ML20039A026

Forwards Proposed Rule 10CFR50, Stds for Reduction of Risk from ATWS for Light Water Cooled Nuclear Power Plants

ML20039A026
Person / Time
Site:	Palo Verde
Issue date:	12/08/1981
From:	Sohinki S NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Callahan D, Cole R, Lazo R Atomic Safety and Licensing Board Panel
References
NUDOCS 8112160153
Download: ML20039A026 (14)

Text

{{#Wiki_filter:h 7g 7 ~ 4 + 9 4 4 t, w y December 8, 1981 Robert 11. Lazo, Esq., Chairman Dr. Richard F. Cole Administrative Judge Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board U. S. Nuclear Regulatory Coamission U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Washington, D. C. 20555 Dr. Dixon Callahan Administrative Judge Union Carbide Corporation [dyt-b\\ P.O. Box Y c3N Oak Ridge, TN 37830 A < ' w /7 / 'Q s -O Cl 179gb In the flatter of """3N*",,'if ARIZONA PUBLIC SERVICE COMPANY, ET AL. gd' (Palo Verde Nuclear Generating / ' reg. s d Station, Units 1, 2 and 3) \\ Docket Nos. STN 50-528, STN 50-529 and STN 50-530

Dear Administrative Judges:

Enclosed for your inforuation is a copy of the recently published proposed rule with regard to " Standards for the Reduction of Risk From Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants". Sincerely, Stephen it. Sohinki Counsel for NRC Staff

Enclosures:

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y' r 57,21 "= = m-m > = _._.m _ re w l m tsin. s N' M LXIQ!iCS voi a No. :s Tucshy. Nosember 2s.,1981 Tr s sect.on of the FEDERAL I GIST ER nction has b't:u determined to be "not 23.1982. wl!! be considered if practical to do so.but only those commente conta.ns not<cs to the rotAc of its maior." lhe Regulatory flexibility Act(Pub.L receised on or before this date can be proposed issuance of ruks and tsgJatons. Tte purfesa of these retices 06-351)is not applicab!c to this action: assured of consideration. is to g.va interested persons an therefore, a Regulatory flexibihty aooatssts: Comments shou!d be ' Ana s will not be prepared.. submitted in writing to the Secretary of kb p or to a ton f e f.nal Th s proposed actionis intended to the Commluton, U.S. Nuclear eliminate an unnecessary bulletin. Regulatory Commission. Washington. L3 , thereby s,asing the Cosernment the cost D C. 20355. Attention: Docketing and of periodic resisions. Sersice Branch. All comments received DEPARTMENT OF AGRICULTURE This program is listed in the Catalog and all referenced and other NRC Rurat Efectrification Administration of FederalDomestic Assistance as dxuments relevant to the NIWS issue 10 850-Rural Electrification leans and will be ayallable for ublic inspection in 7 CFR Part 1701 Loan Cuarantees.10.851-Rural the Commission's Pu lic Document Te!cphone Imans and Loan Guarantees Room at 1717 H Strut NW, - Proposed Rescission of REA Buttetin and 10 852-RuralTelephone Bank Washington. D C. Copics of teferenced g1 7:3g1-11 go, NRC reports may be purchased from the utNcv: Rural Electrification A1 ritten submis'sions made D Administration. USDA. pursuant to this action will be made D en u le as allable for public inspection during Regulatory Commission. Washington. c.cTioN: Proposed mle. - regular business hours at the above 0555. suuuaRY:The Rural Electrification address. r mR mTNtR mRuanos contact Administration (REA) proposes to Usted:Nosember te, taat. David W.Pyatt. Office of Nuclear emend Appendia A-REA Du!!etins to III4 V.1tur. tee, Regulatory Research.U.S. Nuclear proside fur the rescission of REA Adma., trator. Regulatory Commission. Washington. s Bulletm 81-7-381-11. " Changes or Currections in une Construction.- tra om si-am,ma use awl D C. 20555. (301) 443-5900.

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sUPPttMcNTARY mFORM ATION" Concern which has become obsolete.The " regardmg protection against anticipated primary purpose of REA Bulletin 81-' NUCLEAR REGULATORY es ents has long been a, subject of ^ ~- "~ transients without scram (ATWS) n clion hange 0 er ince COMMISSION REA Form 216 was rescinded in an extensive and continumg study by the NRC staff.The significance of ATWS for A 10 CFR Part 50 rms. I ut in 8 -7 381-I ls reactor safety is that some ATWS considered to be unnecessary. Standards for the Reduction of Risk esents could result in melting of the Pubhc comments must be received From AnticipatedTransientsWithout reactor fuel and the release of a large cart: by REA no later than jantary 25,1982. Scram (ATWS) Events for Ught. Water' amount of radioactise fission producta. ApoRess: Submit written comments to Cooted Nuclear Power Planta The principal benchmark for deciding the Director. Engineering Standards ActNev: Nuclear Regulatory whether and to what en tent nuclear power plants should be modified Division. Rural Electrification Commission. bccause of ATWS related safety Administration. Room 1270. South Ac n oN: Proposed rule. concerna is set forth in subsection Building. U.S Department of Agriculture. Washington. D C. 20250. $UMMARY:De Commission is 1611(3[of the Atomic Energy Act.That e,ection grants to the Commission the FoR FURTHtR INFoRMATioN CONTACT: considering three alternatives for Mr. Edwin N. Umberger, telephone (202) amending its regulations to require authority to " prescribe such regulations - 447-7040. A Draft Irnpact Anal sis has improvements in the design and or orders as it may deem necessary 3 been prepared and is asalfable from the operation of light. water cooled nuclear

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• in order to protect health and to Director. Engineering Standards,

power plants to reduce the hkehhood of minimize danger to life or property." Disision, at the above address. faihire of the reactor protection system Throughout the history of regulating to shut down the reactor (scram) nuclear reactors. the dual concept of surPctutNT ARY mroRu ATioN: Pursuant following anticipated transients and to pres enting accidents and mitigating to the Rural Electrification Act, as mitigate the consequences of anticipatea their consequences should they occur. amended (7 U.S C. 901 et seg ). REA transients without scram (ATWS) 1.e.. defense in depth, has been used to proposes to amend Appendix A-REA events.This will reduce the oserall risk achieve this ojective. Thus, conservative Bulletins to provide for the rescission of of nuclear power plant operation.The design. cor.st*uction. testing. REA Bulletin 81-7:381-11.*' Chang'es or consequences of this regulation willbe. required so that accidents wal no maintenance and operation of plants are Corrections in Line Construction." Since to require electric utilities to install no significant effect on the economy will certain equipment in nuclear power - happen (i.e. have a low probability of occur, since no signiricant increase in P ants and. possibly, to implement a occurrence).Then, to provide defense in l cost for consumers, subscribers. industries or Cosernment will result, reliability assurance program. depth, the capability to mitigate their and since no significant impact on OAftS: Comment period expires April consequences is required for accidents economic conditions will be caused. this 23,1982. Comments received after April that are postulated to occur even though m 6 6 9 L w ww---

syh TVoT.To.~NE n5TTwzTy.mmuum-m==.m==.==== =m ~ ~. ~ - ~ = = = = ~ ~ ~ ' submitted is the Comrnissfes foe cAhcr d'han rA.enetivT posed * - i nm = -:.:. w =. = ~- the des!;n is reefred to Instude consideration fri an carfy nerstonl] rates ws.... prou,/ W. uvs a sfpifiant rmno.%:Imact on a meawir s to prc. vent them. CY C0 400. September 4.1%0. end in final form in ShGY C0 eC. Nos ember outestantial number os small entities.*lb SI: /ilWS aald..nts are a cause for concern because a mhmatch can

7. tota ne second NRC proposed rule alternatise proposed rufes offect only.

dacicp betw een the power gerwrated in is a receni proposal by former NRC the licensing and operation of nuclear the reator and the power dissipated la Chair rnan pseph M. llendrie.8 Dr. power plants.ne companfes that own untrolh d ways il ifie scram system ifendric's alm in starting afresh was to these plants do not fa!! within the scope of the definition of"small entitles" set felf s to shut down the reactor fo!!owing try an approach that would make forth in the itegulatory Mesibility Act or i o fault in the nmmat heat dissipation licensees look carefully at their plants the Small Dusiness Sire Standards set functions (transtent events) ne power for Anys related vulnerabihtles and out in regulations issued by the Small - mtsmatch can threaten tfie integrity of then fie these vulnerable areas, !!astness Administration at 13 CFR Part the barriers that confine the fission ernpfoying systems analysis ce 21.Since these compantes are dominant. L3 products. A core meltdown accident.in reliabihty techniques. %e Commission be!Ieves that the in their sersice areas.this proposed rule soms cases accompanied by a failure of does not fall within the purview of the containmerit and a sery large refease of likelihood of sesere consequences radioactivity. is a possible outcome of arising from an AnvS event during the Act. some ADVS accident sequences Dus. swo b four year period required to Hrst NRC. Proposed Rule (the Staff th consequences of some postulated Impfement a rule is acceptably small. Rule) AnVS accidents are unacceptable. %ls judgment is based on (a) the nere have been roughty one favorable experience with the operating %e review and evaluation by the thousand reactor n ears of emperience ' reactors. (b) the limited number of NRC staff of the Information that has cccumulated in foreign and domestic operating nucfear power reactors,(c) the been des eloped over the past ten years on AnVS csents and of the mannerla commercial!fght water.coofed reactors inherent capability of some of the operating WRs to partially or fully which they should be considered in the without an AnVS accident.%is p mitigate the consequences of ATWs design and safety evaluation of nuclear - capertence suggests that the frequency cf AnVS accidents is less than or of the events.(d)the partialcapability of the power plants is contained in the report, crdir of once in a thousand reactor recirculation pump tilp feature to " Anticipated Transients Without Scram / ye:rs.nere have been several mitigate ATWS es ents that has been for Light Water Reactors," NUREG-prscursor events,Ie fautto detected Impicmented on all DWRs of high power 0100. Volumes 1 through 4.There are th:t could have given rise to ADVS level, and (e) the interim steps taken to swo primary factors in the staffe. cvsnts Dis suggests that the frequency develop procedures and train operators evaluation.The first is the degree of cf AnVS accidents, though ! css than to further reduce the risk from some assurance that AnVS esente can be enco ln a thousand reactor years, may ATWS events.On the basis of these presented, which depends on the not be very much less. Such frequencies considerations, the Commission believes reliability of current reactor protection era too high for accidents of the severity that there is reasonable assurance of systems.The second is the capabihtbof distribed above.%us the NRC has safety for continued operation until esisting reactor designs to mitigate t ditermined that reductions must be implementation of a rule is complete, conuqmees g ATWS events. med in the frequency, severity,orboth De implementation schedule contahed no rehaWhty of cunent nector th) frequency and seserity of ADVS In this rule balances the need for careful otec'i n systems has been esumated The Nuc! car Regidatory Commission anal sls and plant modifications with {ased on the operating esperience to cccid:nts. has under consideration thice proposed the estre to carry out the objectives of date and reliability analyses. lfowever, cliernath e rules, each intended to the rule as soon as possible. the very high level of reliability required is difficult to demonstrate with. f' reduce the risk posed by ATWS Paperwork Reduction Act confidence because it depends on cccidents.Two of these originated A request for clearance of any accurately determining the rate of within the NRC, and are described application and reporting requirements common cause failures. Common caum below.ne third la set out in a petition of the alternative finally selected willbe failures involve fallares of multiple for rufemaking filed by twenty utilities submitted to the Office of hfanagement components resulting from a single (" Electric Utihties Petition." PRM SM9 and Dudget under the paperwork cause or event. Reactor protection 45 FR 73000. November 4,1930 and the Reduction Act (Pub.L15511). At the systems are carefully reviewed to supplement to the petition published on time, the SF-83 "Requent for Clearance.- identify and eliminate all but the most february 3,1981, da FR 10501) The utilities' petition will not be reproduced Supporting Statement, and related unlikely common cause fattares. hers, how es er, the cunent period for the documentation submitted to OMB will Ilowever,one connon cause failure in be available for Inspection and copying the reactor trip portion of the protection r ettility petition is hereby reopened to run for a fee in the NRC Public Document system of a commercial nuclear powee concurrently with that of the two NRC Room at 171711 Street NW reactor has occurred during proposed rules for the purpose of comparing and contrasting the utility Washington, D.C, ' approximately 1000 reactor. years of petition with the two pro {the NRC-Regulatory Mesibility Certification operating eaperience.The failure was osed rules detected during normal surveillance and published herein. Both o In accordance with the Regulatory corrected before any event requiring a proposed rufes mandate improvements Mexibihty Act of 1980,5 U.S C. 605(b). reactor scram occurred.%ere has also in ATWS prevention and mitigation. the Commission bereby certifies that been one partial failure to scram in a' ney differ in scope, approach, and commercial power reactor, which criteria. 's me amarando a or cunnan !=pti s4 ne first NRC-proposed rule Is known Heah w cw.n m canay. ar.afo.it, and occurred at low power and resulted in no core damage or radiation release. ce the staff rule and is a direct outgrowth of NUREC4600," Anticipated [,%^y,K$',dl]"y'j%.a^n,$'h Common cause failures have also c,mn,..;on,,%, %,aea. mn occurred in other systems in nuclear Tr:nsients Without Scrum for Ught Sueet NW, WesMagem BC. Water Reactore," Volumes 1-4. It was e e s W L

I rederal Re:;1 ster / Vol.IE, No. 220 /,L_esday n.~.-- ,,..n.,,~,.~.- Alternative 2 or 4 ten obter 91 ants. f . power plants and other potential sl%htly resised fmm in Volume 4.W commn cause (Mfures in reactor intent of th proposed ruleis to adopt o that begar. operu:n belt re late t?G9. protection systems lave been identified. comtsinaO of the Mternatives Eecau e of their on:que characm13 tics, pecause of the low rate of occurrence of recommended in Volume 4 (except for the staff bleved that more extensive comrnon cause failurcs, operating one chant;e for reactors designed by modifications would not be appropriate c.sperience ! not, and cannot be. Westinghouse and licenrod to operate for these plants.He proposed rule does officient to conclush ely determine on a before 1GSI).%e propysed ru!c would not explicitly address these plants statistical basis whether reactor imp!cment the requirements In a (except in the imp?cmentation schedule), protection systems are reliable enough different manner from that described ic but the intent is to consider any to mak e the probability of unacceptable Volume 4 of NUREC-0100 %e form of exemptions from the acceptance criteria, conscquences from AnVS events the requirements in the proposed rule is of the proposed rule for these older. also different from that recommended in plants based on analyses by the.... ecceptably s-nall.The prediction of NUREG60 in that the proposed rule licensees and evaluations similar to common cause fai!ures is as much art as it is science. System reliability analyses specifies acceptance criteria for ABVS those conducted under the that attempt to predict the nature and mitigating systems while the required Commission's systematic evaluation frequency of common cause failures mitigating systems are specified la program (SECY-77-561 October 1977) In l context with the overall safety of these suffer from probicms of comp!cteness Volume 4. and accuracy, particularly when the Alternative t is to make no. facilities. desired failure rate is extremely small. rnodifications at all. As discussed the Alternative 3, as modified in the While quant tative estimates of NRC has concluded that the reliability proposed rufe, would increase the protectior. system reliability provide of current reactor protection systems is reliability of the reactor trip portion of importat t information the conclusion as insufficient witu respect to AnVS and the reactor protection system for some to the adequacy of protection system that the probability of AnVS events is plants and provide for the mitigation of reliability must be based on engineering sufficiently great to warrant most ATWS events.ne reliability of judgment.The NRC has concluded that improvements.Therefore, this the protection system would be the reliability of current reactor alternative is not represented in the Increased in the same manner as in Alternative 2. liowever, this increased protection systems has not been proposed rule. demonstrated to be adequate and most Alternative 2, as modified in the reliability of the reactor protection likely is root adequate. proposed rule,would increase the system would not be required in plants The probability of severe reliability of the reactor trip portion of that have a greater capability to mitigate reactor protection systems and improve ATWS events.The mitigation of most consequences resulting from ATWS events is also affected by the capability the capability of existing systems to ATWS events in pWRs was expected to of nuclear power plants to mitigate mitigate some ATWS events. Reliability be accomplished as in Alternative 2. ATWS events.nis capability varies of the reactor trip systems would be except that means would be required to depending on the design of the reactor increased by the addition of isolate the containment early in an s3 stem and the status of systems and supplementary protection systems that ATWS event upon detection of radiation the values of system process variables would be independent and diverse from released from failed fuel.The mitigation at the time the event occurs.The the reactor trip port on of the current capability of BWRs was expected to be capability of a plant to mitigate ABVS reactor protection ystems. Diversity increased by providing automatic events can be assessed by analysis. would be achieved by the use of initiation of the Standby I.! quid Control flowever, uncertainties in the design components from fifferent System and increase its flow capacity. characteristics of the reactor, the manufacturers, by the use of Considering the state of design and probability of failure of the mitigating components having different principles. construction. and a balancing of public systems and the probability that the of operation or power sources, and by safety benefits against economic cost values of system process variables will the use of components in different the Commission proposes in this first be different from those assumed in the operating modes (normally energized vs. rule that plants receiving an operating analysis all combine to produce normally deenergized). This alternative license before 1984 should be required to uncertainty in the results.Therefore, the would not provide increased reliability implement Alternative 3 as modified in difficulty of demonstrating a capability of the reactivity control portion of the the proposed rule. to adequately mitigate ATWS events is protection system, i e., the control rods Alternative 4, as modified in the similar to the difficulty of demonstrating and control rod drives.llowever,in the proposed rule,would increase the that ATWS events can be prevented. case of reactors designed by General reliability of the reactor trip portion of Based on analyses performed to date, Electric it was proposed to increase the the reactor protection system of all however, it is clear that, in most cases, reliability of a portion of the control rod plants and provide for the mitigation of present reactor designs have inadequate drive system,1.e., the control rod drive almost all ATWS events.The reliability capability to mitigate the consequences scram discharge volume.The capability of the protection systems would be increased in the same manner as in of many postulated ATWS events to mitigate ATWS events would be - Alternative 2.The mitigation of virtually should they occur. improved by providing actuation llaving concluded that Improvements circuitry that is separate from the all ATWS events was expected to be are needed to reduce the probability of reactor protection system for some substantially increased by additional severe consequences from ATWS existing systems such as primary system pressure relief capacity in the reactor events, the staff developed fue relief valves, turbine trip, and auxiliary coolant system.The mitigation alternatives, three of which would feedwater in pWRs and the recirculation capability of BWRs was expected to be reduce this probability by increasirc pump trip in BWRs.%is alternative is increased by the addition of high increments and would require increasing very similar to the proposed rule offered capacity neutron poison injection amounts of modifications.ne by the utility group. systems.In balancing public safety alternatives were first described in ne staff proposed in Volume 4 of benefits against economic cost, the Volume 3 of NUREG-4460 and again in NUREC-0460 to implement only Commission proposes in this first rule N l e l es : 1;1,gQp;g{f / %gM4 CMTSMTai$MQfGl$$'J%$

IVopose us L 1 2-1, 1931 l'ederal I?r1;ister / Vol. 46 No. 226 / Tuesday, Novernber v. ~~.,.v~.~.m F s I 37 21 where theIcn of safety b cftcadyIdgh,y' ' k' .h th/Adsiscry Committer ca Reacto-criteria for recep:asc. that tl.csc e sttnsive dcto danges models. Since the parnca Safq;uards [ACRS) tccc:nmended in-couht en!y be practical!y inorporated in evaluation triodel are uncenain to some omittica the requitcment for - + plar.ts not near comp!ction and not to be degree and some may vary over the Impros ements in the protection systern lifetime of the p! ant, the level of safety is rebbility.Hus, the prcposed rule licensed before 1984. Re proposed requirements In Vofume determined to a la'ge extent by the m allows the pro'ection systern 4 of htfiEC-4800 werein the form of degree of conservatism in the Improvemenis to be o.aitted if more P specific design changes.%e proposed parameters used in the evaluation conservative values of the parameters,' rule a!so specifics the des gn changes models,which 6ffect the conservatisrn such as moderator temperature i 4, a. required to improve the rehability of the of the calculated consequences of coefficient are used in the evaluation protection spfem and the response for postulated AnVS events.%e proposed models and the capability to cornply,. contafnment isolation, but the chanaes rule specifies that realistic values of with the acceptance criteria is -y In mitigation capability are required parameters may be used when the value demonstrated. In plants licensed after O through the necification of acceptance is known with reasonabfe accuracy, bat january 1,1938 orlater the time - criteria, critoria for evaluation models, that parameters with latSe uncertainties available to design and install the I and mitigating system design criteria, must be conservatively treated.He mod $ cations to the protection system is Re specification of criteria requires Intent is to obtain realistic analyses of sufficicot to ensure that the design ' licensees and applicants to demonstrale the course of AnVS events.yet predict process would not be compromised and that the desi ns of their plants are in the consequences conservatively. In improvements in the protection systems d compliance and thus provides more order to ensure that the consequences of of allof these plantsis required by the.: cssurance that the safety objective is most AnVS events will be within the ..n proposed rule.. being ettained. This form afso a!!ows the acceptance criteria. the proposed ruleOne plant modification that would be designzr more flexibility in design and a specifies that the value used forparameters that vary over the life d greater potential for minimizing costs. being Implemented onboiling watee Although the ultimate safety objective the plant (the most sign 3 cant of these in ' reactors. In an order dated February 21.- Is to limit the release of radioactivity to the moderator temperature coefl1cient) 1980,licensecs of BWR plants were the environment, the acceptance criteda must be a value that is not exceeded directed not to operate after December In the proposed rule are d;tected toward over most or virtually all of the plant 31,1980, without a recirculation pump ansuring the integrity of the reactor lifetime.In the case of the moderator trip installed. l.icensees have also been coolant system and the reactor core temperature coefficient, the value used directed [lE Bulletin No. 80-17 da'.ed. following AnVS events.ne staff in the evaluation modelthat was lessJuly 3,1980, and NUREC-0737, recognizes that failure to satisfy these negative than the value expected to be " Clarification of Dil Action Plan acceptance criteria does not necessarily. experienced during 90 or 99 percent of Requirements") to ensure that opera' ting result in seve;e radiological the designlifetime of the p! ant would procedures and operator training consequences and has considered the ensure that the consequences of most or address the actions to be taken in the ed&tional safety margin in developing virtually all ADVS events would not plants as now designed if an ADVS did the proposed rule.In formulating the violate the acceptance criteria. occur.These requirements are prudent. proposed rule,the Commission has Although improvements in the measures that will reduce the risks from considered the need to compare for each capability to mitigate AnVS events ABVS events during the interim period p! ant the offsite doses that might resultprovide a significant increase in the before the plant modifications from ABVS events with 10 CFR Part 100 level of safety,there is some uncertainty determined by the Commission to be.,, guidelines. Based on conservative associated with this conclusion.%is necessary can be installed.. generic calculations performed by the uncertainty derives from the uncertainty In particular cases, additional ' st:ff, there is reasonable assurance that in the reliability of mitigating systems requirements or earlier implementation. c:Iculited offsite doses from AnVS will and in the evaluation models used to may be appropriate.For exampfe,.,, be within the Part 100 dose guidelineo if define them.Because of this uncertainty candidates would be those existlag 3 the ccceptance criteria of the proposed the staff believesthatimprovementsin nuclear power plants that are { rule tre rnet. Accordingly, the - reactor protection system reliability considered to be at high risk sites owing C:mmission has decided that applicants should also be required.%ese Io a combinationof population density. end licensees wi!! not be required to' mo&fications to present 2eactor -

• meteorological conditions and othee.

calculate the potential offsite protection systems, as with any

factors, r:diological doses resulting from an modifications to a nuclear plant, have

%e proposed rule would provide fee. ADVS event under i 100.11.1f only the potentialfor introducing these guidelines for calculated offsite unrecognized failure modes that could implementationof the requirementsis, stages in order to gain the greatest doses were specified, the flexibility for result in a decrease in the level of increase in safety in the shortest time thi designer would be' increased, but the safety. A careful design processin and at theleast cost.The modifications ctttinment of the safety objective would conjunction with the quality assururm. to improve the reliability of the be more difficult to demonstrate,il verification, and test programs la protection system and the mitigating i systems designs were specified, the necessary to ensure that this willna! system actuation circuitry would be flexibility of the designer would be occur. However, the implementation of required within two years of the reduced, and the demonstration that the these' improvements'in reliability in effective date of the rule.In order to m c fety objective had been attained some plants is to be accomplished accomplish this, descriptions of the f,- g would be generic rather than for within two years, and such a abart -, modifications are to be submitted foe. '. specified plants.Pdor attempts at such a design and installation schedule might review by the NRC within one yearaf 3:neric demonstration have been Emmptorlise the design program.in the effective date of the rule. unsuccessful, as discussed above. plants ach as those designed by Pursuant to the Atomic Energy Act of The level of safety,thatis, whether Westinghouse, which have a capability 1954, as amended,the Energy most or virtually all ADVS esents can to mitigate nearly all ATWS crents and be mitigated.is specified through the 4 r e

Federal I?rg,fr.ter / Vol 40. N'. ".M / kr day. November..,. ~.. - -~ _..-.- - - -,.- - - ~ A - - ~ 57525 24, 1981 / Proposed Rules ~_~ l . tfeorg deation dYi of 1974, as amen'ded, the RCS prcc...e boundry does not models must represt nt the eff.cct of th) and section 553 of title 5 of the United exceed Ifut permitted by the " Level C failures In miti;;iting systems that are a States Code. notice is hereby given that SeMr.e I.irr.it" as defined in Article NB-direct consequence of the ATWS event cdoption of the following amendments 3000 of Section !!! of the ASME Boiler being modeled. For facilities issued to to CFR Part 50 is contemplated. and Pressure Vessel Code'and the operating licenses on or after January 1 calculate ! deformation of RCS. 1981, and not standardized to a facility. PART 50-dot.*CSTIC LICENSING OF components is limited so that the at the same site that was issued an PRODUCTIOfd AND UTILIZATION operability of components necessary to operating license before January 1,1984 FACILITIES safely bring the reactor to and maintain evaluation modcIs must also represent 1.The authority citation for to CFR it at a cold shutdown condition is not the effect of the likely random single Part 50 reads as fo!!ows: impaired, or (B) the integrity or failures of active components in g ra W d c mp Ms M h mWgeg systems. Authority: eccs.101.104.161.182.183. 88 Stat. 936. 9J7. 948. 953. 954,as amended (42 demonstrated based on Conservative (ii) The value of parameters that vary U.S C. 2133. 2134. 2201. 2232. u33); secs. 2o2, assessments of tests conducted to 20s. m8 Stat.1244.1240 (42 U.S C. 5842. 5846). determine the integrity or operability of. over the lifetime of the facility or represent the chara,:teristics of unicss otherwise noted. Section 5018 also components under the conditions mitigating systems that are permitted by issued under sec.122. 68 Stat. 939. 42 U.S C accompanying postulated ATWS events procedure to be inoperable for any 2252) Sections 50 80-50 81 afso issued under and based on the likely condition of the period during operation must be ~ Sec.164 c8 Stat. 954, as arnended. Seca, 50.100-50102 issued under sec.188. 68 Stat. components over their design life. selected so that values that would result (ii) fuelintegrity. The calculated in violation of the acceptance criteria Sa 2 amended 4 U.'S d " " damage to the reactor core as a would not be expected to occur during 243). I 50XI) issued under sec.1611. 68 consequence of postulated ATWS (A) hiost of the design lifetime of

• Stat. 949 [42 U.S C 2201ii)) and il 50.70-events, including oscillations of power 50 71 and [ 50.78 issued under sec.161o. 68 and flow, must be limited to ensure that facilities issued operating licenses before january 1,1984 or of facilities S!at. 950. as arnended; (42 U S C. 2201(o)) and the core geometry is not distorted to an the Laws referred to in Appendices.

extent that would impair core cooling or standardized to a facility at the some

2. A new i 50.60 is added to read as safe shutdown.

site that was issued an operating license before January 1,1984. follows: Dil) Rodeooctivity release.%e calculated release of radioactivity from (B) Almost all of the design lifetime of facilities issued operating licenses i 50 60 Acceptance criteria for protection the fuel rods to the reactor coolant a7ter january 1,1984, except 7,cg)its.on or esainst anticipated transient without scram system during postufated ATWS events es e events for Ilght. water-cooled nuclear must not exceed one ercent of the staMaued to a facNy at Ge same Power Plants. radioactivity within i e fuel rods of a site that was issued an operating license (a) Definitions. (1)" Anticipated pressurized water reactor or ten percent m }ammary 1. M. Transient Without Scram"(ATWS) of the radioactivity within the fM tala (3)Mugaung % stem CrRem, AMS means an anticipated operational of a boiling water reactor. nutigating systems must be independent, occurrence as defined in Appendix A of (iv) Containment The calculated separate and diverse from the reactor this part fo!! owed by the failure of the containment pressure, temperature, and Protection system. ATWS mitigating resctor protection system specified in humidity resulting fe postulated systems must be designed, qualified.- General Design Criterion 20 of Appendix ATWS events must not exceed the monitored and penodically tested to A of this part. design values of the containment ensure continuing functional capability (2) "ATWS evaluation model" means structure and components or the under the conditions accampanying the calculational framework for contained mitigating systems, equipment postulated A*lWS events, including evaluating the behavior of the nuclear and components. For boiling water natural phenomena such as power plant during a postulated ATWS reactor pressure suppression earthquakes, storms, tornadoes. sv:nt. containments, the relief or safety valve hurricanes, and floods expected to occur (3)"ATWS mitigating systems" means discharge line flow rates and during the design life of the plant. those systems including associated suppression pool water temperatures. ATWS mitigating systems must be controls. instruments, power supplies must be limited so that steam quenching automatically initiating when the cnd other systems assumed to function instability will not result in destructive conditions monitored reach when evaluating the behavior of the vibrations. predetermined levels and contm, ue to nuclear power plant following an ATWS (v} fong-term shutdown andcooling. Perform their function without operator event.

• lhe reactor design must permit the action unless it can be demonstrated (b)(1) Acceptance Criteria. Each light-reactor to be safely brought to and that an operator would have adequtte water-cooled nuclear power plant must maintained at a cold shutdown information and would reasonably be be designed, constructed, and operated condition following postulated ATWS expected within the time available to so that the consequences of postulated events without insertion of control rods, take the proper corrective action.

cnticipated transient without scram (2) Evoluotion ModelCriteria. (i) (4) Evoluotion models. Each applicant (ATWS) events calculated in ATWS evaluation models must, with orlicensee shall submit evaluation l cccordance with an ATWS evaluation reasonable accuracy or acknowledged models as defined in paragraph (b)(2) of model approved pursuant to paragraph conservatism, represent the actual this section, together with the (b)(4) of this section conform to the characteristics of the facility modeled description and results of the analyses f;11owing criteria: and each significant physical and test necessary to verify the validity (i) Primary systempressum. De phenomenon that would occurin the of the assumptions made in preparing l calculated reactor coolant system (RCS) reactor and related systems during the such evaluation models to the Nuclear pressure and temperature resulting from - course of the modeled event. Evaluation Regulatory Commission for approval by postulated ATWS events must be (within six months of the effective date limited so that either (A) the calculated 'see i so ssa for opprovst of this inco'rporation of the rule) or prior to issuance of an mtsimum primary stress anywhere in br refmme. operating license, whichever is later. sk h

b 1981/ Propose u enn-7526_..l'.eder_al Rcd.. ster / Vol. 46. No. 220 / Tuesday, Novem er 24. ~~-, 3- = ~. - rri' ability defHenres in those functi:ns .-w (iii)TEow m'o.% tions richssary to and systeme bt prevent or m'thats (*,) P/cas for compliamo ike'. reduce t!ie common niode fatture ATWS accidents.Ta cover th: wplicaut or licensee shat! submit a potential of the confrot rod screm possibility that the reliability assurance

foolption of all measures to be tden discharge volume in plants desfgned by to emure compliance with the criteria the General Electric Company including programs might fail to correct on obscure reliability defect. some 3ct forth in paragraph (b)(t). (b)(2) and diverse scram discharge volumelevel

' additional requirements for ATWS. ib)(3) of this section together with suchsensing desices: and mitigation would be selectively propored changes in technical (iv) Those modifications necessary to mandated.ncse improvementsin specifications and license amendmente provide a supplementary reactor trip ATWS tolerance of reactor plants have s may la necessary to ensure system that is diverse from the reactor bee.s chosen to afford an opportunity to r.cmpliance with these criteria to the trip portion of the current reactor learn from experience without incurring Nuclear Regulatory Commission as a substantiallikelibood of aa protection system. (2) Lemption. Pressurized light-unacceptable radiological release..- follows: (i) For alllight wster. cooled nuclear water. cooled nuclear power plants De NRCis exploring the possibility E issued operating licenses before January that the regulation of reactor safety may power plants for which operatinglicenses have been issued on or before 1.1DM or standardized to a facility at evolve toward regulating the processby August 22.1969. no later than (eighteen the same site that was issued an whichl censees ensure public heath and months efter the effective date of the ' operating license before January !,1984safety and away fromlicensing the need not comply with the requirements det' ails of plant design and operation. rul;). (ii) For cli light. water-cooled nudear of paragraph (c)(1)(ivlif the facility Programs like the reliability assurance power pltnts for which operating conforms to the requirements of program in this proposed rute offer l licenses has e been issued after August paragraphib) of this section except thatpromise of growing into a formal. 22.1969. no tater than (one near after the the fraction of the design lifetime used effective date of the rule) or prior to to determine the value of parameters auditable way the NRC can determine thatlicensees are doing a satisfactory Issuance of an operatinglicense, must be greater than that specified in of this section. Job of ensuring public health and safety. whicheser is later, paragraph (b)(2)(i) description of the A number of diverse regulatory (6) /mplementation. Each applicant or licensee shallimplement those measures measures together with such proposed initiatives are supportive of this tread. (3) Submittal. A Among them are the requirements on necesstry to ensure compliance with the changes in technical specifications or criteria sit forth in paragraph (b)(1) of license amendments as may be licensee staffing and organization, the proposal that licensees employ this section on the following schedule: necessary to ensure compliance with the (i) For alllight. water nuclear reactor criteria set forth in paragraph (c)(1) of probabilistic risk assessment methods powzr plants for which operating this section must be submitted to the as design and operations management licenses have been issued on or before Nuclear Regulatory Commission no later tools, and the pilot studies of independent design reviews.: August 22.1969.by dates agreed upon than (nina months after the effective The necessity for and conte'nt of the with th2 NRC.These dates must be date of the rule) or pnor to issuance of submittid for approva! not later than an operatinglicense, whichever is later. proposed rule is based on (1) operating experience to date with power reactor (threa ytars after the effective date of (4)Im lementation.Those measures under paragraphs (c)(1) of this scram e) stems. (2) s) stem reliability the rula). (ii) For alllight. water cooled nuclear section must be completed:

• "*I SI(3) the qualitative findings of require Y.

reretor power plants for which (i) For alllight-water cooled nuclear reactor nsk assessment, and (4) ATWS licenses have been or ma be reactor power plants for which accident analysis. operatinfter August 22,1969 but before operating ;icenses have been or may be There has been one partial failure of issued a (three years after effective date of the issued after August 22.%69but before the scram system in a commercial power rule), ell modifications shall be (two years after effective date of the react r.It occurred at Browns Ferry Unit complrted prior to startup fo!!owing the rule), all modifications shall be 3 on June 2a m AMongh ee ins (three years particular scram system failure mode first refueling that befthe rule)d nuc7 ear ' completed prior to startup following the - tha citer affective date o first refueling that begins (two years (iii) For alllight. water.coole after effective date of the rule). cause a severe radiologicai release rer.ctor power plants licensed on or after (ii) For aillight. water cooled nuclear accident, the event and the reviews (thre2 years after effective date of the .vactor power plants licensed on or after resulting fromit revealed a numberef rule). cll modifications shallbe (two years after effective date of the reliability deficiencies in the BWR completed prior to issuance of an rule). all modifications shallbe scram systmspese are now bemg. completed prior toissuance of an rectified by the mdustry subject to the cperztinglicense. review and approval of the NRC staff. (c) Additionaltequirements-{1) operating license. (d) Dose coleulotlocs. Applicants er. One objective of the proposed reliability Actuotion:In addition to those requirements set forth in paragraph (b) licensees are not required to calculate assurance program is to institutionah:4 cf this section,eachlight. water-cooled the potential offsite ra diological doses within the liceased industy the nuclear power plant except as provided resultingfrom an anticipated transient thoreugh evaluation aad implementation in paragraph (c)(2) of this section, must without scram event under i 100.21 of 'See.for eumple.TRC AcGon Han Devstaped be provided with: this chapter, (i) Actuation circuitry for ATWS As A Result of the BU 2 AcGonPian"KUREG-mitigating systems that is independent Sec nd NRC-Proposed Rule (the oseo.-Polg on bceeding Eth Pendsgu Coost,uc, ion ve,mit sod u.nur,,,,, ins c.nse end diverse from the reactor protecfion ((endrie Rule) Appisenons. SECY-al-:oD. and "Use d he essence of the second NRC. Independent Des'sn Reue ses IIDRsl in t!w system: Co e of [li) Prompt sulomatic conteinmern proposed rule is thal power reacter isdztion initiated by a significant source licensees would be required to 8js hnesMrkY Commission's Pubhc Document Raam.vi711 Street (f r diation in The conisinment resulting implement a reliability assurance NW, Wnkng'oa.D C. ~ from failute of the fuelrods following program to seek out and rectify ~ postulated ATWS eventr. t l 1 s N $43*MJ_k hhj [ .h h sh(( "N( h - %M L N duf *d Y N N M11

rederal Ity;ister / Vol. 46. No. 220 / Tuesday. November 21, 1981 / Proposed RuVes 6?/3W ~ ~ ~ n_._.__.___.:.n~~.m-~~~~~~-=~-m tsf t'neledons of esperience nith compromise the nvailability of one of studies 1.ko tho e now beirq made in ' functions important to ATw' the s3 stems required to mitigate cn response to t?ie Brow ns Ferry Incident. prner. tit,a or mitigation. NI WS et ent. or both. These c.m be counted on to make o. Reli4 bitty deficiencies in safety nus, a third object:ve of the ' recurrence of that failure mode mudi s3 stems ddier subst.intial!y in the Ltnd reliability assurance program is to less likely In the future = and frequency of opportunities to detect search out and evaluate the potential Calculations of the espected and repair them. Some faults are self-common cause failurcs that might consequences of very severe reactor announcing and thus elicit prompt cuntribute to failure in ino or more accidents hase been made in the repair. Others show up in each systems whose reliability is important to Reactor Scf. ty Study (WASit-1400)' sun ci!!ance test.Some faults may not AnVS accident sequences.nts scarth and other studies.ne raults Indicate - be res caled by routine surveillance should embrace not only auxifiary that the accidents that could tests. For Instance, the reliability defect systems but also human factors via test. realistically be expected to result in. responsible for the partial scrum failure maintenance, and operations, technicci lethal radiation doses outside the plant at Brow ns Ferry could not have been specifications dealing with equipment site m e those denoted as release detected in routine sun eillance tests of availat;ility and environmental category 1,2 or 3 accidents in the the scram system. c,,nditions in the plant. notation of WASil-1400.%cse are also System reliability calculations by the A fourth objective of the relintEity the accidents that are expected to cause Electric power Research institute and assurance program is to search out and substantial offsite property damage. others have shown ihat component . valuate the susceptibility of the Studies of ADVS accidents in fzilures of reactor scram systems that redundant divisions of each safety pressurized water reactors (PWRs) are detected and corrected in each system important to ADVS prevention suggest that only a small pertentage of surveillance test are very unlikely to or mitigation to common cause failure. reactor scrams are limiting transients. cause ADVS events. Other systern Concern with common cause failur* Dat is. only a smaU fraction of the failure modes can only be detected in modes of the scram system has been opportunities for AUYS accidents occur some but not all surveill,ince tests.Still central to the history of the ATWS under circumstances that most severely others show up only in some or all contros ersy. challenge the ABVS tolerance of the genuine demands upon the system. A common cause failure of an plant. In addition, the qualitative Some reliability defects cannot be electrical natars has already occurred in findings of PWR risk assessment studies detected even in genuine system a reactor scram system In a commercial suggest that even the tnost limiting demands unless triggered by other nuclear power plant (Kahl reactor) that classes of ADVS accidents in PWRs are failures. Examp!cs of the latter category could have resulted in its failure to unlikely to produce a release category 1. are the hydraulic design deficiencies in operate on demand.That failure was 2 or 3 radiological outcome, the UWR scram discharge system detected during normal surveillance and in boiling water reactors (BWRs) u res caled by the incident at Browns rectified. A similar common cause substantial fraction of scrums take place Ferry. Such blind spots ir, the esperience failure was detected and corrected in under circumstances that can lead to a tease for safety systems can conceal the startup testing of the MonticeUo limiting transient. BWRs are least serious flaws in reliability.Thus a reactor. Estimates of the upper limits of forgising of those ABVs events in second objective of the reliabihty the frequency of ATWS events for the which the reactor is isolated.Even tf assurance program is to conduct a commercial power reactor Industry are reactor isolation does not cause the thorough analysis of the startup test of the order of10-8per reactor year.ne transient in the first place, the effects of program, the surveillance test program. NRC staff has concluded that operstmg a failure to scram are likely to trigger and the record of system functional experience is not sufficient to determine reactor isolation.Furthermore.BWR risk esperience to identify and-where. conclusively on a statistical basis assessment studies suggest that ATWS feasible-close loopholes through which whether reactor scram systems are accidents may give rise to release design deficiencies, construction reliable enough to make the probability categon 1. 2 or 3 (as described in deficiencies. vulnerability to test or of unacceptable consequences from WASH-t400) outcomes. mamtenance error, or component ATWS esents sufficiently small. nese arguments suggest eat Ns failures might escape detection and thus The improvemen;s emanating from may already achieve the minimum correction for considerable periods of ~ the proposed reltabihty assurance ATWS tolerance nessan to regram will make ATWS accidents less supplement the reliability assurance time. Studies initiated in response to the fikely and the systems that mitigate en uld, {'. '*Q..d i jo's ] Browns Ferry partial scram failure ATWS events more reliable. i indicated that two auxiliary systems, Nevertheless, it is necessuiy to ensure their provisions for AT%.S m.tigation. i that serse the scram sytem as weU as that mitigating systems will render the Homer,a more careful analysis of other systems, could have caused partial outcome of most ATWS events ADVS-tolerance is required in the or complete scram failures.This acceptsble.The principle of defesse in Pmposed rule to proside the basis for discovery is suggestive of a class of depth calls for reactor plants to be and form of actions to be taken by common cause failures that might designed and operated 6 such a way UC'" 5" compromise the safety of a reactor. that a rare ATWS accident can he In FWRs. the limlfing transient wtth Failures in auxiliary systems might tolerated. - cause the initiating transient as well as The require'ments for ATWS tolerance respect to ADVS is a complete interruption in the delivery of feedwater degrade the reliability of the scram . In hght. water cooled commercial power to the steam generators at full power. s; stem. or they might contribute to the reactors are intended to afford au Shuuld the scram fall to shut the reactor scram failure and also could opportunity to leam from expenence without placing the public bealth and

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'The enciary systems of note are the went . safety in jeopardy. The first occurrenca from the Dmsion ofTechnicalInformahan and sistem servins the naam d.scherp volumes. and w co,npr,=ed.1, pt, runs tu,wer ted of an ATWSprecursordue toany poc,. ment Control. US Neclur Regulatory particular failure mode wiH result m, c-nimian. w uhmston. on sosss. .cr.. v.2, o

/ Vol. 40, No. 220 / Tues&y. wm wy. ~.~. ~.~. w mm ~ f i t van.~ -. m. u ~. m.3 ~. pressure f$ction system.and b) th3. 5752!!, I ederal 1%.. s er ..-..n gross a',ovep o.md I.**b ' ' ' < c iN si r m. tor coc!,mt pressurs -[ dr..n,'d e wn'inud r>owcr generatio<. containment.nis f r n. cound vy valves (Nough which a LOCd .m 4 4 and the dahnm2 ea2 remo.vf.as the probable outcomes of everi th uost c. would bypass containment and'could,;. '- i ; h snur.d.ny coolant boifs away,causes a

f notbeIsolated. fin some PWRs.the very ra p;

..,4, e c. severe and d imaging pressure ~ surge in pressure of the reacter coolant. - excursions associated with ATWSin ne sesenty of this pressure excurston .~ - autostart of die auxiliary feedwater f PWRs. Is a sensitive function of the moderator Analysis of ATWS transients by th( system foIIowing a fecdwater transient, t temperature coefficient. 0,e capacity of NRC staff and the reactor suppliers can overcoolthe rea torif the scrarnIs successful. in such plants, the rapid start, ,l the rehef vahcs attached to !!.c reactor suggest that Westinf.ouse reactors. coolant system, and the speed with have sufficient reteicapacity so that ', logic may be Interlocked to take place.g(, j only if the scram fails.!!awever, such c whkh the auxiliary fcedwatcr :)bs.Ide pressure excursions expected oflimiting stem, i ATWS transients will not be damaging, ~ interlocks must not degrade the r.. t g ',s l starts.%e pressure surge will su as the power decreases due to the provided that the auxiliary feedwater reliability of the auxillay feedwater p,3 incr easing moderator temperature., system starts promptly. Combustion ' system for the more frequentloss-of.# r CJ reofenishment and reactivity controlis Engineering and Babcock and Wilcox. feedwater transients in which the scram ; - Subsequent reactor coolant reactors may be subject to severe la successful and la which a delayed s,.. - ~, ,5 prosided by the high pressure injection pressure excursions even with prompt ;autcstart of auxiliary feedwateris s e t (IIP!) system, which pumps cooling. start of the auxiliary feedwater system, appropriate.Deidentificationof the 0.1. *

• water cor.taining a reactivity poison into should the ATWS accident take place -

m. requiredinstrumentation and the. p.s - the reactor coolant system.. ; ~. when the moderator temperatureble.The NRC staff, training of operators may %e most severe test of the ATWS part of the reliability assurance tolerance of a PWR lies in its survival of. coefficient is unfavora' has argued in NUREC-0100 that these program, and the verificatiort that the the pressure excursion and in the ~ ~ plants should install additional relief..Instruments and the critical pressure, successful sta t of the auxiliary capacity toimprove their ABYS - boundary valves on the reactor coolant, system have the required tolerance foe, d that feedwater and high pressure infectionsystems.He possible outcomes of the. ; tole the limiting pressure excursions would,; pressure excursion are (1) the reactor wiD produce substantial occupational be part of the ATWS tolerance ?..,. A exposures to radiation to those f requirements. , - G M ';. coolant system and interfacing .s L equipment are undamaged. (2) the., installing them, and are unnecessary Ue moderator temperature f.e. reactor coolant syster. remains intact, because the plants already have coefficient,which strongly inDuences

• O the severity of the reactor coolant but instruments on the pressure -boundary fa0 or the valves for the llPI., sufficient toler has concluded that the minimum ATWS ' dur system are damaged (3) the reactor proprietary reports.,.,,

'l' coolant system is ruptared producing a a with the first fuelload.%e early months ; lossef-coolant accident (LOCA)to,i ;; h-containment. (4) steam generator tubes 'tolerance necessary to complement the, of plant operation are also characterized. rupture causing a primary.to-secondaryreliability assurance program does notdictate additional . i.OCA or a LOCA to other interfacing,. ,4I syste ms, or (5) combinations of (2). (3) or. capacity in CE and B&W plants inlight - the plant is shaken down and the plant of the several mitigating factors noted (4). The first outcome is clearly...... personnel gain experience with the preferred.He second outcome makes It above.Ilowever, there are a number of equipment.nerefore,much of the risk clear that care must be taken to ensureother safety-related incentives to alter associated with ATWS accidents is a the provisions for reactor ccolant.. expected to be concentrated in the first that the operaton have sufficient .m,, ir. formation about the status of thepressure reduction or relief in PWRs.' - months of plant operation.One reactor to manage the recovery. ShouldDeseinclude deliberate...,.. ~ mitigating factor is the less than. the llPI pressure boundary valves allstize in the c!csed alignment, the core,,. d< pre equilibduminventory of fission products a f. P sarety injectionin smaD LOCAs and . accumulated in the fuel at this time.. d! melt.nis is one of several paths. feedwater transients with scram,to 7Nevertheless,PWRreactorlicensees % from ADVS to a contained core melt. avoid the melt-through of reactor vessels N: w would be required to pmpose and imp!eme'nt particularly stringentlimiting. ' cccident. A LOCA to containment is likely to be mitigated by the Emergency, wh!!e at elevated pressure,and to r.,.i .. conditions of operation in the technical - fd: ,, enable the ECCS accumulators to extend the point of no return for the ?l.* Core Coolin; System (ECCS) even ti restoration of AC power in station ~. 9. specifications to constrain though the initial pressure conditions..,,, blackout accidents.%e NRC expects to'. ', when combin .j cre outside the design envelope for.. ;- 'r. ECCS analysis.nus no core meltis ;. 3J take up the case for and ainst altered the prevailing moderator temperature r. :r,W-expected (although a contained core

j. -

melt is a remote possibility), and a core".. pressure relief provisions or PW coefficient, and the powerlevel.. , f., - . forthcoming rulemakings on aevare ;; -accidenta.. ; * ~ > b . r. 4. melt with missile damage to j ' containment is a still more remote, possibility. Steam generator tube ruptureDe required ADVS toleranidof lalargeimodern boiling water o.1 '1 can provide aleakage path to the.,,'.'. PWRs rests:(1) upon the prompt start of reactors, a transient with failure ta'~ '.' ; { the auxiliary feedwater system (2) the ' scram from full power is very likely to.'. cutside atmosphere that bypasses. contalament.liowever, ECCS is likely tobe successful, so the c' ore would not.2 a l follow,the isolation of the. erstors to diagnose the ATWS reactor, nota the o[ent sequence and successfuDy i melt. Allbut one steam generator can ~~ occi isolation valves.If the reactor coolant 4 }'. .very probably beIsolated,thus. ,,,o maneuver the plant to minimtze the ~ recirculation pumps continue to run, the j release of radiation (3) the tralmns of high and a ; terminating a mlnor release. He sevare release category 1, L ors. clperators,{4) the avaIIability of the high jcwerlev jj events occur only for a core melt and s'. ..,.. j.., g ~...c 4.; ;.. f' tN edL 'b *.. r s 7 '2 '., X . t _. ~ ?' i ' [w.yG',n ' M. :. ".W:' K...:..f:.:5*L'.T1;'. ".t, ' ;._ ':n ;,. _l, :; r ? a ~... ' k t".~ **,.r "*U (b. / ^ .. s..,. :. - ' 's + --.i ., p<*

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5 Ihferal itenister %A. 40. No. 220 / Tuesday, Nravember 21, 1931 / Proposed ' Rules 57529 w??".:?.~r' M '~~~~.~~-- ~ -~= ~ ~ ~ ~ ~ ~ ~ = = roetc gu(sr.ure c' ;it,lon wdl take. ReSctor Core Isolativi Cooling s3 stem A~t WS crents, thus ll reateriing ef I successful ndti;ption. In some sequence

• s place. IT en if the teactor coolant system' (RCIC) should be expected to autostartvmiants, operators might be temped to '

surc es lle pressure surge, the very, and run, delivering coolant to the * " reactor.%e flow rate delivered by the depressurize the reactor to enable foW.- high steum flow will rapidly heat the RCIC is lower thart that of the !!PCI. lf.,' pressure reactor coolant injection bd.in' suppression pool and pressurfte the ' the RCIC is the sole operative means of so doing, disable turbine-drh en coola,1 s ' a containment. In addition, the high! pressure coolant injection (IIPCI) m'ay replenishing reactor coolant, the injection systems or otherwise., + not suffice to cool the core; overheating - adequacyofcore cooling ratherthan compromise possible avenues et '.W > S the heat deposited in the suppression ' successful AB)S mitigatiort.%eJ5b and core damage may fulfow. tJ1timately' pool. is likely to be the factor limiting reliabihty assurance program must we 1 the time allowed to shut down the entail a thorough 1ns estigation of sucO " the containrr.ent is ' expected to rupture due to os erpressure whi!c the core rcactor without unacceptable ~9 ADVS accident sequences,of the a M,. + instrument indications available, and of * ' sustains damage. Continued core consequences. De RCIC can ' coolant replenishment is questionable 1 - successfully cool the reactor once it is the possible range of operator actions.- s after containment rupture. A large ' ' shut down, and it can slow the boiloff of Operator training should familiarize ' outcome. A necestary mitigating feature reactor coolant in the reactor. c-operators with the optium strategies and N radiological release is a plausible De NRC has concluded that the alert them to serious errors that could is thus a prompt automatic trip of the liquid reacthily poison injection system occur in dealing with ADVS accidents f. recirculation pumps to avoid the in large, modern BWRs must have a subject to a strong disincentive to ' ' ~5' BWR reactor operators may be pressure excursion and diminish the start time and poison injection rate such power and the consequent steam flow to-that either of two redundant trains of,, actuate the Standby Liquid Contml,'b, the suppression pool. Given a trip of the (SLC) systembecause of the coscy ' high,-pressure reactor coolant replenishment systems. either of which, nature of spurious SLC actuations.ney - recirculation pumps,the rea'clor power will stabilize at roughly 30% power until may be expected to be available under may alsobe inclined to override an *. N the reactor coolant boils down and ATWS ounditions, can successfully, autostart of the SLC Lf they doubt that. - steam bubbles tvold formation)ln the - mitigate ATWS transients.%e two. an AnVSindicationis genuine or the.. core throttle the chain reaction.J - failure of the scram system is trains may be the llPCI and RCIC. He criterla of successful mitigation' Irreparable. %e NRC recognizeas the b.. f' .Hereafter, a static or oscillatory ' equilibrium willbe maintained in which ' areq1)b containment temperature ' ~ - ligitimacy of the concern with the cost of the reactor sustains the average power and pressure must remain within the. spurious SLC actuation necessary 1o boil o!Ihowever, much j. To deal with these conflicting '^.M reactor coolant is delisered. up to aboct, design envelope.{2) the core must retain concerns, the NRC proposes to regidr coolable geometry, and (3) neither 3'y'i pow er. Analysis shows that liPCI or Prompt fatalities not sedous offalte ' the automatic start of the SLC system ' - main feedwater can adequately cool the - core to avoid estensive core damage. Property damege are predicted by (

• under circumstances diagnosed tobe analyses whose conservatism 1s ADVS seque'nces. Ucensees are free to Ilowever the power delivered to the compatible with that enzployedin ' '.

employ rehability engineering methods. suppression pool will be greater than the to minimize the likelihood of spurious )l WASH-H00.* . pool cooling system can dissipate. Concern has bee.n expressed that the actuations under non.ATWS Herefore, contalnment overpressure.. RCIC. though capable of meeting these circumstances provided these pro'visloas friture remains a distinct possibiltty success criteria, does not prevent the do not compromise the reliability of the

• unless the reactor is shut down. either "' automatic depressudsation of the essential SIE safety functir n in genuine

. by control rod insertion or by11guld .~- reactor coolant system. Operator etion ATWS sequences.

• ^ '

reactivity poison injection. Well before. the containment is significantly" '5-', 'Is necessary la less than ten minuted t2'Inlight of the analysis and operator ,1 training associated with the reliability /j ". override the automatic depressurization pressurized the suppression poolw! or to throttle low pressure ECCS should assurance program,it is not d-emad cpproach saturation, and b steam. the depressuriza' ion occur.& NRC U necessary 80 preclude provisiona foe c condensation willbecome unstaHe. ' Chugging steam condensation may ~ staff does not wish to force an afteration ' manually overriding the autostart of tho' ' of thelogicgoverningthe automatic" l' SLC. As part of the reliability assurance threaten containment interpty or .o pressure suppression and thus shorten depresurization system (ADS)whi& 1" program, a thorough analysis is to be ii made of the circumstances in which am the time available to shut down the might compromise the reliability of the - operator might be tempted to overt ADS in non-ATWS events. Options to retctor without unaccepteNe s resolve these competing concems wM1

• genuinely needed SLC actuation..". 6 consequences. Inlimiting transients,the ;be evaluatedby the NRC staff during Consideration should be given to

..Y fillure of the main feedwater system ~ improved instrumentation if the correct mry be the initiator of, or companion of. the comnent period. We are interested ' In receiving comments on the potentid *s diagnosis of such sequences is e a.,- the initiating event.De HPCIis s singe-train system.De fauh orhuman ( effects of the a ree posed ruice on 't ambiguous. Operators must be tratnad 1 [, crror that precipitates the initia8 i ' ,t this suSsystem th -pressure makeig) to gh e first priority to safety rather than Transient might also disable thelfPCL1s of the BWRdesign. ' - .J.r ., v to the availability of the plant for power generation.& anticipation that ..m Severalfactors complicate Sun '* cddition, system reliability analpas ' 1 analysis of the AIWS-tolerance of BWR repested manual scrams or quick han [- h:ve Indicated that l{PCI may fab orbe ' P ants. & delivery of mala feedwater. - in the control cabinets may succeed ta 1 l unavailable in as many as from M to ' 10T of the casesin'which a demandis [ whia may be available la some ATWS ' inserting b contml mds would be ans"- made of the syster'n. dis maybe~ j. accident sequeoms, may ddute liquid "*. unacceptable justification foe oven (Sng insufTicient reliability for the m!tigetlen Potson and increase h powerlevdin SLC actuation. ..'*.-m. " - + la conjunction with this fodn qf das M ' 1 Of a potentially serious acddenthavirig" g$ 1 rule the NRC does not deein It ^ -M ga j o frequency of occurrence that might be ( necessary that b SIE meet h 31:$e. ' ts high as once in a htrsand reacter,* I cw " washington.Dn sannsJhmern ted MNurww* ^. f ' failure criterion as well as the years. A second diverse system.9se ~ j 7-j [*. e.3, y. - 3;4 y

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(a)fnitio/rellobili. osrurancb ~ 22') / Tuesh.~ - rAl Register / Vol. 40. No. ~ pwpoq.%ehitialreliability,', m.oss .I l f sit-we ad to the 4e NRC stcff.o. j* I'c n.- g y g. m.y,. JD cnalysis and classification of t e.,,.h L ; ', sm4h of the petaifed invelamentIn... i of [ f.. ss criteria. ia tle very unlikelyexpcHence revie.v ante're uclect on principaldeterminants of tho.. f h procedural or hardware backlits in t econtext of AUVS t of an AUVS esent and a failure o radiolog* cal seserity of each clas L matic and manual starts of the St.C n- /~ of l ble i f( d the proposed rufe emphasizes criter a, the Initial plant conditions, the type 2cm, a fa!!back strategy !s avai a s forthe soundimpler.entationof thereliability assuranc augh manualrod Inseition and 1 d the, ', i its t ction the reactor protection system, an f l1 2rvention in the reactor pro e = stem control cabinets. Nevertheless.SLC must not depend upon a singfe state of operability orinoperabiht .,,p' the staff review to these criteria.together with the con Fj r h f dments sision of an aux!!iary systemthe tbe and approvalof thelicense amenassociated with changes . outcome.%is analysis mus 3 ilure of which would also compromise f employed in each of the following i of ' u e reliability of the scram system ore recirculation pump trip m prec p a j .<<r. i it te Pursuant to the AtomicEnergy Act of ... -.. ';.. J.

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operation.. ~T.- (1) Training oflicensed reactor is. / prograan 1954, as amended, the Energy n 1 einitiatingtransient.BWRs must also operate under. Reorganization Act of 1974,as amended, operators in the diagnosis and p'rognos-7 i k; d f and section 553 of title S of the UniteStates Code, of the several AnVS accident.

pecified Umlung Conditions o

.ti d to ~ sequences. Operators must be tra r e @peration that constrala power during j l i, dments generationundercircumstances n. s make productive use of their time adoption of the following amen l which equipment unavailabiity d - to10 CFRPart 50is contemplate... j t compromises the reliability of sys em Consideration mustbe given to 'ays. '..{ j H Important to AnVS prevention or.PART 50 DOMESTIC IJCENSING OF improvingInstrumentation. disp! 1 est i mize ZATION mitigation. The older lower-power-level reactors. PRODUCTION AND UTILI 1 / o eg y r, p3 {d l f FACILITIES..,, <...s. ;tation for10 CFR ~ r mzy differ significantly in theleve s o debyed diagnc, sis of AnVS sequences AnVS. tolerance rovided.These plants inay substantially increase the 1.De eu horit ows:. y. $y'the radiologicalseverity of the outcotDe. - Authority Seca.193.104,161.t82,183.es.. Par (50 rea s as ~ folera ce1o ei (2) An analp.s of hypothetical errors Stat.93e,937.9fS 953.954,as amended (42US C 21 in or erroneous departures from proper. th ir A l NRC sitff.Thedualapproachof ATWS test and maintenance proce 2 5646), t to

j 206. 88 Stat 1244.U40 [42 U.S C. 584,

,tolersnce and the reliability assarance 7a also h Each entess otherwise noted.Section so. issued und program provides defense in dept. i ' US.C ~ ATWS prevention or mitigat on.. ,1' l c!Iows the other to be implementedwithout highly conservative marg n Consideration must be 8 ven to ' i i 2152. Sections 50 80-5 sec.184. ea Stat.954. as amended. (42 q rovidedby ATWS-fit' 2:341. Secs. 50.100-50.102 usued under see-1as,es procedures, and personnel traloing to ? Iolerance al ows tealistic cost-bene %e margi mmimize thelikelihood that the i considerations to govern the select on reliability of these systems w 8 Stat.958, as t ed ' purposes of sec. 223.6 cnd schedule of implementation for' amended. (42 US C 2273). I 50.54{0 asu - SC. fixes suggestedby the reliability *under sec.1811. 6a Stat.947.(42 U22n(01, and Ii 50.70- ........J 8 hsued b experience base with syst nded;(42 cssurance program.The very costly accident at%ree Mile . maintenance. under occ.16to.68 Stat oso, as ameUS C 220t(o)) and the Island has demonstrated that the i Important to ATWS prevention or... protection of alicensee'sinvestn ent n (' ' mitigation through which reli Appendices.

• o reactor plant provides a powerfuleconomic Incentive to search out 1

2. A new $ 50.66 is added Io r'ead as.7 d.

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~ -'..U? correct' reliability defects in thefunctions that pro.ect a reactor core considerableperiodsof time. lollows: - ?* $ 50.60 Standarde for the reduction of risk ' must be classified by (i) kin W from Antic 3patedTransientsWithout Scram .4 ' from damage.These economic - (ATWS! events for Dght water. cooled.

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deficiency, construction deficiency,- - . considerations. togeth'er w

.3 vulnerability to test or maintenance svaluation of offsite risks affectingpublichealth and safety are suffi

"""I*'##***'8 *" 9 I power reactor licensee'aball establishEach ) error, active or passive failure).(ll t (lii) f ' ts determine the scope and schedule o - affected compone".ts or sub and maintain a reliability assurance the more expensive or intrusiv, design, h df program for functions associated witd itigationof .- W - and(iv} the frequency andkin o ;..d ficienc c!!erationsin plant operation oremergingfrom the reliability assurance j . /,, ' opportunity to detect the e

• test program covering startup or on

. tha prevention an m j Anticipated Transients Without Scraas timeenly tests, tests ass j e relisbility assurance programla {ATWS) employing state-of-the-art a am. i p methods and procedures toidentify n ' got to be a paper study to demonstrate .I ustbe developed dy to the NRC staff that the plantis alrea I t tionof t i and Implemented so that 4 l ti m safe enough.The role of probabiis c ' cvaluations is secondary to the. ? -@ f ~ is responsible for the impleme J d u , cost-effective improvem ATWS risk.Defensein depth mustbe the extent reasonably ac i o qualitative search for and evaluat on - De maintainedby operatingcommerdal. -powe f.,. f specifictypes of reliability defecta. f reliability assurance programla notIntended primarily to assist NRC the plant to common caus = h t afford an opportunity to teamfrosa eexperience dinto t i review.Rathee. itis to beintegrate li bility ef.7 ' aevere radioactivity releases. Specific f.". cause degrades the re asM red y *; the conduct of plant management.. f r personnel trotning, and the conduct o acceptanoc criteria are delineated thea' 2 ~ operations.Itislatended to strengthe responsibility for safe design an l ; . h,q $y.f J.w - ~~ d a e j ~ ),F 'n,. .g ..~ ;. i below. h the g i~.. .f. 1 eperationof theplantrestingwit. ~. w.' i. ) : ....- a.:.. ...n. f" - :,z. c. 4 .~..:.g:, f ~, -r t T. :~.':>%...p-: :* -g .r f-y.. Oc'* . m.... ~ w ~~ ; J n .6E

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h.. I'ce!...il Regster / Vol. 43. No. 220 / Teesday, November 24,19M / I reposed Rules 57531 important to ABVS prevention or (2) Pressurtied water re.ictor licensees and approval. Iloiders of operating ) mitiption.and thos. in which a single receising an operating license after licenses, applicants for operating root cause degrades the reliab!!ity of Aupt 22,10G9. shall:. licenses, and those expecting to file an . two or more s3 stems whose concurrent (i) Pioside for the prompt, automatic application for an operating license f,iilure contributes ta a sesere ABVS start of the auxiliary feedwater system within one year of (the effective date of accident sequence.The kinds of root under circumstances indicative of a the rulej shall file the reliability gauses to be consid, red are those listed transient entailing loss of main assurance program plan at a time to be in paragraph (a)(3)(i) of this section. feedwater and a failure to scram. agreed upon by the NRC staff.%e time Consideration must be given to (ii) Ensure that the instruments afforded for plan development will be im roved design or eperation to reduce necessary for the diagnosis of and not less than one year (from the - vu nerability to common cause failures, recovery from AnVS accident effective date of the rule].%ose holders sequences will not be disabled by the of construction permits who f;!e an. (b) Continuing rel,ubilit assurance pmgram. Each commercia power effects of such accidents, and application for an operating license on (iii) Enwre that those reactor coolant. or after [one year from the effective date reactor licensee shall maintain a system pressure boundary valves of the rule] shall file the reliability continuing reliability assurance program for functions important to ADVS through which high. pressure injection assurance p.ogram plan at the time of prevention and mitigation that includes can reach,the reactor remain functional operating license application.He plans the following: after limitmg ADVS transients and (1) Configuration control for designs, Jhose valves whose integrity is essential must identify (i) the ways the reliability gg, g o ea 8 procedures, and technical specifications d Io c dents into the engineering and operations to assure consistency with the initial retain their integrity throughout limiting management of the plant (ii) the reliability assurance ana (ting affected ses. ABVS transients. reporting and approval requirements (2) Procedures for upda assurance analysis for, and prior to, power reactor licensees not covered in - interna' to the licensee *s org portions of the initial reliability (3) Commerciallight water. cooled (iii) thr plans for information evaluation paragra hs(c)(1)or(c)(2)of this section and es ;hange among licensees as part departures from the controlled design, procedures, or technical specifications. shall su mit an analysis of the ADVS of the experience feedback function (iv) Applications for license amendments to tolerance of their plants. the criteria for reporting to the NRC,(v) (4) Each commercial power react'or the criteria for the adoption and implement these changes must include a licensee shall prepare, submit for review scheduling of alterations to plant des!gn brief analysis of the impact of the and approval and implement Limiting or operation emerging from the change on the reliability of systems Conditions of Operation that proscribe reliability assurance program, and (vi) important to ADVS nsk. operation in, and mandate expeditious the date at which the initial reliability (3) An experience feedback system to retreat from, operation under conditions assurance studies can be completed. A review, operational and test data on that compromise the AnVS tolerance of brief summary of findings and plans for relevant systems in the licensed plant the plant. Limiting Conditions of the resolution of reliability deficiencies and the relevant experience,at plants Operation should also minimize must be filed with the NRC upon having a similar system fugn. Each operation under conditions in which the. completion of the initial reliability operational occurrence nest be ABVS tolerance of the plant would be assurance studies. Subsequent resiewed for clues h ersig'it or errort severely tested by a limiting ATWS discoveries of reliability deficiencies in' in the reliability asuranet enalyses. event. Consideration of the prevailing the plant must be reported in accord The initial reliability essurance analyses plant parameters as well as equipment with prevailing practices for reporting and cost benefit analyes based thereon operability is appropriate in the Limitin8 licensee events.The reliability are to be updated when tu experience Conditions of Operation. assurance program will be subject to feedback system reveals oversights or (5) For the purposes of this pargraph, audit by the NRC. It is not expected that limitations in these studies. the ADVS tolerance of a plant is the NRC will engage in routine review (c) Design ondoperationforATWS inadequate if any of the more limitin8 and approval of the program unless a - tolerance. (1) Boiling water reactor transien's, followed by a total failure of pattern suggestive of noncompliance is licensees receiving an operating licens* the scram system, result in any one of b "

• d.

after August 22,1969,shall: the follo * (2) Applicants for or holders of (i) Provide equipment to trip (i) Con nment pressure or automatically the reactor coolant temperature above the design values, operating licenses subject to paragraph recirculation pumps under conditions (ii) Loss of coolable geometry in the-(c)(1) or (c)(2) of this section shall file with the NRg p!ans for the indicative of an ABVS event. core, or (ii) Provide equipment to (iii) Releases of radioactive material implementation of the requirements of automatically deliver liquid reactivity that may realistically cause any offsite paragraph (c) of this section [within one poison so that either of two independent prompt fatalities or serious offsita year of the effective date of the rule] or reactor coolant replenishment system property damage. upon license application, whichever is trains expected to be available during (6) Applicants or licensees are not later.The fullimplementation of the ABYS events can srecessfully bring the required to calculate the potential offsite requirements of paragraphs (c)(1),(c)(2), 1 reacter to stable hot shutdown.The radiological doses resulting from an and (c)(4) of this section must be poison inlection system must not depend ABVS event under i100.11 of this completed: 1 for its function on a single division of an (i) For all light-water cooled nuclear chap)ter.(d Schedule ofimplementation and reactor power plants for which / auxiliary system whose failure could I precipitate the transient, degrade the reporting requirements. (1) Plans for the operating licenses have been or may be reliability of the scram system. or defeat implementation of the reliability issued after August 22,1969. but before the recirculation pump trip, and assurance program called for in (three years after effective date of the (iii) Provide a reliable scram discharge paragraphs (a) and (b) of this section rule), all modifications shall be volume system. must be filed wit! the NRC for review completed prior to startup following the

[ _. l eder,,1 Renhter /,Vol. 40 No, 22M N,,,.m.--g% l ~~l uotations, and (3) sufficient issuc rinclos ~.~ .....,~ -.w - ouId be / Ann:bl cW r,ia Securities Dea!crs Automafr.) Quobthm. ez.da relaxe to mote dom) scscmbiaSptem ("NASDAQ") now in cpet.2 tion rJefin[ 2.ri licMr. (i.eu "eaa scquirements establishni by trajor. for ten ) ears,has gically Imptusd th3 r effective date cIti.e wls).d nuclear efficiency of the OIC markct and hasa i)l'er a!!li;;?,t water coofe it r e xchanges.. t or paer plantslicensed on or s e O ATC Comments should be received by cars after cifes.tive date of the January 29.1982. Comments, whkh should refer 1 Board.The SEC, over t!>e past few years,hasimproved and strengthened. he )ll mod,fications shallbe to Docket No. R-0372, may be maifed to AconEss: snpfeted prior to issuance of an its disclosure rules, so that financial r), a ~ the Secretary, Board of Covernors of theFederal Reserve information on foreign ea well as ratirg lic ense. 3)lloiders of operatirglicenses domesticIssues is available to the Constitution Avenue,N.W., Washington, ject to p.sragraph (c)(3) of this public In a co,nprehensive and timely tion shall file with the NRC p1ans for D C. 20551, or de!!vered to Room B-2223 fashion.In addition, the National g e accomplishment of the ATWS between 8.45 a m. and 5-15 y m, Comments recehed may be Inspec Association of Securitics Dealers ferance assessment called for in ("NASD") requires that its domestic and 15 Room B-1122 between a 45 a.m. and 5: .ragraph (c} of th!s section[within one foieign issucts file financial data with it p m., except as provided in i 261.6(a) of. err of the effective date of the rule). W, the Board's Rules Regarding Availability as a prerequisite for trading on uchlicensees shall file the results of ese studies, together with proposed of Information [12 CFR 261.6(a)]. NASDAQ. None of the approximately one-hanges,if any,in design, procedures, ron runfuta turoRMAfloN CONTACT: hundred ei hty (180) foreign stocksc nd technical specifications to assure Robert S.plotkin, Assistant Director, 8 ATWS-toferance, and a proposedOmplementation schedule shallbe filed I aura Homer Securities Credit Officer, placed on the OTC Ust, as they do not or Jamie tenoci, Financial Analyst. Division of Banking Super meet the existing criterion,which !e requires all OTC Ust candidate to be a f ef c Regulation [202-452-2781). " organized under thelaws of the t) e SUPPLEMENT ARY INFORf.tATIOPc In July I ' "gd D: tid at Washinston. D C., this 19th day of 1969,the Board adopted criteria for requests have been received from both including stocks on the OTC Ust.indiscussions leading to th investor groups and the general public to Novemtser,19el. por the Nucicar Regutatory Comrnission. include foreign OTC stocks on the OTC SN IN such criteda,the Board indicated Ust.When the Board first adopted its , Secretary ofr3e CommissiorL generally that (a) stocks to be included I criteria for inclusion on the Ust, ther on the Ust should have market pt on s -nua ra.4 wei.eas**i characteristica similar to exchange-listed securitie s. (b) manip foreign issues.nis problem has now ad>o Cooe r586-4W__ _ - _ m been remedied. Furthermore, foreign issuers to be included or excluded fromthe OT c.- FEDERAL. RESERVE SYSTEM as possible, and (c) fluctuations in thenumber of stocks o exchanges and are thereforeautomatic 12 CFR Parts 207.220, and 221 in this connection, the Board also minimized.ne changes now proposedin the proposes to allow AmericanDe ository IDocket No.R-03T21 Receipta ("ADRs") to be eligible for OTC Ust criteria are the result of a proponiTo Rav se Criteria for initialcnd Centinued inctusion on the Ust of redew of the OTC margin stocklisting t inCIusion on the OTC Ust. ADRs are and continued listing Tequirements IDreceipts issued against securities of OT C Mirgin Stocks light of recent developments in thesecurities markets in gen foreign issuen deposited in an ccENCY; Board of Cobernors of the American depository, and are exe Federcl Reserve System. htarket in particular, and staff Proposed amendments 34 Act.There are approximately sixty experience with administering the AcTaoN: requirements. It is believed that revising (60) ADRs currently in NASDAQ.He The Board propos-s to amend the criteria is especially appropriate atBoard would allow ADRs to be the requirements set forth in Regulations $UMMAnY: (1,T cnd U for inclusion and continued this time because of a recent decision to considered for inclusion on the OTC ust provided the foreign securities revise the Ust three times a year . snclusion on the Ust of OTC h!anpncommencing in 1982. rather than twice a against which the ADRs are issued are ~ Stocks ("OTC Ust"t Brokers anddealers may not extend credit on stocks f6 as is the current practice.This has d en a frequent recommendation of the 'c[wh im o ce ein rti g which are tratied over the-counterunless such stocks appear nn the OTC accurities industry, ne following is a g,g 5 $;[p Ust. Loans by banka and other lenden discussion of the specific proposals to ahs nsiste t wit th:t tre used to purchase stocks thatamend OTC Ust criteria. policy currently employed by stock' exchanges with respect to exchange . eppect on the OTC Ust are subject to A. Deleting Requirement nel Issuer be Organized Under the Laws of the United listings ' and wi'h the S the at a "' nts w uld Exchange Commission's current gd m n modify three areas in the existing rules p, States or a State pMposal to aUow ADRs tobedesig As'early as 1064, when the SEC first for initial and continued OTC Ust recommended a broadening of the cligibility.First, they would permit Federal Reserve's margin authority toencompass over th

%,

• I'$,'M,a securities squity securities of foreign issuers andAmerican Depository Receipts ("AD

,. ~. Board indicated that accurities, to be in be considered for OTC Ustinclusion. eligible for credit at a broker. should. thmror..,,,mma ra,, s,ce,an is. Second,the proposals would replace.cartain criteria which must currently be meet the prerequisites of (1)marke 'sEc melem No. 34-isus. depth,(2) a reliable system of mit in the af ternative and replace themwith mandatory requirements. Final ~ y...}}