ML20207D976

Informs That Commission Approved Early Release of Rulemaking Language for Part 70 Once Language Approved by EDO

ML20207D976
Person / Time
Issue date:	05/18/1999
From:	Hart K NRC
To:	Martin D, Raddatz C NRC
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ML20207D973	List: ML20207D966 ML20207D976
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NUDOCS 9906040125
Download: ML20207D976 (61)

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{{#Wiki_filter:- l Barry "Oadelsohn - Part 70 Proposed Rule Language Paggil c; =~ t

• From:

Ken Hart - To: Charleen Raddatz, Dan Martin Date: Tue, May 18,199911:27 AM

Subject:

Part 70 Proposed Rule Language The Commission has approved early release of this rulemaking language once it is approved by the EDO. At that point, it should be posted on the Web with a notice that it is the staffs proposal and yet to be reviewed by the Commission.

Thanks, Ken 9906040125 990602 PDR ORG NE ED PDR x.

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[7590-01-P] NUCLEAR REGULATORY COMMISSION 10 CFR Part 70 RIN 3150 - AF22 Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material AGENCY: Nuclear Regulatory Commission. f4,,4/ 1i1% ACTION: Proposed rule. y + -my ,to amend its $

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) is regulations governing the domestic licensing of special nuclear matbed (8thA) for licensees authorized to possess a critical mass of SNM, that are engaged ir(one of the fo% Bowing pyy ~ c activities: enriched uranium processing; fabrication of uranium fuel or fuel as. + g' uranium sy pq ww; enrichment; enriched uranium hexafluoride conversion;p sing; f tion of mixed-oxide fuel or fuel assemblies; scrap recovery o[/ pecial nucteer meterial; or any other o

g p ' M activity involving a critical mass of SNM that the Commissionddterm%;pbid significantly ines co affect public health and safety or the environrnald.hhe pro bsed am ents would identify jmr* m M appropriate consequence criteria and the level lof proteollon needed;to prevent or mitigate accidents that exceed these criteria; req 0 ire ected I rform an integrated safety A A M3 E.s,# analysis (ISA) to identify %y ~ potential accidents'at.the facility:and the ite .g myys necessary to prevent theos potential acektents and/or mitigate their consequences; require the NagA v._, A implementation of measures to ensure that the'ltems relied on for safety are available and .A gy reliable to perform their fuI& ' including a summa ;A cIlon when"needed; require the inclusion of the ISA,Nva.4with the Econse application; and allow for licensees to make y nu certain changes,40' heir safety prograhYpd facilities without prior NRC approval. j$ $[$a] Ef s ..[ fk by ATTACHMENT 1 $siw+1 < MM i 2 )

e DATES: The comment period expires (insert 75 days after publication in the Federal Reaister). Comments received after this date will be considered if it is practical to do so, but, the Commission is able to ensure consideration only for comments received on or before this date. ADDRESSES: Submit comments to: Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC,20555-0001, Attention: Rulemakings and Adjudications Staff. Deliver comments to: 11555 Rockville Pike, Rockville, Maryland, between 7:30 a.m. and 4:15 p.m. on Federal workdays. You may also provide comments via NRC's interactive rulemakir}g website: "the sm W NRC home page (http /www.nrc. gov). From the home page, select "Sdemaking" tool 1 k bar at the bottom of the page. The interactive rulemaking website then be a s selecting "Rulemaking Forum." This site provides the ability to uploid.comrne/ynts as files ( ~ ~ v% p W,; format), if your web browser supports that function. For 'nformatiok he interactive Nw rulemaking website, contact Ms. Carol Gallagher by telephone at (30j)heg 154005 or e-mail { cag @ nrc. gov. A% %$V i FOR FURTHER INFORMATION, CONTACT: Theodore 8;8bN,'Oh of Nuclear Material a y ug as*DCf,20555-0001, Safety and Safeguards, U.S. Nuclear Regulatory Cornmission, ashington, telephone (301) 415-7260; e-mail tss @nrc. gov. r /%. /a SUPPLEMENTARY INFORMATION: 5 l.

Background

hf g h 4 e;n Action,ky Oh ll. Description Qth , ' ' g( '[G% [kbh %._ s ty incident atkhMi hed fuel fabrication facility in May 1991 prompted A near-crit NRC to review afety regulations rNnsees that possess and process large quantities of SNM. [See G-1324, " Prop Method for Regulating Major Materials Licensees" (U.S. ,a Nuclear Re Commis ' 992) for additional details on the review.) As a result of this review, the Jadke staff recognized the need for revision of the regulatory base for WwnMy these licensee'speepeelaff for those possessing a critical mass of SNM. Further, the NRC staff 3

1 i l concluded that to increase confidence in the margin of safety at a facility possessing this type and amount of material, a licensee should perform an ISA. An ISA is a systematic analysis that identifies: (1) Plant and external hazards and their potential for initiating accident sequences; j (2) The potential accident sequences, their likelihood, and consequences; and (3) The structures, systems, equipment, components, and activities of personnel relied on to prevent or mitigate potential accidents at a facility. 1 NRC held public meetings with the nuclear industry on this issue during Mayand ( November 1995. The Nuclear Energy institute (NEI) explained, to the Commissiok 's g emw position on the need for revision of NRC regulations, in 10 CFR Part 7)0'sfa a y spy wm meeting, and in a subsequent filing of a Petition for Rulemaking (PRM-70-7) in SeptembarqA;m% tp p yn 1996. NRC published in the Federal Reaister a notice of receipt of tNe PRM and p W%g KQ public comments on August 21,1996 (61 FR 60057). The PRM requested that NRC amend N$dbh Part 70 to: g a$ (1) Add a definition for a uranium processing and fuel fabfiE tion piant;[$ y .3 1 (2) Require the performance of an ISA, or acceptable aliirriative, at uhunluinf processing, fuel fabrication, and enrichment plants; an ( ) .Y, (3) Include a requirement for backfit analysis,'under certain circumstances, within Part 70. &a A in SECY-97-137, dated June 30,19 lthe staff proposed a resolution to the NEl PRM Iwith rulemaking. The staff's and recommended that the Commission t the sta p recommended approhhdemakk the bahEiI ments of the PRM, with some modification. In brief, hsed sas to revise Part 70 to include the following major elements: kkhb gk (1) Perforrnance of a formal $8A would form the basis for a licensee's safety // WW \\ program. This/equirement would applyto alllicensed facilities or activities, subject to NRC ./%M M regulation, thatare authorized to possess SNM in quantities sufficient to constitute a potential p;m pp for nuclear celticality (except ower reactors and the gaseous diffusion plants regulated under 10 CFR Part;76)[M g-4

- (2) Establishment of criteria to identify the adverse consequences that licensees must protect against; (3) inclusion of the safety bases in a license application (i.e., the identification of the potential accidents, the items relied on for safety to prevent these accidents and/or mitigate their consequences, and the measures needed to ensure the availability and reliability of these items); (4) Ability of licensees, based on the results of an ISA, to make certain changes without f NRC prior approval; and = ' (5) Consideration by the Commission, after licensees' initial con' duct and implementation of the ISA, of a qualitative backfitting mechanism to enhance re / j stability. Mk 5' 's. /f

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M In an SRM dated August 22,1997,-the Commission "... ap d th 's pro fh M revise Part 70" and directed the NRC staff to "... submit a draft p ...by July 31, N 1998." A draft proposed rule was provided to the Commissio

• Y-98-1 ed Rulemaking - Revised Requirements for the Domestic L ial Nucl terial,"

dated July 30,1998. The draft proposed rule reflect e ap ed in SECY-97-137. In particular, the safety basis for a. ility, i ing i suits, would be borpor submitted as part of an application to NRC, f ,ew, a in the license. Also in SECY 98-185, the staff recommended that i litativ kfit m nism should be considered for implem ion only a fetyb the results of the ISA,is established and inc n the after es and staff have gained experience with the i n of t' irement. In response to om ion issued an SRM dated December 1,1998, ~ hich. directed th not to publ J ~ w proposed rule for public comment. Instead, the j I eholder input and revise the draft proposed rule. In Commission dir d the staff to obt that SRM, th mission also di ted the staff to: (1) ^[ mental for NRC's regulatory purposes for inclusion as part of the license or t can be Justified from a public health and safety and cost-benefit i 5 s L

1 basis, and assure that Part 70 captures for submittal those few significant changes that currently would require license amendments; (2) Require licensees / applicants to address baseline design criteria and develop a preliminary ISA for new processes and new facilities; (3) Justify, on a health and safety or cost-benefit basis, any requirement to conduct a decommissioning ISA; (4) Require that any new backfit pass a cost-benefit test, without the " substantial" increase in safety test; (5) Require the reporting of certain significant events because of their potential to in worker or public health and safety; f)( (6) Clarify the basis for use of chemical safety and chemical consequence' h ,M particularly within the context of the Memoranda of Understanding he Occ and Health Administration (OSHA) and other government agencie $I (7) Critically review the Standard Review Plan (SRP) to en [ hintfly providing spe acceptance criteria, it does not inadvertently prevent licensees or a y N suggesting alternate means of demonstra*ing compliance with the rule; and [ (8) Request input on how applicable ISA methodologies should be em inthe j licensing of new technologies for use within new or existing [ N5! [4 As directed in the SRM, stakeholder input was solicited and obtalped at public meetings AM ff R held in December 1998 and January and March A 999. A website was; established to facilitate communication with stakeholders and to solik[r her i . The ear industry submitted comments by letters andpostings on the white. ThikshNiiiidh$ posed rule incorporates gn ~ A we ; y much of the DecemberX;np;:1998 SRMMreedl0S and reflects tsnguage responsive to many of the comments received.yr;' 4It pithat moitM9te major concerns with the earlier draft gn:ug rule have been resolved. kkk, hk gu j

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11. Description of Proposed Action The proposed rule grants the NEl September 1996 PRM in part and modifies the i

petitioner's proposal as indicated in the following discussion. { i The Commission is proposing to modify Part 70 to provide increased confidence in the margin of safety at certain facilities authorized to process a critical mass of SNM. The Commission believes that this objective can be best accomplished through a risk-informed and performance-based regulatory approach that includes: (1) The identification of appropriate risk levels, considering consequence criteria and tk level of protection needed to prevent accidents that could exceed such c iteria; ( j (2) The performance of an ISA to identify potential accidents at 5djacilitp$nd 9ditems relied on for safety; [ [h j (3) The implementation of measures to ensure that the iterns t.elied onk safety areby@;7/ available and reliable to perform their function when needed; Ty b (4) The inclusion of the safety bases, including the ISA summarylln"the license application; and (5) The allowance for licensees to make certain changes llo their safety peo0 ram and gfb h [ facilities without prior NRC approval. The Commission's approach agrees in princi with the NEM However, in contrast to the petition's suggestion that the ISA requirement limit hNf",Nranium A N N processing, fuel fabrication, and uranium enrichme%nt plantildensees, require the performance of an ISA for a broadef range of Part 70 l My a nsees that are authorized W f:;mb Af to possess a critical mass of SNM. The'Part.70 licensees that would be affected include licensees engaged in'one of the follo#GwAwing'ac6vities: enriched uranium processing; AA W y n.p ' A A I uranium fuel or fuel assemblies! uranium enrich 6ent; enriched uranium hexafluoride wa gg4 4 conversion; plutonium prodessin0[ fabrication of mbed-oxide fuel or fuel assemblies; scrap recovery of special nuclea,r-meterial;Aor any othercactivity involving a critical mass of SNM that wy ,y y-m the Commission determines could 'significantly affect public health and safety. The proposed rule would not aphto licensees auh"thto possess SNM under 10 CFR Parts 50,60,72, iffb and 76. Furthennere, the Commission is not currently proposing, as suggested in the NEl B, A A/ petition, to i a backfit provision in Part 70. Based on the discussions at public meetings held on May 29,gfand March 23,1999, the purpose of the NEl-proposed backfit provision 7

is to ensure that NRC staff does not impose safety controls that are not necessary to satisfy the performance requirements of Part 70, unless a quantitative cost-benefit analysis justifies this action. The Cornmission believes that once the safety basis, including the ISA summary, is incorporated in the license application, and the NRC staff has gained sufficient experience with implementation of the ISA requirements, a qualitative backfit mechanism could be considered. Without a baseline determination of risk, as provided by the initial ISA process, it is not clear how a determir,ation of incremental risk, as needed for a backfit analysis, would be accomplished. Furthermore, although NEl previously stated that a quantitative backfit approach is currently feasible, it would appear that a quantitative determination of incrementa(risk wouh ta. Af require a Probabilistic Risk Assessment, to which the industry has been rongly Given the differences of opinion on this subject, the Commission requesdepublid'aomment on its intent to defer consideration of a qualitative backfit provision in P (0,and hk suggestions for backfit provisions that would specifically address fkcycle ba[Ilfit need)k the information that would be available to conduct the associated h dh The majority of the proposed modifications to Part 70 are fourdh a new Subpart H, jy % a ; A " Additional Requirements for Certain Licensees Authorized to Pospss a'CddeglMass of %:r. Mk Special Nuclear Material," that consists of 10 CFR 70.60 through70.74. Theespeeposed modifications to Part '70, discussed in detail below, are r $ hse con)ihein the margin of safety and are in general accordance with. ' approach'w g e Commission ny 4y aya in its SRMs of August 22,1997, and December, 998. Section 70.4 Definitions. 4 b x Definitions of theiollowing 12 teen 6Would be section to provide a clear ja m:nm v -y understanding of the of the new Subpart H: "AouW, "Available and reliable to perform mp'A was their function when neem %onfigurahm'menagement", " Critical mass of SNM, " Double contingency", " Hazardous 'produc f Ibensed materials", " Integrated safety analysis", " Integrated a ems relied on for safety", " Management ' measures", "Unaccakfk$le perforhIignes hencies", and " Worker." Npy }

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Section 70.14 Foreion militarv aircraf t. This paragraph reflects an administrative change to renumber the paragraph from' 70.13a. Section 70.17 Soecific exemotions. This paragraph reflects an administrative change to renumber the paragraph from 70.14. Section 70.50 Reoortina reauirements. Paragraph (c) would be reworded to include information to be itte g verbal or written reports to NRC. The new information derives from[ specifi Subpart H, such as sequence of events and whether the event wa luat the IS. the extent the new information is also applicable to licensees not s ubpart H, the information was added with no differentiation noted. The new inform uld only apply to Subpart H licensees is noted. N. Y, Section 70.60 Aoolicability.' This section lists the types of NRC license r appii "ts wh e subject to the new Part 70, Subpart H. The Commission ha , ed t, he new uirements should not apply to all licensees authorized to possess Ima of S nstead, the Commission has identified a subset hese licen ,-based . ssociated with operations at these facilities, sho 4 'ect to' uirem is change would exclude certain .w facilities (e.g., those a ly to or use SNM in sealed form for research and sq educational purposes) fr ' gpuire use of the relatively low level of risk at rt is i ded to ensure that the significant accidents these facilities. In ge t; that are possible i fabrication . nd the other listed facility types) have been 19 analyzed in advj , and that appr i controls or measures are established to ensure adequate pro lon of workers,' ic, and the environment. The requirements and provisions lib ?N 'A worker, making, is defined as an individual whose assigned duties in the course of employment inv diation and/or radioactive material from licensed and unlicensed sources of radiation (i.e., an in l is subject to an occupational dose as in 10 CFR 20.1003). 9

in Subpart H are in addition to, and not a substitute for, other applicable requirements, including those of the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Labor, OSHA. The requirements being added by NRC only apply to NRC's areas of responsibility (radiological safety and chemical safety directly related to licensed radioactive material). In this regard, the requirements for hazards and accident analyses that NRC is adding are intended to complement and be consistent with the parallel OSHA and EPA regulations. The regulation states that Subpart H does not apply to decommissioning activities. NRC notes that the existing regulation [670.38(g)(4)(iii)] requires an approved decommissioning plan (DP) that includes " a description of methods used to ensure protection of work g environment against radiation hazards during decommissioning." Because the DPheued for NRC approval before initiation of "... procedures and activities nec to cdh decommicsioning of the site or separate building or outdoor area," t $P will contbMNibe vehicle for regulatory approval of the licensee's practices for proteh of headi nd s f h? during decommissioning. The ISA should provide valuable informahhfIspect to mN developing the DP and the use of the ISA in this manner is encoura 'gg if Section 70.61 Performance Reauirements. in the past, the regulation of licensees authoriahl to posOM undIr 10 CFR Parts 20 and 70, has concentrated on radiation protectio (yfor persor)ts invo%Apyl0ed) nuclear activiti nts to h 70 would explicitly conducted under normal operations. The pro e amen address potential exposures to workers or n$ s of public a er vironmental releases as a result of accidentsg art 20 contirdes eNRCh r protection of workers and Ap,n ymy public frorn radiatior) 'hormal operations, anticipated upsets (e.g., minor process upsets i that are likely to occurb e tim)h he life of the facility), and accidents. Although NNh regulathiNNirf rt 20 also be observed to the extent it is the Commission's int 9 m Oh,. %f practicable during an erne,rgeney{;wiehderd for all p thhot the Commission's intent that the Part 20 jy ph requirements apply;es the design s pf v7 of the likelihood of those accidents, cause accidents are unanticipated events that usually /N occur over a relatively short perio time, the Part 70 changes seek to assure adequate y9 At protection of Weltiers, members of the public, and the environment by limiting the risk p % D' (combined li

end consequence) of such accidents.

L 10

m There are three risk-informed performance requirements for the rule, each of which is set out in 10 CFR 70.61: (1) section 70.61(b) states that high-consequence events must meet a likelihood standard of highly unlikely; (2) section 70.61(c) requires that intermediate-consequence events must meet a likelihood standard of unlikely; and (3) section 70.61(d) requires that risk of nuclear criticality be limited by assuring that all processes must remain subcritical under any normal or credible abnormal conditions. The term " performance requirements" thus considers together consequences and likelihood. For regulatory purposes, each performance requirement is considered an equivalent level of risk. For example, the acceptable likelihood of intermediate-consequence events is allowed to be greater than th acceptable likelihood for high-consequence events. /- A .~ _, A risk-informed approach must consider not only the consequences of p i l accidents, but also their likelihood of occurrence. As mentioned ab he perf hk

• requirements rely on the terms "unlikely" and " highly unlikely" to focus on the ddiof acci J

However, the Commission has decided not to include quantitative unlikely" and " highly unlikely"in the proposed rule, because a single definition for k hat would apply

• ei-k - 4g..Au to all the facilities regulated by Part 70, may not be appropriate.

endingen,, type of s@ m facility and its complexity, the number of potential accidents and their consequehoeseould differ ~ markedly. Therefore, to ensure that the overall facility riskh boINnts is acc e for I y

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different types of facilities, the rule requires applicant - devel,br $N val (see A ym 970.65), the meaning of "unlikely" and " highly unlikdy" spec ~. to their,pses and facility. To accommodate this development, the Com fon belie that thehSRP is the appropriate y M document to include guidelines for licensees' tse. A "St d Review Plan for the Review of a License Application for a le Faci y p y developed. The draft SRP m mys vmy provides one accep oach meaning of y" and " highly unlikely" that can m' ue be applied to existing fuel lecilitie ?A The general appr hehtying ' rformance requirements is that, at the e time of licensing, each hnI$ikhchem hablectrical, industrial) that can potentially affect radiological s isidentifikh ated, in an ISA, by the licensee. The impact of y %g accidents, bot rnal and externaha9pciated with these hazards is compared with the three performance irements. Any(NaNall) structures, systems, components, or hum i for mitigating (reducing the consequence of) or preventing for which cr (reducing thh. taken in th of)tlie accident such that all three performance requirements are satisfied, mu kdge9hed s an " item relied on for safety." " Items relied on for safety"is a 11

term that is defined in 10 CFR 70.4, and in this approach, the applicant has a great deal of flexibility in selecting and identifying the actual " items." For example, they can be defined at the systems-level, component-level, or sub-component level. " Management measures" [see discussion in 10 CFR 70.62(d)] are applied to each item in a graded fashion to ensure that it will perform its safety function when needed. The combination of the set of " items relied on for safety" and the " management measures" applied to each item will determine the extent of the licensee's programmatic and design requirements, consistent with the facility risk, and will ensure that at any given time, the facility risk is maintained safe and protected from accidents (viz., satisfies the performance requirements). The proposed performance requirements also address certain chemical . t. result from the processing of licensed nuclear material. The questio ext authority to regulate chemical hazards at its fuel cycle facilities wa d after as - 1986 at a Part 40 licensed facility, in which a cylinder of uranium h luori tured a resulted in a worker fatality. The cause of the worker's death was ion of hydrogen fluoride gas, which was produced from the chemical reaction of uran. ride and water (humidity in air).' Partly as a result of the coordinated Federal re, nsehl Congressional investigation into that accident, NRC and the QSH$ entered i ,,in 1988, that clarified the agencies' interpretations of their Insibilitiikthe regulation of chemical hazards at nuclear facilities. MOU wing four areas of responsibility. Generally, NRC covers the tthre

eas,

~ OSHA covers the fourth area: (1) Radiation risk produced by r ctive ials; (2) . Chemica produce oactive -(3) Plan that fety ive materials; and ] (4)- Plant . t res pational risk, but do not affect the safety of {}W2 licensed ra terial M One goal of the ' merkin 970.61 is to be consistent with the NRC-OSHA MOU. The e, the perfo irements in 970.61 include explicit standards for the MOU's first areas of respon in addition, the third MOU area of responsibility is specifically e led by licensee r the ISA requirements of 970.62(c)(1)(iii). As an example of MOU are e failure of a chemical system adjacent to a nuclear system could affect lear system such that the radiation dose (and associated likelihood of xceeded a performance requirement, the chemical system failure 12

f } I l j would be within the scope of the ISA and the means to prevent the chemical system failure from 1 impacting the nuclear system would be within NRC's regulatory purview. OSHA provided comments, by a letter dated February 1,1999, on a draft of the rule that had been revised to be consistent with the MOU. In that letter, OSHA expressed concerns that the rule language would preempt OSHA from enforcing any of its standards, rules or other requirements with respect to chemical hazards at the facilities covered by the NRC draft rule. This concern is based on case law under the OSH Act. The pertinent provision in the OSH Act states: "(b)(1) Nothing in this chapter shall apply to working conditions of employees with respect to which other Federal agencies, and r M. State agencies acting under section 2021 of title 42, exercise h j g}p 1, Ag statutory authority to prescribe os enforce standards or regulations affecting occupational safety or health."[29 U.S.C29653(b )f YQD. 1 4 .x NRC staff subsequently met with OSHA officials on FebruaryA,-a.1999?and some j clarifications and further information were provided at that meeting /gy vAs d'feeult chhe meeting j s. \\ ww i discussions, some changes were made to the rule language {o fnote clearly sp%ywecify the scop gweq of NRC involvement. However, these changes do not fuhfesche tes basic preapption issue. j The problems identified with the rule are not unique, i.nkhe pr $sueis generic and may already exist for any NRC-licensed facilities wdr there 5 req to analyze hazards. At the February 25 meeting, OSHA oAfIrbed thsNhe rule guage is consistent NY h M with the October 21,1988 MOU; indicated theithey have vio suggested changes to the MOU; N nn and indicated that they are not oppor edh the proposed rulili"n/y'e Commission's view is that Th ps nt% g;f the proposed rule is coneleternt with NRC responsibilities arkf authority under the Atomic Energy y ywr va Act, and consistent with,the OSHA MOusThe_ only resolution of the preemption issue appears q to be a legislative modifi thir$ OSH kotki b comments would be appreciated on any p ma options that may have,b,een overlooked.h m .m Within eacg rformance r , NRC recognizes that the proposed radiological standards are more restrictive, in termsdacute health effects to workers or the public, than the f 11 ~ chemical sta s for a given consequence (high or intermediate) and that this is consistent 4 a 4 with current ory practice Tfie choice of each criterion is discussed below in a paragraph-by1pragraphpescussion of 970.61. hsh 13

a, The use of any of the performance requirements is not intended to imply that the specified worker or public radiation dose or chemical exposure constitutes an acceptable criterion for an emergency dose to a worker or the public. Rather, these values have been proposed in this section as a reference value, to be used by licensees in the ISA (a forward-looking analysis) to establish controls (i.e., items relied on for safety and associated management measures) necessary to protect workers from potential accidents with low or exceedingly low probabilities of occurrence ihat are not expected to occur during the operating life of the facility. Section 70.61(b). This section addresses performance requirements for consequence events. The consequences identified in $70.61(b) of the proposed r re referr consequence events" and include accidental exposure of a worker in Iiocate trg outside of the controlled area to high levels of radiation or hazardo, Is. These accidents, if they occurred, would represent radiation doses to a wo dividuallocated j outside of the controlled area at levels with clinically observable r-concentrations of hazardous chemicals produced from licen rial at or life-threatening injury could occur. The goal is to ensure an of risk iting the combination of the likelihoo' of occurrence and the ied s, high-d consequence events must be sufficiently mitigated a low nseq j prevented such that the event is highly unlikely (or lower). T licati f " items, ed on for safety" provides this prevention or mitigation fuacti j; f of a walilhhi$ a radiation dose of 1 Sv (100 ac 1 rem) or greater total _ e equ E) is considered to be a high-consequence event. ' According to the N$ pcil on Protection and Measurements (NCRP, 1971)', life-saving acti .. sea for and removal of injured persons, or entry to prevent conditi t would pr e numbers of people"-- should be undertaken only when'the " nned dose to thi ( body shall not exceed 100 rems." This is consistent ' 'terlNCRP posit l (NCRP,1987) on emergency occupational exposures, that states'" the exposu ay approach or exceed 1 Gy (100 rad) of low-LET [ linear energy tran equivalent high LET exposure) to a large portion of the body, in j l a short time, s to understand not only the potential for acute effects but he or I 14 (

r (; i she should also have an appreciation of the substantial increase in his or her lifetime risk of cancer." Section 70.61(b)(2). The exposure of an individuallocated outside of the controlled area to a radiation dose of 0.25 Sv (25 rem) or greater TEDE is considered a high-consequence event. This is generally consistent with the criterion established in 10 CFR 100.11, " Determination of exclusion area, low population zone, and population center

distance," and 10 CFR 50.34, " Contents of applications; technical information," where a whole-body' dose of 0.25 Sv (25 rem) is used to determine the dimensions of the exclusio area ar low-population zone required for siting nuclear power reactors, gk h

Section 70.61(b)(3).- The intake of 30 mg of soluble uraniu ~an indivih outside of the controlled area is considered a high-consequence .t. Thishoice, 5 { l based on a review of the available literature [ Pacific Northwest La PNL),1994), is consistent with the selection of 30 mg of uranium as a criterion that .ed during the Part 76 rulemaking," Certification of Gaseous Diffusion Plants." I rtic I rule that we ~ ; { established Part 76 (59 FR 48944; September 23,1994) stat "The N ider whether the potential consequences of a reasonable sp ed ace, t scenarios exceed... uranium intakes of 30 milligrams...." The fi . le al , Commission's intended use of chemical toxicity considerations in. 76 is. nsist s practice A ~ j: elsewhere [e.g.,10 CFR 20.1201(e)], and pr , ny,. tial reg. tory gap in public protection against toxic effects of soluble ura exposumNe azardous chemicals produced Section 70-n ac l' from licensed materia ations' .1) could cause death or life-threatening injuries to a worker; or (2) Irrev Ith effects to an individual located outside i of the controlled area, -con uence event. Chemical consequence criteria corresponding to pated adver . ects to humans from acute exposures (i.e., a single exposure multiple exposur rring within a short time -- 24 hours or less) have i a x i been develo r are under dev ment, by a number of organizations. Of particular interest, the al Advisory mittee for Acute Guideline Levels for Hazardous Substances l is developin j> ]8uideline Limits (AEGLs) that will eventually cover approximately 400 industrial 4 ]W, mg'd pesticides. The committee, which works under the auspices o1 i 15 l

g 1 the EPA and the National Academy of Sciences, has identified a priority list of approximately 85 chemicals. Consequence criteria for 12 of these have currently been developed and criteria for approximately 30 additional chemicals per year are expected. Another set of chemical consequence criteria, the Emergency Response Planning Guidelines (ERPGs), has been developed by the American Industrial Hygiene Associtation to provide estimates of concentration

ranges where defined adverse health effects might be observed because of short exposures to hazardous chemicals.' ERPG criteria are widely used by those involved in assessing or responding to the release of hazardous chemicals including "... community emergency planners and response specialists, air dispersion modelers, industrial process safety engine

' implementers of environmental regulations such as the Superfund Amendment a Reauthorization Act, industrial hygienists, and toxicologists, transport fe gfire protection specialists, and government agencies...." (DOE Risk Y- + ment O' Despite their general acceptance, there are currently only approxi y 80 crite)!N available, and some chemicals of importance (e.g., nitric acid) are ' The qualitative language in the performance requirement all icant/ licensee to propose and adopt an appropriate standard, which may be an ndard, or where there is no AEGL or ERPG value available, the appii vel criterion that is comparable in severity to those that hav for o emicals. For example, for the worker performance requireme . istin used by licensees to define appropriate concentration leve satist e pe requirement are j . the AEGL-3 and ERPG-3. AEGL-3 is defined , e air ne con cation (expressed in 8 ppm or mg/m ) of a substance at or above it is p cted t e general population, including susceptible, xduding h

ceptible, ould expvience life-hd as ximum airborne concentration j

~ hreatening effects RP t

below which it is belie ly all ould be exposed for up to 1 hour without experiencing or developin ing Hb s." Similarly, for the public, AEGL-2 is 8

defined as "The airbo res in opm or mg/m ) of a substance at or. above which it is p ted that th ulation, including susceptible, but excluding hypersusceptib ' f ividuals, could nce irreversible or other serious, long-lasting effects orimpaired _ to escape," and PG-2 is defined as "The maximum airborne concentration below which lieved that r1 d individuals could be exposed for up to 1 hour without experiencin rersible or other health effects or symptoms that could impair an I individual's a otective action." 16 i 4

L l. j Section 70.61(c). This secti$n addresses performance requirements for intermediate-L consequence events. The' consequences identified in $70.61(c) of the proposed rule are referred to as ' l " intermediate-consequence events" and include accidental exposure of a worker or an t: . 1 Individual outside of the controlled area to levels of radiation or hazardous chemicals that generally correspond to permanent injury to a worker, transient injury to a non-worker, or - significant releases of radioactive material to the environment. The goal is to ensure an. acceptable level of risk by limiting the combination of the likelihood of occurrence and the. . identified consequences. Thus, " intermediate-consequence events" must be suffici ntly mitigated to a lower consequence or prevented such that the event is unlikely (or e . application of " items relied on for safety" provides this prevention or f + Section 70.61(c)(1). A worker radiation dose between 0.2 25 r nd 1 S rem) TEDE is considered an intermediate-consequence event [ov rem)is a big consequence event). This value was chosen because of the use of rem) as a criterion in existing NRC regulations. For example, in 10 CFR 20 of incidents," immediate notification is required of a licensee if ' ual r otal - effective dose equivalent of 0.25 Sv (25 rem) or more." 20.1 lanned . special exposures," a licensee may authorize an adut rker t excess'of normal occupational exposure limits if a dose of th agnit does ed 5 times the annual dose limits (i.e.,0.25 Sv (25 rem)] duri , indivi 's lifetirt in addition, EPA's Protective Action Guides (U.S. Environmen ecti en

2) and NRC's regulatory guidance (Regulatory 8.29,19 ify 0.25 s the whole-body dose limit to workers for life-s ps an of larg ations. NCRP has also stated that a TEDE of 0.25 porres once-in-a-lifetime accidental er emergency dose for workers.

y_ j Section 70.Sfe)(2).,A dos , ividuallocated outside of the controlled area between 0.05 rem) and 0.25 em) is considered an intermediate-consequence event. ' NRC sed a 0.05-Sv ( rh) exposure criterion in a number of its existing regulations. mple,10 C 2.106, " Controlled area of an ISFSI or MRS," states that ' "Any individ nd the nearest boundary of the controlled area shall not receive a do 5 rem to the whole body or any organ from any design basis i 17 4 L

accident." In addition, in the regulation of the above-ground portion of the geologic repository, 10 CFR 60.136, states that "...for [ accidents), no individual located on or beyond any point on the boundary of the preclosure controlled area will receive...a total effective dose equivalent of 5 rem...." A TEDE of 0.05 Sv (5 rem) is also the upper limit of EPA's Protective Action Guides of between 0.01 to 0.05 SV (1 to 5 rem) for emergency evacuation of members of the public in the event of an accidental release that could result in inhalation, ingestion, or absorption of radioactive materials. Section 70.61(c)(3). The release of radioactive material to the environment putside restricted area in concentrations that, if averaged over a period of 24 hours, exceesSB00 the values specified in Table 2 of Appendix B to Part 20, is consi ya jw consequence event. In contrast to the other consequences criteria $wtht directlygetiMt wohy ars f., and members of the public, the intent of this criterion is to ensure etion e envi S from the occurrence of accidents at certain facilities authorized to g(?A // gr mass quantities of SNM. This implements NRC's respo he environment, in accordance with the Atomic Energy Act of 1954, et seo,, and thehht mental Policy Act of 1969, et seq. The value established for the environmental cons n is i i to the D "'u A NRC Abnormal Occurrence (AO) criterion that addresses the sal of ty w radioactive material from its intended place of confinement tion 2Sn Energy A4 "? N Reorganization Act of 1974, as amended, requtes'? y y ' to Congress annually). that A be rep In particular, AO reporting criterion 1.B.1 re ed areahs the rgegrting of,Iions which event that involves "...the release of radioactive material to an urdet a period of 24 hourseh5000 tf blues sp\\cIlledYn Table 2 of Appendix B to 10 %w ag* wa CFR Part 20, unless the has demengtsbed compliance with 10 CFR 20.1301 using 10 m.A %4 Debember 19,1996; 61 FB 67072]. The CFR 20.1302(b)(1) or 10 CPSMt3Q2(b)(2)(lF(Nrt 20 apply to concentrationslisted ig; gBfd and water effluents,0s unrestricted gp established these concentrations based on an implicit effective se equivalent limh6f8.5 mSv/yr (50 mrem /yr) for each medium, assuming continuously expked to the listed concentrations present in an unrestricted an individual area for a yew % / ifan ' wesegekiinuously exposed for 1 day to concentrations of radioactive material 500h han the values listed in Appendix B to Part 20, the projected dose 18

r ) l would be about 6.8 mSv (680 mrem), or 5000 x 0.5 mSv/yr x 1 day x 1 yr/365 days. In addition, a release of radioactive material, from a facility, resulting in these concentrations, would be expected to cause some environmental contamination in the area affected by the release. This contamination would pose a longer-term hazard to the environment and members of the public untilit was properly remediated. Depending on the extent of environmental contamination caused by such a release, the contamination could require considerable licensee resources to remediate. For these reasons, NRC considered the existing AO reporting criterion for discharge or dispersal of radioactive material as an appropriate consequence criterion in this gff rulemaking. Section 70.61(c)(4). An acute chemical exposure to hazardous chemicals produced A /3 ~< from licensed material at concentrations that either; a) to a worker, coul$cause Irraw;oeralds. health effects (but at concentrations below those which could causeb or ps a

a effects); or b) to an individual located outside of the controlled area,'could cause;notabi ME discomfort (but at concentrations below those which could cause i effects), is considered an intermediate-consequence event. Chemical consequ%c %enoscrReria corresponding to anticipated adverse health effects to humans from acute expos

([hkal exposure V'~m or multiple exposures occurring within a short time -- 24 hours or)ms) have beertde.hveloped, or are under development, by a number of organizations. Ot pseticular% Merest, two existing pm Nwy standards, AEGL-2 and ERPG-2, can be used to defind corm halkrirreversible py fy %Df ' health effects, and two existing standards, AEGL-1/end ER G21, can be.nped to define the W 7 concentration level for notable discomfort. Th alitative nguage,the performance i N r requirement allows the applicant / licensee to, pt and propose ameppropriate standard, which A W may be an AEGL or ERPG standard, or%ere there is n. vh AEGLo?r ERPG value available, tj o applicant may develE ht a e th is comhb i severity to those that have m been established for othiri heMucals. Chh c %Qf:Q "^t 2?Y& kh) Section 70.61(dtyTNs seM addresses performance requirements for an accidental w g%gp nuclear criticality, The third. riormance requirernent states that the risk of nuclear criticality accidents must be limited assuring that ur r normal and credible abnormal conditions, all nuclear to M processes apm6 critical, including use of an approved margin of subcriticality for safety. It ya gy also requiresStatpreverhe'pntrols and measures shall be the primary means of protection against nucle $ a$idents. Although detecting and mitigating the consequences of a 19

y lN nuclear criticality are important objectives (e.g., for establishing alarm systems), the prevention of a criticality is a primary NRC objective. The basis for this provision is the NRC strategic plan (NUREG-1614, Vol.1), which, for ) nuclear materials safety, states NRC's performance goal of "...no accidental criticality involving I licensed material." The language chosen for this performance requirement closely follows the language of the applicable industry standard, ANSI /ANS Standard 8.1-1983," Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors." Section 70.61(e). This section addresses items relied on for safety and management 5 measures. Paragraph 70.61(e) would require that each engineered or adr rativ $r control system that is needed to meet the performance requirementd designahd relied on for safety. This means that any control or control system $pqlet is negessary to rna$y xy wo l i the acceptable combination of consequence and likelihood for an designated an ltem relied on for safety. The importance of this section is that, once %glWehis designated as W w an item relied on for safety, it falls into the envelope of the safety pfegramL by section c.,, g 70.62.' For example, records will be kept regarding the item, andmanagemettatoespes such as the configuration control program are applied to the iteeht es that hb the item, - to ensure that the item will be available and reliable to ormih$hedidshen eeded. AV WH The failure of an item relied on for safety d A"not necessarily1negethat an accident will occur which will cause one of the consequencesiisbd in perform e requirements to be M 7 (r dant or dverse) control systems that exceeded, Some control systems may have rallel would continue to preverAthe accident 8hh eed for defhb-in-depth and single-failure hiessed odh ty and bof the potential accident. In resistance would id weA y, other cases, the failuredan%. +may meshNw tut the particular accident sequence is no longer w wnys " highly unlikely", or "unlikelygJganise cases?teggerformance requirement is not met, and the e hs Wr expectation would be that,s geasure would exist (possibly in the form of an operating procedur at ensured docility would not operate in a condition that exceeds the performancehquirement. For eenhe, a facility that relies on emergency power could not operate for arnhended time in thehsence of an emergency power source even if grid power manner, the.ibs relied on for safety and the management measures is available. AV complement 4agnettre adequate protection from accidents at any given time. d 20

p E Section 70.61(fL This section u.Oresses the term " controlled area" used in the performance requirements. Section 70.61(f) requires licensees to identify a controlled area consistent with the use of that term in Part 20, and provides clarification regarding the activities that may occur inside the controlled area. The function of this term is to delimit an area over which the licensee exercises control of activities. Controlincludes the power to exclude ladividuals,if necessary. The size of the controlled area is not specified in the regulation because it will be dependent upon the particular activities that are conducted at the site and their relationship to the licensed activities. [Within the controlled area will be a restricted area (as defined in 620.1003), accessbk to which is controlled by the licensee for purposes of radiation safety.] k Individuals who do not receive an occupational dose (as that ter is usedhPart20)in y y mm the controlled area will be subject to the dose limits for members o public in 10 CPRV,$? s g W.y However, the Commission recognizes that certain licens$es may haNyongo 20.1301. activities at their site (i.e., within the controlled area) that are not r toh licensed activities. For example, a non-nuclear facility may be adjacent to thk ' facility but both are within the controlled area (which may be defined similar to the b h)Rhis raises a question regarding the appropriate accident standard for thessindhriduals. Pelec6c%'of the w n individuals at the non-nuclear facility must consider that jhofteture;ef many potenlist;y pc A yg accidents at a fuel cycle facility is such that there may not be sufffhkInt tinh b take action to exclude individuals from the controlled area. Therab, for hosehM$k accident evaluation, the rule explicitly contains two options br these. individuals s well as an implicit p3 r ff M third option). In the first option, the licensee _ evaluates,Jn the ISA,.the risk at its location (as opposed to that at any point at or beyon8 th ontrollehaktbary) and determines that it pm na w3w meets the performapos requirementsjormorsbers of the public. In the second option, j performance requirerbrNehrorkerskthlied to individuals in the controlled area if the provisions of Section 70$hb)h satisfiekhionditions ensure that the individuals are aw$nts at the nuclear facility and have received m aware of the risks to themf$v +Mpolintial acl: rdr5 Id y y ,+g appropriate training s6d access to 'information. This parallels and is consistent with the use of g u__y the term," Exclusion area", by 10 CFR Pads 50 and 100, which states, " Activities unrelated to pg JM operation of t,he peactor may be pegitted in an exclusion area under appropriate limitations, provided thabgehgnificant haza#6s to the public health and safety win result." The implied third option is to h controlled area such that within it only activdes associated W' sy with the licensedeuslaar facility are permitted. i ss 21

[. i' ) L The Commission's intent is that the ISA does not evaluate compliance with the accident ' standards for individuals who make infrequent _ visits to the controlled area and restricted area (e.g., visitors).1 Use of the ISA to determine the risks to these individuals would need to consider second-order effects such as the probability of the individual being present at the time ' that the unlikely (or highly unlikely) accident occurred. This level of detail is unnecessary to i accomplish the purpose of this rule (viz., to document and maintain the safety basis of the facility design and operations). Application of the Part 20 regulations provides adequate protection for these individuals. -in addition, the provisions (i.e., performance requirements) to protect workers and non-workers during accidents should, implicitly, provide a degr e of f protection to the infrequently present individuals. 4 E { j Section 70.62 Safety Proaram and Intearated Safety Analysis, j g j This paragraph addresses the safety program, that include .ocess y infor l 1 ISA, and management measures. The performance of an ISA, an lishment of { measures to ensure the availability and reliability of items relied on en needed, are the means by which licensees demonstrate an adequate level of ecti ilities. The ISA is a systematic analysis to identify plant and external andt Ifor initiating accident sequences; the potential accident seg ons " es; and ) the site, structures, systems, equipment, componen acti relied on for safety. As used here, " integrated" means joint co eratio and from, all relevant hazards, including radiological, criti fire, a emical e structure of the safety program recognizes the critical role t ISA s in i ying potential accidents ^ . and the items relied on safety. H , also re t the performance of the ISA, by itself, will not en te p

tead, ive management system is needed to ensure that.

lied re available and reliable to perform their function when needed. D freme part of the safety program are included in this section. 3 Section 1 2(a). Each lice . uld be required to establish and maintain a safety n y program that I sonstrates compi with the performance requirements of $70.61. Although the ISA wou [ e primary t identifying the potential accidents requiring consequence mitigation a n, process safety information would be used to develop the - ISA, and man 'sures would be used to ensure the availability and reliability of items j 22

e i, . relied on for safety identified through the ISA. The management measures may be graded

a'ccording to the risk importance associated with an item relied on for safety.

The licensee is also required to establish and maintain records demonstrating that it . has, and continues to meet,'the requirement of this section. These records serve two major purposes. First, they can supplement information that has been submitted as part of the license application. Second, records are often needed to demonstrate licensee compliance . with applicable regulations and license commitments. It is important, therefore, that an. appropriate system of recordkeeping be implemented to allow easy retrieval of required information, j Finally, each licensee would also be required to establish and maintain a i j ~ documenting each discovery that an item relied on for safety has fail rfo 1 l 6e

either in the context of the performance requirements of $70.61 or

, mand. I ~ e2 " in the context of the performance requirements of 970.61" mear 11at it lied on i safety that fail would require logging even if their failures did not re ss upsets or ' accidents but could have resulted in the accident conditions they are ggainst, had all conditions been optimum for the accident. This would not include lure" s, such as routine maintenance on an item, when the item or measure y docu not be available. The log must contain: (a) the identity of the it the function affectee'; (b) date of discovery of the failure; (c) durat time; unable to perform its function; (d) any other affected items r onfo ety safety function; (e) affected processes; (f) the cause of the f , g) wh _ r the fa ' e was in the context of . performance requirements, or on demand, th; an any ' tive or compensatory actions taken.~ The log uld be initi e time and updated promptly at the completion of each n of n item n for safety. The purpose of the - log is to assist NRC i who _ selied on for safety are, in fact, available and . reliable and in detecting s mst pact ISA evaluations. Section 70. ). This par Id require the licensee to maintain process-safety M information pe !p ng to the hazards materials used or produced in the process, the technology ! process, and the uipment in the process. NRC confidence in the margin of ' safety at its hd facilities s, in part, on the ability of licensees to maintain a set of - current, acc j records available for NRC inspection. The process-safety in support of development of an ISA. information s 23 m

m l 'Section 70.62(c). This paragraph proposes requirements for conducting an ISA. There . are four major steps in performing an ISA: (1) Identify all hazards at the facility, including both radiological and non-radiological. hazards. ' Hazardous materials, their location, and quantities, should be identified, as well as all hazardous conditions, such as high temperature and high pressure. In addition, any interactions that could result in the generation of hazardous materials or conditions should be identified. (2) Analyze the hazards to identify'how they might result in potential accidents. These accidents could be caused by process deviations or other events internal to the pla, or b g M credible external events, including natural phenomena such as floods, earthquak accomplish the task of identifying potential accidents, the licensee n ens iled and accurate information about plant processes is maintained and ' personnel performing the ISA. I, avalla ? 1 (3) Determine the consequences of each accident that ha tified. For an i 4 ) accident with consequences at a "high" or " intermediate level," as d FR 70.61, the likelihood of such an accident must be shown to be commensura hl ences,as required in 10 CFR 70.61. I i L (4) Identify the items relied on for safety (i.e., t p relied. o prevent ' accidents or to mitigate their consequences, identifie heI f re needed to reduce the consequences or likelihood of the ace' ts to a ptableI[. YThe identification of items relied on for safety is required only f , ents cons [; nces at a high or intermediate level, as defined in 10 CFR'70 it is expected thi the license cant The ISA using a " team" of individuals with exp gine ocess ns related to the system being evaluated; the team s i pe erience in nuclear criticality safety, radiation safety, fire safety ical p ty, as warranted by the materials and

potential hazards ass ss g evaluated. At least one member of the ISA team should be'an

dual who ce and knowledge that is specific to the process

] being evaluate nally, at least on ual in the team must be knowledgeable in the l specific ISA logy being u Cur 70 licens r whom the rule applies, would be required to develop plans and s ~ thin 6 months of the effective date of the rule. Each plan would ident that would be subject to an ISA, the ISA approach that would be 24 ~ x

r kIb implemented for each process, and the schedule for completing the analysis of each process. Licensees would be expected to complete their ISA within'4 years of the effective date of the rule; correct any unacceptable vulnerabilities identified; and submit the results to NRC for, . approval in the form of an ISA summary that contains the information required by 10 CFR 70.65(b). Pending the correction of any unacceptable vulnerabilities, licensees would be expected to implement appropriate compensatory measures to ensure adequate protection until the vulnerability can be more appropriately corrected. Applicants for licenses to operate new facilities or new processes at existing facilities : L would be expected to design their fscilities or processes to protect against the occu rence ok ' _ the adverse consequences identified in _10 CFR 70.61, using the baseline design CFR 70.64(a). _ 'Before operation,' applicants would be ' expected to u eiri on as-built conditions and submit the results to NRC as ISA summarie ng with applications, following the requirements in 10 CFR 70.65(b). The Commission believes that sufficient flexibility is permitti . A methodolog j . chosen to be able to accommodate a wide range of technologies. H ure that - ' sufficient flexibility exists, the Commission is requesting commen t Section 70.62(d). _ Although the ISA would play entifyi ential accidents and the items relied on for safety, the perf nce t, by itself, ensure adequate protection. In addition, as woul provid for in 0.62(d), an effective management system would be nee , nsur at the it s relied on for safety are available and reliable to perform their functi en n . As d before, management V. measures may be gra to better im the res A. Managemen are riorme licensee,in general on a continuing basis, that ,to ite for safety. Management measures include: a) configuration managem tenan ing and qualifications; d) procedures; e) audits and assessme ati g) records management; and h) other quality assurance elemen V hangesin . ation of the facility need to be carefully controlled to ensure consi l[ cy among the faf sign and operational requirements, the physical configuration f the facility doc ntation. Maintenance measures must be in place to ensure the ity and relia of all hardware, identified as items relied on for safety, to i perform thei ~ ed. Training measures must be established to ensure that all personnel rel ty are appropriately trained to perform their safety functions. 25

Periodic audits and assessments of licensee safety programs must be performed to ensure that facility operations are conducted in compliance with NRC regulations and protect the worker and the public health and safety and the environment. When abnormal events occur, investigations of those events must be carried out to determine the root cause and identify ~ corrective actions to prevent their recurrence and to ensure that they do not lead to more I serious consequences. Finally, to demonstrate compliance with NRC regulations, records that ] document safety program activities must be maintained for the life of the facility. This section also would require that the safety program ensure that each item relied on ) for safety would perform its intended function when needed and in the context of th performance requirements of this section. The utility of the two modifyi requir en , needed," and "in the context of the performance requirements of this ,i follows: The phrase "when needed" is used to acknowledge that a cular ty contr not be continuously functioning. For example, it may not be opera g maintenance calibration testing, or may not be required when the process is not o er when special nuclear material is not present. However, the phrase, when nee d e a licensee from compliance with the performance requirements. For ex f a parti nent is out for maintenance, the licensee must consider credible s in d ing the ISA and identifying items relied on for safety - a high seque_' ce still has to be highly unlikely. Compliance with the performan requir ntsi ses can be established by various means including ident ion of ionalite relied on for safety (and j application of safety program management sures t em), limiting operations or placing the plant in a d nt operati 'during < nee of the item relied on for safety. To illustrate, a po one-week maintenance outage of the emergency diesel generat ied o would still be a credible event sequence. -If the loss of power, c erat inoperable status, could result in a combination of d likelihood s a performance requirement, then the licensee J would not be in pliance with the ance requirements of 670.61. A licensee cannot claim, after t l intenance, that e the power was not lost, the generator was available when ne L conceptis ~the ISA is used as a risk-informed, forward-look at the credible faci effects on plant systems and modes of operation. The rule j would require necessary to comply with the performance requirements be 26 j

identified as important to safety and placed under the safety program management controls. In-identifying each item, the ISA must consider various modes of operation and the likelihood that L a given safety control will be inoperable (e.g., because of being off-line for maintenance) during credible event sequences. The section would also require that the safety control perform its function "...in the context of the performance requirements of this section " This phrase indicates that the function of interest is the one credited in the ISA to meet certain consequence criteria with a certain frequency. Second, this phrase would require that additional safety controls be defined in cases where one control does not result in compliance with the performance requiremen has periods when it is inoperable. Using the loss of offsite power exam e again, would still be required to meet the risk-informed performance require of th an ' operable h' ' emergency diesel generator used as an item relied on for safety is service for maintenance. y WM p. SEtion 70.64 Reauirements for new facilities or new orocesses at eI '" les. This section deals with baseline design criteria for new faods o sses at existing facilities. A major feature of the proposed amendments to uirem at . licensees and applicants for a license perform an IS use . develop risk-informed decisions regarding facility safety. The I proce appli ting' designs to - identify risk insights on those areas that warra , itional eventive mitigative measures. hthe ISA before For new facilities, the proposed rule would r. e the rman a construction (see the existing $70.21 - 0.23(a) updating of the ISA before beginning operation pr facilities mmission recognizes that good engineering practice certbh gquirements be applied as design and safety considerations for ar1y lear p facility. In addition, a fundamental element of NRC's saf ' n'i. r designs and operations should provide for defense-in-depth protection inst accide . re, the Commission has specified baseline j design criteria 1 0.64 that are si . =use to the general design criteria in Part 50 Appendix . A; Part 72, S rt F; and 10 CFR. .131. The baseline design criteria identify 10 initial safety design cons ns,includin " quality standards and records; b) natural phenomena 27

m hazards; c) fire protection; d) environmental and dynamic effects"; f) chemical protection; g) emergency capability; h) utility services; i) inspection, testing, and maintenance; j) criticality control; and k) instrumentation and controls. The baseline design criteria do not provide relief from compliance with the safety performance requirements of 670.61. The baseline design criteria are generally an acceptable set of initial design safety considerations, which may not be sufficient to ensure adequate safety for all new processes and facilities. The ISA process is intended to identify additional safety features that may be needed. On the other hand, the Commission recognizes that there may be processes or facilities for which some of the baseline design criteria may not be necessary or appropriate, based on the results of the ISA. For t processes and facilities, any design features that are inconsistent with the baseline /'desi A 7 criteria should be identified and justified. k Using the baseline design criteria and considering defen g, j j levelb'% :D design should result in a new facility design that is based on provi success j %,k p protection such that health and safety will not be who!Iy dependent on any single element of theNyy l design, construction, maintenance, or operation of the facility. The hNof, incorporating defense-in-depth practices is a conservatively designed facility and sys%u4x y tern thatmili exhibit greater tolerance for failures and external challenges. The risk insights obta\\ h performance of the ISA can be then used to supplement tHkfirni h.by focbirhttention py me g ~ on the prevention and mitigation of the potential accidents havirkp 3 Section 70,65 Additional content of acolicatio_r[ In addition to the information that cur y must submitte'd'to NRC, under 970.22, for a license application, this section requires a itionalinkmnitti6fn e submitted to demonstrate jn pm A V A compliance with the, proposed new subpart.iln particubjthis additional information would need - to include a description of the% applicant's safety progam established u y< w description of the manag%ement enessures, and an ISA summary. s wm The ISA summar[bc a) a e iption of the site and the facility; b) a N %wA description of the team qualifications and ISA methodology; c) the processes analyzed in the j ISA and the ma # mym9 um consequences lof,each; d) a demonstration of how the licensee meets p a a Environrn[enteland dynamic effects are effects that could be caused by ambient co %ft gf item relied on forW will need to hinction within its expected environmeret (i.e., under normal operating conditions, expected accidsdeendlionsiets).)These conditions could include high temperatures, or a corrosive environment. It could also inchale$namicehanges in surrounding conditions caused by an accident (e.g., the bursting of a high-pressure pipe). Mg&iV 28

n the requirements for criticality monitoring and alarms in $70.24; e) a demonstration of how the licensee meets the performance requirements of 670.61 and, if applicable,970.64; f) a list of ltems relied on for safety and a description of their safety function; g) a description of the proposed standards used to assess the consequences from acute chemical exposures; and h) . the definitions of "likely", "unlikely", " highly unlikely", and " credible" as used in the ISA. The plant and process descriptions, ISA team qualifications and methods, and definitions of terms used in the ISA, are all needed to fully understand the facility and the ISA and how it was developed. Although some of the facility information is also requested in 970.22, there may be information about the facility which would be too detailed for i,clusion py the general site description, but would be needed to be included here to understa ,18A 10.61 h0h4 is a and ISA results. The demonstration of how the licensee meets 7 un vs fr 1 critical element in determining whether the applicant understands a complies pthogy4 . // { M x, =m regulations and can operate the facility safety. Another critical elerftent is the applicant' W4 WA Af v' identification of the items relied on for safety. Through the ISA prooses,P e applicant shouldN 8 have identified potential accidents that can occur in individual proce indin the facility as a AVNM&h, \\ .vhole. As discussed earlier, these accidents are prevented or thelf consequences mitigated i using controls that are identified in the ISA summary as items,ralled on for sa b> swc:A gf important for NRC staff to review the items relied on for safety,1that were identifi as such by the applicant or licensee, to determine whether potentbcidehnkunOfy prevented or mitigated. Since items relied on for safety play a ke//y role in as#suring thetthe performance hap . requirements are met, and because the applica $$ns greatIexibili selecting and identifying py m what the actual " items" are (as discussed injalation to j70.61), tt}e ms relied on for safety would be clearly and unambiguously ids $tifi on a lis[Nhili[ items is then managed and controlled by the appilo/ :&ent through the. management measures in 970.61 to ensure that they M Rn Wif y-p g, 4 continue to perform the safety function r TBy evaluating the ISA methodology, and the vwa n rw ISA summary, supplemented by feviewing the ISA and other information, as needed, at the g;. 4 yw licensee's facility, the stan een better u6derstand the potential hazards at the facility, how the dy applicant plans to address these % ' daterjd thereby have confidence in the safety basis on hazat k which the liceny I be issued. The ISA summary would be required to be submitted on the docket in conjunction with the license tion but would be considered part of the license. The ISA, on which the ISA summahNie4 wouldkmaintained current at the licensve's facility and available for NRC review,b_$huld be submitted and docketed. The information and commitments 29

7-1 L j L l l contained in the license application that are incorporated into the license conditions cannot be changed without prior review and approval'of NRC staff, at which time a license amendment is issued. Although the ISA summary will be on the docket, since it is not part of the license it can 1 be changed without a license amendment, unless it reflects a change that cannot be made without prior approval per $70.72(c).. However, the information used to perform the ISA, and the ISA summary, both form integral parts of the safety basis for issuance of the license and therefore must be maintained to adequately represent'the current status of the' facility. So that l NRC knows the current status of the facility, changes to these documents, on which NRC based its safety conclusion, are to be submitted to NRC, as discussed in 670.72. [ Section 70.66 Additional reauirements for the aooroval of license aoo A' r,s. M ^? 7 ur In addition to the requirements found in the existing rule (i. CFR 70 Commission must determine that the requirements in the new sub 10 C D0.60 th ? 70.66, will be satisfied. e Section 70.72 Facility chances and chanae process. This section deals with changes to site, structures,6 gguipme

ents, and activities of personnel after a license application has '

Past incidents at fuel cycle facilities have ofte suite fully analyzed, not authorized by licensee management, or not ately rst lity personnel.

. Therefore, effective control of changes to a fa , site,

ctures, tems, equipment, components, and activities of personnelis a ;

eleme assur' fety at that facility. This section would require t, icensee to I and us ' $ evaluate changes and the potentialimpacts of ges g menti . By using this system to evaluate, implement a ges ] , the licensee can make certain changes without NRC pre approval. ea ation contained in the ISA summary, the licensee would be req ' _ within days of the change by submitting updated . ISA summary pa that time. s that affect the on-site documentation, such as the ISA, manag nt measures org -safety information, the licensee would be required to notify NRC in 12 months of tp change. This update frequency would allow NRC staff to review the c being mada the facility in enough time to ensure that the licensee's evaluations 4, o health and safety were accurate. It also allows NRC staff to maintain relat cility and safety information on the docket at all times. In addition, 30

maintaining the license and ISA summary so that they reflect the current configuration of the facility would facilitate a relatively simple, cost-effective license renewal process. Some changes, however, would require NRC pre-approval before they can be implemented. These are changes that are considered major and could have a significant impact I on health and safety. The staff considered two options for the types of changes that would require NRC pre-approval. Option 1 is consistent with the types of changes that have required pre-approval at Part 70 licensees in the past, and which the staff believes would require NRC pre-approval for only a relatively few significant changes. Option 2 is consistent with the change control process required for Part 50 licensees (power reactors) and which the staff believes would require more requests for NRC pre-approval. g# The advantages of Option 1 are that it focuses on the most sig nt chMges 6the facility and is equivalent to looking at the highest risk changes. It conns very . d(( criteria and is therefore easier to implement and inspect. It also woh likely onifres license amendments a year which is generally consistent with the h

A I e et these b

) v;' facilities. Since Option 1 would permit more changes without NRC p, a relatively short timeframe (90 days) for submitting updated ISA summary popes is'requiresfin, order for and tom corhbont that NRC to have information that reflects the current status of the teci p as % 27 adequate protection is still provided with the changes, as reflected in s. u o r. WA A' advantages of Option 2 are that NRC would have more#7 control over%o' changes at the th a fy Wy facilities, i.e., staff expects that more changes would be reviewed by th's 14sff before being 6f jy y-implemented; thus, it would be less likely that NRC,would_have a concern with a change after hf' fr M the fact; and it is consistent with the change control procoes at power reactors, where changes are reported only after 12gonths. The proposed h isnguageierloots Option 1. %# NA Ng[ Section 70.73 Renewal of Ilonnaan.h~ % sv c + %4 Under the proposed amendments to Parti 70, changes to site, structures, systems, Jr W ^ equipment, compogents, and activitie% c%s of personnel made by the licensee pursuant to G70 s g~q would be docurr4rited on a continuing basis on-site. A description of those changes would also tg (;; be sent to N . periodically. This pecess is intended to keep the documents, which support ,4 jy the license, t and thereby@, establish a "living" license, in the past, the license renewal p g process was tardensome tohRC and the licensee because all changes made to the facility y n ~g since the last'g, renewal would be reviewed at one time. However, with the proposed 31

""Wng iicense," changes to the facility will be reviewed by NRC either before changes are made, . or relatively shortly thereafter. As a result, review of the license renewal application is expected 1

to be performed with minimal additional review of the licensee's safety program.' This approval.

would be contingent on the licensee satisfying any requirements associated with the National Environmental Policy Act of 1969 as implemented in 10 CFR Part 51. ] 1 Section 70.74 Additional reportina reauirements. The new requirements that would be incorporated in the proposed amendments to Part 70 would revise the reporting of ' events to NRC. This new approach, based on consideratio the risk and consequences established in 10 CFR 70.61(b) is intended to replace ld on the approach licensees have currently been using for reporting cri eve Bulletin 91-01. The new approach would cover all types of events, ust criti - and establish a timeframe for reporting that is scaled according to Th reporti requirements are intended to supplement the requirements in the e rts 20 and 70 a elsewhere in the regulations. A more detailed discussion of the new '. ts is found in the following discussion of Appendix A to Part 70. [ 4 Anoendix A' Reportable Events. The reporting of events supports NRC's need _ ea at could result in an imminent danger to the worker or to public h h and ty or ti vironment. In i particular, NRC needs to be aware of licensea , s to ess po ial emergencies. Further, once safe conditions have been re afte

vent, has an interest in disseminating informat on the eve nuclear other interested parties, to reduce the likelih event the fu Iso, in the event of an accident, NRC must be able to uratel for information by the public and the media. Finally, NRC must perf f individual licensees and the industry as a whole to fulfillits sta tect health and safety of the worker and the

. public and the env'. ment. Licens porting of events consist of two reporting classes based on the hazard -- reports tha t be made in 1 h and those to be reported within 24 hours. According to this approa sees woul rt events based on two criteria: 1) whether actual consequen . whether a potential for such consequences exists; and 2) the [ seriousness' ences. The events that must be reported within the shortest 32

timeframe (1 hour) are high-consequence events.' These events encompass unintended criticalitics and loss of criticality controls, and loss of chemical controls or the occurrence of chemical exposures that exceed the performance; requirements in 970.61(b). Less serious events or failure to meet the performance requirements for reasons not otherwise specifically stated, that have occurred shall be reported within 24 hours. These include chemical exposure to licensed material or hazardous chemicals that exceed the lower threshold limits in 970.61(c)(4), and events that were dismissed in the ISA based on likelihood. Events that could potentially lead to exceeding the performance requirements in 70.61 should also be reported. External events, such as a hurricane, tornado, earthquake, flood, ork n a fire, either internal or external to the plant, that affected or could have affected a fachily, must be reported within 24 hours. This reporting requirement would captur.shexaksbdo that strikes a facility, an earthquake motion experienced by a facility [myor any type filej{Sinces w a these events could have affected a facility, NRC would want to kna bout suhievents tbe d" 9 p1 assess a licensee's conclusion of whether any detrimental effects did in occur, or could wp have occurred in the absence of controls that were present but not part of 9e safety basis. sm m m Another category of potential events that would be reported is one%at invohimillkexistence of v o an unsafe condition that is not identified in the ISA. This conddson could be caused' a sym sw deviation from established safe operating conditions, b an pnange$ mated and unamaIyzed set of p W'uld he%' reported within 24 circumstances, or by an improper analysis. This type. event hours. k$ A N The proposed rule also would require conc rent ren' orting of events when a news pg MT release is made or if other Government agencies are noillied, as is'y done under 10 CFR Part 8/ g lay 50.72, to support NRC's ability to be ( e to questions concerning the safety of NRC-k bW licensed facilities. k jcm %h 4 kW 33 1

r REFERENCES Graig, D.K., plal,, " Alternative Guideline Limits for Chemicals Without Environmental Response Planning Guidelines," American Industrial Hygiene Association Journal,1995. Fisher, D.R., Hui, T.E., Yurconic, M., and Johnson, J.R., " Uranium Hexafluoride Public Risk," Pacific Northwest National Laboratory, PNL-10065, Richland, WA, August 1994. National Council on Radiation Protection and Measurements (NCRP), " Basic Radiatioh Protection Criteria," NCRP Report No. 39, Washington, DC,1971. Ahb) [ dhx National Council on Radiation Protection and MeasurementsiNCRP), g/ [% j " Recommendations on Limits for Exposure to lonizing Radibn," NChReport NO[9hh" Washington, DC,1987. If Upk Nd U.S. Nuclear Regulatory Commission," Proposed Me. adsfo Re'gula% jor s.; g Ma tin .,m Materials Licensees," NUREG-1324, Washington, DC 5etNuary 1992 ' jji wh,q esgh UkA U.S. Nuclear Regulatory Commission / Occupatleilal Safety wul Heinish Administration py a w,,4 (OSHA)," Memorandum of Understanding Between NRC and OSHAiWorker Protection su M at NRC-Licensed Facilities" (53 FR 43950; Octobe/31,1988 cy 19 f kh U.S. Nuclear Regulatory ComrM'"Certifiofution ofheous Diffusion Plants"(59 FR k3 1994) #q.a% b 48944; Septen %,...,;<,g)A kih5 9 4 U.S. Nuclear RegulatorplCoinpaissionily yAbnormal Occurrence Reports: Implementa og + n of Section 208 of Enemy p,ation Act of 1974"(61 FR 67072; December 19, 1996). /, /b gqP U.S. ar Regulatory Cornmission, " Site Decommissioning Management Plan," e@ g NUREG-3444, Washington! DC, October 1993. 34 E

m 6 U.S. Nuclear Regulatory Commission, " Strategic Plan, Fiscal Year 1997 - Fiscal Year 2002," NUREG-1614, Washington, DC, Septembcr 1997. U.S. Environmental Protection Agency," Manual of Protective Action Guides and Protective Actions for Nuclear incidents," EPA-400-R-92-001, May 1992. U.S. Nuclear Regulatory Commission, " Instruction Concerning Risks from Occupational Radiation Exposure," Regulatory Guide 8.29, Rev.1, February 1996. A Theide, L., " Emergency Information Where it's Needed," DOE Risk Management Quarterly, Vol 5, No 2, Richland, WA, May 1997. ki h $W kWMA These documents are available for inspection and copying fore fee at the NRC Public gy v - A Document Room,2120 L Street, N.W. (Lower Level), Washington,00 20506000t> Copies of NUREG-1324, NUREG-1614, and NUREG 1444way alsk ed from du n wy 1 the Superintendent of Documents, U.S. Government Prirding OfHoeM,0 Box 37002, Washington DC 20402-9328. Copies are also availablI om t bager k.'To cal information Service,5285 Port Royal Road, Sprin ld VA 22.161. ' r m#f M pr g.a e rr Regulatory Guide 8.29 may be purchas6+d fom the-Governm. Printing Office (GPO) M9'

1. 1 at the current GPO price. Information on cument GPO s may be'obtained by contacting ff t*W ny the Superintendent of Documents, U.S/Mvernment lhindng~~ Office, P.O. Box 37082, n

yq %,i gy Washington DC 204030888., Issued giaides may also be' purchased from the National y;u v @w Technical Information Sendee %'on a stan g r, basis. Details on this service may be ~ vra obtained by writing NTIS,5205 Port floyal Reed,L8prlhgfield, VA 22161. Copies of the folloi,ewsdng draft regulatory guidance documents may be requested by ww sy W % writing to U.S. Nuclear Regulatory Comivilpelon, Reproduction and Distribution Services, Av yveg Washington, DC20555-0001:"Standsilifteview Plan for the Review of a License Application /pr N for a Fuel Cyole{acility" (Draft NUMEG-1520); and " Integrated Safety Analysis Guidance py W Document"(OnRNUREG-151 Q[ 9 z -v y Flmppg of No Significant EnvironmentalImpact: Availability 35

n l The Commission has determined, under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment, and therefore an environmental impact statement is not required. The proposed amendments to Part 70 are intended to provide increased confidence in the margin of safety at certain facilities that possess a critical mass of SNM. To accomplish this objective, the amendments: (1) identify appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed such criteria; (2) require affected licensees to perform an integrated safety analysis (ISA) to identify potential accide at th facility and the items relied on for safety; (3) require the implementation of measuree $ ensure that the items relied on for safety are available and reliable to perforn/}, h p m A. needed; and (4) require the inclusion of the safety bases, as reflected n the ISA,aummaryM A the license application. The language, in the proposed rule, that dhes an ashironmen)$1 h consequence of concern,is relevant to the question of environme Licensees wo be required to provide an adequate level of protection against a "...releasehsadioactive ,cyyu & material to the environment outside the restricter' area in concentrations pausraged over pi +% 24 hours, exceed 5000 times the values spedied in Table 2 of Appendix B td40CPR)Part 20." V cy c +$ ' otect a implementation of the new amendments. ',actuding the r t t events that could damage the environment, ir expected to r. Inash ovement in er y wg licensees'(and NRC's) understanding of the risks aftheir faciSties and81efability to ensure AU A? ts that those risks are acceptable. For existing licensees, cdeficiencies identified in the ISA M9' N would need to be promptly addressed. For licens operations would not begin unless licensees demonstrated an adequate le[el " protection agelnddVpotential accidents identified in av gg a f swem w 1hp%selsty and tal imh 5 the new amendments is positive. the ISA. As a result,wg 4 g s g There will be less advefesLimpact on th 'em4cnment from operations carried out in accordance W WMA with the proposed rule than1f" Stops operations were cam. d out.in accordance w.th the existing e i Part 70 regulation. . Qhk* w,y n,9 g V The dete ion of this E' al Assessment is that there will be no significant offsite impact on public from thi However, the general public should note that NRC ~ welcomes pu articipation. N as also committed to complying with Executive Order m; (EO) 12898, al Actions to Address Environmental Justice in Minority Populations and Low-income "#dFebruary 11,1994, in all its actions. Therefore, NRC has also determined thkMme Idisproportionate, high, and adverse impacts on minority and low-36

T: I l F. Income populations. In the letter and spirit of EO 12898, NRC is requesting public comment on l any environmentaljustice considerations or questions that the public thinks may t.e related to this proposed rule, but somehow were not addressed. Comments on any aspect of the Environmental Assessment, including environmental justice, may be submitted to NRC, as indicated under the ADDRESSES heading. NRC has sent a copy of the Environmental Assessment and this proposed rule to all State Liaison Officers and requested their comments on the Environmental Assessment. The Environmental Assessment is available for inspection at the NRC Public Document Room,2120 L Street NW, (Lower Level), Washington, D.C. and the Part 70 website. Single copies of t environmental assessment are available from Barry Mendelsohn, Office of NucleabbatarN Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washi , DC,g20555 0001, A telephone (301) 415-7262; e-mail: btm1 @nrc. gov. Paperwork Reduction Act Statement MM Ah&??. h, This proposed rule amends information collect on requirements that are subject to the pg a Paperwork Reduction Act of 1995 (44 U.S.C. 3501, et sea.). Thisye has been submitted to the Office of Management and Budget (OMB) for review a hf,the pa k requirements. [ f;gg The public reporting burden for this information collecdon is eshmeted to average 70 hours per resoonse. and the recordkeeping burbs estir Yed to a age 500 hours per 47 py - 1 licensee, including the time for reviewing instructions, searching existing data sources, pf n%jg gathering and maintaining the data needed,tand completing and reviewing the information collection. NRC is seelbn&g public comment ch the potim6alimpact of the information p A ;; p w - 37 % a m.g A vnA collections contained In he proposed rule;and on the following issues: 1. Is the pro sind btion d Mecessary for the proper performance of NRC'sfunction? WillJheiN h'hve pra ! utility? p n 3A 2. Is t burden estimate accurate? ym # 3. Is ere a way to enharpce,the quality, utility, and clarity of the information to be M collected? j 4. N can the b of the information collection be minimized, including the use p mtechnif s? of automated. t j ep khjb t 37

I Send comments on any aspect of this proposed information collection, including suggestions for reducing the burden, to the Records Management Branch (T-6-F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail at bjs1 @nrc. gov; and to the Desk Officer, Office of information and Regulatory Affairs, NEOB-10202 (3150-0009), Office of Management and Budget, Washington, DC 20503. Comments to OMB on the information collections or on the above issues should be submitted by (inscrt 30 days after publication in the Federal Reaister). Comments received I after this date will be considered if it is practical to do so, but assurance of consideration cannot be given to comments received after this date. f Y Public Protection Notification 4 sn. If a means used to impose an information collection does n$tbispla $u'rrently lilid% EA NM P OMB control number, the NRC may not conduct nor sponsor, and a person, s not required to% respond to, the information collection. ihN> wA 1 k Regulatory Analysis g 'ygyg) fd %f The Commission has prepared a draft Regulatory Analyels o% 15 /

x.

1 n . sed regulation. jy jg y :^ c The analysis examines the benefits and costs of the alternatives considepd y the A44 W ra Commission. The draft Regulatory Analysis is available f. Inspectionin the NRC Public pf Ad Document Room,2120 L Street N.W. (Lower tsvel), W ington, D.C. and the Part 70 website. Single copies (the analysis dayhobtainah T. Mendelsohn, Office of j~m 46mq p g Nuclear Material Sa ~and Safeguania, U.S! Nuclear Regulatory Commission, Washington, . na n %:w % DC, telephone (301) 15i730E e-mail:btnienic. gov. The Commission r% el%'ublic comment on the draft Reg W% equests p ws yy on the draft analysis mapNg, to NRC ss" indicated under the ADDRESSES heading. e p ,~ g ?.L if Regdisf5g/ Flexibility Certification y Asr W % y the Regulatory Flexibility Act, as amended,5 U.S.C. 605(b), the b Ay CommissionMalhain that thisproposed rule, if adopted, would not have a significant economic impact on a s er of small entities. This proposed rule would affect facilities that 38

c 1 are authorized to possess a critical mass of SNM and who are engaged in one of the following j activities: a) enriched uranium processing; b) fabrication of uranium fuel or fuel assemblies; c) uranium onrichment; d) enriched uranium hexafluoride conversion; e) plutonium processing; f) fabrication of mixed-oxide fuel or fuel assemblies; g) scrap recovery of special nuclear material; or h) any other activity involving a critical mass of SNM that the Commission determines could significantly affect public health and safety or the environment. These licensees do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act, nor 1 the size standards published by NRC (10 CFR 2.810). Voluntary Consensus Standards g jh h; The National Technology Transfer Act of 1995, Pub. L.104-8, requiresfNt g Agencies use technical stardards that are developed or adopted bpvoluntary/consen standards bodies unless the use of such a standard is inconsistentW ap/ en pf Ny g mum,icable law or otherwise impractical, in this proposed rule, the NRC proposes to usethe fotowing voluntary pms consensus standard, ANSI /ANS Standard 8.1-1983, " Nuclear Criticality Safelyin Operations y;jpug with Fissionable Material Outside Reactors," developed by the American Nuclear Soolety. Portions of the standard were used in the definition of doutile conting%ency and in}70.61(d). wp gw ,r v ww .g gy The NRC invites comment on the applicability and use[of other stendenkM*f f ' %Q'y Sw Backfit lysis y NRC has determined that the ba$kfit rule does notM this proposed rule; therefore, 4% AL M V1'i e a backfit analysis is not required fory$ proposed rule beoedse these amendments do not involve any provisionEbumbici impos 'as defined in 10 CFR Chapter I. 7 k %Q b gf[JII ( List Subjects in 10 CFR Part 70 $9' Criminalpenalties, Hazardoud materials transportation, Material control and accounting, Nuclear matkPackaging and ntainers, Radiation protection, Reporting and recordkeepi }h! ntific equipment, Security measures, Special nuclear material. 39

i ( For the reasons set out in the preamble and under the authority of the Atomic Energy L Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 553, NRC is proposing to adopt the following amendments to Part 70. 1 yw. germ

I fk

.a g ' ' adh,, y YYbi &.

%IC O 3kry 3.

3 T.m) Midi I ,s .-).U. 4ww %,,L pfhik; 3lllW Y u, 'Y ' ' n? l 3 .f?,:j ;y;qp Sh gg.y 40 l l

Part 70 - DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL

1..The authority citation for Part 70 continues to read as follows:

AUTHORITY: Secs. 51,53,161,182,183,68 Stat. 929,930,948,953,954, as amended, sec. 234,83 Stat. 444, as amended (42 U.S.C. 2071,2073,2201,2232,2233,2282,2297f); secs. 201, as amended,202,204,206,88 Stat.1242, as amended,1244,1245,1246 (42 U.S.C. 5841,5842, 5845,5846). Sec.193,104 Stat. 2835, as amended by Pub. L.104-134,110 Stat.1321,1321-349 (42 U.S.C. 2243). A Sections 70.1(c) and 70.20a(b) also issued under secs. 135,141, Pub. L. 97-425,96 Stat. m gy 2232,2241 (42 U.S.C.10155,10161). Section 70.7 also issued under,ub. L. 95401, sec.10, P 92 Stat. 2951 (42 U.S.C. 5851). Section 70.21(g) also issued under sk122, O8)w w StadiS9 sw U.S.C. 2152). Section 70.31 also issued under sec. 57d, Pub. L. 93 ,88 St 75 ((2 U 8.C. 2077) Sections 70.36 and 70.44 also issued under sec.184,68 Stat. 964, as amended (42 U.S.Cd um n %w 2234). Section 70.61 also issued under secs. 186,187, 68 Stat. 955142 .S.C. 2236, 2237)F w il3A Section 70.62 also issued under sec.108,68 Stat. 939, as amended (42 U.8;C,2138). b

2. The undesignated center heading " GENERAL PROVISION 8'Us redesignated Ni"Subpart I

A -- General Provisions." hIk

3. In 970.4, the definitions of Acute, Available and rei e to deform their function when Mf M

n needed, Configuration Management, Critical mass of specialeuclear material, Double contingency, W H n Hazardous chemicals produced f rom licensedmaterial, Integrated safety analysis (ISA), Integrated safety analysis summa items relied +onfwior safetinManagement measures, Unacceptable i pwsgy ) performance deficie h Workhhhbd, in aIh$k$al order, as follows: [qq[k;e%g% %g$h:y% Q70.4 Definitions. M S

a eJQ;pfef&

Q j.y fiQ A_cu e as U in this Part rneens s' single radiation dose or chemical exposure event or l PnM multiple radiation /Wdose or chemical exposure events occurring within a short time (24 hours or less). } A uA Avah;h-'Mikioerform their function when needed as used in Subpart H of this Part means tkhuh$b the analyzed, credible conditions in the integrated safety analysis, 41

items relied on for safety will pedorm their intended safety function when needed and management measures will be implemented that ensure continuous compliance with the pedormance requirements of 70.61 of this Part, considering factors such as necessary maintenance, operating limits, common cause failures, and the likelihood and consequences of failure or degradation of the items and measures. Confiauration manaaement (CM) means ensuring, as part of the safety program, oversight and control of design information, safety information, and modifications (both temporary and permanent) that might impact the ability of items relied on for safety to perform their function b %[hk needed. Critical mass of soecial nuclear material (SNM) means special nuclear materInf in aquenuty q y y m9 exceeding 700 grams of contained uranium-235; 520 grams of uranium-2d5;' 450 grameMW m /r WW plutonium; 1500 grams of contained uranium 235,if no uranium en, to;more than 4 percentM by weight of uranium-235 is present; 450 grams of any combinatiokthenef; pr one-half such Ap%;v.:m quantities if massive moderators or reflectors made of graphite, heavy wate qpy jfhk( Double continaency means a process design tis'#^63$A i tincorporatessulliment actorsof safety M 17 WW# to require at least two unlikely, independent, and concurrent changes i ~.. s conditions betore W N a criticality accident is possible, jj gp' 3 y Hazardouschemicalsoroducedd.riEcensedJ ^ '"iIneans substances having licensed as a o r material as precursor;oompound(s)a 1Moubstances thstiph,hsically or chemically interact with Npmeg gw licensedmaterials;thataretcNic, explosive,hhontec[lNd Th se i flammable, corrosive, or reactive to the extent that they can endanger life or healtknkhuate e nclude substances commingled with licensed material, and mM$tances as hydrogen fluoride that is produced by the reaction of uranium bluorideat b do not include substances prior to process addition to licensed mate ifor after proces bion from licensed material. inter ~ " safety analysiaTISA) means a systematic analysis to identify plant and external hazardsandkh@iothIbting accident sequences, the potential accident sequence w vv likelihood and consegsences, and the items relied on for safety. As used here, integrated means 42

joint consideration of, and protection from, all relevant hazards, including radiological, nuclear criticality, fire, and chemics!. However, with respect to compliance with the regulations of this Part, the NRC requirement 'a limited to consideration of the effects of all relevant hazards on radiological safety, prevention of nuclear criticality accidents, or chemical hazards directly associated with NRC licensed radioactive material. Intearated safety analysis summary means the document ' submitted with the license application, license amendment application, or license renewal application that provides a synopsis A of the results of the integrated safety analysis and contains the information specified in 670.65(b). Mf g 1[ ems relied on for safety means structures, systems, equipment <componentshotivities of personnel that are relied on to prevent potential accidents at a f ty that coth Ihe performance requirements in 670.61 or to mitigate their potential c/4onsequences. This does~not 47 WW gm jf y. limit the licensee from identifying additional structures, systems nt, componentsfor"w %-A i activities of personnel (i.e., beyond those in the minimum set necess for compliance with the j performance requirements) as items relied on for safety. [ h Manaaement measures mean the functions perjolmedt e erally on a g

wmppy, continuing basis, that are applied to items relied upon icr safety, to ensure the ilmes are available tv Ar W L ^r and reliable to perform their functions when neededs Managerpent measureslNelude configuration management, maintenance, training and qualbhons, pbedurehtkudits an imr M

incident investigations, records manageme fand othegality assurance elements. Unacceptable"k'%,YY ance J """ mean dsficien*cies in the items relied on for safety v % aA or the management measures thet need tcibe corrected to ensure an adequate level of protection v;m as defined in 10 CFR 70.61g),(a),3(d). ~ '- s a fg c A Worker means an individual % exposure to radiahn and/or radioacthh;sesigned duties in the cou rial from licensed and unlicensed sources of radiation K4 $4 (i.e., an individual who is subject to en occuoational dose as in 20 CFR 20.1003).

4. In s revised to read as follows.

43

r-670.8 Information collection requirements: OMB approval. l (b) The approved information collection requirements contained in this part appear in j 6 70.9, 70.14, 70.19, 70.20a, 70.20b, 70.21, 70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, I 70.39,70.42,70.50,70.51,70.52,70.53,70.57,70.58,70.59,70.60,70.61,70.62,70.64,70.65, l 70.66,70.72, and Appendix A.

5. The undesignated center heading " EXEMPTIONS" is redesignated as "Subpart B --

Exemptions." j[/ $6 70.13a and 70.14 [ Redesignated) .hbQf# jf* .k% y'

6. Sections 70.13a and 70.14 are redesignated as 9 70.14 and 70.17, respectively22fh fr

7. The undesignated center heading " GENERAL LICENSES"Is&,b redesignated as"Subpart:#

i) ${%@A. C -- General Licenses." %.y; \\qu,

8. The undesignated center heading " LICENSE APPLICATIOr;S"9iM ateo as em;pp W,

"Subpart D - License Applications." h4gMA / %w f l

9. The undesignated center heading " LICENSES"Iredeshtsd as "Subpart E --

Licenses."

10. The undesignated center heading (ACOL'ISITION, USE AND TRANSFER OF SPECIAL mi pa sp NUCLEAR MATERIAL GREDITORS' RIGHTS,"is redesignstid as "Subpart F - Acquisition, Use, and Transfer of Specialbar Mab,Nitors' Rfhtitsh NA N@A

11. The undesignanoenter heading?8PECIAL NUCLEAR MATERIAL CONTROL RECORDS, REPORTS AND IN8PECTl9NS"quyis redesignated as "Subpart G -- Special Nuclear m-Material Control Records, Reports, and in%spections."

Ar ya i l

12. Ir$3 50 paragraph (c Js revised to read as follows.

f iMA y Djbgy. % $-p/ qs 44 I \\ l 1

$70.50 Reporting Requirements (c) Preparation and submission of reports. Reports made by licensees in response to the requirements of this section must be made as follows: (1) Licensees shall make reports required by paragraphs (a) and (b) of this section, and by ] section 70.74 and Appendix A of this Part if applicable, by telephone to the NRC Operations I 8 - Center. To the extent that the information is available at the time of notification, the information provided in these reports must include: (i) Caller's name, position title and call back telephone number; fF (ii) Date, time, and exact location of the event; Mb (iii) Description of the event, including; ) g. (A) Radiological or chemical hazards involved including isotopes, quantitiesiand ~ chemical and physical form of any material released; k (B) Actual or potential health and safety consequences to the woikers, the public, and'ths. ~ y.g environment, including relevant chemical and radiation data for actdsd personnel exposures to fy Visa radiation or radioactive materials or chemicals (e.g., level of radiation exposurefooncentration of chemicals, and duration of exposure); gygg y$g7g%

u fp (C) The sequence of occurrences leading to the epincludhg degraditledor failure of structures, systems, equipment, components, and ac ties othersonr$lre on to prevent potential accidents or mitigate their consequences d

d (D) Whether the remaining structures, ems, equipment, corviponents, and activities of M f! personnel relied on to prevent potential accid. or mitigiute their consequences are available and reliable to perform theirJunction. (iv) External-s affecting tie event; 'M W 'a h Os (v) Additional acGons hken by the 90enses in response to the event; (vi) Status of the Ikh; whet NNht is on-going or was terminated); (vii) Current and,h ok s, i ng any declared emergency class; (viii) NotificaWons related thebthat were made or are planned to any local, State, y y mVy or other Federalagencies; E Qr iM f ' (ix) Status of any press releases related to the event that were made or are planned. P"h j AG th A / g

Vp' 8 The bomineralef ' telephone number for the NRC Operations Center is (301) 816-5100.

45

L _(2) Written report. Each licensee who makes a report required by paragraph (a) or (b) of this section, or by 970.74 and Appendis A of this Part if applicable, shall submit a written follow-up report within 30 days of the initial report. Written reports prepared pursuant to other regulations may be submitted to fulfill this requirement if the report contains all of the necessary information and the appropriate distribution is made. These written reports must be sent to the U.S. Nuclear ' - Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the appropriate NRC regional office listed in Appendix D of 10 CFR Part 20. The reports must include ' the following: pk (i) Complete applicable information required by 670.50(c)(1); (ii) The probable cause of the event, including all factors that contributed to e and g Q the manufacturer and model number (if applicable) of any equipment t , ed or'p f"t ned; - (iii) Corrective actions taken or planned to prevent occurrenc similar or in the future and the results of any evaluations or assessments; a k (iv) For licensees subject to Subpart H of this Part, wheth t was identified evaluated in the Integrated Safety Analysis. ( - (d) The provisions of 970.50 do not apply to licensees subj to 69k33,@iisy do apply to those Part 50 licensees possessing material licensed under who a% t to the re notification requirements in $50.72. + /

13. The undesignated center heading ODIFl lON EVOCATION OF LICENSES" is redesignated as "Subpart I -- M tion ev n of Licenses."

$$ 70.61 and 70.62 [R ignated]

f.{

-{

14. Sections 70.69

,ignate Ys9f70.81 and 70.82, respectively. . [- %j 15.The undesignat eadind EMENT" is redesignated as "Subpart J -- Enforcement." k ' $$ 70.71 and 7p [ Redesignated) 18.S ns 70.71 and 70 are redesignated as l70.91 and 70.92, respectively. i

19. II.

BPART H" ($$ 70.60 - 70.74) is added to read as follows: w 46

r. -

) . Subpart H - Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass l of Special Nuclear Material. ) Sec. 70.60 Applicability. 70.61 Performance requirements. i 70.62 Safety program and integrated safety analysis. 70.64 Requirements for new facilities or new processes at existing facilities. ] 70.65 Additional content of applications. 70.66 Additional requirements for approval of license application. e 70.72 Facility changes and change process. kl D. 70.73 Renewal of licenses. f 70.74 Additional reporting requirements. Ihhf[ $$W r T j!! j tW $ 670.60 Acolicability. %4 frQ % The regulations in Q70.61 through 97'O.74 apply,in addition toother applicable Commission regulations, to each applicant or licensee that is or plans to be- (1);authorizeb greater Mc~A Wg than a critical mass of special nuclear material, and (2) engaged in endched uranium processing, fabrication of uranium fuel or fuel assemblies, uranium $!chmeNheM warfm hexafluoride conversion, plutonium processing, fabrication of fxed-oxid fuel h$$$semblies, scrap A G' bi N recovery of special nuclear material, or any other activity thht the Commission determines could 27 41 M significantly affect public health and safety. The regulations in 70.61through 670.74 do not apply to decommissioning activities perforrnoff p@uant to "a'kable Commission regulations p ed ga;M su 4 including 70.25 and 570.38 of thistert&Also, the re ns in 70.61 through @70.74 do not y; r sy;;, ;.x apply to activities that'are;' certified by the Commission pursuant to Part 76 of this chapter or Q W-;% licensed by the Commission pu% muent to other ,of this chapter. M i 670.61Performano6 Reauirements. [2hV e4 (a) Each opplicant or licensed shall evaluate, in the integrated safety analysis performed An M in accordanoisiellh 70.62, its compliance with the performance requirements in paragraphs (b), (c), and (d) hhd[ jf %ks 47

o L (b) The risk of each credible high'-consequence event must be limited, unless the event is highly unlikely, through the application of engineered controls, administrative controls, or both, that reduce the likelihood of occurrence of the event or its consequence. Application of additional controls is not required for those high-consequence events demonstrated to be highly unlikely. High-consequence events are those internally or externally initiated events that result in: ~ (1) An acute' worker dose of 1 Sv (100 rem) or greater total effective dose equivalent; (2) An' acute dose of 0.25 Sv (25 rem) or greater total effective dose equivalent to any . individual located outside the controlled area identified pursuant to paragraph (f) of this section; (3) An intake of 30 mg or greater of uranium in soluble form by any individual I outside the controlled area identified pursuant to paragraph (f) of this section; or (4) An acute chemical exposure to an individual from licen ateri ous chemicals produced from licensed material that: [ [ f (i) Could endanger the life of a worker, or [ (ii) Could lead to irreversible or other serious, long-lasting s to any indivi located outside the controlled area identified pursuant to paragra section. If an applicant possesses or plans to possess quantities of mater, cap h chemical ' exposures, then the applicant shall propose appropriate qua standa ' e health effects, as part of the information submitted pursuant to rt. q l (c) The risk of each credible intermediate-c eque evem p limited, unless the i event is unlikely, through the application of en , ed is,ad trative controls, or both, that reduce the likelihood of occurrence of th ntori nseg . Applicationof additional controls is not required hose inte conseg demonctrated to be unlikely. . Intermediate-conse ' - ' _ents ernali ernally initiated events, that are not y,, - high-consequence eve , Jult i (1) An acute worke " r greater total effective dose equivalent; 5 Sv (2) An acute d m) or/ greater total effective dose equivalent to any individuallocated [ide the contI$ ntified pursuant to paragraph (f) of this section; (3) A 2 ur averaged relt f radioactive material outside the restricted area in concentratio ,ceeding 5000 ti the values in Table 2 of Appendix B to 10 CFR Part 20; or (4) e' chemical sure to an individual from licensed material or hazardous chemicals p d material that: (i) Co versible or other serious, long-lasting health effects to a worker, or 48 4 .i

[ l J (ii) Could cause mild transient health effects to any individuallocated outside the controlled area as specified in paragraph (f) of this section. If an applicant possesses or plans to possess quantities of material capable of such chemical exposures, then the applicant shall propose appropriate quantitative standards for these health effects, as part of the information submitted pursuant to $70.65 of this Part. (d) In addition to complying with paragraphs (b) and (c) of this section, the risk of nuclear criticality accidents must be limited by assuring that under normal and credible abnormal conditions, all nuclear processes are subcritical, including use of an approved mar ' subcriticality for safety. Preventive controls and measures must be the pnmary mea ion' against nuclear criticality accidents. + ^ (e) Each engineered or administrative control or control sy nec to com [ paragraphs (b), (c), or (d) of this section shall be designated as an on for safety. safety program, established and maintained pursuant to 970.62 of thi ensure that each item relied on for safety will be available and reliable to perform its i d i en needed and in the context of the performance requirements of this se hI h w$s ey' 3, in which the ^~ (f) Each licensee must establish a controlled a,as. licensee retains the authority to determine all activiti includi exclu ' "movalof personnel A and property from the area. For the purpose , plying. Ih the p rmance requirements of this section, individuals who are not worke defineN 70.4 eay be permitted to perform v ongoing activities (e.g. a facility no to the I ities) in the controlled area, if the licensee: b y '(1) Demonstra ment ' grated safety analysis, that the risk for those ~5 m individuals at the location ' ivities exceed the performance requirements of p paragraphs (b)(2), (b and ( )(ii) of this section; or. (2) Provide Training in with 10 CFR 19.12(a)(1)-(5) to these individuals to ensure that the aware of the ri lated with accidents involving the licensed activities asdetermin heintegratedsafp# analysis, and (ii) Conspicuously posts and maintains notices stating wher formation in CFR 19.11(a) may be examined by these individuals. Under these condit e requirements for workers specified in paragraphs (b) and (c) of this section to these individuals. 49

l 670.62 Safety Proaram and Intearated Safety Analysis (a) Safetyprogram. (1) Each licensee shall establish and maintain a safety program that demonstrates compliance with the performance requirements of 70.61. The safety program may be graded such that management measures applied are commensurate with the reduction of the risk attributable to that item. The three elements of the safety program, namely process safety information, integrated safety analysis, and management measures, are described in paragraphs (b) through (d) of this section. (2) Each licensee shall establish and maintain records that demonstrate comphance with the requirements of paragraphs (b) through (d) of this section. !b kf (3) Each licensee shall establish and maintain a log, avagable for NRC' Inspection, gg a7 %wnm documenting each discovery that an item relied on for safety or management measure has failed M*7 to perform its function either in the context of the performance requirements of 670.6 k %., A? upon; e demand. This log must identify the item relied on for safety or manl ' measure that has , % wA failed and the safety function affected, the date of discovery, date o.f thefailure, wa duration (or estimated duration) of the time that the item was urable.to perform lie f p; A yy other affected items relied on for safety or management'measumeland their safety function, affected processes, cause of the failure, whether the fanure was a the;co%ntext of the performance y ya W requirements or upon demand or both, and any corrective or co,ympensatoryaction that was take gr y _W AQ M V: The log must be initiated at the time of discov and updated pro y upon the conclusion of each investigation of a failure of an item reliakon for safety or management measure. A h (b) Process seisfykiformati Each licensee or applicant shall maintain process safety information to enab e hhi'pebmanc krated safety analysis. This process safety wA ym information must include infepmellon pertaining to%thshazards of the materials used or produced in the process,informatic6 he tec gy of the process, and information pertaining to the equipment in the process. We%%N W dy r i Af (c) Integratedsafety ana (1) Each licensee or applicant shall conduct an integrated safety analy At is of appro;Iit detail for the complexity of the process, that identifies: (i) R NNkated to possessing or processing licensed material at its facility; k [ 50

(ii) Chemical hazards of licensed material and hazardous chemicals produced from licensed material; (iii) Facility hazards which could affect the safety of licensed materials and thus present an increased radiological risk; (iv) Potential accident sequences caused by process deviations or other events internal to the plant and credible external events, including natural phenomena; (v) The consequence and the likelihood of occurrence of each potential accident sequence i identified pursuant to paragraph (c)(1)(iv) of this section, and the methods used to determine the consequences and likelihoods; and (vi) Each item relied on for safety identified pursuant to $70.61(e) the characteristics of its preventive, mitigative, or other safety function ' Jhe a and conditions'u'nder which the item is relied upon to support comp ce with tp$ [ requirements of $70.61. (2) Integrated safety analysis team qualifications. In order t he adequacy of integrated safety analysis, the analysis must be performed by a team . e in engineering and process operations. The team shall include at least one palpon k ence and knowledge specific to each process being evaluated, and persoge1$o have e h nuclear a gy gy criticality safety, radiation safety, fire safety, and chemi . Onels, ember of the team must be knowledgeable in the specific integratadhfety a y being used, y y (3) Requirements for existing licensees. Nelkithsta g oth ons regarding the effective date for Part 70 Subpart H requirem

, icen shall cyply with the provisions in paragraphs (c)(3)(i), (ii), and (iii) of this s eginngen . Individuals holding an NRC lice go thedatiN n of the final rule > shall, with j

regard to existing lic hivities h (i)Within 6 moNhb date srx. n. "n of the final rule >, submit for NRC approval, a plan that describes the n . ty an roach that will be used, the processes that a < gy will be analyzed, and t letingthe analysis of each process. (ii)Within rs of <the d . tion of the final rule >, complete an integrated safety analysis, corr I unacceptable ance deficiencies, and submit an integrated safety analysis sum in accordance 970.65 or the approved plan submitted under paragraph (c)(3)(i)of th ion. i jy g-51 i

) i (iii) Pending the correction of unacceptable performance deficiencies identified during the conduct of the integrated safety analysis, the licensee shall implement appropriate compensatory measures to ensure adequate protection. (d); Management measures. Each applicant or licensee shall establish management rneasures to provide continuing assurance of compliance with the performance requirements of . 970.61. The measures' applied to a particular engineered or administrative control or control system may be commensurate with the reduction of the risk attributable to that control or control system. The management measures shall ensure that engineered and administrativ ontrol control systems that are identified as items relied on for safety pursuant to 670.61 Part are designed, implemented, and maintained, as necessary, to ens hya and reliable to perform their function when needed, in the context of com 'nce with requirements of 970.61' of this Part.- f f +4 n Yg

• 7

' 670.64 Reauirements for new facilities or new processes at e::isti ili (a) Baseline design criteria. Each prospective appI' license ress the following baseline design criteria'in the design of new f ) existin ensee shall Ys address the following baseline design criteria in the d n of ng isting facilities that require a license amendment under $70.72. baselirWhesign must be applied to the design of new facilities and new processe notpquire r its to existing facilities or existing processes (e.g., those housing or a nttot " w pr ); however, allfacilities and processes must compi ' h the perf require .61. Licensees shall maintain the application of th ' unle ' ation p ' pursuant to paragraph (c) of this section demonstrates item i for safety or does not require adherence to 2-4 the specified critena. 4 (1) The sign must be developed and implemented in accordance with gement mea ovide adequate assurance that items relied on for safety will be ble and reliable rm their function when needed. Appropriate records of theseitems t be maintained or under the control of the licensee throughout the life of the

facility, j

L(2) . The design must provide for adequate protection against naturalphen ideration of the most severe documented historical events for the site. 52 -. s.: +

(3) Fire orotection. The design must provide for adequate protection against fires and - explosions. (4) Environmental an'd dynamic effects. The design must provide for adequate protection from environmental conditions and dynamic effects associated with normal operations, ' maintenance, testing, and postulated accidents that could lead to loss of safety functions. (5).GLhemical orotection. The design must provide for adequate protection against chemical risks produced from licensed material, plant conditions which affect the safety of licensed material, and hazardous chemicals produced from licensed material. (6) Emeroency capabilitv.- The design must provide for emergency capabili to mai. { control of: { (i) Licensed material; I (ii) Evacuation of personnel; and T j (iii) Onsite emergency facilities and services that facilitate [ us _ availabl services. (7) Utility services. The design must provide for continued wf essential utility 4m /g gi services. (8) Insoection. testina. and maintenance. The des ogs relie , ty must provide for adequate inspection, testing, and maintenance availabi and reliability to perform their function when needed. s. (9) Criticality control. The design must prov for crit ity co ding adherence to the double contingency principle, f (10) Instrumentation and controlg. T sign miMprovide nclusion ofinstrumentation behavik ied on for safety. and control systems to nitor and co i s y-(b) Facility an' esign slayout must be based on defense-in-depth practices". The design pr inco (he extent practicabio: Al

• As use 170.64, defense-i

' practices means a design philosophy, applied from the j outset and thro ompletion of the n, that is based on providing successive levels of protection l such that healp d safety will not be olly dependent upon any single element of the design, construction, nance, or oper of the facility. The net effect of incorporating defense-in-depth practices is a tively des facility and system that will exhibit greater tolerance to failures i and extemal 'nsights obtained through performance of the integrated safety analysis can ~ lement the final design by focusing attention on the prevention and mitigation of th olential accidents. i 53 h

(1) Preference for the selection of engineered controls over administrative controls to increase overall system reliability; and (2) Features that enhance safety by reducing challenges to items relied on for safety. 670.65 Additional content of aoolications. (a) In addition to the contents required by Q70.22, each application must include a description of the applicant's safety program established under Q70.62, including the integrated safety analysis summary and a description of the management measures. (b) The integrated safety analysis summary must be submitted with the license ottenewal A p a:x application (and amendment application as necessary), but shall not be incorporatesinthe'lleense. y i har However, changes to the integrated safety analysis summary shall mset the conddi'wp :s.onsof $70.7l The integrated safety analysis summary must contain: [ N l y (1) A general description of the site with emphasis on those factors that could affect safetyM. WiiMN (i.e., meteorology, seismology); ggj (2) A general description of the facility with emphasis on;j$g$se h r-a . safety, including an identification of the controlled area boundanos; i (3) A description of each process (defined as a single reasonably simpi integrated unit operation within an overall production line) analyzed in'the integb sehsIy sis in sufficient Ar ff y ' mw detail to understand the theory of operation; and, for.each proces~s, the hazanis fhat were identified AC? / p in the integrated safety analysis pursuant to $70.62(c)(1 iii) and a general description of the fy' 6# types of accident sequences; g (4) Information that demonstrates {$be lice hc6Npliance with the performance ^ M f um g my requirements of @70A1; t{e requirements for' criticality monitoring and alarms in 70.24; and, if

• w ng yu a

applicable, the requirementelef'g70.64; W M % mA %m (5) Adescription of thetsem;qualifica'hons,g Y, the methods used to perform the integr safety analysis; [b b (6) A list briary describing a'l Asms folied on for safety which are identified pursuant to 70.61(e) in su nt detail to u r d their functions in relation to the performance pm

g requirements.cf $70.61; f'

(7) Ahiption of the p sed quantitative standards used to assess the consequencen p

e Wibplicensed material or chemicals produced f rom licensed materialt from acute

,.my:y which are on-jmpected to be on-site as described in 70.61(b)(4) and (c)(4); 54

I l (8) A descriptive list that identifies all items relied on for safety that are the sole item l preventing or mitigating an accident sequence that exceeds the performance requirements of $70.61; and (9) A description of the definitions of likely, unlikely, highly unlikely, and credible as used . In the evaluations in the integrated safety analysis. 670.66 Additional reauirements for aooroval of license aoolication. An application for a license from an applicant subject to Subpart H will be approved if the A Commission determines that the applicant has complied with the requirements 'of $70.21, 670,32,, 670.23 and 670.60 through 670.65. 6 70.72 Facility chanaes and chanae process. s' (a) The licensee shall establish a configuration management m to ' ate,im W and track each change to the site, structures, processes, syste. ment, componerfts; computer programs, and activities of personnel. This system must; ented in written procedures and must assure that the following are addressed prioth y change: (1) The technical basis for the change; (2) Impact of the change on safety and health or (3) Modifications to existing operating proce bncl mat Aegessary training or ^ retraining before operation; (4) Authorization requirements for the Ange; (5) For temporary changes, the approhduratiog., exp ndate)of thechange;and (6) The impacts ormodificationsa$th tegrat fsis, integrated safetyanalysis 'develhhaccordance with 670.62. summary, or other s ram i k A (b) Any change to sifeggueses, processeels9 stems, equipment, components, computer programs, and activities bbe e ted by the licensee as specified in paragraph py y-m (a) of this section,Mfore the ch^ - lipplemented. The evaluation of the change must .At WW determine, bef .the change is im , ted, if an amendment to the license is required to be submitted in adDprdance with 670.34. E' 55

(c) The licensee may make changes to the site, structures, processes, systems, equipment, 1 1 components, computer programs, and activities of personnel, without prior Commission approval, if the change: (1) does not: 5 (i) Create new types of accident sequences that, unless mitigated or prevented, would exceed the performance requirements of 670.61 and that have not previously been described in the integrated safety analysis summary; or (ii) Use new processes, technologies, or control systems for which the licensee has no prior experience; h (2) Does not remove, without at least an equivalent replacement of the safety (Unotion, an ($ item relied on for safety that is listed in the integrated safety analysis ary; (3) Does not alter any item relied on for safety, listed in ti tegrated Afkanh t p W m f summary,- that is the sole item preventing or mitigating an accide. :s,equence th,at exceeds the - performance requirements of 670.61; and yk?MM/y w% # n j 2 :: %i % (4) is not otherwise prohibited by this section, license conditio'n, Lor order. N (d)(1) For any changes that affect the integrated safety, analysis summary les e0bmitted in pu, A + ?? accordance with 70.65, but do not require NRC pre-approvelithe lloonsee shall submit revised ny y,=g y pages to the integrated safety analysis summary, to NRC, within 90 lcl thq change. q A N

- V (2) For changes that require pre-approvalhnder 670t72, th illoonsee shall submit an

,t cy Ay cf amendment request to the NRC in accordanceiih;th 70 34 and 670.65. Ay rd Ar (3) A brief summary of all changes to the records;retiluired by$70.62(a)(2) of this Part, that must be hubbilthkiNRC every 12 months. ^ are made without prior Commission &g &hw %7 (e) If a change' covered by 670.72;ie made,,the affected on-site documentation must be . Jw j updated promptly. w% W w%g%f/ dh ?h (f) The lic pose shall maintalnjeoonis of changes to its facility carried out under this section. These; records must inclMhritten evaluation that provides the bases for the f ms PQ f^ y change the "tyl%. 5 Any In the defining characteristics of the elements of an accident sequence may p# cl Sne accident sequence for a given process. For example, a new type of accident could involve a Steventinillator,'significant changes in the consequence, or a change in the safety function of a cohteolb.gdteinperature limiting device versus a flow limiting device). i 56 l s

r s determination that the changes do not require prior Commission approval under paragraph (c or - d) of this section. These records must be maintained until termination of the license. l - 670.73 Renewal of licenses. Applications for renewal of a license must be filed in accordance with 662.109,70.21,70.22, 70.33,70.38, and 70.65. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference, provided that these f . re erences are clear and specific. e s E 670.74 Additional reportina reouirements. [ 3. (a) Reports to NRC Operations Center. (1) Each licensee shall report to the NRC Operations C vents describe N Appendix A to Part 70. (2) Reports must be made by a knowledgeable licensee repqosent ny method that will ensure compliance with the required time period for r hvent a, (3) The information provided must include a d her related information as described in $70.50(c)(1) jkIs# ppg!" (4) Follow-up information to the reports mustbe provl unti hhmation required to Mij L4 be reported in 970.50(c)(1) of this Part is com ef // (5) Each licensee shall provide reasomWle assu ~ tha pblecommunicationwiththe NRC Operations Centeps available hevent!$$$$f l kA (b)WrittenReddsigh ense a report required by paragraph (a)(1) of this section shallsubmit a wrib eport days of the initial report. The written report $b$b(c)(2). must contain the infor p 57

Appendix A to Part 70- Reportable Safety Events As required by 10 CFR 70.74, licensees subject to the requirements in Subpart H of Part 70, shall report: . (a) One hour reports. Events to be reported to the NRC Operations Center within 1 hour of discovery, supplemented with the information in 10 CFR 70.50(c)(1) as it becomes available, followed by a written report within 30 days: [h / (1) An inadvertent nuclear criticality. t (2) An acute intake by an individual of 30 mg or greater of u m in a s I g 7 kY w{ (3)_ An acute chemical exposure to an individual from licM. erial or hazar { wm chemicals produced from licensed material that exceeds the quantitat. s established to satisfy the requirements in $70.61(b)(4). fL 8,ePha as doc $$ (4) An event or condition such that no items relied; ort $r tapented in the - Integrated Safety Analysis summary, remain avalla nd rohk ja ent sequence bion) h evaluated in the Integrated Safety Analysis, to perf thei (i) in the context of the performance r ,mentsgl70.61(.and $70.61(c), or (ii) Prevent a nuclear criticality accid .e., los/$f all c ' in a particular sequence). (5) Loss of p. k f n. 3 geh th item relled"on for safety, as documented in the Integrated Safety Anaik , remeMi ble and reliable to prevent a nuclear criticality accident, and has been in greb hght hours. } (b) Twenty 4ourh its. Even rted to the NRC Operations Center within 24 hours of discovery, s mented with th. ation in 10 CFR 70.50(c)(1) as it becomes available, ' followed by en report within days-bD a 58

E -(1) Any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of $70.61. ' (2) Loss or degradation of items relied on for safety that results in failure to meet the performance requirement of $70.61. (3) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed materials that ' exceeds the quantitative standards that s the requirements of $70.61(c)(4). (4) Any natural phenomenon or other external event, includi fires inter. to the facility, that has affected or may have affected the intended fun or avail reliability.of one or more items relied on for safety. d' -(5) An occurrence of an event or process deviation that wap on hge, integrated Safety Analysis and: YM4Kh (i) Was dismissed due to its likelihood; or hT (ii) Was categorized as unlikely and whose a ated quences would have exceeded those in 670.61(b) had the item (s) Wied on for safe ormed their safety i funct!on(s). , s. [2624g ff g; \\ + 59

rc l (c) Concurrent Reports. Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made, shall be reported to the NRC Operations Center concurrent to the news release or other notification. Dated at Rockville, Maryland, this day of .1999. 2 For the Nuclear Regulatory Commission. or hk#s jhb ['?%hii;h/i \\ 4 Annette Vietti-Cook, I;, } Secretary of the Com ) Nd d 6 [/ jj # [A% \\ e_ asm b' ig. N wax , ydfW.,

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