to " adequately documented" l Does this refer to staff l Q. and " deficiencies in documentation".What is the general nature of the l records or licensees' records? i inadequacy -- failure to record rationale or techni:al bases for staff evaluations? Deficiencies in documentation exist in both the records of'the l A. licensees and the staff. In the older plants many of the current technical positions did not exist, therefore neither the licensee Newer license applications may have nor the staff addressed them. evaluated current issues, but the staff evaluation has not always adequately documented the technical considerations and bases for acceptance of all the issues considered in the review. l The discussion refers to staff time spent in " justification for the Q. Oces safety of existing reactors" and in "ad hoc responses." What is source of the recuest "ad hoc" refer to issues or cases?Are they externally initiated or self-init for justification? Or does the discussion refer exclusively to the Pollard /4.E. episode? Also, won't ad hoc reviews continue in any case, i.e., even if a systema evaluation plan is adopted? i "Ad hoc" evaluations refer primarily to.the method by A. a. There has which various safety issues are considered. not been a systematic evaluation made of all issues that need to be considered. l l

? 't' t '~ .- 2 u The source of the reouest for justification or the stimulus b. that prompts us to eval. ate a-safety' question is. generalizing. but u i not always, internally initiated. However, thisLseems to be academic in.so far as the point being.made in this portion of the: discussion.. The point isLad hoc reviews:are i inefficient and do' not result inJ the most effective overall-1 j

decision, j

ThePo11ard/r6E.eoisode.had?nothingitodowiththis.geneN1' q c. point. There.will always be.some case by case reviews,:however the d. proposed systematic evaluation' program will' minimize ad; hoc reviews and will resultlin better organ.ization of those that l are required. See p. 11 of.theLDiscussion Paper-Pace 2 0 '. The first sentence implies a reversal of the: position-NRC took before JCAE in February. The NRC position presented to the JCAE in February is not being' ~ A. The second paragraph on page 2. states "... theineed reversed. for a more systematic. and comprehensive design = review of operating l reactors does not necessarily imply the presence of.any critical-1 safety defect in currently operating plants." and "the systematic evaluation of. operating reactors is expected to confirm and. document the staff opinion that present-safetyJmargins-are adequater...." 1 Although we are convinced..that operating' plants even the old! i i ^ I ones, are safe.one certainly cannot say :they; meet; current licensing requirement when they, haven't been evaluated;against those criteria, Thus the Systematic Evaluation Program isL proposed to confirm ( i the level of safety and document _' acceptability of. that level of This is not.cor,sidered contradictory to the.NRC position presentec; safety. to the JCAE. What are alternatives to-Systematic. Evaluttion Program?'. Continuous' 9. uodating of SER as ' new or' revised regulatory requirements ;are developed? q A.. See response to the second question onLpage one.above. t i + i

i

i .3-does l The reference to the "least possible impact to the industry"iscussed Q. not seem to be consistent with projected industry workload d later in the paper. ~ Certainly a status quo approach to the consideration of new licensing A. l requirements would have the least impact on the industry but this Other alternatives approach was deemed unacceptable by the Task Force. considered would have had considerably greater impact. The recommended program, therefore, is one which will successfully accomplish the objectives of the program and have the least impact on the industry. This is the context in which the phrase was used. First sentence of second paragraph states that "a more systematic Q. and comprehensive design review of operating reactors" is needed. What does the "more" refer to, i.e., more than what? If the staff believes that it is unlikely that "any critical safety defect" is. present, then why worry? A. "More systematic and comprehensive design review of operating reactors than has been eerformed." Especially with regards to the older plants and the current criterf a. Though we aren't " worried". about the safety, we do have concern about the adequacy of documentation, the degree to which operating plants conform with current criteria and providing a balanced assessment of overall plant safety. O. In second sentence of second paragraph, can the staff list the "certain new safety issues" that have been examined in the past? A. Yes. The total list would be rather impressive, however for the purpose of the staff paper and in the interest of briefness the few typical examples cited are believed to be sufficient. -Q. At end of same parggraph, why is it "likely" that some increase in safety margins may be required? For each plant, unless new information is developed that invalidates earlier safety judgement, the plant should remain " safe". A. All of the new information developed has not been applied to old plants; therefore, it is speculated that some increase in i safety margins may result from applying this information to old plants. ~

m 4 4. Will there be an impact /value assessment to-accompany staff re-review? Q. As indicated on.page 7, there will.be an impact /value assessment.made A. in support of any backfit decisions.with each plant re-review. ) pace 3First sentence should read-(if we understand thel proposal). "The 'Q. following.five criteria were used by the Task Force in developing.- the program..." l A.' A change will be made in the staff paper, y Why should the Systematic Evaluation program " assess the. adequacy. Q. of the design and operation of currently licensed' nuclear power-plants?" Isn't 00R's function now to assess adequacy of operation? l It is this DOR' responsibility that prompted the formation. of a Task Force to develop a-program for systematic evaluation of. A. operating nuclear plants. 'Q..The " criteria" or " objectives" should be~ related to present;NRR Why isn't present practice " systematic"? For example, I practice. isn't the second item a statement of present practice?.Doesn't present system result in integrated (?)'and balanced decisions on backfitting? F The present practice is not totally systematic in that it does not [ A. Individual start at the beginning and consider all-issues collectively. i safety topics are currently evaluated for all operating plants as they arise, but are not integrated'with associated consideration for each plant; therefore, a completely balanced judgement on backfitting cannot be made. In the program discussion it-is stated that'"the basis for all topics -Q deleted" will be documented. Why not document those included as. well? It is stated that licensees "will not^ be involved" in this - phase. Shouldn't they be consulted? Those topics retained for review will be documented by the evaluation' Although the licensees could undoubtedly be helpful in' documenting A. i ts el f. the basis for deletion of topics, it is doubtful that-they.would object. t In the interest of minimizing the impact on the l to deletion, of topics. licensees, it was not considered appropriate or necessary to.ginvolve-i. them in this activity. ---m__-_ m..___ _ _ _ _j

4 . Pace 3 Cont'd, Criteria for deciding on " safety significance".' each topic should Q. be discussed, A. It could be stated that general criteria for categorizing the culling safety significance of topics will be developed prior to of the topic lists however discussion of the criteria for determining safety significance should not be necessary te assess the merits of the program. Page a Reference of " disparity" of technical positions implies there are Q. likely to be " critical safety defects" if the disparity is great. Disparity between current technical positions and existing facility A. designs does not necessarily indicate critical safety defects. Departure from current technical positions must be assessed on a total plant basis to determine if significant safety defects do exist The older the plant the less adeouate the documentation is 1.ikely to O. be. We can't argue with this statement but where does it end? Why will this re-review be conside*ed "adecuate" by those in charge 5 years from now? Different people with different ideas may result in different regulatory requirements even when the plants, the operating conditions (and even the safety) are unchanged. A. The lack of adequate documentation ends with the plants for which SER's are prepared after January 1977. Subsequent re-reviews should not be required because of the procedures established for applying all future technical positions on a systematic basis'to all plants. This process is described on page'5 of the draft paper. In first sentence of second paragraph, who is on NRR review team? 3Q. Does it consists of all DOR personnel or do other divisions participate? A NRR review team would consist.of 5 to 6 technical specialists plus A. the assigned Project Manager and the 01&E inspector, as available. With the exception of the I&E inspector the review team would be NRR personnel. Consultation would be obtained by the rev'iew team from others as reouired.- ( j

. Page 4 Q. The evaluation procedures for the review teams is described as being " developed early in the program". Why not before beginning program? A. It is difficult to develop the procedures without concrete problems to examine. Developmer_t of pmcedures in conjunction with screening of issues and with specific problems in mind is considered to be a more efficient way of proceeding. Q. Isn't the team responsibility described in last t,antence (just before item 3) more properly a responsibility of NRR management? l A. This is a project management type function to assure that all required activities have been properly scheduled and accomplished. Q. What about impact /value assessment in backfitting decisions? This could lead to decisions that are not necessarily technically " consistent". A. A value/ impact assessment will be made pursuant to 50.109 for all backfitting decisions, Q. The reference (item 3) to a " management evaluation" 30 months after program initiation is disturbing. All of the first phase is te take 9 weeks with a maximum of 3 individuals. If another 9 weeks (with 6 individuals maximum) is taken by test case, the initiation of a man-agement evaluation should be possible in 18 weeks, i.e., 4 months from program start-up. To expend 30 months of efforts before doing impact /value assessment, etc., a Why not make it a step-by-step program (ppears to be unneesssary. proc safety issues to be reviewed, test case, etc. with evaluation at ' each step? i A. Management review will be a continuing function. However, it is considered necessary to get a good understanding of the problems which will arise and the value of the review for the first eight / plants before committing to a specific program on the remaining reactors. i) I

---

___m__

M

a q .7 - l j 1 '\\j Pace 5 Q. Certain options are indicated to follow management evaluation. Why f aren't there options to consider earlier in the program? j l A. A number of options have already been considered. The management evaluation will involve a second review of the options in light j of the experience gained in reviewing the first eight plants. i I I Q. Reference to criterion of efficiency (line 2 from top) is not clear. Is efficiency the only concern? And in what tenns is efficiency jl I being defined here? A. The other options listed in that paragraph probably would be.as I effective in meeting the program objectives. However, experience with J the first eight reactors may indicate that the objectives of systematic l review for later reactors can be met with less expenditure -of manpower then case-by-case reviews would require. Q. Secend sentence of second paragraph of item 4 refers to a procedure whic.n we understand is already in operation. A. The procedure is fully in operation for staff safety evaluations to be issued in January 1977. The Task Force concluded that documentation of differences and the nature for their acceptability should be improved for staff safety evaluations already issued. Q. In third sentence of second paragraph under item 4,-doesn't " actions" refer to " procedures"? I A. " Procedures" would be a better word. The sentence'will' be corrected. l l l t l

~ s. .o s 4

• Page 5-(continued)

In first paragraph discussing resources required,sthere is a reference Q. isn' t it 9 months, i.e., 8 plar.:s, 4 plants per team to 3 years, at 9 weeks each plant? (see Table -3 of Task Force Report.) Each reactor review will require about 2. years to complete. However-A. the last plant review ' completed by a team will not be completed until 3. years after formation'of the team. Page 6 The magnitude (and precision);of the numbers is disturbing in view Q. of the many acknowledged uncertainites--which are to' be resolved in - first phase and~ test case of second phase. The magnitude of the numbers is based on previous experiences with A. review of.0L applications, full term operating licenses and generic l The estimates will be verified on.the reviews of the first I issues. .i 1 eight plants prior to proceeding;with additional plants. What does reference to '.'other technical staff" -(7th line from top Q. of page) mean? Other liRR personnel? "Other staff" would be almost entirely NRK staff with.some pcssible A. input from Ol&E. At end of first paragraph, it is stated that schedule may be shorter-if a ecmbination of case-by-case and reactor class reviews is performe j Q. j Why not do it this way? i Reactor class reviews are not considered appropriate for the first: 1 A. eight reactors which are each somewhat'un_ que. However experience with l i the review of the first eight reactors is. expected to provide be'tter;. insightintotheeffectivenessof.performingreviewsonaclatsofreacEors rather than on case-by-case reviews. .j 1a i -_-i____ _.____l__.._____,_

i .g. l Page 6 (continued) The estimate of 63 professional man-years for review of 8 oldest plants Q. i is to be compared with the 170 man-year estimate for the remaining 62 We don't understand the disparity not the relation between facilities. these estimates and those presented in Table 3 of Enclosure 1.

Also, if the oldest plants require so much effort, would it be justified?

A. The oldest plants are those whose design departs most from current In cri teria. They are expected to require a wider scope of review, addition it is expected that the efficiency of later reviews will be Table increased by the experience gained in the first eight reviews. 3 includes only the initial phase of the review effort and is consistent 1 with the manpower and schedule estimates in the Commission paper, i The manpower discussion is not clear. For example, there are references l Q. Don't to generic topics and " plant-specific" topics as separate items. e generic topics apply to specific plants? We also find it difficult to L relate the estimate of Table 3 (Enclosure 1) to the discussion. A. Generic topics apply to specific plants but not all plant related topics are generic. Table 3 is only part of the effort discussed'on page 6. I page 7 Q. The discussion of costs to licensees implies that the re-review.will only increase regulatory requirements. Aren't some requirements likely to be found unnecessary or not cost-effective as a result of the re-evaluation? I The design of the eight oldest plants and their safety' analysis depart' A. most f rom current requirements. If is possible, but not likely, that some existing requirements will be found unnecessary. i l

x..

7S %

s

, l.! ~ ? - 10 ' - l Pace 7(continued) I The cost to the licenske should be ' supported by some analysis.- Q.'. (It appears *to. Be almost as expensive as applying for;a new' license.)- o How much of the licensee's cost will be for activities that are likely. i to lead to no changes in plant or in operating procedures? What is L I basis for believing that: licensee can justify theseicosts to the-customers (or to' the-state PUC) 7,. The estimated costs to.th'e licens'ees: are based _on estimated costs 'to L A.- H apply for a new license.. While the secpe of review is expected to be. 1ess than that for a new license, the difficulty ofl review may be The fraction of the licensee's cost which would'not' lead to greate r. s be assessed without' l changes in plant or. operating procedures cannot i performing the reviews. 0ur justification of the ' costs are based on improved confidence in safety. 'The li'censee's justification to'their 4 customers' is a consideration for the' license'es. How can Comission approve a' program without any estimate of; impact Q. and value? What'. criteria should the"Coc aissi'on use to evaluate the-program? A. We do not believe the Commission should judge the merit of this proposed. program solely on an' estimate'of value and-impact. However, we have provided an. estimate of the upper bound on impact _ and^after completing the evaluation' of the first 8 facilities...;it may be possible to provide the value/ impact assessment'. The " Recommendation" provides no alternatives: to be consideredv i,eL Q. a "take-it-or leave it" proposal - A, Discussed previously,- l +. i 1 - - - - - - - _. - - -. - - -. -,. _ -. -.. - _. - _ -^ .? l

.4 l .i J . l

1 page7_(continued)

If the program or any 3 art of the program were approved, wouldn't' l Q. it be better to publisi proposal in Federal Register with request' l for comments from-industry and the public prior to full implementation? j J l The initial reviews will be done on a case-by-case basis. The scope, A. value and impact is expected to be somewhat different.for different Therefore a publication of:a proposal in the Federal Register f cases. for the entire program does ~not appear to be of value. 1 l s //

o REPRESENTAT1VE ALTERNAT1VE REV1EB -APPROACHES CONS 10ERED - 1. Status Ouo_ This approach will continue what has been NRC practice with respect to operating facilities. It will review individual. issues on.an. ad hoc basis. Detailed review of oldest olants; rely on findings of WASH-1400 for 2. plants from San Onofre 1 on The oldest plants will' require a complete review analogous.to1that performed for a new operating. license. The remaining operating reactors No. review are judged acceptable based on the' fi.ndings of WASH-1400. is required for them. Detailed review of the oldest olants; review of selected criteria or 3. tooics for San Onofre I to Surry 1; rely on findings of WASH-1400 for-later olants The oldest plants willcrequire a complete review l analogous to that performed for a new oparating. license. San Onofre 1 to Surry 1 will-l 5e reviewed with respect to selectv safety tooics. The remaining operating reactors are judged acceptable based on the findings of WASH-1400. No review 1s required for them; 4. Detailed review of oldest olants; review of selected ' criteria or tooics from San 3nofre 1 on The oldest plants will require a complete review analogous to that performed for a new operating license.. All remaining operating -l reactors will be. reviewed against selected safety topics. l } 5. Culled comolete tonical list j .t The staff 1 A complete listing of all safety topics;will..be orepared. will then evaluate each facility against all-topics on thellist and-J 1 provide documentation of the acceptability with respect.to current j cri teria. ) .f 6. Selected list of " Standard Review Plan" criteria J Selected criteria from the SRP which contribus significantly to overall safety will-be identified-and evaluated for'each plant. l 7. WASH-1400 type saf_ety study of. each olant A complete accident risk assessment using the. methodology of, WASH-140'J j will be prepared for each plant.. .] i 8. De Hovo operating license tyoe review q .i All licensees would submit a' new FSAR if appropriate, for 'a complete. E staff review which titlizes the " Standard Review Plan". i 1

4 g,,,,,[ Victor Stello , -2, V b) Marginal utility Tho' Reactor Safety Study has shcvn that reacters designed and operated. to meet HRC safety requirements impose lo'.t risks to the public. Major emphasis in your re-reviev should be on the deter tir.ation of " comp 11anco" with existics NRC standards in significant areas. Availability'/ of numerical risk values is of no particular importance in licensing reviews. c) thwarranted change l The' perfor=ance of quantitative risk assessments as the basis for licensing decisions succests that a numericci l standard for acceptable risk is availe.ble; it is not. l l NRC is unlikely to generate such i numerical standard without policy approval from the Co==ission end extensive public disetu:sion. Also, it cheuld be noted that the Reactor Gafety Study stronCly recon = ended against changing l the' character of the licensing process. With respect to appmach (2), partial risk assessments uculd aleo be difficult to accomplish for the same reasons enumerated above. We recorrend that the engineering insicht: frca the Reactor Safety Study as opposed to partial rick assessments be used to supplement tne standard licensing revious. We believe that such insichts can help narrow the scount of re-reviev effort required, by focusing attention on the more significant , contributors to the risk and can ascist in the resolution of these key subjects whi DOR believes will requiro safety decisions. In st.mmary, we recon =end to DCR that the insights frem tho,Beactor Safety Study be used in conjunction with the more f :ilic. licensing techniques to achieve DOR's re-review progra= objective. Ve appreciate the opportunity to advise you of our views in this matter and request that W be kept current on your further thinking about the re-review progr = and its needs. Subject to manpower constraints, we are villing to provido you vith whatever assistance you desire in this regard. Enclosed are a fut subject: that on a preliminary basis me.y require less re-review emphasis l than others. These initiating events end accident sequences were found to have a relatively small contribution to the public risk as assessed in l the Reactor Safety Study. \\ r <> /.- ... / ; +. Saul Levine, Deputy Director Office of Nuclear Regulatory Research Enclosure . As stated l 1

~.. e' 4 2.j e,. g. '8-W'f l, ~'

s.

Initiating events or accident. sequences having a relatively.small contribution to public risk in the Reacter Safety Study,:(1) 1 Rod Ejection Accidents Turbine Missiles. l Cold Water Accidents-RCP Flywheel Missiles. Reactor Vessel Rupture -Borca Dilution Events u l Steam Generator Ruptures- -Airplane'Crashesi ' Containment Inerting Function Iargest IDCA's L Main Stcam Line Rupture . Containment' Isolation Function i II ote that all of the listed excmples may not be' applicable to N ( a particular lh'R design. s I e e O e a fi e I t l :- n. f i,' e e L ' l / 1 i 5 i}

4 RESPONSE TO COW.ENTS ON THE DISCUSSION PAPER Page 1 The staff states that the basis for the systematic evaluation'is the It is need for i= proved documentation concerning operating plants. stated that there are nu=erous disparities between current' technical I l positions on safety issues and those existing at the' time of OL issuance and that staf f saf ety evaluations have not' been conducted l [ -concerning their acceptability. l Q. Why haven't safety evaluations been made? A. The staff has not thought these individual issues were of great enough saf ety significance to perform Lemediate safety evaluations. Because of manpower limitations, only items of obvious safety signifi-cance have,in generak been investigated for previously. licensed plants. As _ manpower has permitted, some items of lesser safety significance have been evaluated on old plants by topic and many are still in process. i l Q. What is the nature of the disparities? Do they involve potentially significant h'a:ards? A. The judgment of the staff is that most discrepancies do not involve substantial hazards. Older plants are more likely to have disparities ) l than plants licensed in the recent past. ~ I Q. I4uat are the major deficiencies in the documentation needed by the staff? A. The major deficiencies are a lack of definitive bases for the adequacy of certain safety areas and a lack of comparison of the "as-built" plant with current criteria. I L 1 i l ______.________________Q

• b 1

Page 1 (continued) . i I Q. Is the information'needed by.the staff that. information generated by staf f safety reviews 'or information to be s'upplied by the .f 4 licensee?. ) All information used in the decision process vill need to'be'at least- 'l A '. Some documentation will be genarated by the confirmed by the licensee. ( 1 staff and some by the licensee. -I Pate 2_ a L The staf f has resis'ted the ACRS contention for. rereview for' years .why. Q. is the staf f now reversin, its position? Is the rereview program a l e [ result of a reaction' to CY 75 operating problets coupled with the Pollard et al. episode? The CY 76 experience has been significantly 1 better. i The assumption of the question is false.. a. b. The basis for the systematic evaluation proeram is explained in the discussion paper. c. ACRS concerns are discussed in a later part 'of the paper. -? d. Th.! need for more effort on operating, reactors was recognized prior to-l the JCAE hearings. Part of the NRC response to the testi=cny of-Bridenbaugh, et al, was as follows-(Page I-5): .l q 1 l " A recent reorganization of the NRC.0ffice of Nuclear Reactor Regulation i has resulted in the creation of a separate technical-staff which is devotinL its efforts entirely to operating reactor safety and' environmental-reviews. This increase in manpower'available to' work exclusively.on operating-reacte will result in an even more systematic review and will provide thorough-documentation of the status of operating reactor: facilities with respect to current staff guidelines' for all safety and environmental ~ issues. " r -[

l i l ! Page 2 (continued) Q. The staff clains that evaluations on an ad, hoe basis does not result in a.fficient use of professional time, nor does it lead to effective regulatory decisions. Do you have data to support this concention? Each a_d_ hoc review would not involve all the items that a rereview program would encompass.- A* Reviews of individual items may, in the long run, cover as many items as the systematic evaluation program and would not include a review of the safety of the entire plant. Clearly, reviewing each of these items separately vill cause inefficient reviews due to the overlapping nature of many safety issues. i Page 3 It is stated that the rereview program vill dispel'hn atmosphere of uncertainty for the industry". How does the rereview program change this at=esphere? A. By making any changes required in older plants clear to the ut111ty at an early date. Q. The rersview program could be interpreted as a progra: in which the staff tidies up its paper work. What is the anticipated public health and safety benefit of the progran, particularly since " critical safety defects" (discust. ion at bottom of page) are,not expected to be found? ' { A. a. Some ite=s that may require improvement may be found, especially in older plants, b. The degree of confidence in safety margins will be considerably increas ed, c. Any remedial action on current or future saf ety problems can be better weighed because of the increased k=owledge of available safety i e margins. j l

ate 4 What was the past procedure for-jud ing that safety issues vere E 9 "of somewhat lesser significance" and what now makes these issues,.. of more' significance? A The significance of these ite=s has not changed, only the. determination to resolve these issues. The ne'v b~d procedures vill' assure do&u=ented safe'ty eval ~ Q Vas the past procedure so poor that none of the information gen decisions on all future plants. could be used? Past procedure of discuss 1ng only outstanding problems in safety evaluations A has resulted in non-uniform documentation of departures from current acceptane All available information vill be of use and vill be used in the t I criteria. culling and evaluation of safety items. Will the VASH-1400 data and metho'logTTe'used in' identifying s'afety issu What vill the basis for the safety determination for Q to be rerevieved?Will the basis be the sane as it'has been in che past each plant be? (i.e., a saf ety judgment) and how vill this basis stand up' to future Will the rereview program be sufficiently demands for a rareview? well-founded to eliminate the need for future rerevievs? - The knowledge gained during the WASH 1400 effort will be used in determining l A Safety.. deter-l the relative significance of the items. considered for review. 'j I f I minations vill be made by the same decision process as is used'for new plant: ql How well That is, decisions vill be made' through th'e line organization. the proposed evaluation effort will fare in the future vill depend on the-The intent of the evaluation program is quality of the evaluation effort. i to eliminate the need for future reviews on current, safety issues. / 3

1 4. (, , 1 i -l .) .are 7 i Table 3-shows the ti=e for the' general list to be d j 9 j veeks. ~ \\before'makingany'decisionontheprogram? A substantial manpcuer commitnent will be reqstred to develop the general-l .A It is not clear that such a list'vould assist the Commission to' decif list. q 'the ' issues involved in. the proposed. program. l l The program calls for "ea'rly" identificatie::of the list of safety concerns, Why does. it take so long "vithin' the firstLtwo years" to Q to be

• reviewed. -

As indicated in' Table 3 the generic culling takes only develop the list? .nine weeks and the specific culling takes another nine weeks. } The draf t discu'ssion paper was in error.on this point.and has been revised' A by deleting this phrase. 1 1 I

t I

Pace 10 Q. Why couldn't each group of specialists search its records to establish when each type of reviev was begun, when significant changes were cade, I

what steps would new be considered essential and how much safety f Improvement might be achieved by updating? l Suggestions have been solicited from the staf f on the topics to be. j A. considered. However, the objective of the syst ematic evaluation program is to provide a review of the overall plant safety margins l ( before determining that changes are needed. i i Page 11 l The staff indicated that steps have been taken to limit the potential J ^ ) number of facilities that need an updating review. What are these steps? 1 Documentation of the departures from the Standard Review Plan is A. a. i required by NRR Office Letter #9. I b. A systematic evaluation of each new requirement (e.g. SRP change or Reg. Guide revision) 1s required as discussed in the task force report. l \\ Procedures to implement this are under development. l Q. What records are missing from plants that have been reviewed in the last year or so under the Standard Reviev Plans? i A, Systematic documentation of departures from current acceptance criteria i -l are not available. i l 1 i l i i

1 Page 12 Q. What are the problens with seismic design since Newmark and Agabahian have reviewed designs at least back t'o the 1960's? primarily problems with plants in high seismic areas where the estimate A. of the appropriate acceleration values which should be used in plant design have changed substantially. Q. Why not do a step-by-step evaluation of the program? It will be done. Not all management details have been included A. in the discussion. l Page 13 1 l Q. What are the steps of the individual reviews? Will there be Q1's and Q2's.,etc.? The steps in the individual reviews are discussed in the task force report. Interactions with the licensee will emphasize informal meetings but will 1 1 include at least one for=al request for information. I Q. What are the criteria for determining when a change is required? Each licensee vill be required te provide a coopcrison of design against current criteria and to provide a description of any changes believed (by the licensee) to be necessary. This means a large part of the rereview progra= will be borne by the licensee. The licensee is said to need two to five cines more people to manage its portion of rereview. Are these people available to industry? Will dilution of operating personnel vith thosa necessary to support the rereview impact safety? i A. The method for detetuining the need for a change is described in the i With discussion paper and in the answers to the page 4 questions above. respect to manpower, there is currently no known shortage of technical manpowe: l in the nuclear industry. Operating personnel should not be impacted by the requirement for technical analyses. l l

.. Pace 14 Q. That about impact /value assess =ent in backfd tting decisions? This could lead to decisions that are not necessarily technically "censistent". A. It is unlikely that plants of the same vintage would incur substantially different impact. Pace 16 and 17 Q. The cost to the li:ensees is estimated to be from 2 to 4 cillion dollars

cr plant for the first eight plants (not including equipment = edification).

The staff states that a part of this incurred during the nor=al operation of the facility in keeping with ifconsing requirc ents. L' hat part of the one million dollar per year ongoing costs for analysis, changes to Technical Specifiestions, etc. vould be covered by,the rereview program? A. We are unable to make a more precise esticate at this time. .able 1 Q. What is the correlation betweed the age of the plant and the assumed rereview resources? A. The initial manpower investment will be large on the first group of plants reviewed no catter which group is selected. Older plants are assumed to have longer lists of topics that will remain af ter the culling process has been complated. Table 2 e Q. Surry 1 and Peach Bottom 2 vere the basis for VASH-1400. Is it necessary that these two plants be reviewed again? If so, for what reasons? A. Yes. See memo from S. Levine (attached) on the use of WASH-1400 in the licensing process.

Mp cecq% UNt TED stair.s j NUCLEAR REGULAT ORY CO..iMISSION ( '.t w.M. w,f 8 y Ngh S WASHINGTON. O. C. 20555 i. o e 5 %;h./. y ~v 'S*** JUN 2 4 WO MD10RANDT! KR: Victor Scel10, Director Division of Operating Reactors FROM: Saul Levine, Deputy Director Office of liuclear Regulatory Fesearch

SUBJECT:

DOR FI-PSVE! PROGRMI FOR OFEP.ATEIG NUCLEAR FCUER PLAliTG In response to your May 12 temorendum to Dr. Kouts, we cet with you and members of ycur Tack Force Group on the Re-Reviev Pro 6 ram to discu::s the alternate approaches being considered by the Task Force for conduct of t,he DOR Re-Reviev Program. At this meeting, two potential re-review approaches vere discussed: (1) a cocplete rich assessment gecorally similar to the Raac:cr Safety Study for " classes" of operating plants and (2) the combination of a limited risk assessment covering selected criteria and concerns plus the norra.11y used licensing review techniques. As stated at that meeting, it is our opinion that approach (1)'vould (a) f be exceedingly d4 m a" +o accomplich, (b) produce results that vould probably have only narginal utility for licensing purpescs cad (c) represent a currently unwarranted change in licensin6 practices. Inece points are discussed in detail belev. a) Difficulty . N

1) The size of the license application vould increase substantially in order to provide the additional detailed encineerins information needed for a

=eaninG ul rish assessment. f ii) Extensive nanpower would be required to develop and to reviev 'this information a*.d the additional work would undoubtedly e:: tend the review pericd. Further, it is .not evident that there exict sufficient nunbers of trained practitioners in risk ascecoment to adequately perform this task. .( 6 e a %}}