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{{#Wiki_filter:5017 'W Wl,9e%54 %Cb204)7A ... m--me ,7 i POLICY STATEMENTS 00CKETED Ui%C he Commbalon telles upon several there is no genuine controversy. Both of factors in directing the Ucensing Soords those factors weigh in favor of a finding Nuclear Reactor Regulation" that was gg pg 16M4). A6 adva{nc(e notice ofp and, where appropriate the staff to that any deficiencies between present y l conalder carslully the applicability of licensee planning (which complies with l 50 47(c)(1) for the limited pertad the Commission's pre CUAAD proposed rulem ' . Severe Accident necessary to finalhe a response to the late relation of 10 CR 50.47(b)(12)) Design C 'ilena." pu bd a @@ recent CUAAD decision. Because b and uture planning in accordance with 2.1980 (45 F1t 65474)is bei withdre Commission has not determitml bow,,or the finalinterpretadon rlplanning by's notice published else e in th s even whether, to defina what constitutes standard (b)(12) as a response to the bsue. adequate arratigements for offsite CUARD decision. wtll not be safety POR NRNA 8Fo*14AT60*8 Coor7ACT. Individuals who have been exposed to significant for the brief period in which Miller B. Spangjer. Special Auhtant for dangerous levels of rsdiauon the it takes licensee to implement N hal Policy Development. Division of Commlulon bellewas that untilit standard. S) stems !stegrat on. OUlce of Nuclear provides further guidance on this matter. In addition. as a matter of equity. the Reactor Reguistion. U.S. Nuclear Uceuing Boards (or. in uncontested Commission believes that Ucenalng Regulatory runmineloo. Washington matim. the stsff) should Orst conside, Boarda (and. In uncontested cases. the D.C. 20555. Tele %cne: (301) 492-7305, the applicabibty of to CFR 3a47{c)(1) staff) could teasonably find th,at there suppt.gssorf Any grown afhis before considering whether any an other compelhng reasons to avoid addidonal actions an requindte delaying the beensees of those policy statement sets forth the irnplement planning standard (b)(12). applicants who have complied with the Commission's latendom for hg ando h % wh & Such ransiderstlon is particularly Commission a pre CUARD section approprists becaun the CUARD Sa47(bX12) requinments.Whers mojy;ng s @ n nlated to nector decialon leaves cS en N posalbulty that applicants beve acted in good faith accidents more seven bn design basb modification or refr.taretation of reliance on the Commisalon's prior accidents. & main focus of this pla nrdng standard (b)(n) could reevlt in interpretation ofits own lation, h statement is oc dedslon procedures a determination that no prior reasonablenen of this faith involving staff approval or. optionally, arrangement need to be made for og. reliance indicates that it wculd be unfair Commission certification of new site individuah for whom b to delay liceulas whDe h Commissico standard designs for nuclear power consequences of a hypothetical acddent comphtes its responsa to b CUARD plants. It also provides guidance on are limited to exposun to redistion - nmand. decision and analytical proceduru for ~ in considering the applicability of to Finally,if Ucensing Boards find that the resolution of severe acddent istues CTR $0.4?lc)(1). the Ucensing Boards these factors adequately support the for other classes of future plants and for (and. In uncontested cases the staff) apphcation of to CD eX1h then should consider the uncertainty over the those Ucensing Boards d condude existing plants (operating reactors and continued viability of the current that no hearings would be warranted. plants under construction for which an operating licenae has been ap I Severe nuclaar accfdents an bJed)W f sneaning of the phrase " contaminated Wnfore, untd & Comission - injured individuals." Although, that condude ita U rem ad g/ phrtes currendy includu members of Web subhdaldg h h b b g.A reactor cme whether or not Ibers are - the offsite pub!!c exposed to high levels differenti. N Ucensing Boards could c,- ~. segus g jta==reences On-of radiation. b CUARD court has reasonab y find that any hearing clearly left the Commf asion the ngarding comphance with to C71 Ocbber 3.1980 the Commission issued discretion to " revisit" that definition in a 60 47(b)(12) shall be limited to issues an advance notics of proposed fashion that could remove exposed whica could have been heard before the avlemaki,ng. Severe Accident Design ~ individuals from the coven e of Court's decision in CUARD v.NRc Criteria that invited public comment planning standard (b)(121. T erefore. Dated si Washingtons DC this teth day of M Wm pgents for treating Ucensing Boards (and. In uncontested stay, t eas. seven accident issus (45 FR 65474). By cases the sieff) may reasonably For the Cornminiest another notice published elsewhere in conclude that no additional actions this issue the Commission is should be undertisen new on the g,,,g g' g withdra this advance notice of $trength of the present inte pretetion of n eW Cuimis* that term. pr_oposed emaking This policy rtatement is a revision of Moreover. the Commission bel.' eves

/

So FR 32138 i3 the " Proposed Commission Policy that Ucensing Boards (and. In s*$' Petdeaf SN35 ' 8@ 9 % - Staternent oc Senre Accidents add uncontested casu. the staff) could (pg gp g r, g 't;.,, Related Views on Nuclear Reactor esasonably find that any defidency %n 4 c Regulation" publiabed for pblic hich may be found in comp'ying with a s'.a Poecy Stahmerd Sevb RW comment on April 13.1983 (48 FR 18014). si n a (12 i 218 for the Accidents Regardng Future Designs Twenty six letten of coment oc the pu ses of to CE 50.47(c)(1), De low v-n and Existir>g Piants proposed policy statement were received. b nudear industry genera!]y pro abibly of scridents wbcb might cause extensive tadiation exposure - Aeosev: Nuclear Regulatory. supported the proposed policy statement Commission, ACnosc Polley statemb. 1 e j - <. : ands ested several modifica tions. 4iv} 3 E ch the citMam of the pW I ea e on n po to oNib y oIeu'*1a ' tussenan:nis statement de~acribes the E psNo** t rebd pe o d. e accident is less than one in a muho.n per po! Icy the Comrnission intends to use to focused on a qtion of over nif ance a yur of operetion). and the slow mooln safety luun mlakd k noctor on probabilist>c m assesament.. esclution of adveru reactions to accidents more seve e than design basis espedally when coupled with the overexposure to radiation cre generic accidents. Its main focus is on the Commission's Tafety Coal matters applicable to a' plants and criteria and procedures the Commission Development Program" (43 FR 10771 J licensing situations and over which intends to use to certify new designs for March 24.1983).h policy Statement nuclear power plants. %is policy was revised as a result of these A 8 (// -> CorninJssion policy Statement on Severe comments by the Advisory Committee statement is a revision of the " Proposed suggestions and critidsms as well as y-Acddents and Related Views on on Reactor Safeguards. %., Y Y M 4PR-37 87*2140397 851231 December 31,198tureset) PDR ADOCK 05000293 p PDR g gg

J n POUCY STATEMENTS Many changes have already ben Implemented in existing plants as a needed, to ensun that there la no undue result of the %Q Action Pian NUREG-risk to pubhc health and safety.In Sutement is reprinted along with other oseo and NUREG 0737),8infor(mationImplementing such a systematic information and appendices that provide resulting from NRC. and industry-approach, plants under construction thei perepect!ve on the development and le2plementation of this pohey and 10 v It sponsored research, and data arising beve not yet received an Opereting nietes to other features of the Sevu from construction and operating IJcene wiu be treeted euentfaUy the Accident Program. A copy of NUREG-experience.On he beels ofcurready name as the menner by which opereting 1070 will be evahble for laspection et ereitableInformatloa the Commisolen reactors are dealt with.1%et is to asy, e the Commiseloe's Pubtle Document concludes that exisdag plants pote as plant opecific review of severe accident Room.1717 H Street NW., Washinston. undes rial topublic beelth one vulnerabilities maing thle approach is not D.C Copies of NUREC-1070 may be and seet mepresent basis for considered to be necessary to determine te Purchased by oalhng(202) 27F20eo se action oa' generic rutemaking er esber adequate safety or compliance with (202) 27bt171 or by writing to the regulatory dianges for these plaats % NRC safety reguladons ander b Superintendent of Documents UA because ofsevert Facident risk.The Atomic Energy Act, or to be a necessary Covernment Printina Offles, P.O. Box Comnfesfee ha'i ongoing neclear safety at rootine part of en Opereting !Janse programa thet include: ine resolution of review for this class or plants. staat. Washington, W 300t>7082 or the NationalTechnicallnformation new and severalother Unresolved R the decision proceu for Senice, Departant of Comroercm3285 Safety leaues and Generic Safety Isenes: a new standard plant design-b Severe Accident Source Term an approach the Commission strongly Port Royal Road, Springfield. VA 22101. Program: the Severe Accident Remarch encourages for future plante - the Pohey Peacy statessent Program operating experience and data Statement affirms the Commlulon's A Indr86CW88 evalueuon regardmg failure of certain belief that a new design for a nuclear Engineered Safety Featurn and s power plant can be abown to be the Room se severe accidentlesees la nlated equipment. human errors,afety* acceptable for severe accident concerne th and Butament is prompud by to other scarcos of abnormal events; and ifit meets the fouowing criteria and staFs saf est accidents of this scrutiny by the Of5ce ofInspection and procedural requirements:

class, an the embehaud Enforcement to monitor the quality of

Demonstradon of compliance with

.g g,,g, gn, plant construction, operetion. and the procedure! requirements and criteria s7ons tukbmajarrisk g the pubhc maintenance. Should significant new of the current Commission regulations,

d. sad with sedioacgive release safety information become evallatile.

including the Three Mile Island frees podear[)eceive of the i plaat accidents from whatever sortce, to ques, tion therequirements for new plants as reflected 21 conclusion of "no undue risk. then the in the CP Rule [10 CFR 50.34(f). 47 FRCoaunheic'e s nevere accident poucy is technical laenes thus identified would be 220ej: that es Comunlealoe intends to take di ruolved by the NRC underits beekfit . Demonstnuon of kehnical rusonabh steps to reduce es chances policy and other exhting procedures. twolution of all app!! cable Unresolved d-of a wvem accidet Safety leeues and the medium-and high-involving subesentialdamage to the wb fs rule f,p#d "g",' reasser anse and es mitigate the ti le. n8 One important source of ww rullability of decay best removal g,nou endian acci h [,",eg, ".'g*3pgh, systems and the rollebility of both AC On Apr013,1giB3, the UA Nuclest (-)) spec 15c probabillstic risk assenments. and DC electncal supply systems; Regulatory Connaleston issued foe Each of these analyses, which provide s

Completion of a Probabuistic Risk Commhalu Pouc Statement on Seven pubbe comuneet e " Proposed detailed awessment of posible accident Assessment (PRA) and considendon of Accidenta and Re ted Views ce

~ acenarios, has exposed relatively unique the servere accident vulnerabillues b vulnerabibues to avere accidents. PRA exposes along with da 6*ighu NuclearReactorReguletion" 48F3 these unique features has been reduced that it may add to the eseurence of no 18014) The public comments ave been Generally, the undeelreble risk from to an acceptable level by low. cost undue risk to puhuc beslth and safety, mvfewed. and, on the baala of furbt e and cono n. m abanges in proceduns or minor design

Completion of a staff review of the H a gul g

p, modifications. Accordingly, when NRC design wtth e conclusion of safety to nguletory dwielon making and industrylateractions on uvere 88W "' 1 accident lasses have progressed acceptability using an approach that 7 ,ung,nd fu ucl at stresses deter ,,,gn,g C'M'"ministic ens (neering le%d""Cr'dL"."For",*id'o d ""~'***"'*"''d $ro a d"'" $ d 7 @ ~ f 2 rnsulate an inteested systematic approech to sa e;iminatloo of each Cristaen designe that are veristions of the prueent generetice of 1.WRs wtD be in has with fts legiale anclear power plant now opereting or mandate to under construction for possibly reviewed in future construction permit ensum het oudear poww pianu should signifleant risk contributors that stight ge plications under the guidelines entified for a val or certification of e m, q,,6,,,,,, be plant specific and might be missed

ddman * *= a sesser r ps%.ne. win, a e er===ra air obsent a systematic search. Following standard plant e4=ssaar 8===

the development of such an approach. Becanae &ls pobey statementis just or.=m UN.,i 4 -t onepart of a progreen. locluding en analysis wul be made of any plant the severs Acci i Rosearch Program, da _ m g.,,,,,,4,4 ev that has not yet andergone an appropriate examination and cost-for msolving severe accident lasues, the "'**m' **appamm a ' e anny w aseded a.susr effective changes will be made. if NRC staffis publishing concurren0y 8,' ( " ' M * *..ency - i.. with thh Policy Statement a report on am,sn pii. m h g.poa. ro y, j*@' gyp 'o ew eis rede c d is e Patty sw.mm "NRC Policy em Puture Reeetor Designa: *=ad 8ame .is n Decialone on Seven Accident luces in**d "'dd"'s 64 =arino er se mere m, ee. em sta$able for Q~"'"*""hopeco.a et he NRCs P bus 8* N N ***hdara Nudear power plant Regu!adon" "b

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(NUREC.-2WD). In this report the pobey m maedessi h n., s"'.D"6"um==uidb i Deeember 31,1985 (reset) PS-PR 38

4' POUCY STATEMDWTS pose no undue risk to public beelth and / safety, the Commission has esamined commitment (NUREG-coso. Task (l.B.8) (EpRI) on their " LWR standardized on degraded core accidents currendy an extenelve range of technicalissues referred to as severe nuclear reactor Future Plant Design Evaluation rela to sevm occident risk that have accidenta: Program." been ntified since the accident at

To evold unnecessary delays of It is suumed in this Polley Staternent hm Mile Island.FoHowing plants now under construetlen:

that, over the next to to 15 years. utility implementation of numerm and commercialinterest in the Urdted [. modifications of lent design and r, _x,,,,;,,,,,, r,,,,,,,= 1 States will focus on advanced light i yce p n % 3, m m. olents bose water reactors that involve Mgulato{ rres as dueloped m eerstion aru andar ens'um improvements but are essentially based l Acuon plan (NUREC-j e&er navuumpos further Dechts salees on the technology that was i jgg g.j by soy 4. FNj demonstretad in the destgn. e l-c.w 4,- - - O, 7 -

To'hchieve taproved stability and 100 of these plants in the Urdted States.

i i M"r #4 construction. and operetion of more than e M. {$[MM predictabuity of reactor sogaletion in a This polley should not be viewed as ~ manner that wpuld emerit impmved prejudicial to more extensive changes in r,i-i i. des fuhWrdeg t s-puhuc confidenas ir ourregulatory reactor designs that might be for immedisk m.h a decistor snakies. demonstrated during or beyond that rulemaking orobr wy denps k policies arosented ki this time period. Indeed, the Commission for these plants because d oe,m statement wGllead to amendment of encoursps the development and occident risk. Howewr. the occurrence NRC regulations, standard review plans commercialization of any standard of a severe seddent is ma likely at fw Hmasing ha or o&er h der bm$s est adsht make safey &mugh ease plants than at others. At each procefow and artleda es part d NRC's 6ts, such as em acidpod plant there wiD be systems, componeng %W makes aHowance forSevert Accident Program. Mis greater simplicity; slower dynande er procedures that are the moog re8Ponse to speet conditions irrvolving i significant contributors to severe each a the mult of the accident proarso,r events; passive heat I accident Ask.& intent of this policy development of new safety information removal for less of coolant accidents: statement is to provide stilities with of signincanos for design and operating and other charecieristice that promote basis for development of Corsedesio, procedessa. smore ofBelent construction, operation. guidance that wGI aBow ident1Bcadon of in accordance with the activities, and maintenance

- I*'Y'"D*N'procedurps to enhance these contributors and dmlopment of

'*d **"*"F-the appropriate course of oction, as views, and poucy developments discussed in this policy Statement the B. h/icyforNewMont App // cat /one D,.!= gesceptable . ham Commlulon believes that it la poulble to complete its ongoing reviews of new ihmduedon y plant designs with an expectation of No new commercial nuclear reactors __ M"Me,,ftlye"gy M. fully snool the severe socident have been ordered in the United States noT quesdoes in course of the review. einer December 1373. However, the ~jnenageant" includes occident his beliefis predicated on the Comminion has received uveral prevention, socident management to availability of results from the ongoing applications for reference design certail or retard its progression, and NRC. Industry Degraded Core approvals that are currently andar consequence mitigation to forther limit Rulemaking Program (IDCOR). and review. A reference design is one of the its eNects on public health and safety, vendor remich and insights from the options in the Conuaission's. De Comedesion plans to formalete en Zion. Indian point. Ilmerick, and other standardization policy. When approved approach for a systema 6e safety risk analyses.N review of standard by the NRC stag, a reference design examination of existing plants to dealgas for future Os provideo could be incorporated by reference in a determine whether particular accident incentive to ladustry to addrass sevm new Cp app!! cation and. ultimately, in vulnerabilities are present end who accident phenomena. Indeed, since July an Opera ting IJcense (01.) application. coes effectin changes are desirable to 1983. the star has completed the During the corresponding Cp and OI. ensure that there is no undee risk to public health end safety.in reviewe and has issued Final Dnign reviews, the NRC star would not Approvals (FDAs) for two standard dupUcate thatportion ofits review imp ementing sed a tyrtematic designe (General Electric Company's encompamd orits reference design l ePyroach, plenu unovr construction that BWR/6 Nuclear Island Design. CESSAR approval. %erefore, even in the absena have not yet received an Opmting II; and Combustion Engineering of new Cp applications,in order to j IJeanse will be treated essentially the Incorporate (s System 30 Design, provide guidelinee for the current same as the manner by wideh opersUng CESSAR). A severs accident review by reference design reviews, the reactors are dealt with, ht is to ny. a the NRC star of the CESSAR D design Commission has recognised the need to plant.spedfbc *sview of uvm accident ~ vulnerabilities using this a for forward referenceability is nearly promptly estabush the criteria by which considered to be ascenar;pproech is not complete.The reytew included snew desa' ns can be shown to be ' to determine asseument of alternatin dulgn acceptable la meeting severe accident odeguate sefety or compliance with changes for severe accident risk concerns. De Commisaloa now believes NRC safety regulations ender the reduction. In addition, the staH has been that there exists an adequate basis from Atomic Energy Ad.or to be e necessary involved with pretendering review of an which to establish an appropriate set of or routine part of an Opersung IJoense appUcation for Westinghouse Electric criteria.%is beliefis supported by review for this cleu or plants. Corporation's advanced presourtsed current opereting reactor experienos, h main ourooses of this Polley water reactor design RESAR-6p/90. In ongoing uvere accident rmarch. and Statemem mow-January 1984, the NRC found the insights from a variety of risk analyses.

To clartfy the procedures and RESAR-SP/90 appUcation for a De twultant criteria and rocedural requirements for licensing a new nuclear Preliminary Des Approval acceptable requirements m listed be ow.

plant; for docketing in May 1984 the .lt. Critstia and Procedural Requirements the inst 1 aa at p an PS-PR 33 Decernber 31,1985 (rout)

~ POUCY STATEMENTS web as a propoud custom plant) can 9 requirement may differ considerable Y shown to be acceptable for severe from one review to another. In addition. destped to secommodate all of the acchdent concarus !!!t meets the thelicenue la requind to ensure that host 21e environments resulting from the tonowing criteria and procedural the intent of the ufety requirements la complew spectrum of seven ace dents, regarements: accomplished during procurernent. they can contain a large faction of the

s. Demonstration of complianca construction and operation-radioloskalinventory from e portion of b proceduralrequiremenu am crite a It is rece-that there are a the spectrum of such eevm accidents,

ons, 6

o PRA MWo Nu d For example. large dry containments of the curnnt Commissie Inchadag the nr e Isl continue to undergo evolutionary @ g e n, g w M A reguartmenta la new plants as fg development as the results of research b consegeences of a wide epectrum of in the CP Rule [to programs and reliabulty dats from cor,-melt acddents; hence, fur ) mona

operating reactors become avaQable and rec ^eenents may be annecces$er ary or et resob o dd]

bb Unrudna as innovative uses of PRA in safety , dn epgrading current requirements to Safatyissum e aM high-decision conttxts suggest better ways to gain liznited neprevetnents of their priarityGenedeSafety1 ugs, dudas achieve the benefits of these methods existing espabiDty may be necessary. a opedd focus on suuring while guarding aseinst thdr limitations no Cosedulon expects bt thew e nbabW4 o y or impro uses. Wh!!e learning curves mettm wm continw to be sobbets for i hefectrif[sulpply systems; bihty o both AC of these ds w1U likely continue for a study (e.g., in the NRC reseerd) decade or more, it weald nevertheless W in turbe plant-epecfSe sto

c. Completion of a Probabilistic Risk be construction to cocedidak ele u the Z)co and Inaan Point Apesament (PRA) and consideration of expedance at varias stages of PRA probabms6e dak ownsentsh me sevm accident vulnerabilities the development and utilization. At the 1rrtegrated eyetems analysis will be PRA expons doog wie & Indshu present stage of devdopment, a number und to explore whether other 68 "'

of posidn uses of PRAs have been contstmnent types exhibit a functional dA oPu b e ty., demonstrated, espedally in identifying-containment capabWty equfvalent to and (1) nose contributors to servare ht oflarge, dry containment. d.Completloo of a staff review of the acddent dak that are clearty dominant Although containment strength is an design with a conduloc of safety and hence need to be examined for cost-important feature to be considend in

cceFtabQity using an approach that effectfw dak nduction measurw and (2)such an analysis, credits should also be senmees deterministic engineering those eeddee? wquences that are gins to the inherent anargy and analysis and }udgment complemented dearly insignificant risk contrhters radionuclides absorption capabultles of by PRA~

W m &mfon k @ndy the various designs as weD as other ne fundamatalcdteda listed abon d%wd in between cases are more % wm M M wrd apply to the staff's mim of any am problematic. combustible gues. design. In addnasing cdtada N aWeb According! It is dear that core-melt accident geshcantfw pubucation of.within 13 monb of the evaluations and enstainment failure ce shis uvm accident a, [* d ' ahd! statement, the steff willissue svidence evaluatiou should continue to be U ddtarnadves and on the forza, purpose and role that PRAs performed for a repreuntative sample ( '. emhatio alternatives to addiwa an to ptsy in severe acddent analysts d operating plants and plants under the annsolved and generic safety lasues and dedsloo making for both existing constructive and for all future plant and w watch im cost 45 econ and future plant designs and what designs. mee stuees abouldimpron nd kee minimum critada of edequacy PRAs our understanding of the containtoent sev should meet. Prom experience to date loading and failure characteristics for la evident that PRAs could serve as a,it the various classes oflaculties. %e Q d[

- 'UO' highly weful toolin assessing the dsk-analysu should be as reahstic as g bough QI reduction potential and cost-ponible and shouldinclude, where effectiveness of a number ofimaginati" sppropriata, dpamic and static
- ]s8j8Q 8j8 "g,* ogy g" design options for new planu in loechngs fme combusdoo of hydrogen comparison with desfgn festi;.rea of and othercombustibles, etatic propure co 6*

Q "" descrh b appropriate combination of and noo-coodcasibles, basemat existing plants. %e PRA guidance will and temperature loadings from steam social algniacance ed cannot naddy determinfsde and probabilistic penetnetion by coneelt materials, and be quanded in commensurate unna. considerations as a besta for eevere abcu on aerosols on engineered eafety "C" 0*t*U h"g sec6ans explain in accident decisions. futures. A dari$catioe of contemnt %e foDowin

- - - *dI*'I* "" I' ne proposed Commission Policy performance expectations will be made appbed to se vadas types dmiews Statement oc Severe Accidents tasued including e decision on whether to est he staRmay amounter. It is on Apr013.1983 recognises the need for utabuA new performance crtteria for
b a striking a balance between accident containment eystems and. If so, what these should be.

u a t n teria vention and cons uence mitigation- { exploring b needgor eddltional listad above beforeffad a valor

Ik Commissica.N. neoplses the casadCt is owent' deefgn or opere tfonal features in the importance of such potential that a new design can & rough next generetion of plants to mitigsto the contribwors to uvm acddent risk as

&n==i a s - d = befm ncetring 61: Baal ap= peal w couequences of corecelt accidents. human performance and ubota.%e c 6e e-deeion wm,t,b a wience Issues of both taalder and outs! cardcados. between socidentprevention sn i aawta,,&,esu a w m duny

e. a n anal and, to the extent procticeble.

g,7 slp consequence mitigetion encompast Appegn@A) and 6e a M Mgn actions that improve understanding o will empbesised as spedal A roval(TDA).%e unique containment bull failure conMdentions to tbe design and in the 9 circumstances of each dulgn review chmetedsdes and Hign futurn or ohrating procedures developed for new wil therefora. nquin flexib ty in w sedom ht decnue & mm 6 &cdveness d gg3[ood of contatnrnent building Am wm W empbstzed

O

(( g,' t d failures Althoogh not specifically in design and opersting procedure development. A balanced focus w1D be Denmber 31,1985 (rnet) PS.P440

POUCY STATEMENTS peld as te negative impact of human demonstrable progrees la safety he use of PRA in a twcretep review f /

./

+ on severe aceldent risk as performance. including the reduction in process also raises a number of ' wet as its potenuaDy poeltive frequency of accident precursor events questions. Of particular concern is the someribution to baldne or kmiting the as well as a Al=InMed controversy tindag of the PRA requiressent becaum acasequeman of severe occident among experts as to the ade proprussion. Design feMuros abould be nuclear safety technology. quacy of the completion of a comprehensive and detaued PRA may not be achievable in emptedmed that reduce the risk of earty

Further progress in severe accident the ebeence of essentiaDy complete and

~=ad====t faAnte, thus providing more risk reduction is a hedge against the Saal detailed deelen information. tiene Imr he positive contributions of possibility that current risk estimates brefore, to require a menplete PBA at operaturperformance in curtsRing with their broad ranges of uncertainty the PDA stage would not be esahatic. setuse assident consequences. Also, might unwittingly have been he '-=iaaton's recent r design immeures Aould be given special OPtimleti biased. howmr. Indicates IM a tial seemeen &at serve to decrease the role

Al the severe accident risk amount of detal! thet would gf hamnem error la the sequence af events of an plant may be perndt llalted. quantitative Iseems to the Initiation nr agravatico acceptable in terms ofits direct oNsite risk analysis exist at the PDA et asse abursdaden. In pardcula,'

regional consequences for public health stage. Because the Co==niadaa believes metodo d analysis and associetad data and safMy, the aggregate pmbebillty that risk analysis of this type would be a basse are ander devdopment by es I*ey, over a 30 year period) that one useful design tool. the ws.wa. r severe accident emm accident wtB occurin a large expects eat H would be osaplMed as Cassedsska's ,s.g g @oh,,oe negaove and populauta d macte bold.s a.separete the analyses and part d 6e PDA cados prueen. A and addios eiga . o. eveni mumpw. m y.is wo.id noi be a .o,,e,, .c ons n would yield adverse spillover prerogeisite forissuance of a PDA, N2 m ild or consequences for innocent parties in riowever,if this risk analyels is not g,,g,,g,, other regions (l.a nuclear. oriented performed to the pDA process,it wG1 R is amend that same d6e severe stilities sad their omstomere), not to neve to be provided as part of any CP mandon a changed poudcal applicadon refqthe dedp. eccansat sammarios result in ' 7N savironment for anclear tion itself If the scope of the rpA reference of eGalte consegusaces. e5ecting resource costs design applicati.m is limited to an extent con *=w fectiveneescIn o programmatic activities. that would preclude the completice of a his aframtion, there may be ao sleer hems Imr segulatory acuos bocesse there

8. Application of Criteria for Diferent maaalagful, "we PRA, the is as embetantial efect sa public bulth

. Typw of OL and CP Applications reqdrened foricompieb PRA may be waived. However, the applicant abould or sedety. However, the implementation

a. Applications of Certification of stiU perform and submit supplementary of requirumments to control occupational Ae/erenceDesigns with No previous risk analysis, to the extent practical, to exposure abould be considered along FDA. In accordance with the d==aastrate the adequacy of the wie sie relatively small e5ects on Commlasion's standardizauon design. If a n J-- "ve pubEc heefth and safety for these types regulations and policy, a new reference is not submitted for an FDA, a G/

, of areare socidents.The resolution of design can be subadtted for approval OL applicant referencing the approved t sost-bens 8t iseuse in severe accident Aret as a preliminary da,lgn and then as design would be toquired to subesit a dedulos saaktag is part of the NltC's Baal design. Ca,~sdq!y, the stag plant-specific PRA. For standard destgo Safesy Coal Eralosuon Program. will1aene a Prefirminary Design approvals of restricted scope, additional Alhoid in the licanaln, of axisting Approval and a Final Design Approval. limitations beyond the PRA as r may plames thr==laalon has determined A PDAis not bowmr, a premiuisig existSu of sud a standard ign by &M tese anh pow no undw riak 2 for an PDA. An applicant has the option the bonnes applicant may be limited by (bac tb and salsty. ele dodd ad to submit FDA4evel information initiaDy its very nature to a twHtep Bosesing virwed as imply 2g a Comadulon and proceed directly with an PDA procnes, namely, a Construction Pernait pahey that safety improvements in new review.%ese options remain and on Operating Ucense issued pism demens doeld not be asevely anchanged by this Policy Statement. esparately. This would negate some of After a PDA a Ucation is docketed, the benefits envisioned for en approved sought, no Commission fuuy expects the preliminary bign can be est sendore engaged in dulgning new or certified design wherein a previcua}y es.-a.,d (or omstom) plants will achieve referenced in a new Cp application. De approved site could be matched with it a b@ee standard of severe accident corresponding OL sppuestion would in a one-step. combined @/OL process. safer p=ad-mance than their prior than refmace the approved final design The reference design most settsfy dseigns. This expectanos is based on: (yDA). Of couros. an approved deelsn each of the criteria stated is Section B.2 could also be referenced in a new CP before an FDA can be issued.For

The growing volume ofinformation application.

fa. ward referencoobGity of a new Irms industry and goverumment. ne use of an approved standard standard dalen. the applicant is being apa===ed research and opereting design in new CP/OL applications has aHorded in this Policy Statement the reacser emperienos bee improved our received considerable attention under flexibuity of choosing between a h* L of speci8e severe accident the Commnaslan's legislative initiatives Pr=hminary Design Approval (PDA), a valasrobeties and oflowmoet methods on singlHtep licensing It abould be FtnalDesign Approval (FDA), or a for Smir mattigation. Pether learning on noted that a two-step review procnes for Design Certifloation (DCL The safesy vehnerebQldes and innovative a standard design approvalis not. in approvals (La., a IVA or PDA) be meemds is to be'aspected, itself,loconsistent with singlwtep leeued foDowing the completion of the

The taberent BedbGity of thle IJeonsing.To be tnost effective, single.

stafra review and would be subled to pober Statement [that perm!ts risk risk step licensing presumes the existence of chaDense in individuallimosing tredenSs la eyetems and sub. systems a previously approved design bearings. ne Design Certi6 cation design) encourages thereby innovative essentiaDy an FDA This deelgn could would tie issued by the -enhaion e ways of ashleving an improved overell still be approved in a two-step procou following a rulemaking and systemma sotiabulty et a reasonable cost. as long as both steps were completed in could not be cea!!ensed in ladi e pubac acceptance, and hence advance of the single 4tep licensing bearings. Os or Ota, based on a j invasear acceptancs, of nuclear applica6an, reference deelen that has not been y,s todmakuyis Amp ad at on e's.PR 41 Decemba 31,1985 (rout)

4 POUCY STATEMENTS 7 r e- - ' through rulemaking, abs 0 be b requot to permit the design to be be a matter of separate considerstiae eM4 met to any deelgo changes arising referenced in new Cp and OL spart from this Severs Accident Pobey he the rulemaking proceeding in applications for a Exed period of time. Statement accordanaa with the t'a==laaion's such as five years, & amended FDA

d. A New Custom Plant Constructed bachat poucy and reguistiosa.The wiu be conditioned as appropriate to permit Applicatiort Itis the design cert 1Scadon'would be leeued for ensare that new Cp and OL applications a longer duration than a dovign referencing the design will utisfy sech Commisalon's policy to encourope the use of reimace designs in future CP appewral & speci8e requirements and of the criteria in Section B.2. N seven applicauona. This does not, however.

g % for obtaining design accident review most be completed prior eartfacetions or appvevals wiu be to the issuance of the new CP or Ola precide b um of a custom deagn. Custom designs shad also be reetewed

==+amakad in a forthcoming mvision to (2) Criterion B.2.c requires the against the criteria identified in Sachen to %==laataa's Standardization completion of a comprehensive PRA.!f a r Pokey Statement. comprehensive PRA cannot be B.2. As a result of the circumstances and h Spreenlar Certi/icot/on of completed owing the the limited scope timing involved in the ongoing p*=adard taference Designe Previoue/y Granassf of the design, b applicant aball design review processes, the asIn4. In 1983, the NRC staffleemed perform supplementary risk analyses to Commission expects that most. It not all, two Ptnal Dalen Approvals for the extent practicalin support of the ww CP application incorporating a refeream designa. These were approval or rulemaking process. As sference design woold be besed on perm #tned to be incorporated noted above. the !!mited scope of plant essentiaDy final design information. This refarance in OL applications e. the design and PRA analysis would lead to will result in !:nproved safety and ammaponding CP application had a partialloss of benefits in that a two. regulatory precdces, as weU as reduced referenced the PDA. However, the step CP/OL ucensing process would be time to license and construct a socisar desiumis were not approved for required in Doo of a one. step process. powerplant.To obtain as much of this incorporation in new G applica6oes. (3) With regard to completion of a benefit se beticable for a custom design app tion, the Commission will N r% = = tam w now baueves that comprehensive PRA for a mfereues these designs are suitable for use in new design the Commission recognizes that require a CP applicedoo for a custom a PRA would be more meaningfulifit dalen 2 incide design idorma6am G and OL appbcations under tbs maammaa. M below. Amy were based on a substanual portion of that is sufficiency Real and complete to t to these designa, the complete facility des. W refore, permit completion o't an adequate plant. &an thoes resuldag from the if Justined to the NRC sta. compledon optciBc PRA. It is possible, however. that an severe accident review, wiD require the of the PRA by the FDA appucant may be aproja plicant referencing an or cordAd design in Bos da desismo to be considad under tbs waived. !f a comprehensive PRA is not provtsions d Sec6cm B.3.a, t.a., as new submitted by the FDA applicant for the cwkm plad wmld have in prospect a y designs. FDA a CP/OL appucant refmacing the sisshdy rdad llun&g Ims since s (1) Each of the two reference desipe daign would be required to submit a staff sffort would not be required-er I. ' ') appacants with existing FDAs most plant opeciSc PRA. much less would be required-gar a regemet that their FDAs be amended to A rererence design app!! cant rmykw d b {d or certi6d (-] g gp ,,, gg, param their designs to be referenced in pavinely granted an mA can pumw new O and OL applicadone.N the same options of design a valor detailed changes to socommodate rogamut must sibe 0) taciude the design certification as desen in the gg ,,g,,p,gj tidormation needed to estisfy each of pmeeding metion fw mfennce des circumstances (e.g.,innovetive the artteria stated in Sec6cn 5.1. er DI) with no pmvins MA. N NA w equipment designs to meet new ASME or IEEE codes, etc.) ,,,,,' [I s7bject # C M cfJ 6 3 0iSr @ d ite - - CPad app 6eas the ta a re e and o referendog the deelgri wiD satisfy each (chauen n dMd cans e, 1.Some General Pnnciples of PoEcy of the ar(teria in Section B.2. Requests in we imud by Commluton D'Yelopment 8

- d db doise cadame following a tulemaking proc and h Crmaleslon bas !! censed about to te haMud Review Plan DM could not be challenged in indivi ual 90 cuchar plants and expects to process 8U8CEE hearings. Cps or 014. besed on a app % cations to license approximately 30 s

h he Bret osas, the staf wiU amend reference design that has not been ad 11tional plants. The Commtamem has j the exisdag FDA spa moeipt d es apfed rule =akw haD be coasidad at length the questiam of s m9 esse to,pennit the daign to be from)ect to any changel arising whebt ge: eric rulemaking ebanid be su in new &and OL the rulemaking proceeding in undertaken or additional regulations -app ootions until the amm accidset accordance with the Commission's should be issued at this time to require h mv6ew is completed. N amm backfit polley and regulations.b more capability in opereting plants er mared==d review must be --- m._ daign certification would be issued for plants under construction to improve ommpleted pr!w to the lesuanos d'"any a loeger durecon than a design sevm a,cedent prevention. 4 new G or OL whom appucauces approval The spec.ific requirements and conuquence mitigation, or accidmet referosos the desLgnD the procedures for obtaening design management that would halt or deley e===='u! completin of the ervem certi$ cations or approvals will be furbr com degradation. ear 4d==t mytew,the staf will further establiskd in a forthcoming revision to & TMl accident led to a number of amend the FDA to pwmit the design to b t' Mon's Standardization investigations of the adequacy of design b refersnood in new CP and OL Policy Statement. features, opere ting procedures, and appace6ces for a fixed period of time.

e. A a,,cijver,d Construction Permit personnel of nuclear power plants e such as five years.

Application. Becean of the many provide assurer.cs of no undue risk be the somed case, the sta5 wiD complex factore involved, the criteria regarding sever $ mector accidents. & and procedures for reguletory treatment amend the existing FDA spon receipt of of reactivated Construction Permits will reporPNRC Action Plan Developed as a Result of the TMI-g Accident"(NUREC-December 31,1985 (reset! -_-________________-____a

POUCY STATEMENTS cosa Wy 1980) describes a i comprehensive and integrated plan reliab!!ity of both AC and DC electrical source, which brings into quution the ~ supply systems;the Severe Accident i involving many actions that serve to Source Term Program; the Severe Commission's condusion that existing increase safety when implemented by plants pose no undue risk. then at that operating plants and plants under Accident Remarch Program; operating time the speciBc technicalissues. experience and data evaluation construction. The Commisalon approved regarding equipment fauure, human suggesting undue vulnerability wW - items for implementation and these are errors, and other sources of abnormal undergo dose examinedon and be i

TM] Action Plan Requirements" events; and soutiny by the OfBee of heridled by the NRC under existing identi6ed in a nport, ** Clarification of Inspection and Enforcement to monitor procedures for laeue resolution including the possibWty of generic rulemakin (NUREG&37. November 1980). N the quality of plant construction, when this is justifiable.However g staffissued further criteria on operation, and maintenance.De Commission will maintain its vigilance NRCs amperience sugests that aalsty emargency operetionalfacilities (NUREC.as7, Rey,1), auxiliary in these programs to offset the issues disevered through operating feedwetar system improvements experience programs, quality assuranos leertved from NURECAse?). and uncertainty of whether significant safety programs or easety analyses often instrenestation (Regulatory Guide 1.87 luun remain to be disclosed. Industry pertain to unique charsetertatics of a Revision 21 research and foreign reactor experience spectSc plant dealso and, therefore, are are also meaningful sources of dealt with through t-specine he TMI Action plan led to the information.

modiBcations of to tively modest oost m.J. --ts of over s.400 seperate One important nooros of new rather than mejorseneric design action itaans for opereting reactore and Bye Near Tern Opereting Ucenses. information is the experience of NRC changes. About so percent of the action items and the nuclear ladustry with plant. h Seve Accident Research approved for opereting reactors are now speciSc probabWetic risk assessments is progrom as weU as NRC's extensive mesplete and the remainder are that each of these analyses, which severe soddent studies of certala provide a more detailed assessment of todividual plants wm aid in determining ved to be Snished by the end of poulble accident scenarios, has te extant to which carefuDy analysad Sacelyear 1985. There were sat exposed relatively enique vulnerabilities nfmace plants can appropdataly serve diDerent types of action items approved to severe accidents. GeneraDy. the as surrogates for a chas of simGar _ in the Assion pian (an average of so unduirable risk from them naique plants as the haals fw any generic actions plant).Of this total,as featwee has been reduced to an conclusions. %sse studies wiD also aid. equipment beckatitema 31 acceptable level by low-cost changes in h h % b h M k ecope W tavolved procedurel changes, and at procedures or adnor des' y*P. mach fwloDowsp safety stadue of rogamed analyses and reports.it is modifications. A . when NRC v6du plants. Any generic changes impractical to quantify all of the safety and industry interactions on severs kaprovements obtained by these many chang== Neverthelus, the cumulative acddent lasues have progressed pubBeb d an b wW be,e, quired re 'efect is undoubtedly a signi5 cant sufficiently to denne the methods of trough rulemaking analysis, the %a=I==ta= plans to r e is e improvement in safety. Other information from NRC.and formulate an tatsgroted systematic gckB P'U*I' 5 industry-sponsored research along with approach to an 9= amination of each fauure data from construction and nuclear power plant now operating or ~'2. policy for Operatina Reactors ting experience have led to under construction for possible in hght of he above pdadples and la existing plants. Also, the significant risk contributors (sometimes condesions, te Commission's policy for NR has sponsored 11 plant-called " outliers") thet might be plant opereung ructore ladude the gegjowtog

z. e oper, guidance' ear"~'ple'is 1

specific pRAs and the industry has spedfic and might be missed ebeent a ung,,cg a sponsored many more.h evaluation of systematic search. FoDowing the severe acddent risk by the interrelated development of such an approach, an frequire ao thrtbar story sedo to deterministic and probabilistic methods analysis will be meds of any plant that go ),gth arewe a entluum udeu signt$ cant new safety information arises has identified many refinements of bas not yet undergone an appropriate x to quesues wheter &we is a to current design and operating practice examination. The enmination will that are worthwhus, but has identified include speci6c attention to containment L assurance of no ende risk to gepl)In the(latter event, a careful aa " '

~ % 5= : "

ao need for fundamental (or major) perfonnance in streing a balance ( changes in design. between accident prevention and . On the basis of currently evallahia consequence mitigation. la assessment shad be made of the severe implementing such a systematic acddent velnerebWty posed by the " n5 arm.un. u. r-musion concludes approach, plans under construction that imue and whethm this vulnesbWtyis let amtine e anta anu ne unha M have not yet rewived an Operating ier site ararde or of generic to pubbe besin and saferv and aa** an nr.

a. k-far wme6ete metion on Ucense will be treated essentially the portanna.

seneric rulemmHna or other resul** same as the spanner by which operating e he most cost effective options for . cbaneen for these plants because of reactors are dealt with.%st is to reducing this valnerebQity shad be plart specifle review of severe identiBed and a dedelen shad be " * ' + ,e sewere ocabdent nn-mw , the vulnerebuttles using this approach is not reach 6d ocasistant with the onet s 4 - ~ (described in NUREC-1070 and issued considered to be necessary to determine effectiveness critada of tbs ~8" w - -"e has ongoing programs concurrently with this policy Statement) adequate safety or compliance with Cammiaataa's beckfit policy as to which theiinchule: the ruolution of NRC ufety regulations under the optfon er set of options (if any) are futillable and required to be Unresolved Sefety lasues and othat Atomic Energy Act, or to be a necessary implemented. Ceneric Safety lasues, including a or routine part of an Opereting IJcense ve* la those lastances where the". special focus on assuring the reliability revlew for this clan of plants. W.4=e4wentissue beyced currest of decay beat removat systems and the Should signiacant new safety /], tory req ta, generlei.y a information develop, from whatever (,, edU be the preferrvd.F v PS-PR-43 Decernber 31,1985 Ireset!

_ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - ~ - - ~- POUCf STATEMENTS pas-sh= In okt cam, b lune should severe accidenta. N Commlulon whether mean or median value, is ao b be Espond of through the convenbecal belhves that considerations which so order of 1 chance in 10.000 per reactor ' pesetics of luulng Bulletins and Orders beyond bt to the poulble need for year. For most plants, only a fracton of or Cameric letters when modiacations safety messurn to control or mitigate the calculated sevan core damage an justiSed through backDt policy, or severe accidents in addition to thou throosh plant-epectBc decision maMnf required for conformance with the sequences an likely to progress to large scale core melt Uctil now, few analysts along b haes of the Integrated Safety Commission's safety regulations or hava even tried to take that fraction into Assenenent Program (ISAF) conformance with the Clarification of m6on.* 30 Action plan Requirements,'should wparato consideration, preferring even i Recopising that plant.epedSc not be addnssed in case related safety to refer to the pnviously calculated PRAs have yielded nbbleimight to hearbss. value as the core melt frequency. Of the sedque plant vainenb(D6a to sevm & Beparate Remarks of Chairman core melt sequences, ty-ically only 1in ,,,* tents feeding to low cost PaDadino and the Dlasentig Views of to, or los, an expectelto yield large .nemendons, Baneen oloach Commisaloner Anodatine are attached. n! eases of redjoactive material. On virtually every reactor sita in the United opersting reactor wiD be expected to Dated et Wuhington, D.C., this 30th day of States conditions are such that, even per$ urns a hasited-ocope, acddeot safety W tea 5 with a larp release, then la only 1 analysis designed to mecover inetances For the Neelur Regulatory Commiulon. chance in 10 of any early fatauty-and (La. outliers) of partienlar vulnerebulty seemet f sk, so on.nua, the wealth of risk estiinatu to core melt or to enuesaDy poor Secretary ofthe Comaussion before us indicate that the risk is quite containment performance, given core-low-melt accidents. new plant-specioc Remarks by Chairman it is often said that one should bewan studies wiD serve to verify that of too much trust in the point estimates conclusions developed from latensive I bebeve b Commission is on the of probabilistic risk assenments, that severe seddent safety analyses of right course with this decision. & om should consider the uncertainties, reference or surrogate plants can be severe accident polley statement nia we do. But some then go on to appBed to each of the individual presented bere is based on the demand exact quantitative de5aitions of operating planta. During b next two triuments contained within it, the the uncertainty.nis demand la a form years, b Comminion wiu formulate a additional support of more detailed of bottom line fallacy' f uncertainty erstematic approach, inclo the analysis in its companloo document Prectee statements o g,,;-43; of guhunes NUREG-1070, the massive support of the come ocly with large amounta af data. gm L.! criteria, with an expectation inany other related works of this agency At the very low levels of risk with which that auch an approach wfD be and others in this field, and a logical we are dealing the occurrence of actual implemented by licemen of b conalstency with other actions of the remaining reactors not yet Commission. events is, thankfuDy, very rare indeed. 8rstemati y inan in simple terms this poucy statement hs, we cannot have exact quantitative that existing plants pose no undue estimates of unce.rtainty. But we can and equivalent or superior mann.e gto pub!!c health and safety, and that must, continuaUy. explon the sensitivity ( of our cetimates and our dedstons to b 'l

3. Pobey for Opersting IJoense there is no present basis for regulatory AppBcetiens for Plants currently Under changes for these plants due to severe gape in our knowledge.We have been accident risk.nis condusion on reactor doing that and we will keep at it.

) Construction ne same wvere accident plicy safety does not lead us to diamantle our in summary, preunt reactors pose no guMawa appua to appucabom for regulatory program; rather we are undue risk to public health and safety. operating Ucenas (Ota) u stated above maintaining a vigorous program of his poucy statement acknowledges ~ fee opendg nuclear power is % surveillance, analysia, and evaluatjon to that and indicates a willingness to with b foDowhg addjbou iteMnis for'*** Posa!ble causes of accidents and permit continued operation of existing reactors as weD as to license new item also apphas to any hearing prevent them. In this perspective, the p,,a,,1bes est mi Ccamission has ongoing nuclear safety reactors. This policy statement has been opmung nector.) ght ads for an 7es:ms bt indude: unresolved safIt has been reviewed canfuuy and studied intensively for over three years,

Myidad hbg promdbes are severe occident, source term a endorsed by the Advisory Committee on research p

operating experience Reactor Saf ards.! has not been

,D7uon a d data on, and the scrutiny of lightly consi$ red nor lightly rt=4A I a plant construction, tion and am confident that the Commluion has evalostion, t and tigation of r 'informa enunciated a sound regulatory poucy. eeddents amore severe than b design a conclusfon of no undue riskestion th's Maunting Views of Commisat-- from whatever source, to basas (Class sk %e rew==taalm has then the Asselstine enacanced a pouey ngarding Class 9 technical issues thus identified would be scytr=mantal reviews and hearigs in resolved by the NRC under its backfit S its Stateinent oflaterim Poucy on

Noc3*ar Power Plant Accident poucy or ohr exhting proceduns.

b foremost risk to the pubbe frces Con =Ma stiene Under the National W 1evel of risk found to be b opustion of nucleat nacton dertns Environssental Policy Act of1900"(46 acceptable is wsU documented in b from wre meltdown seddents which basic works of the agency on these can, through b relean of substantial FR w1DL. June 13,1980), and expects to related sub}ecta. N calculated quantitin of rsdioective materiala, coctinue inis polley. W ecrironmental taanse deal essentially with the frequency of severe con damage, result in the injury and death of a utimation and description of the dak of catastrophe number of people.This 's to cra s.wn and "Swement of Pdcy, pohCy statement. whlch $$tablkhes raruwr commwan cwance tar Pwer a.. cia, Commission policies on these severe .S e 3.noreW Sdey AasenswW Progree Opmdrq tJceno*C 4e rt as2x Decentwr M. accident risks, represents one of the psAP5 sacy u.tn um6 n tasa tsan most fundamental regulatory decmons December 31,1985 (reset) PS.PR-44

POLICY STATEMENTS ever made by this agency, nis the technical support for ht conclusion and ecceptabiUty of ht risk is "6 stat =r==s togehr with thne other based on scientifically accepted si allable. reissed regulstory decialons. will chart principles and methodology. Second, the poucy statement does not the future coune of this agency and the Absent a detaued discunion of the go far enough in lasisting upon + anctsar industry on nudear safety severe accident rkk posed by existing redoctions to the seven accident risk of issese ler many years to come. De plants and of the reesoning and futwo plant designs. Soch nductions an three oeur dedstons are the scientific besis supporting the e 's decisico on the Commluton's conclusion on the much more readily achieveble in new

==ys=h8Hty of the severe accident risk acceptability of that risk, that designs for as yet unbuilt plants than for 1 st tw two opereting Indian point plants, condeston mest be viewed as nothing ex] sting plants. Whue b Commisalon's { te development of a backfitting rule more than en unsebetantiated essert6on pobey statement arges reactor designers a substantial safety ^m/ oflittle weight. to make safety haprovements to the for the impoeldon of new the Commiselon's polley destems of future plants, it does nothing wgmireummets together with heavy statement falls to provide any to requin that improvements be mede. roammas em quantitative coet/ benefit explanatice of the emnmlulon's 3rd, the Coedulon's poucy samtyees, and the dewtopment of a treatment of uncertainties in evaluating statesset retains the option of provisimal, and ultimate a final. the risk of sevm occidents.ne subrtning b Man of construedon d esamey goal with numeri standards for absence of virtusUy any explanation of futuvo u bend apce only HmitW eruhmsting b acceptability of nudear how uncertainties have been treated in h'.nl informadon,induding the ted dulge informadan which would occadset e---se= risk. Taken together. these four this poucy statement further andermines an actions will set the the nbdity of the Corn =laston's broad be needed to support issuance of a pp,u-in.,y design approval pDA fremmewak for dedding whether the conclusions on the acceptabuity of the PGtC and b industry will pursue risk posed by severe accidents. experienw with nuclear pow (erplan). Post t exandag and fumre significant safety Wrd, the Commissim falls to address du' ocostruction and regulation has h t ne b ' man i==== whetherfurther vements in in a clear and conslatant manner the design.as-yoo.befbpitfans W b old wEl be pursued for existing need to prevent forht seven reacto' approach. and pleu, and how sod accidents. Although b Commlulon a conunohg to eBow b Man d t m Gl be made. Policy statement pays lip service to this construction with only !!mited sign w ~. work lete, b Commission seems umfartmastely, the first two of these h to kcdude se means b ,,,,Di to repeating the mistakes of dedmans by the==fanion lead me to c est we an on the wron Foure, & r fulon's poh the post-mfstates which have led to r__.dh_ views opposing the g statement places undue rebancein ne defernl d etsnIBeant dulgn Isam c a se s a,na.,oht dedsi probabilistic risk aueuments (PRA's) unto the constnedon and pre.opmdon wome est forth in considerable detaff in ven cddent sa, and a need io.mdify wwk alind@ propw w ompM q the r==-taelon's written dedston (see qu on r a l. 1 CILalMis), and I wiD not rebaarse th" nUence faus to recognha prmat Takaa together, these flaws b the ' views hers.SalSee a to sey that th*7 weakneues in these assenments due to =alon's anm accident pokey p Comedesion's unsebetaanated and. > the limited number of PRA's avausble statement cast doubt upon the adequacy ' omty aptinnistic assumpuans en the he far, b variations among b of the Omn=tastop's omall approach to scooptabatty of the spese existing PRA's, the absence of accepted deahng with one soddent risk and stak posed to the psblic guidelines on how to conduct PRA's and undermine the validity of the tese plants how now been to evaluate them in making severe Connataalon's swwplag lodgmats of by hispolley statenset to owera5

accident risk Judgments, and b the acceptabulty of that risk for existing

- admeag and future noclear WM uncertainties inherent in attempting to and future plants. la his esentry.In my judgmset b. extrapolate plant specific PRA results to D8 ar-d== rm's action today fads to other plants. 1 "r-Before elaborsthat on the major provide even the most rudimentary /btun Monk inBraides d thh poucy statesment, it is

==yR==anam of, or justificetlee for, these %e Commission's policy statement is useful to explain what we know about [ eweeping condusions. As a basis for equally flewed in its treatment of severe the severe occidset risks to the pubhe, f etimaaldodeloa=aking the--a. dan's severe accident policy policy statement promises ht the ju,y ra accident risk for future plants. First, the staasmaat is a comp 1ste faDee. Commission wiu make Sael decisione la Risks am common}y defined as the N N#8 b near term on the acceptabuity of product of the probability that an event new plant designs for severe accident will occur and the consequences of the Isee at least four fundamental flaws in te mnminolon's poDey statement as purposes. At h same time, the policy event happening. In regulating the e statement acknowledges ht key nuclear industry, b Commlulon makes j M appbes to existing plants. Ftret, whue elements in evaluating the ecceptabluty extensive use of a methodology celled 1 the pou f = 4=.cy statement naches a positive of severe ecc6 dent risk criteria for the probabahtic risk auessment (FRA). In l = ce the seceptabulty of the preparetice and evaluation of PRA's. ""d-r'laf a PRA the analyst calculatee seimme amoident risk posed by existing containment performance criterie, and the core meltdown probab111ty and., ' 'plask B hDs to artienlete winst that - cHieria for evaluating b risk sivun a Particular core meltdows ; e irisk lu; R fads to ldentify the relevant ' contributions due to sabotage and . e 4-s,-a issow evaluated hiamessing human performance-will not be scenado, the analyst the instimatas the conageances to the pubIlc.no. Se b!!!ty of that risk:it faus to available for some time. Thus, b Commiselce sees the bottomline of explaim those technicalissues war, Commission's approach is to egree to these PRA's in dedding whether to e r==A-ed and resolved by b make final decisions on uvere accident impeove reactor safety or to relax the r==d-daa in naching its positive risk for future plants before the esiety standards even though such W andit fads to demonstrate technical basis for evaluating the nature FRA's do not n=dA=* all contdbute to n. V PS PR 45' DM 31,1985lroset) J

POUCY STATEMENTS core meltdows risks or quantify aD of N sprud in b estimated core (' the micertalaties. metadown probebuities for a typical b tune. reprewatation of our undentanding of A typical result of a PRA which le plant range from approximately one used by NRC in maching safety chance in one hamad (1ri per yur A serious consideration of the omre decisions is the estimated core to one chance in one hundred thousand nulldown risks would consider thus fell mahdown probabt!f ty of about one in (10 9 per year, with a unedien value of range of calculated risks and wand tee buand (or 1ri per reactor year. one chance in tan thouand (10-*) per address forthrightly the question of 3 However this probability estimate is year, give or take a few. However, there whether h risk in acceptable or chan based on what is esDed the is no f that the median of the unacceptable, both for the immedhate

===tian" value. it is important to cal led values reflects the actual risk fetwo and over the long ter m. '1%e a adentand inet wkt the meanlag of

DY more than do the estimates of Ir*

th-f aaloe's oaosideration of nesww accident risks instead focuses as a this bottomline aanbar really is, per year or 1r8 per year. median number, ignoring h actual Becsees of anajor inadequacies in the Another typical result of PRA's is the range of of values and the uncertamaties den bem, because of the vest prediction that about 1 out of to core inherent in es!ng a median number for complaxity of nuclear plants, becesse e meltdowns likely will result in lethel deciala 1% trummendous nunber of aneumpoone redietion doses to about 1.000 people. Since the foremost risk to the pubbe must be made in calculating oore Such consequences of core meltdown from the commercial nuclear todustry mahdown probabulum, and becsese accidents are attributable to degraded derives from severe accidents, adopting laryr scale com meltdown phenomena performance of the containment, which a policy that seeks to twoolve severe. am poorly understood, no e, can come about in a variety of ways that accident ismee to a definitive===a=r is catemtetion will yield a remotely are not precisely quantifiable. Because & most baalc duty which can be meaningful probabuity of cataMrophic of theu uncertainties in quantification. ,,,,,,,,, g,g,,,g pgg the fraction of core meltdown accidents undertaken by the Commluloa km ,,3 g,,g p g,,,em,, .a.,g which would lead to catastrophic anting its rupoesibility to decide what 7 constitutm acceptable risk to the Wual MmMas of the one consequences is actua y a range of & Commissico claims in this Mwn probabih & W values. range could be two or three statement to have examined as M times greater than the above estimate: or it could be two or three times lesa. extensive range of technicalissues i Petaas which themselm am not - pmanely known individaa! an=paa*=1 Picking the slaimum factor of 2 sad reladng to severe accident riska to t folke probabulues, basa error mtes, assuming there are too operating reaciu'ng its judynant "that existles and boretical models that are thought reactors, the approximate range of plants do not pose an undue level of risk to dancrb most of the important chances of a catastrophic accident to the public." The Commission's policy 4 between now and the year 2000 would statement does not, however, reicalprocesses or engineering be anywhere between tL2 [1 chances le incorporate an axplanatia or for that , Arry one of thne ladividual matter even a description, of the annet tho) sand). ten and 0. cot tone chance in a signmeaatissues that have been esthmates is as likely to be vahd as the ether thousands of calculations. There is Nrefore, b information before the Tuolved and b mannae in which they (. -I u estunate moulting from any one ofll* were resolved.Nor does it include a tw.i.alon indicates that there could description of the snethods of analyses a crectal, but untenable, andartying be anywbere between a ao perosat used to resolving the issues or decision assumption that au core meltdown chance and a at t chance of an critaria that were and for reach aaq===r a have been accounted for in accident at a at reacter in the next ultimate judgment. It is, therefore.s the the antimatse. N analyst then scans aD 15 yeam that would rwult la lethal lapossible to discern the bases far the of the estimates and picks the doses to about 1.000 peopla.The range of Commiulon's decision. probab0f ty value at which half the chances could be larger than this if one settamates are above the half are below. %ie sumber is caDad the median. It la, considers eB contributors to b core pgg maltdown probabihty and all A paramount concern regarding the according to the N==5asion, the "but uncertainties. Likewise, the amnber of wtimats". When calculated in this way, deaths could be larger or manaller. acceptabuity of the riska to the public that must be resolved is how to reach a however, one cannot say with any Admittedly, there are many ways of judgment on this issue in the face of ream =nce that this median value is the going about estimating the ranse of enormous uncertainties which are op to tree som meltdown probabuity, riska. However,if them la validated too tisws the audian value used by the Noneth6less, the Co==fa=4a= quantitative information on care Commission. Depending on how much i airbetrarey choosee this mediaa sunber meltdown risks that is better,it has not uncertainties are factored into the } to one la making its regulatory yet been d==aaatreted. be, becaem decision. judgments could range toe d.,6. sana i of the amany uncertaintin involved in req substantial e!! orts to reduce calculating both the probabuitia and tdows rinks to doing nothing core sw %,,,,,,,a,,,,,,,,,,, the consequences of core maltdowns, about them. Scientifically accepted data a meer meeses.e by e. Aa.tmer commsem en one number dow not give a true pictum and methodology are not available at n mm,aar res entse nepairat som mesame wie she cammina m The Acas roommmmend me of the actual riak. A range of this thne to red" 4 substan toes poealbusties le a more accurete nacertainties.o eat, as the "",,", QSyg*,gM",,h staff of the NRC has repeatedly told the escammene ens rummh in a substamalaa modoresenate Game PRA andyees hear noir estmense se to CoEunissioit it is " mandatory" to w n. es of uws. m s now. e-- h i.tes madam h that pelat en a spectriam et which half of endereed one of te mesas,shesL The test uma comaldar them la any application of dak

==a=# reek esametes As hadecetud a es m esame tan eh.= and har ein honow.The was whom the Cammhsk s

dweed WASH leoD assessments..-

h being informed of b "' ~ > the e.wess valm of as seanner of rinha and to ga ener s.reer sesdrl a tv s and en me d em. L

== when eer - ap uncertainties in the risk estimates'.'the ah smand he W enks? amorey cant pouce sensesar (er d the pewtwood h=idad~ almply ignores theen.The Nunarassa cam d ant ila tsaa tww im gans to provide any basis y-t Deceanber 31,1985 (reset) PS-pM6

4 POUCY STATEMENTS for Rs decision to ignore tha* estimates" of the core meltdown risks calculeted core meltdown risks by half' r uncertannues. Absent some totional without any consideraUon of the effects Unless such a reduction can be i treatment of thm uncertainties or a of the uneartatnues.nis hpproach can " demonstrated", the Commisalon will 1 canymdng justificadon for why they can leed to a decielon to dolns nothing to not consider requiring b change.nle be ored. the pebbe can han little mduce core meltdown risks. Factoring is a much higher barrier to requiring once in ee Commission's into the decision the uncertaintin in improvement in reactor safety than b I conducion that b risha to the pobhc ntimating the level of core meltdown poucy statement would have a belfeve frase a severs accident et a nudeer risks would lead to a decision to search is the Commisalon's policy. powerplaat are neceptable, ne only for ways to reduce the riska. However. Further, the th=wlan's provisional evnllaoss explanation of the NRCs - given the current political climata, there safety goal is not intended to regulate on approach to making decisions in the la little sympathy for beckfitting existing the basis of preventing core damage Goes of these ab=h sacertainties is planta. Ms, the Commisalon chooses to accidents, as implied in the above given om pages las through 140 of rely on a faulty number which supports purported fundamental objective. NUREG-tetVNRC pobcy en Fueue the outcome they prefer and to lenore Rather, b safety goal assumes that the Reactor Deelgne Decislams on Severe the uncertainties those that are hnown containment is an independent bulwark Accident leeuse in Nuclear power plant and quantified and thoos that are not capable oflimiting the external relene Regulation". October tes4. About half of quantinable. of todioactivity to modat amounts for - the pages am blank and os rand =' Whot level of confidence does the most oore meltdown accidents.ha, are not much better.hls discussica of Commluion have in its hdgment bt according te b Commission, them is no uncertainties is inedequate and fails to core meltdown accidents preunt no need to mgulate on the basis of provide a sufficient basis to justify the undue risks to the public? ne preventing core meltdowns. I am not as r==tasion's sweeping wacloaloas on Commission nowhere emprases b sanguine as the Commisalon on b the acceptabluty of the severe acddent degree of confidence it seeks to ensure acceptability of core maltdown Fieb-that catastrophic accidents do not accioants.Evenif the containment Another fund==,-tal ls'ous happen. Yet, the hmtaalon's chlaf happens to retain moet of the r emselutles is the level af risk to 'e safety officat recently wrots "In view'"p[ redjoactive fleefon products in the neut pubtle that r===anably abound be femad the uncertainties severe accident, another accident equal Beyond making a sweeping of secessing severe t' to or more severe than that which uslon bt the severe meddest risk -"rlek,the levelofassarumos (or occurnd at %ree MDe faland would be st die existing plaats does est pees an confidence) of no undee risk to b unseceptable to the pubhc and b undne risk to the public, tbs Commission ' public is regarded as no less important Congress and would be diastrous for anile to addrus this fundamental than the estimated levelof risk itself the nuclear industry and b NRC. question. la fact, the Commiaalon's (emphasis in the original)." Latter from But more importantly, the technical staff is t now embarking on H.R. Denton. NRR. to A.E. Scherer. Commission's belief that the f.- a program of is that "wiD form Combustion Englosering. be dated containment w!D retain aD but madant part of the baals for a ammiaa" December 33.1984, subject "SECY.as. amounts of todioactivity during most r judgment oc the level of safety presently am Severe Accident policy =4 core meltdowns is not yet espportable adieved by existing plants for sever

Another problem with the based on scientificeDy accepted occadents."'Since the a==3=ah is Commlulon's polley statement is that it Principles and methodology.There r

just begi this program,it cannot clearly contradicts what the simply is no actuarial experience er earve to jus the Commission's Commtmalon is dolns in other areas. For direct experimental data ca large scale }edgment on saceptability of the example. in this pobey statement the core meltdown'phaaa-ena or sevwe accident risk. Commlulon states: "A fundamental containment performance i am its Indiin point dedsloa, the objecuve of the Commission's severe charectaristics given a core nultdown. r==taalan adopted specific point accident policy is that the Commlulon In the past, estimates of the quantitiu of estuostas of core meltdown risks for the intends to take au reasonable steps to radioactive releases to the environment i Indian Point reactore and found them to reduce the chances of occurrence of a beve been based on not much more than represent an acceptable level of risk. In uvere accident involving substantial interpola tions of extrapolations of the coures of developing this policy damage to the reactor core and to approximations. It la for this reason the statement the Commission expressed mitigate the consequences of such an Commission has an ongoing program, muca laterest in the bottom line results accident should one occur." However, which has cost a quarter of a biluon of al/ completed PRA*a, whether the compare this statement with b dollars in b last few years,in an reported point estimates were the mean Commission's proposed backfittirig attempt to bring some science to or snedian,ne technical staN has standard:"fte Commission shall estimating the core meltdown risks. repnatedly cautlooed the th=iados require the backf!tting of a facility only However even in this program the data that such bottom line numbers are not when it determines, based on a being generated are from limJted smou credible. What then is the basis for the systematic and documented analysis scale testa. Comuniasion's position that the level of

*
that there is a substantialinaease has, a readieg of this po!!cy severe accident risk posed by the in the overaD protection of the public statemenfindicates that the existing plants is acceptablet o -

bealth and safety * *

to be derived Commission's. claim that in developing he Commluion's decioloo-making from b *>ackfit andthat the direct and this poucy statament it has examletd an process in developtrqs this poucy indirect cost of implementation for that extenalve range of issues is incorrect. It statement is simply to rely spoe " point facuity are justified in view of this shows tother fbet the ammtaston either r

increased protection."(emphasis added) examined b wrong lasues or gave short ,,,, w,,,g.,me p%, r,u,, The Commluton has already defined a shrift to the fundamental fuues. p ,o e w p.e.wn a se Acew substantial increau in protection as in failing io define accuretely the level bem 6a Nesient Pow.r Plaat Anguienm6* Oces' meaning a backfit that would at least of severe acejdent risk at the exfsting sent p ar. reduce the " point estimate" of b lP ants and to addren the need for ) PS-PFI-47 Deosmber 31,1985 (reset) ..t

l POUCY STATEMENTS addaonal changa to b plants to make on acceptable wa & dak accepteW fu b long krm. benent analyses. ys to perfa:m coet. continus to se ca. Wbst actuarial b c-mlaston is speadas pad Further, guidance from b experience we n is sevmly hmited laDune to deal effecunty wie se Comodulon is needed on whether to by ourlack of detailed understanding of b puformance of b plants, their sever, Ant quadon.& concept d emphasise core meltdown prevention daigns, their weak spota. and because the reactor containment ortsineDy measurn or con meltdown mitigation evolved as a vweel to contain a fuD core musum. Of courw. La ordu to dmlop of the wide variations in b designs and seeltdown. But in the mid.19eo's, the a pobey on b latter (whether for k utWty capabW6u. Furbt, b usefulnus of actunial experience in reactor de.igners began lacing high existing plants or future plants), one drewing broad conclusions about cores into the u ne must first identify the root cauen of of containment.The y best of core meltdown riaka. One must also commercial nuciaar reactors is highly those higher powered ooree was so high develop a policy on containroent controvelal and fraught wtth uncertainties. tiset the acetal==t vessel could so performance Bona. longer be considand as an effecthe Unfortunatal, the Comm4aion - 7k e-mlaalon argues that credit pfuses to to addren these can be taken for the improvements nedspendent barrier to the release of the flesian products evolved during a core issum. Aa eNect!ve guide to regulatory implemented to addrosa specifle close meltdown. At that time, the Ateede decision-making on the treatment of calis such as the DC accident, the severe accident Lesoes requires an Browns Ferry fire and the Rancho Seco Energy t%=miaalon's Advisory understanding of what is expected by trenaient. Each of these were previously th-nittee on ReectorSafeguards way of containment performance, of the unrecognaed (or 41 best inadequately (ACRS) began urging the development root eenass of core meltdown riska. and also true of, for example,quences. This is appreciated) accident se and implementadon. in about two years. the of safety features to protect against a of the methods for performing sound Susquehanna station blackout event cost benefit analysee. Yet all of these loss of coolant accidentin which the elements are missing from the from a single faDure, the Indian Point vulnerabili to a single failure of a emergency core cooling system did not Commission's policy statement.%e

battery, the scrcalled interfacing week.%e ABCand theindustry bebevt1 that sufBcient data were e-6 ton's actual decision making guidance la this policy statement is system LOCA's for bolling water eva3able to lus;ify with a high degree of umsted to th.e statement that a new reactors. Noos of these latter events conBdence tb adequacy of the then, requirement might be imposed ifit were identined or high!Ishted through edsung safety standards. Derefort the involves " low-cost changes in PRA's nor were they expected to be.

AECignored the advice of the ACRS. Procedame or minor design given the level of detail that typicaUy Om the yean, es AEC and the MC anodifications." gas into a PRA and gfun the sub}ective after it have retterated these sweeping Da Commission claims thet PRA's natum of PRA's. Whether tbsee latter and optimisuc statemente on ave identify b plant specific vulnerablutin events should be called close calls is acddent risk. At the same time, tbs that dominate the core meltdown riska. argushle but bit occurrences certainly aumerous techn! cal Bews la the it is true that PRA's can idenufyy some of suggat a need to considu the root hada='s judgments have become the vulnerabilldes to catastroptue causes of significant operating events i readily apparent as more in'orma6an accidents. But the Commluton's and the conective meaning of those i. ) and data nearding the Is" *i of Safety of ratiooale for retying upon PRA's in events before pening judgment on the \\ I \\ acceptability of the level of safety l b reactors has become avadable.' assening coe r meltdown risks begs the When aD of the avellable data are questions: what of the uncertainties in schieved at existing power reactore. t e:.I believe it fair to say that PRA's? What of oversights in the Common unae also suggests completing the eettmated uncertalades in the risk analyust What of b multitude of such an analysh before developing i assumodons and approximations in the guidelines for the design of future e.Wla nons today are as large as they PRA'ef What of the residual aisks once reactors. Yet all of these concerns an were at least tan years agc. Yet, the the specific vulnerability has been swept aalde in the Commission's policy tww.t..lan is once again sweeping fLxed? nose questions are germane to statement, aside thou acertaintias in order to moolving sevm accident inuw. Yet ne DC Action Plan called for a large make b same unsubstantiated and they are not addrewed in b number of modification to the opere ting overty optimistic ganaralizations about Commission's policy statement. plauta,in addition to those the acceptabuity of the current level of Operational experience gives modifications. the Action Plan wvere noddent risk which han bees ,additionalinsight into the level of coaunitted to a rukmaking to consider p,,,em wrong in b past, safety, Actuarial experience with to what extent,if at all, existing nuclear Needed1stpor=anaa'a reactor accidents ladicates that the power plants should be required to deal averses core meltdown inquency is not effectively with damased core and core A disciplined approech to deciding above the opper hmit of the PRA results. meltdown accidents. bre was to be a whether to requim core meltdown risk Core meltdown accidents involve demarcation between thoes plants alrudy operating er under construction reduction measures should not only multiple feuures and a p lon of and the next Because b y~amiasWa perceind 2nerstion o specey b Commluion' tions mots eat make clow somewhat ,,n ad,dma,hg accul,o,.,,,s,hould b core mondown f,og y average of 19eo bt there would be a long hiatus in o identiBable. lf the industr ,,,,,,,e_c cy wm - c, high as 1tr* per reactor year, one would new,iani orderm,h use eesad io 'Dr dew carmt(Ma* 6an mte d expect more clow calls on con reconsider the General Design Criteria, d meltdowns than appear to have b daign baus, and the other 18,87$a h% on manaw ne,. %4 O,, occurred within the more than 800 nguhtlom in Hsht of aD that had been in e >desums s Dew okreet Nudase reactor years of UA nuclear power lurned through b years of experience pec,,,,,,,, m_ s.fgy.o,i rmwrof awarewr experience. But such actuarial with large powat reaetors, including the g un#nner d Whe== Prem N interences must be made cautiously in DC sccident. From this in-depth part because the operating reactors assenment of the strengths and weakneup of b large power reactor i December 31,1985 (reset) PS.PR.48 s )

POLICY STATEMENTE dulgns and the apprwth taken by A /toriono/Appmoch to S,y,, atibtiee loward constructing the plants. Accident Decisionmaking condluons are inside the reacter building. NRC would then be la a stion to a dent m* ' 't*8' /, articulate safety princip a that it What the Commission should have ,,g,und W tad to be inwrporatodinto des (gns done in its policy ststement is to set puts ha controuins the eacsdent into as l' ex expertowntal mode lh happeasd durtne b futum appucadona. Mus, the forth precluly and in understandable thret dayL the uncertainty of whether se Cowalasion in W signaled tbm terms what ourpresent estimetion of the accident could result in mehr reisenes of would be a significant ste forward in risk of severe accidents is, whether the neoecder is too Wdens e e the edvancing the protection the public, bmmlulon believes that rid to be enormm dannese to es plant, es expensive Re Ceaunission la this policy statement acceptable or not, what specific

- d *UF * *8" ""' P"**

P d takes several steps backwards-technical support can be offered in - One backward step discussed above support of that judgment, and how the den we's es euver la the Comuniaslan's decision to accept relevant uncertainties have been www aseid hm hs-- " iaccident thw more meltdown risks as Ibey exist in treated.he Comuniasion thould also had already sons soo'lar se make it esterebie. the current generation of plants without have come to gripe with a central While throughout h settre document we emphastas that fundamental chasons are even addressing some of the most quesUon in our regulatory am:that someary to prevent occidents es serices as fhadamental issues. Another backward is, given our present state of owl ThtL we mest not smene that es escident of step is aP _^ of the expressed concerning severe accident risks, sho d als or smetw sertonmen omasat happen. destre for a fresh look at light water we continue to pursue posalble asala, me if te changes a renamend m reactor esisty for future deelps and the improvements in severe accident 1 ,,,, %,,,"y,,[,ygg,,'"P,",h,,j'M.a

D

' *d* insistance on haprovements is tbs lent prevendon and mitigation?If the of severe accident risks for any futum Commission does not believe that the yp g.w plants. A third backward step is this pruent level of sevm accident rid is potentid impact of sech as acesden on acceptable for the remaining 40.yur life public beelth and ufety, should see occur is policy statement is the return to the the future. of some exis plants. then the pbliosophy of the 1seo's and tefire that Commission uld outline its p %pM of es prwideM's CommlMoe on Re construction permits can be lassed for this lon e Acddent at Aree MoeIsland based on only partial deelen bounds.g-term risk wi in Only through such a In order to reduce the severe acc..14. informstles. accepta i ent process can the technical community, de ova Mme k accepublavels,I For future roector orders, nuclear other public makare and the wWd han mdwiden few spdac utilities vu have expressed a. public undere and accept th* inibunk Firet.Iwould hn reqdid a desire for plant designs that are simpler. Cornmission's judgment on the severe detalld ward for plaW4pdoc safer, and mon forgivin6. Both the Bectric power Research Insutsta (EPRI accident rid quesdon. Unfortuna tely, equipment and design vulnerebHities at such an analysis is nowhere to be found ud Wshe plad 2 b and Edison Electric Institute (12T) have) in de Commission's poll statement. ,,n,ect eau wukneues w% and - impressed on the Commiselce he need Wd Based upon the m discussion' constitutes significant omtribotors to /,, for a fresh look at I water reactor I would have rea ed the o ee rid d a severe soddent. ' s ,. technology.%ese n ty sponsored conclusion. First, the risk to the p Seed 1w posed by severe occidents at the, lic ,,,,g g,dd han bibW a organizations have also todicated that g g, ,,,j plant construction for new plaats should sn"a*g plants is not acceptable for the pedem m & m,,g existin O'l'o'o'* "Jm*f"e 2*gni.&e sd A%e*:"Jfn'E2*" special emphasis on areas weakness plant. Yet none of these forward should continue to pursue cost effecuve a g,2;o$;';,74t'i'::=l,- { thinking requirements are to be found la risk reduction measuru for thue plants. good examples) and on speci8c stilldes I the Commiaafon's policy statement. I would ply the as. low as. reasonably. 4 a Wory Mmar@M hm lasteed, tbs Commission states that it achievab e (Al. ARA) principle to De W 9.1988 opmhg put M b will be satis 8ed with mere rannaments la the old designs and bt it is willing to reducing severe accident risk. subject DW h er mph a

y*,O[gd ge(,y inherent in the combination of a again demonstrated the dangere 9

a cubuk pmn pardal daigns for re improvements need not be pursued. I I cannot leave this latter point without would have slanply acknowledged the Inarginal plant design and a utility with marginal opm pedmnance a sad comunentary oc h Comminion's obvious: that the public and the hird. I would ve initiated a priorities. One lasue la this that Congress willnot tolerete. and the compmhensive assenment of the lent commanded great latemet wi ee industry and the NRC cannol allow of safe and the exia lants have Ccammission was kw to circument itsanobr uvm accident as serious as achine ne object of e5 art would suon eat requita a comparison of the %ree Mlle Island accident or worn' be to identify the root ceases of severe a esign to the staff's Standard Review My views in this ard are identical to amident risks.his eNort would also Commission nea,[,,,,,sa,, a,o. rian. n is e - w - m u sia by &e go,,p,mg g,y,,,,y identify possible measures which offer ons of one reactor vender. ee - mise d eigM8ca m>y redudag the Commissios's efforts to use Whetherin this rticular com we eens severe accident risk by overcoming the this policy statement as a webids to dose a esta ociden ornot,this adverse efects of equipment, t the reactor vendor to dreenvent bmdh haan umrl defidencies and areas of @[ Comunisalon's regulations took precedence over any Commlasine (*"g' e(The should not be allowed w ' g, consideration _of such fun <tamantal The acesdent got suf5cian out of hand m despita our best eforts to Issues as the actuallevelof eenre that thom ettempting to con it wm first two initiatives. Indeed, as the my accident risk to the public. the opmting somewhat in the dark. Whue today Commission's' chief safety officar noted the cauus are mU undmtood, e months acceptability of that risk and measures to reduce that risk, potendal eher the eoddent it is still diffloult to know in a June 27.1985 memorendum to the the precise state of the com and what the Executive Director for Operatione: i (. PS-PR.49 December 31,1985 (reset) )

POUCY STATEMENTS t heheve that'the r$ cent Devis-ilie~es event (STA) poeltion by allowing licensees to intehded that use of the dedicated STA alastrates that, in tlw rea! world, erstem and combine one of the required Senior would be an interim measure only until esoponent rehabulda can degrede below Reactor Operator (SRO) positions with these longer. term goals were achieved. those we and the todustry routinely assuene in wtimating oon melt freguancsw Our the STA position into a dual role (SRO/ These long. term initia tives STA) position. Option 2 provides that a collectively result in an improvement in ' engulatory abould marsine agelset degredenen palso w moset licensee may continue to use en NRC. the capabilities and qualificauons of the &e aceMalnda in nr RA sensates. approved STA program, with certain shift crew and their ability to diagnose modification s while meeting licensed and respond to accidents.Dese Finally, for future plants, I would have opersior stamns requirements. Inidatives include shift atamns explicitly required measwes to improve arrscTive oats: October 28,1965. increases. training and qualiikation the todgin ci safety agalast severe ' Post runTMan unposessAT>oel CONTACT program improvements, hardware accidents idl future plants and to address Clare Goodman. Office'of Nuclear modifications, amphasis on human the aldtabe of the past Buch measurve Reactor Regulation. U.S. Nuclear factors consideredone. procedural sould lachsde requirements for greater simpucity in plant design. improved Regulatory Commiselon Washington. Upgrades. and development of extensive maintainability, and a requirement for DC 20655. Telephone: 301/492-4894. eineqncy wganizauons to essentially complete klant designs prior surnasesm&RY pfPORMATIOsc Capabilities during M c M - to the issuance of NRC ppprovalier the

Background

start of plant construction-Draft Policy Statement Following the accident et Three Mile I believe that these measures would Island in March 1979, a number of On July 25.1963. the Commission be sumcient to bring the risk of severe studies were conducted to determine published in the Federal Register (48 FR accidents within acceptable bounds for why the accident occurred, what factors 337e1) e Draft Policy Statement on the remaining operating hves of the might have contributed to its severity. Enitineering Expertise, on, Shift to exis plants and for the operating and what the industry and the NRC hvos any future plants. Moreover. sech an approach would de much to could do to prevent tne recurrence of the eng eer ng and accident a sst ent restore pubuc conndence in nuclear same or a similar accident ihes' expertise must be available to the power and in the effectiveness of the studies concluded. emong other thirfgs* ETny'ti 8 t all nucle pow" that a number of actione should be NRC's regulatoryyrocess. It is taken to improve the ability of shift Engineen.sft Poucy Statement on unfortunate that me th=laston bas, operating personnel to recognize. ns Expertise on Shift offered chosen another path. However. key decisions remain to be made by the diagnose, and effectively deal with plant licensees of nuclear power plants and Commiselon in adopting a Baal transients or other abnormal condition.s. applicants for operating licenses two . To address these recommended options for meeting the stamng backfitting rule and a final safety goal. Improvements, the NRC initiated both requirements of to CFR 50 54(m][2] and noes decialons represent a Baal short term and long. term efforts. The the requirement in NUREG-0"37. Ilem opportunity to come to with many short term effort required that as of 1.A.1.1 for a Shift Technical Advisor of the pivotallassesav la tble f policy statement la that regard, it is January 1.1980 each nuclear power (STA). Option 2 gave them the 4 plant have on duty a Shift Technical opportunity to co Techru.Opuntors,mbine the licens encouraging that there appears to be an Advisor (STA) whose function was to 6"I" (SRO) and Shift emerging aa-asus within the NRC provide engineering and accident cal Advisors (STA) functions. senior technical staff and within the assessment advice to the Shift Under Option 1 licensees that did not ACRS in favoe of safety improvements Supervisor in the event of abnormal or want to combine the SRO e,nd STA i to reduce severe accident risk both for accident conditions. ne STA was functions could continue with their { existing and for future planta, required to have a bachelor's degree in approved STA program in accordance { engineering or the equivalent and with the description in NUREC-0737, ,,,, g,p specific training in plant response to " Clarification of TMI Action Plan Puushed totes /es transienta and accidents.The STA Requirements. Eners>.1easms requirement wu identified to licensees Interested persons, applicants. and via NUREG-0578 (July 1979) ' and licensees were invited to submit wn; ten NUREG 0737 (November 1980) and was commen's to the Secretary of the Commission Poney Statement on later mandaled by plant. specific Commission. Fouowing consideration of Engineering Experties m;SW Confirmatory Ordere. the comments. the Commission Concurrently. the NRC and industry amended the Draft Policy Statement, as Amtwer NuclearRegulatory embarked on a longer. term effort aimed discussed in the following sections. Comenission. at upgred staffing levels and the Commenta on the Draft Policy Statement 'Y "8 i8 ting stake i Sh r vir na n-A total of 34 responses were received machtne mterface and increasing and evaluated. The pubhc comments euesstAam This Policy Statement capabilities for responding to related primarily to the combined SRO/ presents the policy of the Nuclear emergencies. At the time the STA STA position. The following discussion Regulatory Commission (NRC) with requirement was imposed.it was highlights the major points reised in the respect to ensuring that adequate engineering and socident assessment comments and the resolution of those comments. A detailed enslysis of all

t. %e. sin m.

,,,,,ci.w,t, % no, public coinments and their resolution , expertise is E M by the operstmg r staff at a nuc!(at power plant.This er :=erens for e tw mi ow Nac evw onumene was also prepared. (Copies of those Polley Statement offers licensees two Reesu hr H Suse NW. wasMnsm DC. The letters and the detailed analysis of all options for providing engineerirq C,"'F(Qgggq g the public comments are available for expertise on shift and meeting licensed rn.aame w t,y.rrmee w cPo. r.o. sei staar public inspection and copying for a fee o operator staffing requirements. Wee %eien. Dc anmwona. my mer el o tw at the NRC Public Document Room at Option 1 provides for elimination of ,T"*,7',jg'"yg*"c*. s2ss p'" 8

d Imm the NewnelTecWat Inimuan 1717 H Street NW., Washington. DCl the separete Shift Technical Advisor Of the 34 letters received.18 included December 31,1985 (reset)

PS.PR 50 ]}}

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