ML23151A713

PR-070 - 64FR41338 - Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material

ML23151A713
Person / Time
Issue date:	07/30/1999
From:	Annette Vietti-Cook NRC/SECY
To:
References
PR-070, 64FR41338
Download: ML23151A713 (1)

Text

ADAMS Template: SECY-067 DOCUMENT DATE: 07/30/1999 TITLE: PR-070 - 64FR41338 - Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material CASE

REFERENCE:

PR-070 64FR41338 KEYWORD: RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

Docket No.: PR-070 10/30/2000 FR Cite: 64FR41338 In the Matter of Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material Comment Comment Docket Document Miscellaneous Accession Number Submitted by Representing Date Date Description Number 07/26/1999 07/23/1999 Federal Register Notice -

Proposed Rule 10/12/1999 10/12/1999 Letter from Eldon V. C.

Greenberg, Esq., counsel to Nuclear Control Institute requesting 60-day extension of comment period.

Richard L. Black Department ofEnergy 10/14/1999 10/13/1999 2 Peter S. Hastings Duke Cogema Stone & 10/14/1999 10/13/1999 Webster, LLC 3 Ame F. Olsen BWX Technologies, Inc. 10/14/1999 10/11/1999 1

Docket No.: PR-070 10/30/2000 FR Cite: 64FR41338 Io the Matter of Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material Comment Comment Docket Document Miscellaneous Accession Number Submitted by Representing Date Date Description Number 4 J. William Bennett United States Enrichment 10/14/1999 10/12/1999 Corporation 5 Marvin S. Fertel Nuclear Energy Institute 10/18/1999 10/13/1999 6 C. M. Vaughan General Electric Company 10/18/1999 10/13/1999

• 7 Jim Makris Chemical Emergency Preparedness and Prevention Office, United 10/18/1999 10/12/1999 States Environmental Protection Agency 8 James B. Edgar Siemens Power 10/18/1999 10/13/1999 Corporation 2

Docket No.: PR-070 10/30/2000 FR Cite: 64FR41338 In the Matter of Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material Comment Comment Docket Document Miscellaneous Accession Number Submitted by Representing Date Date Description Number 9 Sara C. Yerkes National Fire Protection 10/18/1999 10/15/1999 Association 10 Kay Drey Self 11/29/1999 11/24/1999 09/07/2000 09/06/2000 Federal Register Notice - Final Rule 3

JOCKET NUMBERnn 10 _

PROPOSED RULE rn - 3'2 0 )

C~qP!i 'f I ~

NUCLEAR REGULATORY COMMISSION 10 CFR Part 70

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RIN 3150 - AF22 AC Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material AGENCY: Nuclear Regulatory Commission .

• ACTION: Final rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations governing the domestic licensing of special nuclear material (SNM) for licensees authorized to possess a critical mass of SNM, that are engaged in one of the following activities: enriched uranium processing; fabrication of uranium fuel or fuel assemblies; uranium enrichment (other than certified existing gaseous diffusion plants); enriched uranium hexafluoride conversion;

• plutonium processing; fabrication of mixed-oxide fuel or fuel assemblies; scrap recovery of SNM; or any other activity involving a critical mass of SNM that the Commission determines could significantly affect public health and safety or the environment. The amendments establish performance requirements, require affected licensees to perform an integrated safety analysis (ISA) to identify potential accidents at the facility and the items relied on for safety necessary to prevent these potential accidents and/or mitigate their consequences; require the implementation of measures to ensure that the items relied on for safety are available and reliable to perform their function when needed; require the safety bases to be maintained, and changes reported to NRC; allow for licensees to make certain changes to their safety

program and facilities without prior NRC _a pproval; require reporting of certain events; and require the NRC to perform a backfit analysis under specified circumstances.

EFFECTIVE DATE: The final rule, with the exception of§ 70.76, is effective (IAser:t ~g eays .

~ 1~ :iooo after pabllcat1011 of this final rule). Section 70.76 will become effective after the issuance of staff guidance for the implementation of that provision. Once such guidance has been developed, the NRC will publish a Federal Register notice specifying the effective date of

§ 70.76.

FOR FURTHER INFORMATION CONTACT: Theodore S. Sherr, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-7218; e-mail tss@nrc.gov, Heather Astwood, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-5819; e-mail hma@nrc.gov, or Andrew Persinko, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-6522; e-mail axp1@nrc:gov.

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SUPPLEMENTARY INFORMATION:

I. Background II. Public Comments on Proposed Rule Ill. Changes from the Proposed Rule IV. Section-by-Section Analysis of Part 70 Amendments V. Finding of No Significant Environmental Impact: Availability VI. Paperwork Reduction Act Statement VII. Pubilc Protection Notification VIII. Regulatory Analysts IX. Regulatory Flexibility Certification X. Voluntary Consensus Standards XI. Backfit Statement XII. Small Business Regulatory Enforcement Fairness Act XIII. List of Subjects I I. Background On July 30, 1999 (64 FR 41338), the Commission published a proposed rule for public comment that would amend Its regulations governing the domestic licensing of SNM for certain licensees authorized to possess a critical mass of SNM. The Commission's action was in response to a Petition for Rulemaking, (PRM)-70-7, submitted by the Nuclear Energy Institute (NEI), which was published on November 26, 1996 (61 FR 60057). The proposed rule was 3

intended to grant the NEI PRM In part and would modify the petitioner's proposal. The majority of the proposed modlficatlons to Part 70 were included In a proposed new Subpart H, "Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuctear Material." These modifications were proposed in order to increase confidence in the margin of safety at the facilities affected by the rule.

In developing the proposed rule, the Commission sought to achieve its objectives through a risk-lnfonned and perfonnance-based regulatory approach that inctuded: (1) the identification of perfonnance requirements for prevention of accidents or mitigation of their consequences; (2) the perfonnance of an ISA to Identify potential accidents at th'e facility and the items relied on for safety; (3) the implementation of measures to ensure that the Items

• relied on for safety are avallable and reliable to perfonn their function when needed; (4) the maintenance of the safety bases, lnctudlng the reporting of changes to the NRC; and (5) the allowance for licensees to make certain changes to their safety program and facilities without prior NRC approval.

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The 75-day public comment period on the proposed rule ended on October 13, 1999.

During and after the public comment period, the NRC staff posted on the NRC web site revised versions of the draft Standard Review Plan (SAP) that would implement the proposed requirements (I.e., on August 4, 1999, a complete draft SRP was posted, and revised chapters, taking into account comments received, were posted during the period March 16 -

April 3, 2000). In addition, three stakeholder meetings were held to discuss the SAP (September 14-15, 1999, February 9, 2000, and April 18-19, 2000).

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• II. Public Comments on Proposed Rule In preparing the final rule, the NRC staff carefully reviewed and considered more than 90 comments on the rule, included in 9 individual letters filed during the public comment period. In addition, 13 submittals containing more than 200 specific comments were received on the SRP. To simplify the analysis, the NRC staff grouped all written comments on the rule into the following major topic areas: Performance Requirements and Design Criteria; Content of Applications and ISA Summary; Safety Program; Change Process, License Renewal and Backflt; Definitions; and Miscellaneous. A more detailed analysis is also documented as an attachment to SECY-00-0111. A review of the comments and NRC staff's responses follow:

• A. Performance Requirements and Design Criteria Comment A.1 : One commenter stated that the proposed rule should specify dose limits for anticipated occurrences similar to those In§§ 72.104 and 72.106. This part of the rule would then cover the range of likelihood (anticipated, likely, unlikely, and highly unlikely) of potential accidents that could occur at nuclear fuel cycle facilities. This could result In an increase In thcl number of structures, systems, and components (SSCs) relied on for sa{ety and would impact the design, operation, and licensing of the mixed-oxide (MOX) facility.

Response: No change in the rule language has been made. The dose limits for normal operation are contained In 10 CFR Part 20 [viz., 0.05 Sv (5 rem) Total Effective Dose Equivalent (TEDE)/yr for a trained worker]. The NRC staff views "anticipated occurrences" to be conditions of normal operations, and believes that the measures currently used by Part 70 licensees to comply with Part 20 have been and will continue to be successful in protecting workers and the public during normal operations. Thus, there is insufficient justification to 5

require identification of 'items' to demonstrate _compliance with Part 20 during normal operations.

Comment A.2: One commenter proposed that the NRC maintain consistency with past precedent (i.e., the Commission's rationale In Part 60) and eliminate the specific worker dose limits in Part 70.

Response: No change In rule language has been made. The regulatory experience and !ndustry events that initiated the effort to add a systematic accident analysis to Part 70 primarily involved health impacts to workers as opposed to the public. The NRC staff believes that the rule's focus on both the potential impacts on workers and the public is appropriate.

Based on the discussions and correspondence with the industry and public during development of the proposed rule, and all other comments on the proposed rule, there appears to be general consensus on this approach.

Comment A.3: One commenter stated, In response to the Federal Register notice request for comments on the cl~rlty and effectiveness of the language used (per June 1, 1998, Presidential Memorandum), that the language in § 70.61 (b} and (c) could be substantially clearer; the commenter provided an alternative plain language version of this 1 section.

Response: The language In the proposed rule was written in response to public comments to focus on risk (i.e., likelihood times consequence) of accidents. The language has been changed, In response to the comment, to provide additional clarity. The proposed revisions provided by the commenter, however, are not merely editorial but represent 6

substantive changes. They appear to have eliminated the concept of limiting risk, and ,_J instead, focused on the likelihood of accidents. The revised language in the rule attempts to retain the emphasis on controlling accident risks within appropriate performance requirements.

Comment A.4: Three commenters expressed concerns about how the worker dose limits In§ 70.61 (f) would be applied to "co-located workers." One commenter suggested that the performance requirements In§ 70.61 consider the indMduals working in the nearby facilities as public when performing an accident analysis to determine the consequences of the accidents that may occur at the facility. The commenter concluded that this would result In a more stringent application of safety requirements for the prot~ction of workers, (e.g.,

additional Items relied on for safety) at the MOX Fuel Fabrication Facility, Pit Disassembly, Conversion Facility, Immobilization Facility, and any other nearby DOE facilities, and would also have a substantial impact on the-cost of the MOX facility. A second commenter agreed with this assessment, noting that a worker (as defined In§ 70.4) who leaves the controlled area to perform a work-related function would have to be treated as a member of the public when performing an ISA and would be subject to the more stringent public radiation exposure llmits. OutsJde of the controlled area the TEDE limit of 1 mSv (0.1 rem) for members of the public would apply [cf. 10 CFR 20.1301 (a)(1)] rather than the annual TEDE occupational dose limit of 50 mSv (5 rems) (1 O CFR 20.1201 ) .. According to this commenter, this problem has already arisen at the Hanford Tank Waste Remediation System where co-located workers are governed by the appreciably lower public dose limits. A third commenter agreed with the above positions and also stated that the NRC intends to consider i~dividuals outside of the controlled boundary as workers if they are subject to Part 20 requirements. The commenter 7

noted, as did the first commenter, that DOE requirements in 10 CFR Part 835 provide an equivalent level of protection, such that co-located workers -- who are subject to the requirements of either Part 20 or 10 CFR Part 835 - should be considered "workers,"

provided the licensee can demonstrate the abillty to provide management measures (e.g.,

notification, evacuation, etc.) in the event of an emergency.

Response: NRC regulations do not specifically address personnel designated as "co-located" workers. In response to the comments, the first sentence In § 70.61 (f) was changed to read as follows: "Each licensee must establish a controlled area, as defined in § 20.1003.

In addition, the licensee must retain the authority to exclude or remove personnel and property from the area." The licensee can set the controlled area at any location around Its facillty as long as it maintains control of that area as specified in Part 20 and retains the authority to exclude or remove personnel and property from the area. If the controlled area included the nearby Department of Energy (DOE) facilities, then NRC would consider the personnel working at those facilities to be "workers" for the purposes of the performance requirements of § 70.61, provided the conditions of § 70.61 (f)(2) are met. The DOE and its contractors could satisfy these conditions by documenting their compliance with the

• requirements of 10CFR 19.12(a)(1 )-(5). To emphasize that the § 70.61 (f)(2) requirements, regarding 10 CFR Part 19 training, can be satisfied in combination with existing training, rather than separate training solely devoted to 10 CFR Part 19, 10 CFR 70(f)(2) has been changed to read: "Provides training that satisfies 10 CFR 19.12(a)(1)-(5)". To emphasize that the training provided to satisfy§ 70.61 (f)(2) requirements includes making Individuals aware of the risks associated with accidents involving the licensed activities as 9etermined by the 8

ISA, the word "to" was changed to "and," so that it now reads "to these individuals and ensures that they are aware of the risks associated with accidents".

Regarding the concern about the ,worker who leaves the controlled area, the risk levels

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of§ 70.61 for the public pertaln to any individual, includfng workers, outside the controlled area. On the other hand, with respect to the applicability of the Part 20 occupational dose limit of 0.05 Sv (5 rem)/yr TEDE, a worker can receive an occupational dose and be subject to the Part 20 occupation.al limit, regardless of his location - Including activities outside the controlled area. The "assigned duties performed in the course of employmenr Is the distinguishing factor for radiation workers consistent with the definition of "occupational exposure" in 10 CFR 20. The changes to Part 70, including the "worker" definition, do not affect this. In this comment, the relationship between Part 20 annual limits for radiation exposure and the§ 70.61 standards for a forward-looking severe accident assessment have been misinterpreted. Part 70 revisions do not limit doses outside a controlled area to 1 mSv (0.1) rem/yr.

Comment A.5: One commenter recommended that baseline criterion (8) be. rewritten

• as follows: "the design of Items relied on for safety must provide for adequate inspection, testing, and maintenance, or adequate training, testing and qualification for personnel whose activities are relied on for safety, to ensure their availability and reliability to perform their function when needed."

Response: No change In rule language nas been made. The baseline design criteria are applied from the outset of new design work and are primarily focused on physical design and facility features. The intent is to achieve a conservatively designed facility tolerant of 9

both upsets and human errors. Adequate training, testing, and qualification, as noted in the comment, will be required as management measures under§ 70.62, but the NRC does not see a need for the facility physical design to Incorporate such training, testing, and qualification of personnel.

Comment A.6: One commenter stated that the baseline criterion on environmental and dynamic effects[§ 70.64a{4)] Is unclear. For example, the commenter questioned the need

,for a formal Equipment Environmental Qualification Program, similar to that required under 1O I

CFR 50.49 and Regulatory Gulde 1.89. According to the commenter, the NRC should clarify this requirement and should not impose requirements that may not be appropriate or necessary ,because of the nature of the processes at non-reactor nuclear facilities.

Response: No change In rule language has been made. The baseline design criterion on environmental and dynamic effects does not require a formal Equipment Environmental Qualification Program, similar to that required under 10 CFR 50.49 and Regulatory Guide 1.89. This criterion applies only to new facilities and new processes and Is intended to ensure that potential ambient conditions are considered during the design of the facility.

Comment A.7: Two commenters had concerns regarding the defense-in-depth definition in§ 70.64. One commenter stated that the definition does not reflect the defense-in-depth design philosophy as defined in WASH-1250, "The Safety of Power Reactor and Related facilities," which outlined three levels of safety concepts in the design of a nuclear facility. According to the commenter, the definition presented in §§ 70.64(b)(1) and 10

(2) oversimplifies and does not adequately represent the Implementation of the defense-In-depth philosophy in the design. In particular, the commenter noted that the preference for engineered controls over administrative controls and features that reduce challenges to Items relied on for safety are only partially implemented in the' concept [of defense-In-depth]. Another commenter agreed, stating that§ 70.64(b}(1) appeared unnecessarily prescriptive by dlscou-raglng a licensee from using anything but an engineered safety control. According to this commenter, as long as the licensee can satisfactorily I

I demonstrate that an administrative safety control or a system of administrative and engineered controls will enable the performance criteria to be satisfied, the choice of items

' relied on for safety and the nature of 'defense-in-depth' practices that is applied should be flexible. The commenter's view Is that this flexibility, In the grading of defense-in-depth safety I

concepts, would be consistent with the ability granted a licensee to grade all aspects of Its safety program [cf. § 70.62(a)].

Response: With respect to the footnote to§ 70.64(b) that describes defense-in-depth, which applies to new facilities and new processes at existing facilities, the NRC staff believes that It does reflect the defense-In-depth design philosophy as defined In WASH-1250.

Further, It reflects the Commission's current guidance on the relationship between defense-in-depth and risk-informed regulation that Is discussed In the Commission policy white paper, "Risk-Informed and Performance-Based Regulation." With respect to §§ 70.64(b)(1) and (2),

the NRC staff did not mean to imply that thes~ provisions encompassed the defense-in-depth philosophy.

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Comment A.8: One commenter recommended that the emergency capability baseline design criterion in§ 70.64(a)(6)(11) address on-site personnel (rather than all personnel). The commenter suggested that the rule language be rewritten as "Evacuation of on-site personnel; and...."

Response: The NRC agrees with the comment. The proposed change has been made and Is consistent with the intent of the original rule language.

Comment A.9: One commenter stated that the criticality performance objective in

§ 70.61 (d) Is not related to§§ 70.61 (b) or 70.61 (c); yet, the three conditions are all linked together. The commenter suggested that subpart (d) should be segregated from (b) and (c) if (d) Is preserved as an Independent entry (as would seem preferable). Otherwise (d) should be subsumed under* (b) and/or (c), and the regulatory basis for criticality prevention should be predicated on the risks and/or consequences of the accidents, rather than the presence of Initiator precursor, .Qfil se.

Response: No change In the rule langudge has been made. The NRC believes that

• a separate performance requirement for nuclear*criticallty prevention is appropriate. The NRC staff recognizes that many (but not all) nuclear criticality accidents would reasonably be expected to result in worker doses that exceed the high- and intermediate-consequence standards In § 70.61 (b) or (c). However, regardless of the dose directly resulting from the accident, an inadvertent nuclear criticality should be avoided. This is consistent with the Commission's goal to prevent Inadvertent critlcalltles, as reflected In the NRC Strategic Plan (NUREG-1614).

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8. Content of Applications and ISA Summary Comment 8.1: One commenter stated that the rule should not prescribe an acceptable level of detail required in the application, but should defer this Issue to the SAP. The commenter noted that, although progress has been made In certain areas (e.g., use of language such as "... types of accident sequences .... "), in§ 70.65(b)(6), which requires the applicant to list all items relied on for safety for high- and intermediate-consequence

. accidents, the required level of descriptive detail for items relied on for safety ("sufficient detaif) remains vague. The commenter recommends that information at the "systems level" should be required, rather than at the "componenr or "sub-component' level.  :,1 Response: The NRC disagrees with the comment. The current language permits the description of information at a systems level provided that there is enough detail to understand the function of the system In relation to the performance requirements. The degree of detail provided in the ISA Summary, with the other Information available to NRC

• staff, must be sufficient for the NRC staff to make the determination specified In§ 70.66 (I.e.,

that the performance requirements of the regulation are satisfied).

Comment B.2: One commenter stated that the list of items relied on for safety should not include procedures that the personnel must follow. According to the commenter, since procedures are constantly being adjusted, revised, and Improved, their Inclusion in the list of items relied on for safety would necessitate frequent revisions, to the ISA Summary, that may have little if any safety significance.

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Response: § 70.65(b)(6) requires a list, in the ISA Summary, briefly describing each item relied on for safety. It does not require procedures to be listed in the ISA Summary.

Therefore, the rule language permits the approach described In the comment. Typically, the actual personnel action would be regarded as an Item relied on for safety and this would be expected to be addressed In the ISA Summary.

Comment 8.3: Two commenters had concerns about the relationship among the ISA, the ISA Summary, and the safety program. One commenter recommended that the NRC clarify the relationship of the ISA Summary to the license and the safety basis to ensure consistency throughout the rule with the intent expressed _in§ 70.65(b). The commenter was concerned that the wording of § 70.65(a) is inconsistent with the idea, presented In -

§ 70.65(b), that the ISA Summary will not be incorporated in the license. The commenter suggested removing the language in § 70.65(a) that references the inclusion of the ISA Summary In the license application, since that requirement is adequately covered in

§ 70.65(b). The commenter also recommended that a discussion of management measures be included as part of. the ISA Summary. The second commenter stated that the rule implies that the ISA Summary, as part of *the safety program, Is part of the license. Further, the same commenter stated that the "Statement of Considerations" erroneously states that the results of the ISA must be submitted for NRC approval.

Response: The NRC generally agrees with the comment. The rule language in

§ 70.65(a) has been changed to remove the reference to the ISA Summary and management measures. This removes the implication that the ISA Summary Is part of the license. With respect to the relationship of the ISA Summary to the management measures, although under 14

the proposed rule, the elements of the ISA Summary did not explicitly include management measures, one of the elements [70.65(b)(4)J of the ISA Summary required information that demonstrates compliance with the performance requirements. Such a demonstration requires Information about management measures. As suggested in the comment, the language In

§ 70.65(b)(4) has been cla,ified to expllcltJy Include a description of the management *

• measures. With regard to the comment that the "Statement of Considerations" erroneously "states that the results of the ISA must be submitted for approval", the assertion that the "Statement of Considerations" Is erroneous is Incorrect - the "Statement of Considerations" Is accurate. In response to this comment, and to clarify the role of the ISA Summary in licensing determinations, changes have been made to§ 70.62(c)(3)(11) and§ 70.66. In particular, § 70.62(c)(3)(ii) has been modified to specificaJly state that the ISA Summary Is submitted for approval consistent with the "Statement of Considerations" for the proposed rule. Section 70.66,states that this submission will be approved If the Commission determines that "the applicant has complied with the requirements of§ 70.21, § 70.22, § 70.23, and

§ 70.60 through§ 70.65." The degree of detail provided In the ISA Summary [contents of the ISA Summary are described in § 70.65(b)] and the other information available, m_ust be

• sufficient for the NRC staff to make the determinat1on specified in § 70.66. To supplement staff understanding of information submitted, NRC may visit the facility during the licensing review to ensure a sufficient safety basis for operation.

Comment B.4: Two commenters were concerned with the broad nature of the requirement in § 70.65(b)(3) that seeks information on each process analyzed In the ISA, regardless of the risk associated with the process. According to one commenter, the ISA 15

Summary should only address those processes for which accident sequences have been Identified that would produce consequences that exceed the perfonnance criteria of § 70.61.

Response: The NRC staff needs some lnfonnation on each process analyzed In the ISA to assess completeness and quality of the licensee's ISA process and to understand and assess the completeness and functions of the Items relied on for safety. The degree of detail provided in the ISA Summary, together with the other lnfonnation available, must be sufficient for the NRC staff to make the detennination specified In § 70.66. In addition the infonnation is useful In conflnnlng the adequacy of emergency planning.

Comment 8.5: According to one commenter, § 70.65(b) implies that the ISA Summary Is a single document. In practice, the commenter noted that It will be a sequence of documents that cover the facility, and if multiple documents are submitted, they should all be in the same fonnat.

Respohse: The NRC agrees that multiple ISA Summaries are allowed for each facility resulting in the possibility that the ISA Summary will ~e a sequence of documents rather than consisting ?f a single document. The definition of ISA Summary in § 70.4 has been changed to reflect this process. NRC staff approval of individual documents (e.g., on a process basis) will be conditioned to allow for subsequent identification of system interaction effects that may be identified through the review of other ISA summary documents submitted.

Comment 8.6: One commenter stated that the requirement, in§ 70.65(b)(7), to provide infonnation on the locations of onsite chemicals, is unnecessary.

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Response: No change In the rule language has been made. Section 70.65(b)(7) does not require infonnation on the locations of onsite chemicals to be submitted to the NRC. The regulation requires a description of the proposed quantitative standards used to assess the consequences to an individual from acute chemical exposure to licensed material or chemicals produced from licensed material. This information is necessary to ensure safety and is consistent with NRC's Memorandum of Understanding (MOU) with the Occupational Safety and Health Administration (OSHA).

Comment B. 7: One commenter objected to the requirement to provide process descriptions, noting that American Institute of Chemical Engineers (AIChE) guidelines may result in the process being broken into "nodes" or "segments." The commenter suggested that the rule should specify the descriptions of segments or nodes that C?Uld only be combined into a process if the boundaries established for the hazard analysis match.

Response: The NRC agrees with the comment, but does not believe a change in the rule language is needed. The intent of the§ 70.65(b)(3) requirement is to provide process infonnation so that the NRC staff can understand: what activities are perfonned at the site that involve hazardous materials associated with or produced from licensed radiological material, including any use, storage, manufacturing, or handling of those materials; what was analyzed in the ISA; and the hazards identified in the ISA. The AIChE guidelines use the term "process nodes" with respect to Hazard and Operability Analysis (HAZOP} and define it as "sections of equipment with definite boundaries ... within which process parameters are investigated for deviations .... " In HAZOP analyses, the term "node" designates a pipeline or 17

vessel that has a common design intent. In meeting the§ 70.65(b)(3) requirement, several nodes may be combined.

c. Safety Program Comment C.1: Three commenters questioned the narrow definition of the safety program that is presented In§ 70.62(a) and recommended deleting It from the rule language.

According to the commenters, the safety program is broader than the three elements identified in § 70.62(a){1) as process safety information, ISA, and management measures.

The commenters noted that fuel cycle f acillty safety programs encompass the three elements identified plus all the other topics addressed In the license application. This indudes, for example, radiation safety, criticality safety, chemical safety, and fire protection, in addition to the three elements directly associated with the ISA.

Response: The NRC staff agrees In principle with the comment. The term "satety program," as used In§ 70.62 (a), is related to the elements needed to demonstrate compliance with the performance requirements in § 70.61. This safety program consists of

. process safety Information, ISA, and management measures. There is no intent to indicate that these elements represent the total safety program at the facility. Therefore, the rule language was clarified by changing "The three elements of the s_af ety program; namely process safety information, integrated sat ety analysis, and management measures, are described in paragraph (b) through (d) of this section ... " to "Three elements of this safety program; namely process safety Information, integrated safety analysis, and m8:nagement measures, are described in paragraph (b) through (d} of this section."

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Comment C.2: One commenter stated that the current proposed rule offers sufficient

-flexibility in selecting ISA methodoiogy so that a broad spectrum of facilities can be addressed and such that llcensees have flexibility to interface with their site processes, procedures and resources.

Response: The Commission agrees with the comment; theref~re, no change was made to the rule language with respect to ISA methodologies. The finaJ rule offers sufficient flexibllity in selecting an ISA methodology that can be used to analyze a facility's site, processes and procedures.

Comment C.3: Two commenters wera, concerned about the lmplementatio~,of the finaJ rule, and, in particular, the time frame for compliance with those aspects of the rule not related to the completion of the ISA and the submittal of the ISA Summary. One commenter, citing the experience when Part 20 was revised, recommended an effective date sufficiently far into the future so programmatic changes could be implemented at the operating facilities and any necessary conforming license amendments coul9 be completed. RegardJfl9 the

• latter Issue, both commenters cited 10 CFR 20.1008 as an example of how potential contradictions between license applications and regulations could be addressed. One commenter recommended Including an additional provision of this type, especially In light of license conditions that have been added to licenses recently renewed by the NRC.

Response: The NRC agrees with the comment. In§ 70.76(a), it states that "this provision shaJI apply for subpart H requirements as soon as the NRC approves that licensee's ISA Summary pursuant to § 70.66. For requirements other than Subpart H, this provision 19

applies regardless of the status of the approval of a licensee's ISA S~mmary." In addition, Appendix A was revised to Include the following: "Licensees must comply with reporting requirements In this appendix, except for (a)(1 ), (a)(2), and (b)(4), after they have submitted an ISA Summary In accordance with § 70.62(c)(3)(1i). Licensees must comply with (a)(1 ),

(a)(2), and (b)(4) after (Insert 30 days after publication of this final rule)." In addition,

§ 70.62(c)(3)(ll) was revised to further clarify Implementation schedules for existing licensees.

Comment C.4: Two commenters stated that a graded approach should be used In

• determining the management measures that need to be applied to items relied on for safety.

One~commenter recommended that the language in§ 70.62(d) should be changed as follows "The measures applied to a particular engineered or administrative control or control system may be graded commensurate with the reduction of the risk attributable to that control or control system.* The other commenter recommended that other factors besides risk Including consequences, life cycle, and magnitude of hazard involved, should be used to determine appropriate management measures.

Response: The NRC agrees with the cumment and has made the suggested change to the rule language in§ 70.62(d). Regarding the question of considering other factors besides "risk," the NRC notes that the grading of measures to consequences, life cycle, and

• magnitude of hazard, Is part of grading the measures to risk. The phrase used In the rule -

"commensurate with the reduction of risk attributable to that item" -- does not imply requiring a quantitative determination of the risk significance of any particular item relied on for safety.

The rule Is non-prescriptive regarding the grading approach and criteria to be used, allowing applicants to propose such details.

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Comment C.5: One commenter stated that the 4-year period for conducting the ISA and for modifying the facility to address any Identified unacceptable performance deficiencies

(

may be too short and recommended a 5-year period Instead. According to the commenter, a 5-year time-frame would be consistent with the time allowed for existing licensees that have committed, by license condition, to perform ISAs. The commenter also recommended that the period should start on the date when the NRC approves the plan required in § 70.62(3)(1),

noting that If the clock starts on the effective date of the rule and the NRC takes one year to approve the ISA plan, the licensee will be unduly hampered. In addition, the commenter stated that there should be some Incentive for the NRC to complete Its approval process in a timely manner and recommended Imposition of a 90-day limit for NRC to Issue a decision on the acceptability of a licensee's ISA approach. The commenter also recommend~d that appropriate and sufficient time be allowed for the licensee to present a plan to the NRC and to implement the plan to correct any identified unacceptable performance deficiencies.

Response: Regarding the proposal for a 5-year period for conducting the ISA and I

correcting all unacceptable def lciencies, the NRC believes that the 4-year period proposed in

• the proposed rule is reasonable. However, NRC recognizes that there may be some instances where modifications resulting from the ISA cannot be completed within the 4 years specified and has modified§ 70.62(c)(3)(11) to accommodate these instances by clarifying that NRC may approve extensions for reasons that are beyond the control of the licensee.

Regarding the licensee being unduly hampered because of the time required for the NRC staff to approve the plan required by§ 70.62(c)(3)(1), the NRC staff expects to complete the licensing review wl,thin 90 days, assuming that the information submitted is complete.

However, the time it takes the NRC to approve the plan will depend on the quality of the plan 21

submitted by the licensee. In addition, current industry development of an ISA Summary guidance document should facilitate the licensing review process.

Comment C.6: One commenter stated that the plan required in§ 70.62(c) which should be submitted within 6 months of the effective ,,

date of the rule, ~hould pertain only If a licensee has not already completed the actions outlined In§ 70.62(c)(3)(ii).

Response: The implementation plan and the ISA must satisfy the requirements in the final rule. If the actions outlined In§ 70.62(c)(3)(ii) have been completed, then all that would be required to satisfy§ 70.62(c)(3)(i) is submission of a description of any additional work that must be performed to meet the requirements in Subpart H of the rule, or a confirmation that the work submitted meets the requirements In Subpart H of the rule.

Comment C.7: Four commenters disagreed with the requirement in§ 70.62(a)(3) to establish and maintain a log of failures of Items relied on for safety. One commenter stated that the requirement should be re~rltten to be performance-based rather than prescriptive.

The commenter noted that most licensees have an incident reporting and corrective action system, which is used for all activities at the facility. As long as these systems meet the performance objective, it seems unnecessary for the rule language to be prescriptive in how It Is met. Another commenter agreed, stating that It is inappropriate to Impose this extra record-keeping burden on the licensee, because the licensee already has to generate records of this nature to manage its business and another different log is unnecessary work. Another commenter noted that because of the reporting requirements of§ 70.62(a)(2) and 22

§ 70.74(a)(1), the NRC will already possess all of the lnfonnation sought in the "log" of

§ 70.62(a)(3).

Response: The NRC generally agrees with the comment that maintenance of the failure log would be unneC&SSarily prescriptive. Regarding the concern about prescriptlveness, the rule has been revised to eliminate the requirement for licensees to establish and maintain a specific log of infonnation developed *and maintained elsewhere.

However, the final rule requires that data be readily retrievable and available. This Information Is necessary to evaluate the reliability and availability of Items relied on for safety, the likelihood.of failure of the items, and the effectiveness of management measures implemented by the licensee. The NRC also anticipates this information will be reviewed during periodic inspections by NRC as part of the revised oversight process that is being developed. Regarding the redundancy of reporting, the rule currently requires the licensee to rep~rt only any loss or degradation of Items relied on for safety that results in failure to meet the performance requirements of § 70.61. The requirements of§ 70.62(a)(3) include a much broader set of Items, including all Items relied on for safety or management measures that

• have failed to perform their function.

D. Change Process, License Renewal, and Backflt Comment D.1: Five commenters were concerned about the requirement In

§ 70.72(d)(1) to submit changes to the ISA Summary every 90 days. Two commenters stated that an annual update [similar to the annual Final Safety Analysis Report updates for reactors per 10 CFR 50.71 (e)] should suffice, considering that the potential consequences of reactor 23

accidents are significantly greater than those at fuel cycle facilities. One commenter stated that an annual update to the ISA Summary would be consistent with the reporting requirements (for changes to records) of§ 70.72(d)(3). Another commenter stated that the 9Q-day reporting of changes Is entirely too frequent, which would mean that the facility and the NRC would always have change reporting in progress. According to the commenter, there is no need for NRC to have this "real-time" knowledge; rather, It is only important that the licensee have "real-time" knowledge. The commenter noted that the NRC only needs reasonably current knowledge, because the current ISA Is available and accessible at the site. The commenter believes that a 12-month to 24-month update for reporting, as used in other places, is satisfactory and more efficient, noting that this seems clearly justified based

• on the fact that all the lnfonnation Is avallable at the site and accessible to the NRC at any time.

Response: The NRC agrees with the comment that submitting updates to the ISA Summary to the NRC can be less frequent than required in the proposed rule. The final rule requires only annual reporting within 30 days after the end of the calender year during which the changes occurred.

Comment D.2: One commenter noted that, under§ 70.72, the NRC should define "periodically" in the context of reporting of changes made to SSCs etc.

Response: The NRC detennined that no change to the reporting requirements is necessary In response to the comment. The comment referenced language in the "Statement 24

of Conslderationsn not the rule. The specific reporting requirements were defined in the proposed rule and are included, as revised, in the final rule.

Comment D.3: Two commenters were concerned about the footnote in§ 70.72(c) that attempts to explain new types of accident sequences. Both commenters stated that the language in the footnote would require nearly all process changes to be approved by NRC through a license amendment, which would be in conflict with the overall objectlves for the ,

proposed rule. Both commenters recommended that the footnote be deleted.

Response: The NRC agrees that the footnote did not successfully clarify the definition of "new types of accident sequences." Thus, the footnote has been deleted from,the final rule. The NRC staff will develop a guidance document, with Input from stakeholders, to describe an acceptable change process that meets tt.e requirements of the final rule In more detail. The degree of detail provided in the ISA Summary, together with the other Information available, must be sufficient for the NRC staff to make the determination specified in § 70.66.

In addition, the .NRC staff h~d added a discussion to Chapter 3 of the SAP to describe an acceptable level of detail in the Identification of the types of accident sequences.

Comment D.4: Three commenters were concerned about the requirements, in § 70.72.

regarding configuration management and the overly broad process for making changes at

,licensed facilities. One commenter stated that the requirements, as written, apply to all site, structures, processes, systems, equipment, components, computer programs, and activities of personnel, regardless of safety significance. The commenter noted that compliance with

(

these requirements would appear to require configuration management and change control 25 ,

applied to everything on the site of the licensed facility; that could Include the wastewater treatment facility, a laser facility, the administration building, maintenance of the shrubbery, etc. Every change would require an evaluation and a summary submitted to the NRC, even though inclusion In the change control process would make no contribution to the safety of licensed operations and would impose an undue burden on the licensee. To remedy this, the commenter recommended that the configuration and change process be limited to any achanges to the site, processes or Items relied on f~r safety as described in the ISA Summary." Another commenter agreed, stating that the requirement is too broad and all*

encompassing and would require configuration management evaluation of changes having no or absolutely minimal effect on health and safety (e.g., office remodeling, planting of shrubbery, changing paint colors). The qommenter suggested that rather than control every change by means of configuration management, the licensee should first rely on Internal procedures to screen any proposed changes Initially for their potential safety significance.

Response: No change In the rule language has been made. The emphasis of this requirement is clear1y on licensed operations and the associated safety controls. If a licensee I \

has established a configuration management system in accordance with§ 70.72(a), It is Important the licensee use the system to evaluate every change made at a facility that could affect safety (i.e., generally not shrubbery, paint color) to ensure that any Impacts from those changes on the safety of operations Is identified, considered, documented, before implementing the change. In some cases, the analysis would be trivial because no known hazards would be involved In the change (e.g., certain changes in the administration building, or changes to shrubbery). Often it is clear that there are no safety implications associated with the proposed change. However, there may be special cases in which apparently minor 26

changes could adversely affect safety, such as installation of a drinking water fountain in a radiological control area. In addition, every change which is assessed in the configuration management system does not need to be submitted to the NRC. Section 70.72(d)(2) states that only those changes to records required by § 70.62(a)(2) need to be submitted. These would include changes to the process safety information, ISA, and *management measures. In addition, with respect to the use of an "initial screening" mechanism,' the NRC staff considers an initial screening to assess the safety impact of a change to be part of an evaluation, as called for in§ 70.72(a). In some cases, this screening will be sufficient.

Comment D.5: One commenter stated that§§ 70.72(c)(1)(i), (c)(2), and (c)(3) are wrong to use the ISA Summary as the decision-making document. The commenter noted that the ISA, the detailed licensee-generated information and evaluations that the licensee uses to manage its program, comprises the information base for decisions. Summaries only provide a general level of information about the more important elements of the safety system for operations as determined under the licensed program.

Response: No change in the rule language has been made. The ISA Summary is prepared based on the ISA, and contains key information that is directly related to facility safety, such as a list of items relied on for safety, a description of hazards identified in the ISA, and a general description of the types of accident sequences. The contents of the ISA Summary are described in§ 70.65(b). The NRC staff could review the adequacy of changes using the ISA instead of the ISA Summary, but this approach would require submission of a greater amount of information to NRC and would pose an unnecessary burden on the licensee. (Also, see response to Comment B.3.)

27

Comment 0.6: Two commenters are concerned about the annual requirement in

§ 70. 72{d)(3) to submit a brief summary of all changes to the records required by

§ 70.62(a){2). According to one commenter, the submittal would cover process safety information [§ 70.62(b)] including procedures, drawings, and detailed equipment lists. The commenter does not believe the NRC requires a summary of changes to this type information.

A second commenter agreed, stating that the wording of this section will Inadvertently and significantly expand the information that would have to be reported. In particular, the view was expressed that§ 70.72(d) would require the licensees to submit voluminous information that could include the update to process safety information, Including drawings, flow process diagrams, and piping and instrumentation diagrams. The commenter suggested that this section should be reworded to read: "a brief summary of all changes to the integrated safety analysis and /SA summary. that are made without prior Commission approval, must be submitted to the NRG every 12 months...."

Response: No change in the rule language has been made. The regulation currently requires,submisslon of "... a brief summary of all the changes to the records required by

§ 70.62(a)(2) .... " This does not require the submittal of actual charts and drawings but a written summary of the changes made. For the reasons cited in the response to comment 0.1, it Is important that the NRC be knowledgeable of changes made to this information.

Comment D.7: One commenter noted that, unlike§ 50.59, the requirements of§ 70.72 do not call for the submittal of a brief description and summary safety evaluation for each change. The commenter believes that the NRC would benefit from a description of changes made to the ISA Summary. Accordingly,§ 70.72 should require brief descriptions and 28

summary safety evaluations of each change made pursuant to§ 70.72 and require that an updated ISA Summary be provided on a biennial basis.

\

Response: No change in the rule language has been made. The brief summaries of changes submitted under the requirements of § 70. 72(d)(2) would be expected to Include an explanation of each change, the reasons why the change was made, and why it did not require pre-approvaJ. This Information will be included in a guidance document to be developed. The NRC staff views this as sufficient and does not anticipate the need for licensees to submit a summary sat ety evaluation for ~ach change, as Jong as each change has been made In accordance with the final rule and the approved process. -

Comment D.8: Two commenters questioned the current timeframe {10 years) or the need for renewal of licenses, suggesting that the new rule, In effect, resulted In a '1iving license. One commenter stated that If a "living license* Is truly the outcome as described in the "Supplementary Information," renewal periods as long as 20 years would be appropriate.

The other commenter noted that with updates required every 12 months there Is no real need for the NRC to renew the license - it only becomes a maintenance chore to confirm

')

periodically that the licensing basis remains intact. The commenter believes that the livlng license concept provides advantages for the NRC and the licensee.

Response: Although the NRC generally agrees with those comments, no change in the rule language has been made. A specific time period for renewals is not specified In Part 70 and to establish one in the rule would require consideration of many factors, such as compliance with the NatlonaJ EnvironmentaJ Policy Act and the impact of the loss of 29

commitments linked to license renewal, that were not addressed in the current rulemaking (e.g., financial assurance for decommissioning). Establishment of a new tern, for licenses (e.g., 20 years) In 10 CFR 70 would require an analysis of these factors and an opportunity for public comment The NRC staff will evaluate whether a longer tern, for fuel cycle licenses is appropriate in light of the new requirements In Subpart H. In any case, even if NRC ultimately declines to extend the term of the fuel cycle licenses (nominally 1O years), the burden of llcens~ renewal.should be significantly reduced because the licensee. will be required to maJntaln current the ISA Summary, items relied on for safety, and management measures.

Comment D.9: Five commenters recommended that a backflt provision similar to that In 10 CFR 50.109 or 10 CFR 76. 76 should be included In the final rule. One commenter stated that the backfit provision should apply to current proposed changes at existing facilities. Another commenter stated that the backfit provision should be Immediately effective for those processes or parts of an existing facility for which an ISA has been completed. A third commenter favored an Immediately effective backfit provision. However, as an alternative, the commenter would make the provision effective for facllities or systems for which the ISA has been completed and the ISA Summary submitted to the NRC. A fourth commenter stated that deferring consideration of a backfit provision would be evading an extremely important issue, expressing the view that it Is vltal that a fom,al, systematic, and disciplined review of new, changed, or differing positions that could backfit existing facilities be applied to in9rease regulatory certainty. According to the commenter, no change to the backfit language in 10 CFR 50.109, which has been used successfully to control backfits at power reactors in the past, Is needed to allow for qualitative analysis. 10 CFR 50.1 09, which 30

the commenter endorses, is viewed as neither a quantitative nor a qualitative backfit provision. In contrast to the statement made in the Statement of Considerations of the proposed rule, the commenter does not believe that a comprehensive risk baseline is necessary before reasoned judgments can be made on the benefits and risks of a proposed backflt.

Response: The Commission agrees that regulatory stability and certainty can be improved by establishing a b~ckflt provision for fuel cycle facilities covered by Subpart H of the final rule. Consequently, NRC has Included a backfit provision in the rule in§ 70.76. The

• wording of§ 70.76 ls similar to the current language in§ 76.76. For requirements other than Subpart H, this provision will apply immediately after NRC publication of backfit guidance. For Subpart H requirements, this provision shall apply for a licensee as soon as the NRC approves that licensee's ISA Summary pursuant to§ 70.66. The NRC will publish guidance that will address the qualitative versus quantitative analysis issue and consideration of chemical risks. The NRC staff anticipates completing this guidance within six months of the publication of the final rule. Under the§ 70.76 backfit provision, a backfit analysis Is not

• required for modifications necessary to bring the facility into compliance with the rule, including the performance requirements in Subpart H. The subject of backfit is discuss*ed in more detail in an attachment to SECY-00-0111.

E. Definitions Comment E.1: One commenter recommended a change (from 4 percent to 5 percent enrichment) In the definition of a critical mass of SNM to reflect the higher enrichments that are currently in use.

31

Response: The definition of critical mass of SNM In Part 70 is used solely to detennine when Subpart H applies. To emphasize this point, the definition was changed to Include the phrase, "for purposes of subpart H: The definition, including the 4 percent figure, Is identical to that used in § 70.24, which requires criticality accident alanns and other related measures.

Comment E.2: Regarding the Issue of "reasonable assurance," two commenters stated that, in the definition of available and reliable to perform their function when needed, the use of the tenn "ensure" implies a level of certainty that Is unrealistically high. Both commenters recommended replacing the tenn "ensure" with the tenn "provide reasonable assurance." _

One commenter also recommended remmring the word "continuous" from the definition, which.

would now read " ... means that... ltems relied on for safety will perform their Intended safety function when needed and management measures will be Implemented to provide reasonable assurance of compliance with the performance requirements of§ 70.61."

Response: The definition was revised to remove the word "continuous," but no change was made regarding "ensure." With respect to "ensure," the proposed rule language does not Indicate a level of certainty that is unrealistic. The tenn "ensure" is used extensively throughout NRC's regulations in the context of a licensee's obligations to connote "make sure" or "make certain." Specifically, elsewhere in Part 70 alone, the tenn is used in this context eight times:§§ 70.24(a)(3), 70.320), 70.38(g)(4}(iii), 70.51(a)(10), 70.52{c), 70.57(b)(3),

70.57(b)(4), and 70.57(b)(6). Whereas, the tenn "reasonable assurance" is used just once In Part 70, in§ 70.23(b), to describe the level of assurance that the Commission must find In order to approve construction. The use of "ensure" in the definition of "available and reliable to perform their function when needed" in § 70.4 is appropriate. In short, licensees "ensure" 32

and the Commission determines "with reasonable assurance." Regarding the issue of "continuous compliance," the definition of "available and reliable" in § 70.4 has been modified to delete the word "continuous." This change recognizes the concept that a failure of an Item relied on for safety does not automatically Infer a failure to meet the performance requirements of§ 70.61. In addition, the NRC recognizes that Items relied on for safety may temporarily not be available (i.e., not continuous) when taken out of service for maintenance or functional testing; however, the performance requirements must still be met. A discussion has been added to Chapter 3 in the SAP to address the relationship of failures of Items relied on for safety to meeting the performance requirements.

Comment E.3: One commenter stated that there is a "disconnect' regarding_ the definition of the term items relied on for safety and recommends that the term be replaced by the term Measures relied on for safety.

Response: The reason for the comment is not clear, but perhaps the commenter objects to the use of the term "'item" to refer to a personnel action. Part 70 does, In fact, allow human actions to be Items relied on for safety and permits flexibility in determining how the items and measures are defined. Consequently, the Commission has retained the original text in the final rule. (See related Comments 8.2. and E.4.)

Comment E.4: One commenter was concerned that the term items relied on for safety includes "activities of personnel," and proposed changing the definition in 70.4 to limit items relied on for safety to "structures, systems, equipment, and components." According to the commenter, It is reasonably straightforward to classify physical items as being relied upon for 33

'1 safety, and to apply graded quality assurance controls, including management measures, to design, construction, operation, and maintenance, etc., of those physical Items, based on their respective safety functions. The commenter stated that It can be confusing to try and classify and grade Items when they include "personnel activities," since an activity has little Importance absent the context of its influence on a physical Item's safety function. Removing "personnel activities" from the definition of Items relied on for safety w~uld not limit their Importance but rather, would put activities in context with the structures, systems, equipment, or components to which they are related, without necessitating a change In the balance of the proposed rule. The commenter stated that removing personnel activities from the definition of items relied on for safety will also help address the concern raised (In comment 8.2) regardln the treatment of procedures as Items relied on for safety.

Response: No change was made to the rule language. Human actions that are relied on to prevent an accident (I.e., administrative controls) are as important as the "physical Items* needed to prevent an accident. Just as there are measures (e.g., maintenance, configuration management) needed to ensure the availability and reliability of physical controls, there are analogous measures (e.g., training, procedures) needed to ensure the availability and reliability of human actions. Graded approaches that can be applied to the maintenance of a physical control depending on the risk significance of the control could also

/

be applied to the training of workers who perform safety functions, depending on the risk significance of the human's actions. Although the reliability of engineered controls may be higher than administrative controls, the final rule allows licensees the flexibility to employ both engineered as well as administrative controls 34

Comment E.5: One commenter stated that the NRC should define the terms likely, unlikely, highly unlikely, and credible in the rule so that there will be one set of definitions applied to aJI nuclear fuel facilities. The commenter stated that this will minimize the Interpretation and application of these terms in the ISA.

Response: No change in the rule language has been made. Part 70 applies to different types of fuel cycle facilities, some of which are more complex and have more accident sequences than others. Accordingly, since the application of the terms in the rule will be necessarily specific to the individual context in which they are._applied, the development of a definition for these terms In the rule language is impracticable. The Commission, however, will provide general guidance on the application of the terms unlikely and highly unlikely in the SAP to aid licensees In Implementing the provisions of the rule.

Comment E.6: One commenter recommended a change In the definition of worker. In particular, the language "... exposure to radiation and /or radioactive material from licensed and unlicensed sources of radiation* would be replaced with "... exposure to radiation and /or

• radioactive material from licensed sources of radiation, and radiation from man-made non-regulated sources (e.g., an lndlvidual).n As originally defined, persons who are subject to occupational doses from natural sources of radiation, (e.g., airline pilots and astronauts subject to high cosmic background might be included, whereas workers involved with the possession or use of unlicensed radioactive materials might not be). The'commenter stated that the proposed change removes this source of confusion.

35

Response: The NRC staff agrees In principle with the comment. However, the commenter's proposed change does not elimi~ate the confusion (e.g., some man-made unlicensed sources of radiation are part of background or otherwise not included in

• occupational doses as defined in NRC's radiation protection standards in 10 CFR 20).

Instead, in response to the comment, the definition in § 70.4 was changed to: Worl<er, as used in Subpart H, means an individual who receives an occupational dose as defined in 10 CFR 20.1003.

F. Miscellaneous Comment F.1: One commenter recommended that the criticality requirements of

§ 70.24 be revised to permit alternate criticality control provisions to be accepted for DOE facilities without requiring an exemption.

Response: Comments on § 70.24 are outside the scope of the rulemaking.

Comment F.2: Two commenters recommended changes in the decommissioning requirements of §§ 70.22(a)(9) and 70.38. In particular, one commenter recommended that the timeliness and schedule provisions in the decommissioning requirements of § 70.38 be revised to include separate requirements for DOE f acilitles.

Response: Comments on §§ 70.22(a)(9) and 70.38 are outside the scope of the rulemaking.

36

Comment F.3: One commenter expressed concern with the language in§ 70.23(b),

which states that the Commission will approve construction of a plutonium processing and fuel fabrication facility only after determining that the design bases of SSCs, and the attendant quality assurance program are adequate to protect against natural phenomena and the consequences of potential accidents. In particular, the commenter stated that this

)

provision, as written, seems contrary to other changes being proposed under the draft rule, because it addresses consequences of potential accidents, as opposed to the risk associated with credible accidents.

Response: Section 70.23(b) has not been modified In this rulemaking. The reference to "consequences" in the rule language does not preclude a risk-informed approach in satisfying this requirement. The NRC will need to consider the risk, and thus the likelihood of consequences of potential accidents occurring, in order to determine whether there Is reasonable assurance of protection against such consequences. This consideration of risk will be Important In determining the need for (and the ability of) the applicant to reduce the likelihood of accidents and to mitigate their consequences._ .

Comment F.4: One commenter recommended that§ 70.11 be revised to reflect the applicability of NRC authority over a MOX fuel fabrication facility owned by the ~OE, pursuant to changes In law last year.

Response: The NRC agrees with the comment, but believes a separate rulemaking Is required. Since October 17, 1998, when the amendment to Section 202 of the Energy Reorganization Act of 1974 was enacted, §, 70.11 , as well as several other subsections of the 37

regulations, need to be updated to reflect this legislative change. However, to address this subsection and all the other instances, and to avoid the necessity for potential future revisions of this type, the NRC Intends to institute an administrative-type rule amendment to confonn all of the references to Section 202 in the regulations, including § 70.11 , to merely cite Section 202, rather than repeat the text of that section. Because this rule change affects various parts of the regulations, it will be conducted independently of the current Part 70 amendments.

Comment F.5: One commenter stated that as additional DOE facilities are licensed by the NRC under Part 70, the NRC should ensure that the requirements address the full range of fissionable and fissile materials at these facilities.

Response: This Issue Is beyond the scope of the rulemaklng. It will be addressed, if necessary, In the future.

Comment F.6: One commenter agreed that the proposed rule Is entirely consistent with the U.S. Environmental Protection Agency's Risk Management Program regulations and the general duty clause of the Clean Air Act, and contains appropriate complementary safety measures for facilities possessing a critical mass of SNM.

Response: No response necessary.

Comment F.7: One commenter strongly recommended that the NRC adopt, by reference, the 1998 edition of National Fire Protection Association (NFPA) 801, "Facilities 38

Handling Radioactive Materials." NFPA 801 would apply to§ 70.62, "Safety program and integrated safety analysis," that addresses protection from all relevant hazards, including radiologlcaJ, criticality, fire, and chemical. The NFPA standard would also apply to§ 70.~.

"Requirements for new facilities or new processes at existing facilities," that addresses fire protection. The reference tu NFPA 801 is in keeping with the requirements of Public Law 104-113, "National Technology Transfer and Advancement Act," that requires Federal agencies to use private sector-developed national consensus technical standards in carrying out public policy, wherever appropriate.

Response: The suggested change would be an unnecessarily prescriptive rule requirement. Instead, the NRC identifies the standards in NFPA 801 and 600 as an acceptable approach for demonstrating compliance with 10 CFR Part 70 in the SAP.

Comment F.8: One commenter noted that the proposed rule Incorporates the current terms of the MOU between the NRC and OSHA. This should avoid misunderstanding and result In more effectlve Implementation for all concerned parties.

Response: Although the rule Is consistent with the NRC-OSHA MOU, the rule itself does not Incorporate the terms of the MOU. Nevertheless, the NRC agrees with the spirit of

\

the comment.

Comment F.9: Two commenters expressed concern over those portions of §§ 70.22 and 70.23 of the existing rule that address the regulation of plutonium processing and fuel fabrication facilities. One commenter asked If § 70.22 (f) should be coordinated with § 70.65.

39

The commenter noted that it is not clear if the requirements are collateral, complementary, or redundant. The same commenter stated that§ 70.23(b) should be examined to clarify the need for this requirement in light of similar information being submitted pursuant to § 70.65.

The second commenter agreed, stating that § 70.22(f) requires plutonium-related applicants to provide information on the facility site and design basis of principal SSCs, etc., as part of the license application. The commenter believed that this information is also required In other sections of the revised rule, and thus is redundant.

Response: No change in the rule language has been made. The requirements are not viewed as redundant, considering: the timeframe for submittal of information required by the two sections could*be different; and§ 70.23(b) contains a requirement for NRC construction approval before the start of construction.

Comment F.10: Two commenters were concerned about the construction authorization provisions In §§ 70.23(b) and 70.23(a)(7). According to one commenter, Irrespective of

§ 70.65, the construction authorization provision In-§ 70.2.3(b) appears to be an unnecessary step and should be considered for deletion by the NRC. If the_ NRC chooses to *retain

§ 70.23(b), the N*Rc should clarify how the authorization process would be conducted, given that the procedural step has never been exercised. Furthermore, the NRC should identify how the ndesign basis" authorization is defined, why it is necessary, and how it relates to the ISA. The second commenter noted that § 70.23{a)(7), which applies to other Part 70 licensees, allows construction to commence based on a conclusion by the Director, NRC Office of Nuclear Material Safety and Safeguards, that environmental impacts have been appropriately addressed. The commenter stated that this discretion afforded the NRC under

)

40

§ 70.23(a)(7) - i.e., NRC's authority over construction associated with "any ... activity which the Commission determines will significantly affect the quality of the environment" is adequate to ensure the sufficiency of information provided to the NRC to authorize or disallow construction. The commenter proposes that § 70.23(a)(7) be clarified for applicability to plutonium facilities, and that§§ 70.22(f), 70.23(a)(7),and 70.23(b) be eliminated. Doing so would avoid the preconception that, irrespective of design features and material composition, plutonium is "more special" than other SNM.

Response: No change In the rule language has been made. The Atomic Energy Commission specifically established these requirements (see 36 FR 9786; May 28, 1971 and 36 FR 17573; September 2, 1971) for plutonium facilities in recognition of the potential exposures and ground-contamination levels that may result If only a small fraction of the dispersible plutonium in process were released (see SECY-A 188, March 17, 1971). The current revisions to Part 70 do not impact this section and therefore, the suggested change is outside the scope of the rulemaklng. Regarding the authorization process, the NRC staff has darified this process in a letter to Duke, Cogema, Stone & Webster, dated September 10, 1999. The design basis was also Identified in this letter. NRC provided additional guidance on this process In the draft Standard Review Plan for a Mixed Oxide Fuel Fabrication Facility.

In addition, the NRC staff is currently assessing the opportunities for hearings associated with the review of a license application for a plutonium processing facility and may offer additional guidance on this topic later in 2000.

Comment F.11: One commenter noted that, under§ 70.23(a)(8), the NRC will approve a plutonium facility's license application only after construction of principal SSCs has been 41

completed in accordance with the application. Certainly this is not a requirement unique to plutonium facilities. The NRC already has the authority to grant licenses conditional on successful completion of certain actions (such as successful start-up testing, training, etc.).

Completion of construction in accordance with the license application seems such an obvious.

condition that this specific provision seems redundant and therefore unnecessary.

Response: The NRC established § 70.23(a)(B) specifically for plutonium processing facilities. Because the current revisions to Part 70 do not impact this section, comments regarding this section are outside the scope of the rulemaking. See response to Comment F.10.

Comment F.12: One commenter noted that the terminology in Appendix A (b)(1) clearty ties the failure to the performance requirements. The phrase, "and which results in failure to meet the performance requirements of § 70.61" is very clear. This phrase should be consistently included in Appendix A (b)(2)-(5) using the exact same wording.

Response: No change in the rule language has been made. The linkage to the failu to meet the performance requirements is already included in Appendix A (b)(2) and (b)(3).

For the events described in Appendix A (b)(4) and (5), the NRC staff desires to be informed when such events occur, regardless of the licensee's determination with respect to the performance requirements. In these cases, the NRC staff will independently confirm the licensee's assessment of whether*the performance requirements were met, on the basis of the information reported.

42

Comment F.13: One commenter stated that the reporting requirements of § 70.50 continue to misrepresent the principles of the 1988 NRC-OSHA MOU. Section 70.50(c)(1 )(lli)(A) requires the reporting of chemical hazards and § 70.50(c}(1 )(iil}(B) requires the reporting of P,ersonnel exposures to chemicals. According to the commenter, although the MOU principles have been correctly Incorporated Into other proposed revisions to Part 70 (e.g.,§§ 70.4, 70.61 (b), 70.62(c), 70.64(a), 70.74, and Appendix A), they were incorrectly referenced in§ 70.50. MOU principle (2) limits NRC jurisdiction to regulation of chemical hazards of licensed material and hazardous chemicals produced from licensed material. The two aforementioned sections of § 70.50 should be corrected to properly incorporate the MOU

• principles.

Response: The rule was revised in response to the comment to reflect more precisely the language in the NRC-OSHA MOU.

Comment F.14: One commenter noted that applicants for licenses to operate new facilities or new processes at existing facilities would be expected ("Statements of

• Consideration," 64 FR 41346) to update their ISAs, based on as-built conditions, and submit the results to the NRC bef9re operation. The process for uranium enrichment facilities that must comply with§ 70.23a would differ from this description. Uranium enrichment facilities would submit a complete license application, including an ISA Summary, for construction and operation. This application would be the basis for NRC review, and culminate in issuance of a license for construction and operation. After issuance of the license, the licensee would institute' change control under§ 70.72. The licensee would then be required to submit summaries of changes and ISA Summary updates as required by§ 70.72. An inspection 43

would verify that the facility has been constructed in accordance wlttJ the license, betore operation, as required by § 70.32(k). No pre-operational submittal and review of an updated ISA Summary are anticipated for uranium enrichment facilities because their configuration would be controlled after Issuance of the construction and operation license.

Response: As recognized by the commenter, no changes to the rule are necessary.

) '

The dtfferences in the ll~nslng process for enrichment facilities other than the gaseous diffusion plants (regulated under 10 CFR Part 76) reflect the process mandated In Section 193 of the Atomic Energy Act of 1954, et. seq.

Ill. Changes from the Proposed Rule Subpart A - General Provisions AUTHORITY This section has been changed to reflect the redeslgnations of §§ 70.61 and 70.62 as

§§ 70.81 and 70.82, respectively.

§ 70.4 Definitions.

The definition of "available and reliable to perform their function when needed" has been modified to eliminate the need to maintain "continuous" compliance with the performance requirements of § 70.61. The _definition of "configuration management" has been modified to clarify its role as a "management measure." The definition of "critical mass of special nuclear material" has been modified to emphasize that the definition is only for the 44

purposes of Subpart H. The definition of "double contingency" has been changed to provide minor clarification. The definition for "worker" has been clarified and has been revised to emphasize that the definition Is only for purposes of Subpart H. The definitions of "ISA" and "ISA Summary" have been changed to Indicate that the ISA can be performed on a process by process basis and the ISA Summary can be submitted In multiple documents that cover all the portions of the facility.

Subpart G - Special Nuclear Material Control Records, Reports, and Inspections

§ 70.50 Reporting requirements.

The reporting requirements for hazardous chemicals have been revised to be consistent with the language of the 1988 NRC-OSHA MOU (53 FR 43950; November 22, 1988).

Subpart H - Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear Material

§ 70.60 Applicability.

The applicability of the Subpart H requirements has been revised to clarify when the requirements will take effect.

§ 70.61 Performance requirements.

The performance requirements in §§ 70.61 (b) and (c) have been revised to provide clarification. The requirement to establish a controlled area in § 70.61 (f) has been revised to 45

clarify the conditions for establishing the controlled area, and to clarify the applicability of the performance requirements to individuals within the controlled area. The requir~ment in

§ 70.61 (f)(2) to provide training "in accordance with" 10 CFR 19.12(a)(1 )-(5) has been revised to clarify that equivalent training is acceptable. The new language specifies that this training must "satisfy" 10 CFR 19.12{a)(1 )-(5).

§ 70.62 Safety program and Integrated safety analysis.

The requirement to establish and maintain a safety program in § 70.62(a)(1) has been revised to clarify that tlie safety program referred to in this section is focused *on the safety program for satisfying the new Subpart H requirements, and that the application of management measures may be graded according to risk. $action 70.62(a)(3) has been modified to make it performance-based by eliminating the prescriptive requirement to maintain a log of failures for Items relied on for safety. Section 70.62(c)(3) has been revised to clarify* -

the schedule for planning and performing an l~A, correcting all performance* deficiencies, and submitting the ISA Summary to the NRG for approval. The ISA Summary can be submitted as a single document, or as a sequence of documents (e.g., on a process basis). The approval process for the ISA Summary will require the Issuance of a license amendment; however, a license condition will be established that allows the licensee to make changes in accordance with§ 70.72, including certain changes that do not require prior NRG approval. .In addition, a provision has been added to the schedule for complying with the requirements in subpart H for factors beyond control of the licensee. This would allow additional time for correcting a performance deficiency if the NRG approves. Also§ 70.62(d) has been modified to reflect that management measures may be graded commensurate with the reduction in risk.

46

§ 70.64 Requirements for new facilities or new processes at existing facilities ..

Section 70.64(a) has been revised to provide the correct reference, § 70.62(c), to the performance of an ISA. Section 70.64(a)(6)(ii) has been modified to specify that the required emergency capability is concerned with the evacuation of only on-site personnel.

§ 70.65 Additional content of applications.

Section 70.65(a) has been revised to clarify that the ISA Summary Is not part of the safety program description required for inclusion in the license application. Rather, the ISA Summary, that contains a description of management measures, Is submitted with the license application.

§ 70.66 Additional requirements for approval of license application.

Section 70.66(b) has been added to clarify, for existing licensees, the basis for Commission approval of the ISA p:an, submitted under§ 70.62(c)(3)(i), cmd the ISA Summary, submitted under§ 70.62(c)(3)(ii) .

• § 70. 72 Facllity changes and change process.

Section 70.72(c)(1) has been revised to eliminate the footnote, which did not adequately clarify the meaning of "new types of accident sequences." The revised section 70.72(d)(3) replaces proposed rule section 70.72(d)(1) to reflect a modified schedule for submission of revised ISA Summary pages.

47

§ 70.76 Backfitting.

Section 70.76 was added to Include requirements for performing a backfit analysis.

The wording in § 70. 76 is similar to the current language in §76. 76 for gaseous diffusion plants with one exception, 70.76 (a)(4)(i). The exception In§ 70.76(a)(4)(1) relates to the backfit requirements being inapplicable to changes associated with bringing the facility in compliance with the requirements of the new subpart H. The backfit section includes a provision stating that it shall apply for subpart H requirements as soon as NRC approves that licensee's ISA Summary (contents of ISA Summary described in § 70.65(b)}-pursuant to

§ 70.66 and, for requirements other than Subpart H, it shall apply immediately after NRC publication of backfit guidance.

§ 70.92 Criminal Penalties.

This section has been changed to reflect the redesignatlons of §§ 70.13a, 70.14, 70.61, 70.62, 70.71, and 70.72 as§§ 70.14, 70.17, 70.81, 70.82, 70.91, and 70.92, respectively, and the addition of§§ 70.66, 70.73, and 70.76.

IV. Section-by-section Analysis of Part 70 Amendments AUTHORITY This section has been changed to reflect the redeslgnations of§§ 70.61 and 70.62 as

§§ 70.81 and 70.82, respectively.

Subpart A - General Provisions 48

§ 70.4 Definitions.

Definitions of the following 12 terms have been added to this section to clarify the meaning of certain terms and phrases used in the new Subpart H: "Acute," "Available and reliable to perform their function when needed," "Configuratior] management," "Critical mass of SNM," "Double contingency principle," "Hazardous chemicals produced from licensed

• materials," "Integrated safety analysis," "Integrated safety analysis summary," "Items relied on for safety," "Management measures," "Unacceptable performance deficiencies," and "Worker."

§ 70.14 Foreign military aircraft.

This section reflects an administrative change to redesignate this section, formerly

§ 70.13a.

§ 70.17 Specific exemptions.

This section reflects an administrative change to redesignate this section, formerly

§ 70.14.

Subpart G - Special Nuclear Material Control Records, Reports, and Inspections

§ 70.50 Reporting requirements.

Paragraph (c) has been reworded to include information to be transmitted when making verbal or written reports to the NRC. The new information derives from the specifics of the new*Subpart H, such as sequence of events and whether the event was evaluated in the ISA. To the extent the new information is also applicable to licensees not subject to 49

Subpart H, the Information was added with no differentiation noted. The new Information that would only apply to Subpart H licensees is noted.

Subpart H - Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear Material

§ 70.60 Applicability.

This section lists the types of NRC licensees or applicants that are subject to the new Part 70, Subpart H, and describes when the new requirements will be effective.

§ 70.61 Performance requirements.

This section Identifies the performance requirements that licensees subject to Part 70, Subpart H must satisfy. These performance requirements explicitly address the risks to workers or members of the public and the environmental releases caused by accidents.

Because accidents are unanticipated events that usually occur over a relatively short period L

of time, the Part ,70 changes seek to ensure adequate protection of workers, members of the public, and the environment by limiting the risk (product of likelihood and consequence) of such accidents. If, without the implementation of controls, a high consequence event under

§ 70.61 {b) is highly unlikely, then it is not necessary for the licensee to apply the engineered r

or administrative controls mentioned in the rule. Similarly, If, without the implementation of controls an intermediate consequence event under§ 70.61 {c) is unlikely, then it is not necessary for the licensee to apply the engine~red or administrative col")trols mentioned in the rule.

50

§ 70.62 Safety program and integrated safety analysis.

This section describes requirements for establishing and maintaining a safety program that demonstrates compliance with the performance requlren:ients of § 70.61. The elements of this safety program include the compilation of process safety Information, the performance of an ISA, and the application of management measures to ensure the availability and reliability of Items relied on for safety.

§ 70.64 Requirements for new facilities or new processes at existing facilities.

This section describes baseline design criteria for new facilities or new processes at existing facilities. The application of these criteria, which are similar to the general design criteria in Part 50, Appendix A; Part 72, Subpart F; and 10 CFR 60.131, are consistent with good engineering practice, which dictates that certain minimum requirements be applied as design and safety considerations for any new nuclear process or facility. The baseline design Qiterla do not provide relief from compliance with the performance requirements of§ 70.61.

§ 70.65 Additional content of applications.

In addition to the information that currently must be submitted to the NRG under

§ 70.22, for a license application, this section requirE.: additional Information to be submitted to demonstrate compliance with the new subpart H requirements. In particular, this additional I

information includes a description of the applicant's safety program established under

§ 70.62. This information will be incorporated In the license, as appropriate. The ISA Summary must be submitted with the license application or in accordance with 70.62(c)(3)(ii),

but will not be Incorporated In the license.

51

§ 70.66 Additional requirements for approval of license application.

This section contains the provision that the applicant must comply with the requirements of §§ 70.60 through 70.65 (in addition to §§ 70.21 through 70.23, in the existing regulation) before a license will be granted. It also contains the requirements for approving the ISA plan and the ISA Summary for existing licensees. If the ISA Summary is submitted as a sequence of documents (e.g., on a process basis), NRC staff approval of an individual document will be conditioned for system interaction effects that may be identified through the review of other ISA Summary documents submitted.

§ 70. 72 Facility changes and change process.

This section contains requirements that govern changes to site, .structures, systems, equipment, components, and activities of personnel after a license application has been approved. It requires the licensee to establish and use a configuration management system to evaluate changes and the potential impacts of those changes before Implementing them.

The regulation permits the licensee to make certain *changes without NRC pre~approval, but requires the licensee to submit a brief summary of the changes plus updated ISA Summary pages annually within 30 days after the end of the calender year during which the changes occurred.

§ 70.73 Renewal of licenses.

This section contains the requirements for renewing licenses. It references the existing renewal requirements and the additional contents of application in § 70.65.

§ 70.74 Additional reporting requirements.

52

This section contains new requirements, in addition to those in existing Parts 20 and 70, for reporting events to the NRC. The new approach, based on consideration of the perfonnance requirements established in 10 CFR 70.61 (b), is intended to eventually replace and modify the approach licensees have currently been using for reporting criticality events under NRC Bulletin 91-01. The new approach would saver all types of events, not just criticality events, and establish a tlmeframe for reporting that is scaled according to risk.

§ 70. 76 Backfitting .

• This section contains requirements for perf onnlng a backfit analysis that are based on those in 10 CFR 76. 76. It will become effective after NRC publication of backfit guidance.

NRC staff will work with stakeholders to develop backfit guidance which will Include making clear that an adequate demonstration can be based on quantitative or qualitative evaluations of the nature of the increase In the overall health and safety protection of the public. The NRC will publish a separate Federal Register notice upon publication of the guidance to Indicate the effectiveness of§ 70.76. After that notice is published, this provision will not be applied to Subpart H requirements until the NRC approves the licensee's ISA Summary

• pursuant to§ 70.66. If the approved ISA Summary Is one of a sequence of approvals (e.g, on a process basis}, the backfit provision will apply to the portion of the facility covered by that ISA Summary document. However, the backfit provision does not apply to changes to those portions of the facility that are required by the NRC staff to a~dress system Interaction effects identified through the review of other ISA Summary submissions for that facility.

Subpart J-- Enforcement 53

§ 70.92 Criminal penalties This section has been changed to reflect the redesignations of §§ 70.13a, 70.14, 70.61, 70.62, 70.71, and 70.72 as§§ 70.14, 70.17, 70.81, 70.82, 70.91, and 70.92, respectively, and the addition of§§ 70.66, 70.73, and 70.76.

Appendix A Reportable safety events.

This appendix contains a list of events that licensees must report to the NRC. These events are categorized according to their consequences (or potential consequences) and fall into two classes: a 1-hour or 24-hour reporting timeframe. The emphasis on consequences, rather than risk, is appropriate in this case because the event has already occurred. Append.

A also requires concurrent reporting of events when a news release Is made or if other Government agencies are notified, as is done under 10 CFR 50.72, to enhance coordination and support NRC's ability to respond to questions concerning the safety of NRG-licensed facilities.

V. Finding of No Significant Environmental Impact: Availability The Commission has detennined, under the National Environmental Policy ~ct of 1969-as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule is not a major Federal action significantly affecting the quality of the human environment and therefore, an environmental impact statement is not required.

The amendments to 10 CFR 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 performance requirements and the 54

level of protection needed to prevent or mitigate accidents that exceed such requirements; (2) require affected licensees to perform an ISA to identify potential accidents at the facility and the items relied on for safety; (3) require the implementation of measures to ensure that the items relied on for safety are available and reliable to perform their functions when needed; (4) require the safety bases to be maintained, and changes reported to the NRC; (5) allow for licensees to make certain changes to their safety program and facilities without prior NRC approval; (6) require reporting of certain events; and (7) require a backflt analysis under specified conditions.

The rule language that defines the performance requirements is relevant to the question of environmental Impact. Licensees are required to protect against the occurrence of or to mitigate the consequences of accidents that could adversely affect workers; the public, or the environment. For example, licensees are required to provide an adequate level of protection against a "release of radioactive material to the environment outside the restricted area In concentrations that, if averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 5000 times the values specified in Table 2 of Ap;:,endix B to 10 CFR Part 20." Implementation of the new amendments, Including the requirement to protect against events that could damage the environment, is expected to result in a significant improvement In licensees', NRC's, other governmental agencies', and the public's understanding of the risks at these facilities and licensees' ability to ensure that those risks are appropriately controlled. F9r existing licensees, any deficiencies Identified in the ISA (which must be completed within 4 years) will need to be promptly addressed. For new licensees, operations will not begin unless licensees demonstrate an adequate level of protection against potential accidents identified in the ISA.

As a result, the safety and environmental impact of the new amendments is positive. There will be less potential adverse Impact on the environment from licensed operations carried out 55

under the final rule than if those operations were carried out under the existing Part 70 regulation. Thus, the Commission has determined, based on the Environmental Assessment that supports the rule, that there will be no significant impact on the human environment from

(

this action.

The NRC requested public comments on any environmental jusdce considerations that may be related to this rule. No comments were received in response to this request.

The NRC also requested the States' views on the environmental assessment for this rule. No comments were received in response to this request.

The Environmental Assessment is availabl.e for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, D.C. Single copies of the Environmental Assessment are available from Barry Mendelsohn, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone (301) 415-7262; e-mail: btm1@nrc.gov.

VI. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These requirements were approved by the Office of Manag~ment and Budget, approval number 3150-0009.

The public reporting burden for this information collection is estimated to average 92 hours0.00106 days <br />0.0256 hours <br />1.521164e-4 weeks <br />3.5006e-5 months <br /> per response and the recordkeeping burden is estimated to average 548 hours0.00634 days <br />0.152 hours <br />9.060847e-4 weeks <br />2.08514e-4 months <br />, including the time for reviewing instructions, searching existing data sources, gathering and .

maintaining the data needed, and completing and reviewing the information collection. Send 56

comments on any aspect of this information collection, including suggestions for reducing the burden, to the Records Management Branch (T-6 ES), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail at BJS1 @NRG.GOV; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0009), Office of Management and Budget, Washington, DC 20503.

Public Protection Notification If a means used to impose an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the Information collection.

VIII. Regulatory Analysis The Commission has prepared a regulatory analysis on this final regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission.

• The analysis is available for inspection in the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, D.C. Single copies of the environmental assessm~nt are available from Barry Mendelsohn, Office of Nuclear Material Safety and Safeguards, U.S.

Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone (301) 415-7262; e-mail: btm1@nrc.gov.

IX. Regulatory Flexibility Certification 57

As required by the Regulatory Flexibility Act, as amended, 5 U.S.C. 605(b), the Commission certifies that this rule does not have a significant economic impact on a substantial number of small entities. The regulation affects facilities that are authorized to possess a critical mass of SNM and that are engaged In one of the following activities:

enriched uranium processing; fabrication of uranium fuel or fuel assemblies; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; fabrication of mixed-oxide fuel or fuel assemblies; scrap recovery of SNM or 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 entitles" set forth In the RegulatorY Flexibility Act, nor the size standards published by the NRC (10 CFR 2.810).

X. Voluntary Consensus Standards The National Technology Transfer Act of 1995, Pub. L. 104-113, requires that Federal agencies use technical standards developed or adopted by voluntary consensus standards bodies unless the use of such a standard is Inconsistent with applicable law or otherwise impractical. In this regulation, the NRC will use the following voluntary consensus standard --

ANSI/ANS Standard 8.1-1983, "Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactor$" developed by the American Nuclear Society. Portions of the standard were used in the definition of double contingency and in§ 70.61 (d). A consensus standard with the complete scope of the requirements established in this rulemaklng does not exist. The Commission will reference ANS 8.1 and other consensus standards, as appropriate, as acceptable approaches to demonstrate compliance with specific portions of 58

the final rule. This will be addressed in the Standard Review Plan that is being established with the rule.

r XI. Backfit Statement The NRG has determined that the backfit rule does not apply to this final rule; therefore, a backfit analysis Is not required for this final rule because these amendments do not Involve any provisions that would impose backfits as defined In 10 CFR Chapter I.

However, future changes to the requireme,:its In subpart Hor NRC requirements that apply to facilities covered by subpart H will be subject to the backfit requirements in § 70. 76 established In this rule.

XII. Small Business Regulatory Enforcement Fairness Act In accordance with the Small Business Regulatory Enforcement Fairness Act of 1996, the NRC has determined that this action is not a major rule and has verified this determination with the Office of Information and Regulatory Affairs of 0MB.

XIII. List of Subjects Criminal penalties, Hazardous materials transportation, Material control and accounting, Nuclear materials, Packaging and containers, Radiation protection, Reporting and 59

recordkeeplng requirements, Scientific equipment, Security measures, Special nuclear material.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C.

552 and 553; the NRC is adopting the following amendments to Part 70.

PART 70-DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL I

1. The authority citation for part 70 is ~vised 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).

Sections 70.1(c) and 70.20a(b) also issued under secs. 135,141, Pub. L.97-425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161 ). Section 70.7 also Issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Section 70.21(9) also Issued under sec. 122, 68 I

Stat. 939 (42 U.S.C. 2152). Section 70.31 also issued under sec. 57d, Pub. L.93-377, 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36 and 70.44 also issued under sec. 184, 68 Stat.

954, as amended (42 U.S.C. 2234). Section 70.81 also issued under secs. 186, 187, 68 Stat.

955 (42 U.S.C. 2236, 2237). Section 70.82 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).

60

2. The undesignated center heading "GENERAL PROVISIONS" is redesignated as "Subpart A--General Provisions."

3. In § 70.4, the definitions of Acute, Available and reliable to perfonn their function J

when needed, Configuration management, Critical mass of special nuclear material, Double contingency principle, Hazardous chemicals produced from licensed materials, Integrated safety analysis, Integrated safety analysis summary. Items relied on for sat ety. Management measures, Unacceptable perfonnance deficiencies, and Worker are added, in alphabetical order,. as follows:

§ 70.4 Definitions.

Acute, as used in this part, means a single radiation dose or chemical exposure event or multiple radiation dose or chemical exposure events occurring within a short time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less). *****

Available and reliable to perfonn their function when needed, as used in subpart H of this part, means that, based on the analyzed, credible conditions in the Integrated safety analysis, Items relied on for safety will perfonn their intended safety function when needed,

• and management measures will be implemented that ensure compliance with the performance 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.

Configuration management (CM) means a management measure that provides oversight and control of design infonnation, safety infonnation, and records of modifications 61

(both temporary and permanent) that might impact the ability of Items relied on for safety to perform their functions when needed.

• • • fr*

Critical mass of special nuclear material (SNM), as used in Subpart H, means special nuclear material in a quantity exceeding 700 grams of contained uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1500 grams of contained uranium-235, if no uranium enriched to more than 4 percent by weight of uranium-235 is present; 450 grams of any combination thereof; or one-half such quantities if massive moderators or reflectors made of graphite, heavy water, or beryllium may be present.

Double contingency principle means that process designs should Incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes In process conditions before a criticality accident Is possible.

Hazardous chemicals produced from licensed materials means substances having licensed material as precursor compound(s) or substances that physically or chemically Interact with licensed materials; and that are toxic, explosive, flammable, corr~sive, or reactive.

to the extent that they can endanger lif.e or health If not adequately.controlled. These include substances commingled with licensed material, and Include substances such as hydrogen fluoride that is produced by the reaction of uranium hexafluoride and water, but qo not Include substances prior to process addition to licensed material or after process separation from licensed material.

62

Integrated safety analysis {ISA) means a systematic analysls to identify facility and external hazards and their potential for initiating accident sequences, the potential accident sequences, their likelihood and consequences, and the Items relied on for safety. As used here, Integrated means joint consideration of, and protection from, all relevant hazards, including radiological, nuclear criticality, fire, and chemical. However, with respect to compliance with the regulations of this part, the NRC requirement is limited to consideration of the effects of all relevant hazards en radiological safety, prevention of nuclear criticality accidents, or chemical hazards directly associated with NRC licensed radioactive material. An ISA can be performed process by process, but all precesses must be Integrated, and procE,!ss interactions considered.

Integrated safety analysis ~ummary means a document or documents submitted with .,

the license application, license amendment application, license renewal application, or pursuant to§ 70.62(c)(3)(ii) that provides a synopsis of the results of the integrated safety analysis and contains the information specified in§ 70.65(b). The ISA Summary can be submitted as one document for the entire facility, or as multiple documents that cover all portions and processes of the facility.

Items relied on for sat ety mean structures, systems, equipment, components, and J

activities of personnel that are relied on to prevent potential accidents at a facility that could exceed the performance requlrem~nts in § 70.61 or to mitigate their potential consequences.

This does not limit the licensee from identifying additional structures, systems, equipment, components, or activities of personnel (i.e., beyond those in the minimum set necessary for compliance with the performance requirements) as items relied on for safety.

63

Management measures mean the functions performed by the licensee, generally on a continuing basis, that are applied to Items relied on for safety, to ensure the items are available and reliable to perform their functions when needed. Management measures include configuration management, maintenance, training ai:d qualifications, procedures, audits and assessments, incident investigations, records management, and other quality assurance elements.

) *****

Unacceptable performance deficiencies mean deficiencies in the Items relied on for safety or the management measures that need to be corrected to ensure an adequate level o .

protection as defined in 10 CFR 70.61 (b), (c), or (d).

Worker, when used in Subpart H of this Part, means an individual who receives an occupational dose as defined in 10 CFR 20.1003.

4. In§ 70.8 paragraph (b) is revised to read as follows:

§ 70.8 Information collection requirements: 0MB approval.

(b) The approved information collection requirements contained in this part appear in

§§ 70.9, 70.17, 70.19, 70.20a, 70.20b, 70.21, 70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, 70.39, 70.42, 70.50, 70.51, 70.52, 70.53, 70.57, 70.58, 70.59, 70.61, 70.62, 70.64, 70.65, 70.72, 70.73, 70.74, and Appendix A.

64

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

§§ 70.13a and 70.14 [Redesignated]

6. Sections 70.13a and 70.14 are redeslgr:iated as §§ 70.14 and 70.17, respectively.

7. The unde~ignated center heading "GENERAL LICENSES" Is redeslgnated as "Subpart C-General Licenses."

8. The undesignated center heading "LICENSE APPLICATIONS" is redesignated as "Subpart D-Ucense Applications."

9. The undesignated center heading "LICENSES" is redesignated as "Subpart E-Licenses."

r.

10. The undesignated center heading "ACQUISITION, USE AND TRANSFER OF SPECIAL NUCLEAR MATERIAL, CREDITORS' RIGHTS," is redeslgnated as "Subpart F--Acqulsltion, Use, and Transfer of Special Nuclear Material, Creditors' Rights."

11 . The undesignated center heading "SPECIAL NUCLEAR MATE RIAL CONTROi:..

RECORDS, REPORTS AND INSPECTIONS" is redesignated as "Subpart G--Special Nuclear Material Control Records, Reports, and Inspections."

12. In§ 70.50, paragraph (c) is revised and paragraph {d) is added to read as follows.

§ 70.50 Reporting requirements.

65

{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§ 70.74 and Appendix A of this part, if applicable, by telephone to the NRC Operations Center. 1 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; (ii)' Date, time, and exact location of the event; (iii) Description of the event, including; (A) Radiological or chemical hazards involved, Including isotopes, quantities, and chemical and physical form of any material released; (B) Actual or potential health and safety qonsequences to the workers, the public, and the environment, Including relevant chemical and radiation data for actual personnel exposures to radiation or radioactive materials or hazardous chemicals produced-from licensed materials (e.g., level of radiation exposure, concentration of chemicals, and duration of exposure);

(C) The sequence of occurrences leading to the event, including degradation or failure

'at structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences; and (D) Whether the remaining structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences are available and reliable to perform their function.

(iv) External conditions aft ecting the event; 1

The commercial telephone number for the NRC Operations Center is (301) 816-5100.

66

(v) Additional actions taken by the licensee, in response to the event; (vi) Status of the event (e.g., whether the event is on-going or was terminated);

(vii) Current and planned site status, including any declared emergency class; (viii) Notifications, related to the event, that were made or are planned to any local, State, or other Federal agencies; (ix) Status of any press releases, related to the event, that were made or are planned.

(2) Written report. Each licensee that makes a report required by paragraph (a) or (b) of this section, or by§ 70.74 and Appendix A of this part, if applicable, shall submit a written I

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 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,

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with a copy to the appropriate NRG regional office listed in Appendix D of 10 CFR Part 20.

The reports must include the following:

(I) Complete applicable information required by§ 70.50(c)(1 );

(Ii) The probable cause of the event, including all factors that contributed to the event

• and the manufacturer and model npmber (If applicable) of any equipment that failed or malfunctioned; (iii) Corrective actions taken or planned to prevent occurrence of similar or identical events in the future and the results of any evaluations or assessments; and (iv) For licensees subject to Subpart H of this part, whether the event was ldentif:ed and evaluated in the Integrated Safety Analysis.

67

(d) The provisions of§ 70.50 do not apply to licensees subject to§ 50.72. They do apply to those Part 50 licensees possessing material licensed under Part 70 that are not subject to the notification requirements in § 50. 72.

13. The undesignated c~nter heading "MODIFICATION AND REVOCATION OF ,

LICENSES" is redeslgnated as "Subpart !*-Modification and Revocation of Licenses."

§§ 70.61 and 70.62 [Redesignated]

14. Sections 70.61 and 70.62 are redesignated as§§ 70.81 and 70.82, respectively .

15. The undesignated center heading "ENFORCEMENT" is redesignated as "Subpart J-Enforcement:" *

§§ 70. 71 [RedesignatedJ

16. Section 70. 71 is redeslgnated as § 70.91.

17. Section 70.72 is redesignated as§ 70.92 and paragraph (b) is revised to read as follows:

§ 70.92 Criminal Penalties

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(b) The regulations in part 70 that are not issued under sections 161 b, 161 i, or 161 o, for the purposes of section 223 are as follows: §§ 70.1, 70.2, 70.4, 70.5, 70.6, 70.8, 70.11, 70.12, 70.13, 70.14, 70.17, 70.18, 70.23, 70.31, 70.33, 70.34, 70.35, 70.37, 70.66, 70.73, 70.76, 70.81, 70.82, 70.63, 70.91, and 70.92.

18. In part 70, a new subpart H (§§ 70.60*70.74) is added to read as follows:

68

Subpart H--Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear Material Sec.

70.60 Applicability.

70.61 Performance requirements.

70.62 Safety program and integrated safety analysis.

70.64 Requirements for new facilities or n~w processes at existing facllities.

70.65 Additional content of applications.

70.66 Additional requirements for approval of license application.

70. 72 Facility changes and change process.

70.73 Renewal of licenses.

70.74 Additional reporting requirements.

70.76 Backfitting

§ 70.60 Applicability.

The regulations in§ 70.61 through§ 70.76 apply, in addition to other applicable Commission regulations, to each applicant or licensee that is or plans to be: authorized to possess greater than a critical mass of special nuclear material, and engaged in enriched uranium processing, fabrication of uranium fuel or fuel assemblies, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap recovery of special nuclear material, or any other activity that the Commission determines could signlficantly affect public health and safety. The regulations 69

In§ 70.61 through§ 70.76 do not apply to decommissioning activities performed pursuant to other applicable Commission regulations including § 70.25 and § 70.38 of this part. Also, the regulations In § 70.61 through § 70. 76 do not apply to activities that are certified by the Commission pursuant to Part 76 of this chapter or licensed by the Commission pursuant to other parts of this chapter. Unless specifically addressed In§ 70.61 through§ 70.76, implementation by current licensees of the Subpart H requirements shall be completed no later than the time of the ISA Summary submittal required in§ 70.62(c)(3)(1i).

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§ 70.61 Performance requirements.

(a) Each applicant or licensee shall evaluate, in the integrated safety analysis performed In accordance with § 70.62, its compliance with the performance requirements In paragraphs (b), (c), and (d) of this section.

(b} The risk of ~ach credible high*consequence event must be limited. Engineered controls, administrative controls, or both, shall be applied to the extent needed to reduce the likelihood of occurrence of the event so that, upon implementation of such controls, the event is highly unlikely or its consequences are less severe than those In 70.61 (b}(1 )-(4). High consequence eyents 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 located outside the controlled area identified pursuant to paragraph (f) of this section; or 70

(4) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that:

(I) Could endanger the life of a worker, or

{ii) Could lead to irreversible or other serious, long-lasting health effects to any individual located outside the controlled area identified pursuant to paragraph (f) of this section. If an applicant possesses or plans to possess quantities of material capable of such chemical exposures, then the applicant shaJI propose appropriate quantitative standards for these health effects, as part of the information submitted pursuant to§ 70.65 of this part.

{c) The risk of each credible intermediate-consequence event must be limited.

Engineered controls, administrative controls, or both shall be applied to the extent needed so that, upon implementation of such controls, the event is unlikely or its consequences are less than those in §70.61 {c)(1 )-(4). lntennediate consequence events are those internally or externally initiated events that are not high consequence events, that result in:

(1) An acute worker dose of 0.25 Sv (25 rem) or greater total effective dose equivalent; (2) An acute dose of 0.05 Sv (5 rem) or greater total effective dose equivaJent to any individual located outside the controlled area identified pursuant to .paragraph (f) of this section; (3) A 24-hour averaged release of radioactive material outside the restricted area in concentrations exceeding 5000 times the values in Table 2 of Appendix B to Part 20; or (4) An acute chemical exposure to an Individual from licensed material or hazardous chemicals produced from licensed material that:

(i) Could lead to irreversible or other serious, long-lasting health eff acts to a worker, or 71

(ii) Could cause mild transient health etf ects to any individual located 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 margin of subcritlcallty for safety. Preventive controls and measures must be the primary means of protection against nuclear criticality accidents.

(e) Each engineered or administrative control or control system necessary to comply with paragraphs (b), (c), or (d) of this section shall be designated as an item relied on for safety. The safety program, established and maintained pursuant to § 70.62 of this part, shall ensure that each item relied on for safety will be available and reliable to perform its intended function when needed and in the context of the performance requirements of this section.

(f} Each licensee must establish a controlled area, as defined in § 20.1003. In addition, the licensee must retain the authority to exclude or remove personnel and property.

from the area. For the purpose of complying with the performance requirements of this section, individuals who are not workers, as defined in § 70.4, may be permitted to perform ongoing activities (e.g., at a facility not related to the licensed activities) in the controlled area, if the licensee:

(1) Demonstrates and documents, in the integrated safety analysis, that the risk for those Individuals at the location oLtheir activities does not exceed the performance "

requirements of paragraphs (b)(2), (b)(3), (b)(4)(ii}, (c}(2), and (c)(4)(ii) of this section; or 72

(2) Provides training that satisfies 10 CFR 19.12(a)(1 )-(5) to these individuals and ensures that they are aware of the risks associated with accidents Involving the licensed activities as determined by the integrated safety analysis, and conspicuously posts and maintains notices stating where the information in 10 CFR 19.11 (a) may be examined by these intfividuals. Under these conditions, the performance requirements for workers specified in paragraphs {b) and (c) of this section may be applied to these individuals.

§ 70.62 Safety program and integrated satety analysis .

• (a) Safety program. (1) Each licensee or applicant 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 graded commensurate with the reduction of the risk attributable to that item. Three elements of this 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 or applicant shall establish and maintain records that demonstrate compliance with the requirements of paragraphs (b) through (d) of this section.

(3) Each licensee or applicant shall maintain records of failures readily retrievable and available for NRC inspection, documenting each discovery that an item relied on for safety or management measure has failed to perform its function upon demand or has degraded such that the performance requirements of § 70.61 are not satisfied. These records must identify the Item relied on for safety or management measure that has failed and the safety function affected, the date of discovery, date (or estimated date) of the failure, duration (or estimated duration) of the time that the Item was unable to perform Its function, any other affected Items 73

relied on for safety or management measures and their safety function, affected processes, cause of the failure, whether the failure was in the context of the performance requirements or upon demand or both, and any corrective or compensatory action that was taken. A failure must be recorded at the time of discovery and the record of that failure updated promptly upon the conclusion of each failure investigation of an item relied on for safety or management measure.

(b) Process safety Information. Each licensee or applicant shall maintain process safety information to enable the performance and maintenance of an Integrated safety analysis. This process safety information must include information pertaining to the hazards of the materials used or produced in the process, Information pertaining to the technology of the process, and information pertaining to the equipment In the process.

(c) Integrated safety analysls. (1) Each licensee or applicant shall conduct and maintain an integrated safety analysis, that is of appropriate detail for the complexity of the process, that Identifies:

(I) Radiological hazards related to possessing or processing licensed material at Its facility; (ii) Chemical hazards of licensed material and hazardous chemicals produced from licensed material; (Iii) Facility hazards that 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 facility and credible external events, Including natural phenomena; 74

(v) The consequence and the likelihood of occurrence of each potential accident sequence 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) of this part, the characteristics of Its preventive, mitigative, or other safety function, and the assumptions and conditions under which the item is relied upon to support compliance with the performance requirements of§ 70.61.

(2) Integrated safety analysis team qualifications. To assure the adequacy of the Integrated safety analysis, the analysis must be performed by a team with expertise In engineering and process operations. The team shall include at least one person who has experience and knowledge specific to each process being evaluated, and persons, who have experience in nuclear criticalfty safety, radiation safety, fire safety, and chemical process safety. One member of the team must be knowledgeable in the specific integrated safety analysis methodology being used. .,

(3) Requirements for existing licensees. Individuals holding an NRC license on [the date of publication of the final rule] shall, with regard to e)tjsting licensed activities:

• (I) By <the effective date of the final rule plus 6 months>, submit for NRC approval, a plan that describes the Integrated sat ety analysis approach that will be used, the processes that will be analyzed, and the schedule for completing the analysis of each process.

(ii) By <the effective date of the final rule plus 4 years>, or in accordance with the approved plan submitted under§ 70.62(c)(3)(i), complete an Integrated safety analysis, correct all unacceptable performance deficiencies, and submit, for NRC approval, an integrated safety analysis summary, including a description of the management measures, in accordance with § 70.65. The Commission may approve a request for an alternative 75

schedule for completing the correction of unacceptable performance deficiencies If the Commission determines that the alternative is warranted by consideration of the following:

(A) Adequate compensatory measures have been established; (B) Whether it is technically feasible to complete the correction of the unacceptable performance deficiency within the allotted 4-year period; (C) Other site-specific factors which the Commission may conside~ appropriate on a case-by-case basis and that are beyond the control of the licensee.

(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 measures to ensure compliance with the performance requirements of§ 70.61. The measures applied to a particular engineered or administrative control or control system may be graded commensurate with the reduction of the risk attributable to that control or control system. The management measures shall ensure that engineered and administrative controls and control systems that are identified as Items relied on for safety pursuant to§ 70.61 (e) of this part are designed, implemented, and maintained, as necessary, to ensure they are available and reliable to perform their function when needed, to comply with the performance requirements of§ 70.61 of this *part.

§ 70.64 Requirements for new facilities or new pfocesses at existing facilities.

76

(a) Baseline design criteria. Each prospective applicant or licensee shall address the following baseline design criteria In the design of new facilities. Each existing licensee shall address the folio.wing baseline design criteria in the design of new processes at existing facilities that require a license amendment under§ 70.72. The baseline design criteria must l.

be applied to the design of new facilities and new processes, but do not require retrofits to existing facilities or .existing processes (e.g., those housing or adjacent to the new process);

however, all facllltles and processes must comply with the performance requirements in

§ 70.61. Licensees shall maintain the application of these criteria unless the analysis performed pursuant to § 70.62(c) demonstrates that a given item is not relied on for safety or does not require adherence to the specified criteria.

(1) Quality standards and records. The design must be developed and Implemented In accordance with management measures, to provide adequate assurance that Items relied on for safety will be available and reliable to perform their function when needed. Appropriate records of these items must be maintained by or under the control of the licensee throughout the life of the facility.

(2) Natural phenomena hazards. The design must provide for adequate protection against natural phenomena with consideration of the most severe documented tiisto[ical events for the site.

(3) Fire protection. The design must provide for adequate protection against fires and explosions.

(4) Environmental and dynamic effects. The design must provide for adequate

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protection from environmental conditions and 9ynamic effects associated with normal I

operations, maintenance, testing, and postulated accidents that could lead to loss of safety functions.

77

(5) Chemical protection. The design must provide for adequate protection against chemical risks produced from licensed materal, facility conditions which affect the safety of licensed material, and hazardous chemicals produced from licensed material.

(6) Emergency capability. The design must provide for emergency capability to maintain control of:

.J (I) Licensed material and hazardous chemicals produced from licensed material; (ii) Evacuation of on-site personnel; and (iii) Onslte emergency facilities and services that facilitate the use of available offsite services.

(7) Utility services. The design must provide for continued operation of essential utility

• services.

(8) Inspection, testing, and maintenance. The design of items relied on for safety must provide for adequate Inspection, testing, and maintenance, to ensure their availability and reliability to perform their function when needed.

(9) Criticality control. The design must provide tor crltlcallty control including adherence to the double contingency principle.

(1 O) *lnstrumentatlon and controls. The design must provide for inclusion of instrumentation and control systems to monitor and control the behavior of Items relied on for safety.

(b) Facility and system design and facility layout must be based on defense-In-depth practices. 1 The design must incorporate, to the extent practicable:

1 As used in § 70.64, Requirements for new facilities or new processes at existing facilities, defense-in-depth practices means a design philosophy, applied from the outset and through completion of the design, that is based on providing successive levels of protection such that health and safety will not be wholly dependent upon any single element of the design, construction, maintenance, or operation of the facility. The net effect of incorporating defense-in-depth-practices is a conservatively designed f acllity and system that will exhibit 78

(1) Preference fot 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.

§ 70.65 Additional content of applications.

(a) In addition to the contents required by § 70.22, each application must Include a description of the applicant's safety program established under § 70.62.

- (b) The Integrated safety analysis summary must be submitted with the license or renewal application (and amendment application as necessary), but shall not be incorporated in the llcense. However, changes to the Integrated safety analysis summary shall meet the conditions of§ 70.72. The Integrated safety analysis summary must contaih:

(1) A general description of the site with emphasis on those factors that could affect safety (i.e., meteorology, seismology);

(2) A general description of the f acillty with emphasis on those areas that could affect safety, including an Identification of the controlled area boundaries; (3) A description of each process (defined as a single reasonably simple integrated unit operation within an overall production line) analyzed In the integrated safety analysis In sufficient detail to understand the theory of operation; and, for each process, the hazards that were Identified in the integrated safety analysis pursuant to§ 70.62(c)(1 )(i)-(iii) and a general description of the types of accident sequences; greater tolerance to failures and external challenges. The risk insights obtained through performance of the integrated safety analysis can be then used to supplement the final design by focusing attention on the prevention and mitigation of the higher-risk potential accidents.

79

{4) Information that demonstrates the licensee's compliance with the performance requirements of§ 70.61, including a description of the management measures; the requirements for criticality monitoring and alarms in § 70.24; and, if applicable, the requirements of § 70.64;

{5) A description of the team, qualifications, and the methods u::,ed to perform the integrated safety analysis; (6) A list briefly describing each item relied on for safety which is identified pursuant to§ 70.61 (e) in sufficient detail to understand their functions in relation to the performance requirements of § 70.61; (7} A description of the proposed quantitative standards used to assess the consequences to an individual from acute chemicaJ exposure to licensed material or chemicals produced from licensed materials which are on-site, or expected to be on*site as described in§ 70.61 (b)(4) and (c)(4);

(8) A descriptive list that identifies all items relied on for safety that are the sole item I

preventing or mitigating an accident sequence that exceeds the performance requirements of

§ 70.61; and (9) A description of the definitions of unlikely, highly unlikely, and credible as used in the evaluations in the integrated safety analysis.

§ 70.66 Additional requirements for approval of license application.

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(a) An application for a license from an applicant subject to subpart H will be approved if the Commission determines that the applicant has complied with the requirements of

§§ 70.21, 70.22, 70.23, and 70.60 through 70.65.

(b) Submittals by existing licensees in accordance with § 70.62(c)(3)(i) will be approved if the Commission determines that:

(1) the integrated safety analysis approach is In accordance with the requirements of

§§ 70.61, 70.62(c)(1 ), and 70.62(c)(2); and (2) the schedule Is In compliance with § 70.62(c)(3)(1i).

(c) Submittals by existing licensees in accordance with § 70.62(c)(3)(il) will be approved If the Commission determines that:

(1) The requirements of § 70.65(b) are satisfied; and (2) The performance requirements in§ 70.61 (b), (c) and (d) are satisfied,*based on the information In the ISA Summary, together with other information submitted to NRG or available to NRG at the licensee's site.

§ 70.72 Facility changes and change process.

(a) The licensee shall establish a configuration management system to evaluate, Implement, and track each change to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel. This system must be documented in written procedures an~ must assure that the following are addressed prior to implementing any change:

(1) The technical basis for the change; 81

(2} Impact of the change on safety and health or control of licensed material; (3} Modifications to existing operating procedures including any necessary training or retraining before operation; (4) Authorization requirements for the change; (5) For temporary changes, the approved duration (e.g., expiration date) of the change;and (6) The impacts or modifications to tre integrated safety analysis, integrated safety analysis summary, or other safety program infonnation, developed in accordance with

§ 70.62.

(b) Any change to site, structures, processes, systems, equipment, components,

• computer programs, and activities of personnel must be evaluated by the licensee as specified in paragraph (a) of this section, before the change Is implemented. The evaluation of the change must determine, before the change is implemented, if an amendment to the license is required to be submitted in accordance with § 70.34.

(c) The licensee may make changes to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel, 'Without prior ,

Commission approval, If the change:

(1) Does not:

(i) Create new types of accident sequences that, unless mitigated or prevented, would exceed the performance requirements of § 70.61 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; 82

(2) Does not remove, without at least an equivalent replacement of the safety function, an Item relied on for safety that is listed In the Integrated safety analysis summary; (3) Does not alter any item relied on for safety, listed In the i~tegrated safety analysls summary, that Is the sole item preventing or mitigating an accident sequence that exceeds the performance requirements of § 70.61; and (4) Is not otherwise prohibited by this section, license condition, or order.

I (d)(1) For changes that require pre-approval under§ 70.72, the licensee shall submit an amendment request to the NRC In accordance with § 70.34 and § 70.65~

(2) For changes that do not require pre-approval under§ 70.72, the licensee shall submit to NRC annually, within 30 days _after the end of the calendar year during which the changes occurred, a brief summary of all changes to the records required by§ 70.62(a)(2) of this part.

(3) For all changes that affect the integrated safety analysis summary, the licensee

,_J shall submit to NRC annually, within 30 days after the end of the calendar year during which the changes occurred, revised Integrated safety analysis summary pages.

(e) If a change covered by§ 70.72 is made, the affected on-site documentation must be updated promptly.

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(f) The licensee shall maintain records of chanaes to its facility carried out under this section. These records must Include a written evaluation that provides the bases for the determination that the changes do not require prior Commission approval under paragraph (c) or (d) of this section. These reccrds must be maintained until termination of the license.

§ 70.73 Renewal of licenses.

83

Applications for renewal of a license must be filed in accordance with §§ 2.109, 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 references are clear and specific.

§ 70.74 Additional reporting requirements.

(a) Reports to NRC Operations Center. (1) Each licensee shall report to the NRC Operations Center the events described in Appendix A to Part 70.

(2) Reports must be made by a knowledgeable licensee representative and by any method that will ensure compliance with the required time period for reporting.

(3} The information provided must include a description of the event and other related information as described in§ 70.50(c)(1 ).

(4) Follow-up information to the reports must be provided until all Information required to be reported in § 70.50(c)(1) of this part is complete.

(5) Each licensee shall provide reasonable assurance that reliable communication with the NRC Operations Center is available during each event.

(b) Written reports. Each licem,ee that makes a report required by paragraph (a)(1) of this section shall submit a written follow-up report within 30 days of the initial report. The written report must contain the information as described in§ 70.50(c)(2).

§ 70. 76 Backfitting 84

(a) For each licensee, this provision shall apply to Subpart H requirements as soon as the NRC approves that licensee's ISA Summary pursuant to § 70.66. For requirements other than Subpart H, this provision applies regardless of the status of the approval of a licensee's ISA Summary.

(1) Backfitting is defined as the modification of, *or addition to, systems, structures, or

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components of a facility; or to the procedures or organization required to operate a facility;

) '

any of which may result from a new or amended provision in the Commission rules or the imposition of a regulatory staff position interpreting the Commission rules that is either new or different from a previous NRC staff position.

(2) Except as provided In paragraph (a)(4) of this section, the Commission*shall require a systematic and documented analysis pursuant to paragraph (b) of this section for backfits which It seeks to impose.

(3) Except as provided in paragraph (a)(4) of this section, the Commission shall require the backfitting of a facility only when it deter-nines, based on the analysis described in C'"' '

paragraph (b) of this section, that there is a substantial increase In the overall protection of the public health and safety or the common defense and security to be de.rived from the backfit and that the direct and indirect costs of Implementation for that facility are justified in view of this increased protection.

(4) The provisions of paragraphs (a)(2) and (a)(3) of this section are inapplicable and, therefore, backfit analysis is not required and the standards in paragraph (a}(3} of this section 85

do not apply where the Commission finds and declares, with appropriately documented evaluation for Its finding, any of the following:

(i} That a modification is necessary to bring a facility into compliance with Subpart H of this part; or

' (ii} That a modification is necessary to bring a facility Into compliance with a llcense or the rules or orders of the Commission, or into conformance with written commitments by the licensee; or (iii) That regulatory action Is necessary to ensure that the facility provides adequate

• protection to the health -and safety of the public and is in accord with the common defense and security; o"r (iv) That the regulatory action Involves. defining or redefining what level of protection to the public health and safety or common defense and secu~ty. should be regarded as adequate.

(5) The Commission shall always require the backfltting of a faclllty If It determines that the regulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the public and Is in accord witn the common defense and security.

(6) The documente,d evaluation required by paragraph (a)(4} of this section must

- include a statement of the objectives of and reasons for the modification and the basis for 86

invoking the exception. If immediate effective regulatory action is required, then the documented evaluation may follow, rather than precede, the regulatory action.

(7) If there are two or more ways to achieve compliance with a license or the rules or orders of the Commission, or with written license commitments, or there are two or more ways to reach an adequate level of protection, then ordinarily the licensee is free to choose the way that best suits its purposes. However, should It be necessary or appropriate for the

)

Commission to prescribe a specific way to comply with Its requirements or to achieve adequate protection, then cost may be a factor In selecting the way, provided that the objective of compliance or adequate protection is met.

(b) In reaching the determination required by. paragraph (a)(3) of this section,. the Commission will consider how the backfit should be scheduled in light of other ongoing regulatory activities at the facility and, in addition, will consider information available concerning any of the following factors as may be appropriate and any oth!;!r information relevant and material to the proposed backfit:

l (1) Statement of the specific objectives that the proposed backfit is designed to achieve; (2) General description of the activity that would be required by the licensee in order to complete the backfit; 87

(3) Potential change in the risk to the public from the accidental release of radioactive material and hazardous chemicals produced from licensed material; (4) Potential impact on radiological exposure or exposure to hazardous chemicals produced from licensed material of facility employees; (5) Installation and continuing costs associated with the backfit, including the cost of facility downtime; (6) The potential safety impact of changes in faclllty or operational complexity, incf uding the relationship to proposed and existing regulatory requirements; (7) The estimated resource burden on the NRC associated with the proposed backfit and the availability of such resources; (8) The potential impact of differences In facility type, design, or age on the relevancy and practicality of the proposed back.fit; and (9) Whether the proposed backfit is lnte~m or final and, if Interim, the justification for imposing the proposed backf it on an interim basis.

(c) No license will be withheld during the pendency of backfit analyses required by the Commission's rules.

88

(d) The Executive Director for Operations shall be responsible for implementation of this section, and all analyses required by this section shall be approved by the Executive Director for Operations or his or her designee.

19. Appendix A to part 70 Is added to read as follows:

Appendix A to Part 70--Reportable safety events Licensees must comply with reporting requirements in this appendix, except for (a)(1 ),

(a)(2), and (b)(4), after they have submitted an ISA Summary In accordance with

§ 70.62(c)(3)(11). Licensees must comply with (a)(1 ), (a)(2), and (b)(4) after (Insert 30 days after publication of this final rule). As required by 10 CFR 70. 74, licensees subject to the requirements in subpart Hof part 70, shall report:

(a) One hour reports. Events to be reported to the NRC Operations Center within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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:

(1) An inadvertent nuclear criticality.

(2) An acute Intake by an individual of 30 mg or greater of uranium in a soluble form.

(3) An acute chemical exposure to an Individual from licensed material or hazardous chemicals produced from licensed material that exceeds the quantitative standards established to satisfy the requirements in§ 70.61 (b)(4).,

(4) An event or condition such that no items relied on for safety, as documented in the Integrated Safety Analysis summary, remain available and reliable, in an accident sequence evaluated in the Integrated Safety Analysis, to perform their function:

(i) In the context of the performance requirements In§ 70.61 (b) and§ 70.61 (c), or 89

(ii) Prevent a nuclear criticality accident (i.e., loss of all controls in a particular sequence).

(5) Loss of controls such that only one item relied on for safety, as documented in the Integrated Safety Analysis summary, remains available and reliable to prevent a nuclear criticality accident, an9 has been in this state for greater than eight hours.

(b) Twenty-four hour rep.arts. Events to be reported to the NRC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery, supplemented with the infonnation in 10 CFR 70.50{c)(1) as It becomes available, followed .by a written report within 30 days:

(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 perfonnance requirements of§ 70.61.

(2) Loss or degradation of Items relied on for safety that results in failure to meet the perfonnance 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 satisfy the requirements of§ 70.61 (c)(4).

(4) Any natural phenomenon or other external event,\ including fires internal and external to the facility, that has affected or may have affected the intended safety function or availability or reliability of one or more items relied on for safety.

(5) An occurrence of an event or process deviation that was considered in the I

Integrated Safety Analysis and:

(i) Was dismissed due to its likelihood; or 90

(ii) Was categorized as unlikely and whose associated unmitigated consequences would have exceeded those in§ 70.61 (b) had the item(s) relied on for safety not performed their safety function(s).

(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 lofh day of September, 2000.

For the Nuclear Regulatory Commission.

~6~ nnetteVietti-Cok, Secretary of_ the Commission.

91

Mrs. Leo Drey DOC .: 1ED 7o 515 West Point Avenue US 1*1 C University City, MO 63130 P4 :01 November 24, 1999 Secretary of the Commission Ot U.S. Nuclear Regulatory Commisstou Washington, DC 20555-0001 AO."

Attention: Rulemakings and Adjudications Staff

Dear Sir or Madam:

I am writing with two questions related to uranium fuel fabrication plants. I live in St. Louis, about thirty miles from the ABB-Combustion Engineering plant, at Hematite, Missouri, a historic facility licensed in 1956 as the nation's first commerci~l uranium fuel fabrication plant. The plant was shut down in 1974 and because it was so radioactively contaminated, the plan, after decommissioning the plant, was to ship the debris off to a low-level radioactive waste facility for disposal. Instead, United Nuclear Fuels sold the plant to Combustion Engineering, and it has been operating -- and expanding -- ever since.

I. Would you please tell me the status of the rule that was being proposed in July 1999 for those NRC Part 70 licensees authorized to possess an amount of enriched uranium capable of resulti~g in a critical mass accident? The proposed amendments were to require that the licensees (a) perform an integrated safety analysis, and (b) design a safety program to provide protection against accidents that could release radioactive and hazardous materials to the environment. If such a rule has been finalized, may I please have a copy?

2. Although I have had concerns about the Hematite plant for years, I have been even more uneasy about its operation subsequent to its takeover by a foreign company, Asea Brown Boveri, as the relatively recent parent company of Combustion Engineering.

My question: would you please explain how a non-American company is allowed to own and control the Hematite plant and its special nuclear materials under the Atomic Energy Act of 1954, which has as a stated purpose the "Government control of the possession, use, and production of atomic energy and special nuclear material"? I'ye also never understood how foreign companies are now being allowed to purchase and own domestic nuclear power plants which also contain and generate special nuclear materials.

I am sorry to have to burden your office directly with these requests for information, but I am not computer web-site literate and do not have rea~y access to the Federal Register.

I will be happy to pay for any costs involved in retrieving, copying, or mailing this information. If there are any charges, please send your invoice to me at the above address.

Your response will be greatly appreciated.

Sincerely, DEC

• 2 \999

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J International Executive Offices *99 C"T 18 .p~-' *22 t, I 1 Batter'Jjmarch P ark P.O. Box 9101 .....,

Qitincy, Massachusetts 02269-9101 USA Telephone (617) 770-3000 F ax (617) 770-0700 Washington Office Suite 560, 1110 N. Glebe R oad Arlington, VA 22201 Telephone: (708) 516-4846 DOCKET NUMBER Fax: (708) 516-4850 PROPOSED RULE PR 10

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October 15, 1999 Secretary of the Commission Attention: Rulemakings and Adjudications Staff U.S . Nuclear Regulatory Commission Washington, D.C. 20555-0001 Re: 64 FR 41338; 10 CFR Part 70 Proposed rule: Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material/RIN 3150 AF22 To Whom It May Concern:

The National Fire Protection Association (NFPA), founded in 1896, is an international nonprofit membership organization, with over 68,000 members, dedicated to reducing the worldwide burden of fire and other hazards on the quality of life by providing and

• advocating scientifically based codes and standards, research, training, and education.

NFPA's standards making system is based on a consensus process through a democratic, dynamic and expeditious system. NFPA codes and standards are recognized and used around the world as reliable, state-of-the-art information for achieving fire safety.

NFPA commends the NRC for its on-going efforts to increase confidence in the margin of safety governing the domestic licensing of special nuclear material for licensees authorized to possess a critical mass of special nuclear material and that are engaged in special activities.

NFPA strongly recommends the NRC adopt by reference the 1998 edition of NFPA 801, Facilities Handling Radioactive Materials. NFPA 801 would apply to Section 70.62, Safety Program and Integrated Safety Analysis, which addresses protection from all relevant hazards, including radiological, criticality, fire and chemical. The NFPA standard would also apply to Section 70.64, Requirements for New Facilities or New Processes at Existing Facilities addresses fire protection.

OCT 19 1999

~edged by card OIJH Tl II 11111- .,.._.

Publishers of the National Fire Codesand National Electrical Code A non-profit membership organization dedicated to promoting safety from fire, electlicity, and relatea1hazards t hrough research, codes and standards, technical advisory services, and public education since 1896.

u.s. NUCLEARAE8ULATORYOClll8800 RULEMAKNi81~

OFFICE CF'l'Nl-.rARY OFnEallllBION

Secretary of the Commission U.S. Nuclear Regulatory Commission Page Two Reference of NFPA 801 is in keeping with the requirements of Public Law 104-113 "National Technology Transfer and Advancement Act" which requires Federal government agencies to use private sector-developed national consensus technical standards in carrying out public policy wherever appropriate.

Thank you for the opportunity to provide this statement. Please do not hesitate to contact me if I may be of further assistance.

Sincerely,

• ~C-~

Sara C. Yerkes Director, Government Affairs Cc: Anthony R. O'Neill, Vice President, Government Affairs Richard P. Bielen, Chief Systems and Applications Engineer 2

SIEMENS DOCKET NUMBER PROPOSED RULE PR 7D

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October 13, 1999 *99 OCT 18 P3 *21 JBE:99:057 Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Attn: Rulemakings and Adjudications Staff Gentlemen:

Subject:

Siemens Power Corporation's Comments on the Proposed Revision to 10 CRF Part 70 Siemens Power Corporation (SPC) has participated with the Nuclear Energy lnstitute's (NEl 's)

Facility Operations Steering Committee in a review of the subject proposed rule. SPC agrees with the comments forwarded by NEI on behalf of th e industry in its October 13, 1999 letter.

Very truly yours,

-1£-B. Edgar Staff Engineer Licensing

/las cc: Felix Killar, NEI Siemens Power Corporation 21 0 1 Horn Rapids Road Tel: (509) 375-8100 ~ OCT 19 1999 Richland, WA 99352 Fax: (509) 375-8402 ckrlowtedged by card ..........., fl 11111. . . . ...-n,

.~uCLEAR REGUlATORY ~

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O OFFICE OF SOLID WASTE AND EMERGENCY AD i ~ESPONSE Secretary of the Commission U.S. Nuclear Regulatory Commission DOCKET NUMBER Washington, DC, 20555-0001 PROPOSED RULE PB 7o

("'IFR'-/1338)

Attention: Rulemakings and Adjudications Staff

Dear Sir or Madam,

The purpose of this letter is to express the support of the Chemical Emergency Preparedness and Prevention Office (CEPPO) of the United States Environmental Protection Agency (EPA) for the proposed rule titled "Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material."

CEPPO is the EPA headquarters office responsible for developing, coordinating, implementing, and managing EPA's emergency preparedness, accident prevention, and public right-to-know activities. In this capacity, CEPPO implements section 112(r) of the Clean Air Act, which requires EPA to promulgate regulations for chemical accident prevention at U.S.

stationary sources. These regulations are contained in 40 CFR part 68, and require facilities holding more than a threshold quantity of any of 140 listed chemicals in a process to implement a risk management program (RMP). Section 112(r) of the Clean Air Act also contains a "general duty" clause (112(r)(l)), which requires owners and operators of facilities handling any extremely hazardous substance to identify hazards which may result from accidental releases of such substances, to design and maintain a safe facility taking such steps as are necessary to prevent releases, and to minimize the consequences of accidental releases which do occur.

Some NRC-licensed facilities subject to the proposed rule are likely to be subject 40 CFR part 68 or the general duty clause of CAA section 112(r)( 1), or both. The Commj,ssion has made general reference to this fact in the preamble of the proposed rule:

"The requirements and provisions in Subpart Hare in addition to, and not as4.bstitut~ 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." OCT I 9

~ctmowfedged by-card I I I ZIC 1999p -

I Internet Address (URL)

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2 EPA agrees with the Commission that the proposed rule is entirely consistent with EPA's RMP regulations and the general duty clause of the Clean Air Act, and contains appropriate complementary safety measures for facilities possessing a critical mass of special nuclear material. If you have any further questions concerning our comments on the proposed rule, please contact James Belke ofmy staff at (202) 260-7314.

ire r 7

ica Emergency Preparedness and Prevention Office

t' GE Nuclear Energy General Electric Company DOCKET NUMBER PR P.O. Box 780, Wilmington, NC 28402 910 675-5000 PROPOSED RULE -.--__....O ,,__

f/J'IFRo/t 33iJoctober 13, 1999 o - 0~,

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Secretary of the Commission

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• f I U.S. Nuclear Regulatory Commission OJ Attention: Rulemakings and Adjudications Staff :I::>

0 Washington, D. C. 20555-0001 l,,J

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Subject:

Comments on Proposed Rule 10 CFR 70: Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material (Federal Register. Vol. 64, No. 146, pp. 41338-41357, dated July 30, 1999)

Reference:

Docket 70-1113

Dear Madam:

General Electric's Nuclear Energy Production (GE NEP) facility in Wilmington, North Carolina is pleased to have the opportunity to submit the attached comments on the publicly noticed proposed rule making regarding 10 CFR 70 published July 30, 1999.

GE NEP manufactures low enriched BWR fuel, and therefore we are an effected party.

As you know, GE NEP has been heavily involved in work related to implementing improved methodology for fuel fabrication plants. The model utilized is the hazard analysis approach used by the chemical industry successfully for a number of years. The hazards analysis process has been expanded to provide risk based performance based focus to provide a mechanism that would focus the facility and the regulatory attention and efforts on those things that were most important with regard to safety.

Using the risk based performance orientated approach outlined in the proposed revision to 10 CFR 70, the licensee and the Nuclear Regulatory Commission will have available information with which to make more educated decisions in matters of safety. The integration also improves the implementation. The output more clearly communicates to the public the nature of key risks at a facility and identifies how those potential situations are prevented or mitigated as the case may be.

GE volunteered to implement this concept on February 19, 1997 as a part of upgrading a portion of the Wilmington, NC plant. The license renewal at the same time included a complete license definition of the implementing management systems for the conduct and use of the Integrated Safety Analysis (ISA) techniques. GE has provided summaries of all work completed to date and the implementation is continuing according to a schedule provided to the NRC. GE therefore supports the concepts embraced in the proposed Part 70 and the comments attached represent recommendations for improving the rule for implementation and operability.

OCT I 9 991

~cknowledged by card _ _ _.,....

IJ.S. NUCLEAR REGULA! Urt 1 \AJIWO~

RUI.EMMN3S & ~ STAF-1 OFFICE OFlHE SECRETARY OF COMMISSION

Secretary of the Commission October 13, 1999 Page 2 of 2 GE has supported the industry effort through NEI and helped forge the unified industry position on key elements of the proposed rule. We have also shared much of what we have learned in public forums and with the NRC as the proposed rule has been formulated.

GE is reasonably pleased with the current proposed rule. While it has taken considerable time and energy to reach this point, it is this kind of process which leads to a quality rule that works for all.

Sincerely, GE NUCLEAR ENERGY C. M. Vaughan, Manager Facility Licensing cc: CMV-99-051

Secretary of the Commission October 13, 1999 Page 1 of 4 Proposed Revisions to 10 CFR 70 Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material (Federal Register, Vol. 64, No. 146, pp. 41338-41357, dated July 30, 1999)

Comments Submitted by GE Nuclear Energy NRC Solicited Comments

1. Backfit Provisions (FR p 41340) - GE believes that the backfit provisions should be immediately effective for the new rule. The Commission has granted this to virtually all facilities - with the exception of the fuel fabricators.

The NRC has indicated that the basis of safety is not currently clear enough to make a backfit determination; however, that position must be questioned as all the facilities have been licensed, approved for operation and routinely inspected over roughly a 30 year period. Backfit has been granted to Part 76 facilities without anywhere near the formal interrelationship .

• 2. Preemption of OSHA (FR p 41342, 70.61) -As GE understands the proposed rule, it describes a situation wherein the current terms of the MOU between NRC and OSHA are incorporated into the regulations to avoid misunderstanding. This should result in more effective implementation for all concerned parties.

GE supports the proposed rule in this respect.

3. ISA Methodology (FR p 41346, 72.62(c)) - GE believes that the current proposed rule offers sufficient flexibility in selecting ISA methodology so that a broad spectrum of facilities can be addressed and such that licensees have flexibility to interface with their site processes, procedures and resources.

4. ISA Summary Update Frequency (FR p 41348, 72. 72(c)) - GE believes that the 90 day reporting of changes is entirely too frequent.

Secretary of the Commission October 13, 1999 Page 2 of 4 It would mean that the facility and the NRC would always have change reporting in progress. There is no need to have such real time knowledge. It is important that the licensee has real time knowledge.

The NRC only needs reasonably current knowledge, as the current is available and accessible to them at the site.

GE believes that the 12-month to 24 months as used in other places is satisfactory and is more efficient. This seems clearly justified based on the fact that all the information is available at the site and accessible to the NRC at any time .

• GE is also concerned that there has not been more consideration of extended license term along with the discussions of the timing of updates. One of the largest problems the NRC has experienced is the load of license renewals. A large part of this is the fact that facilities are required to completely resubmit their license application and re-demonstrate the safety of the plant even though it is operating and approved by the NRC. This is a highly inefficient process for both the NRC and the licensee.

An alternate that needs to be considered in association with this proposed revision to 10 CFR 70 is a more permanent license. With updates every 12 months for example there is no real need for the NRC to renew the license - it only becomes a maintenance chore to periodically see that all the information is there and acceptable. The NRC has periodically referred to this as a "living license". GE believes that the living license concept provides advantages for the NRC and the licensee.

The following comments address specific sections and features of the proposed revision to 10 CFR 70.

70.4 Definitions -

Critical mass of special nuclear material (SNM) means . . .. This definition uses 4 percent by weight of uranium-235 as one of the benchmarks. Since most of the LWR fuel manufacturing now operates at enrichments, the rule would be better served to be updated and use 5 percent by weight of uranium-235 .

Secretary of the Commission October 13, 1999 Page 3 of 4 Items relied on for safety . . .. The term of items relied on for safety and then defined here is a disconnect. The problem is the use of the term "items". With the definition presented, a better choice would be to use the term "measures relied on for safety". Additionally, where "items relied on for safety" is used in the rule it should be changed to "measures relied on for safety".

70.62 Safety program and integrated safety analysis (a) (3) -

(3) Identifies a log to be maintained. GE believes that it is inappropriate to add this extra record-keeping burden on the licensee, because the licensee already has to generate records of this nature to manage their business and some different log is unnecessary work.

The wording should be changed from "Each licensee shall establish and maintain a log, available for NRC inspection, documenting each discovery ... " to "Records shall be established and maintained by the licensee, and available for NRC inspection, documenting ... ".

The other places the "log" is mentioned needs to be changed to "records".

• 70.65 Additional content of applications -

70.65 (b) is written to represent the ISA summary to be a single document. In practice it will be a sequence of documents that cover the facility. All should be in the same format, but it should be clear that it is not a single summary - but of course in some cases, it could be.

70.65 (b) (3) defines a process description as a requirement. In actual process, the process may well be broken down differently. The AIChE guidelines give guidance about the requirements for applying the hazard analysis techniques to the process being studied. So specifying here it should be the segment or node, and several of these could be combined and called a process if and only if the boundaries established for the hazard analysis match.

Secretary of the Commission October 13, 1999 Page 4 of 4 70.65 (b) ( 6) requires a brief list describing all items relied on for safety.

See GE's comments under 70.4 Definitions. To be clear, this should read "measures relied on for safety".

70.72 Facility changes and change process -

70.72 (c) (1) (I) and (c) (2), (c) (3) uses the term of ISA summary as the decision making document for change and change process. It is incorrect to use the ISA Summary in this manner.

The ISA, which is the detailed licensee generated document that the licensee uses to manage their program, is the reference document.

Summaries are just that - to provide a general level of information about the more important elements of the safety system for operations as determined in accord with the licensed program.

70.72 (c) (2) and (c) (3) use the term of "item relied on for safety". This is in error and should read "measure relied on for safety".

70.72 (d) (1) uses 90 days for making reports of the change. This is discussed elsewhere and should be of a frequency of 12 - 24 months .

• 70.73 Renewal of license -

Since the NRC is considering mandatory reporting of changes to the license making a living license for the facility, it is appropriate for this section to eliminate the requirement for renewal since it is not necessary.

Appendix A to Part 70 - Reportable Safety Events -

(a) (4) Needs to be looked at by the NRC in view of the need to change "items relied on for safety" and change "ISA summary" to "measures relied on for safety" and "ISA".

The most appropriate wording here would seem to be "measures relied on for safety as described in the ISA summary".

October 14, 1999 NOTE TO: Emile Julian Assistant for Rulemakings and Adjudications FROM: Carol Gallagher /1 _.. ,, Li, /} I A,.J ___,

ADM,DAS ~ r~J*~

SUBJECT:

DOCKETING OF COMMENT ON PROPOSED RULE - DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL; POSSESSION OF A CRITICAL MASS OF SPECIAL NUCLEAR MATERIAL

• Attached for docketing is a comment letter related to the subject proposed rule. This comment was received via the rulemaking website on October 13, 1999. The submitter's name is Charles M. Vaughan, GE Nuclear Energy, PO Box 780, MC J26, Wilmington, NC 28402.

Please send a copy of the docketed comment to Ted Sherr (mail stop T-8A-33) for his records.

Attachment:

As stated cc w/attachment:

T. Sherr

NUC L EAR ENERGY INSTITUTE Marvin S. Fertel SENIOR VICE PRESIDENT, NUCLEAR INFRASTRUCTURE SUPPORT

( r-_:_ 000fCErEO & INTERNATIONAL PROGRAMS

\ _: OCT 1 8

• *.S::, I October 13, 1999 \::)

Secretary of the Commission ~

U.S. Nuclear Regulatory Commission DOCKET NUMBER R Attention: Rulemakings and Adjudications Staff PROPOSED RULE p 7/J Washington, D.C. 20555-0001 ( lol/Fl<J/1338)

REFERENCE:

Comments on Proposed Rule 10 CFR 70: Domestic Licensing of Special Nuclear Material ; Possession of a Critical mass of Special Nuclear Material (Federal Register Vol. 64, No. 146, pp. 41338-41357 dated July 30, 1999)

Dear Sir or Madam:

The Nuclear Energy Institute 1 (NEI), on behalf of the nuclear fuel cycle industry, submits the attached comments on the proposed revisions to 10 CFR 70 in response to a reque st for public input in the July 30, 1999 Federal Register notice.

NEI is generally pleased with the proposed revisions to 10 CFR 70. There remain,

• however, a number of provisions in the proposed rule where additional revisions are necessary to achieve the intent of promulgating an affective, safety-focused, performance-based rule. Our comments on these revisions are addressed in the attachment to this letter. NEI additionally provides comments on several issues identified in the Federal Register announcement for which comments were solicited by the Commission.

NEI appreciates the opportunity to work closely with the Commission, NRC Staff and other stakeholders in developing a revised 10 CFR 70. We compliment the NRC on its solicitation of stakeholder participation, the scheduling of public 1 NEI is t he organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect/engineering firms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry.

OCT 19 llU_

1 776 I STREET NW SUITE 400 WASHINGTON DC 20006-3708 PHONE 202 739 81 25 FAX 202 293 3451 www ne1 org

u...... ~UCLEAR REGULATORY COMMISSION RULEMAKINGS &ADJUDICm'ION8 IWF OFFlCEOFlHESECfETARY OF THE COMMISSION Ooculnat'f 8ldlllcl PostmarkDa1e 11;/21/ q, 1

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Rulemakings and Adjudications Staff Nuclear Regulatory Commission October 13, 1999 Page 2 meetings and workshops to facilitate the exchange of ideas and the use of the NRC's Rulemaking Web Page to post comments and draft revisions. We trust that a similar process and commitment of NRC resources will continue to be made to develop a Standard Review Plan (SRP) to guide both license applicants and NRC Staff license reviewers in preparing and assessing license applications and amendments.

Yours Sincerely, Marvin S. Fertel Attachment c.: The Honorable Greta Joy Dicus, Chairman, NRC The Honorable Nils J. Diaz, Commissioner, NRC The Honorable Edward McGaffigan, Jr. , Commissioner, NRC The Honorable Jeffrey S. Merrifield, Commissioner, NRC Dr. William D. Travers, EDO/NRC Dr. Carl J. Paperiello, Deputy Executive Director, Materials, Research and State Programs, NRC Mr. William Kane, Director, NMSS/NRC C:\Files\Part 70\Federal Register Cover Letter.msw

COMMENTS ON PROPOSED REVISIONS TO 10 CFR 70 DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL; POSSESSION OF A CRITICAL MASS OF SPECIAL NUCLEAR MATERIAL (Federal Register Vol.64 No. 146 pp. 41338-41357 dated 30 July 1999)

COMMENTS SUBMITTED BYTHE NUCLEAR ENERGY INSTITUTE Introduction Revisions to 10 CFR 70 are designed to improve confidence in the margin of safety at fuel cycle facilities through application of a safety-focused and performance-based regulatory approach. The Integrated Safety Analysis (ISA), a risk-informed forward-looking assessment of credible facility hazards and their effects on plant systems and modes of operation, will provide information vital to evaluating the safe,ty basis of a facility. Rule revisions will focus licensee and NRC resources on those facility operations that could pose the greatest risk to human health and safety and the environment. The revisions are intended to reduce the regulatory burden on licensees and the NRC by granting the former the right to make changes to the facility or its processes without seeking a license amendment for changes that maintain or improve safety. While the existing Part 70 licensed facilities have an excellent safety record and the NRC's oversight and regulation of these facilities has been effective in protecting public health and safety, the revisions to 10 CFR 70 should enhance confidence in the margin of safety at such facilities.

The safety-focused regulatory approach incorporated in the Part 70 revisions will reduce the regulatory burden on both the NRC and licensee. For example, the proposed revisions should limit license amendment requests to those few facility changes that could have a direct impact on safety. The proposed revisions also introduce specific licensee performance requirements, strP.aroline licensee reporting requirements and formalize incorporation of baseline design criteria for new facilities and for new processes at existing facilities.

NEI is generally pleased with the proposed rule revisions and with the manner in which they address issues raised in NEI's Petition for Rulemaking dated July 2, 1996. These revisions address NRC's current risk-informed regulatory approach, in which risk information is used in concert with operating experience and engineering judgement to ensure safe operation of fuel cycle facilities. NRC's statutory responsibilities in radiological safety and chemical safety directly related to licensed material should be better addressed with the new Part 70.

NEI's major comments on the proposed rule revisions focus on the following:

• Facility Change Process (§70. 72)

• ISA Summary (§70.65)

• Standards for Protection of Co-Located Workers (§70.61)

• Backfit Provision In addition to comments on the above provisions, NEI has identified corrections or comments on the following sections:

• Correct usage of terminology (§70.4)

• Incorporation of NRC-OSHA MOU for chemical safety (§70.50)

• Failure log (§70.62)

• Safety program scope (§70.62)

• ISA (§70.62)

• Baseline design criteria and safety grading (§70.64)

• Incorporation of a rule implementation provision Each of these comments is discussed below in the order in which they appear in 10 CFR 70.

Reasonable Assurance (§70.4)

NEI recommends that the term "reasonable assurance" be used in place of "ensure" in the definition of 'available and reliable to perform their function when needed: in

§70.4. "Ensure" connotes a high degree of certainty - bordering on a guarantee -

that a goal or objective will be met. In the Part 70 context, it may be interpreted to require certainty that an items relied on for safety will be available and reliable when required. Regardless of the thoroughness of personnel training, or the engineering excellence and quality assurance applied to the design and construction of an item relied on for safety, or the redundancy designed into a facility's safety control systems, failures and abnormal events will inevitably occur. Thus, a licensee should be expected to provide reasonable assurance that a particular items relied on for safety will be reliable and available, when required. In the comparable regulations pertaining to items relied on for safety at nuclear power facilities (10 CFR 50, Appendix A, Criterion 1), the licensee must only " ... provide adequate assurance that these structures, systems and components will satisfactorily perform their safety functions" [underlining added]. NEI recommends that comparable language be used in the §70.4 definition of this term.

NEI recommends that a licensee provide reasonable assurance that an item relied on for safety will be available and reliable when it is required to perform its safety function. Thus, during periods when a process is shut down, undergoing maintenance or when Special Nuclear Material is no longer present, compliance 2

with the safety requirements of §70.61 would no longer be required. What is important is that an item relied on for safety be available and reliable, when it is needed. NEI believes this meaning is conferred by the definition by the words "when needed" and that "continuous" be simply deleted. The definition should read:

§70.4 Definitions: Available and reliable to perform their function when needed: " ... means that ... items relied on for safety will perform their intended safety function when needed and management measures will be implemented to provide reasonable assurance of compliance with the performance requirements of §70. 6 .. ."

For consistency in the Part 70 revisions, NEI recommends that the term "adequate assurance" in §70.64(a)(l) be replaced by "reasonable assurance."

Reporting Requirements (§70.50)

The reporting requirements of §70.50 continue to misrepresent the principles of the 1988 NRC-OSHA Memorandum of Understanding (MOU). §70.50(c)(l)(iii)(A) requires the reporting of chemical hazards and §70.50(c)(l)(iii)(B) requires the reporting of personnel exposures to chemicals. Although the MOU principles have been correctly incorporated into other proposed revisions to 10 CFR 70 (e.g. §§70.4, 70.61(b), 70.62(c), 70.64(a), 70.74 Appendix A), they are incorrectly referenced in

§70.50. MOU principle (2) limits NRC jurisdiction to regulation of chemical hazards of licensed material and hazardous chemicals produced from licensed material. The two aforementioned sections of §70.50 should be corrected to properly incorporate the MOU principles.

Controlled Area and Co-Located Workers (§70.61)

§70.61(£) requires a licensee to establish a 'controlled area' for a facility in which it can control the activities of personnel. §70.61 states that any individual located outside of the controlled area is subject to the lower (public) radiation dose limits.

NEI is concerned with the manner in which §70.61 could set radiation exposure limits for co-located workers. We are particularly concerned with the treatment of radiation exposures from an NRC-licensed facility present on a DOE site (e.g. a MOX fabrication facility on a DOE property). As currently written a worker (as defined in §70.4) who leaves the controlled area to perform a work-related function would have to be treated as a member of the public when performing the ISA and would be subject to the more stringent public radiation exposure limits. Outside of the controlled area the TEDE limit of 0.1 rem for members of the public would apply (cf. 10 CFR 20.1301(a)(l)) rather than the annual TEDE occupational dose limit of 5 rems (10 CFR 20.1201). Such a problem has arisen at the Hanford Tank Waste Remediation System-Privatization where NRC subjects 'co-located' workers to the appreciably lower public dose limits.

3

NEI recommends that the NRC apply constant rediation exposure limits to all plant workers, regardless of their presence inside or outside of the controlled area. The 10 CFR 70 regulations should be harmonized with comparable DOE radiation exposure limits.

NEI recommends that the phrase " ... any individual ... " in sections b(2) and c(2) be clarified to exclude facility workers who may have occasion to work outside of the controlled area. This phrase should be amended to read " ... any individual (other than a worker) ... "

Safety Program Definition {§70.62)

There is inconsistent use of the term "safety program" throughout the proposed revisions. For example, sometimes the rule implies that the ISA Summary is part of the safety program (it is not), and thereby part of the license. The explanatory notes in the Federal Register also erroneously describe the safety program; for example, on page 41346, it (correctly) states that the ISA comprises one component of the safety program, but then (erroneously) states that the results of the ISA must be submitted for NRC approval. This is inconsistent with our understanding developed during the NRC workshops and clearly not consistent with the direction given by the Commission in the Staff Requirements memorandum dated December 1, 1998. §70.62(a)(l) defines the licensee's safety program to consist of three components (process safety information, ISA, management measures). This definition is too narrow. The safety program includes these important components, but also includes the commitments and programs addressed in the eleven chapters of the Standard Review Plan (e.g. radiation protection, compliance with 10 CFR 20 occupational radiation exposure limits, etc.). In this regard, NEI recommends that the last sentence in §70.62((a)(l) be deleted. The content of §70.22 adequately defines the requirements for a licensee safety program.

Log o{Failures (§70.62)

The regulatory reporting requirements of §70.62(a)(2) and §70. 7 4(a)(l) direct a licensee to report to NRC Headquarters within one to twenty-four hours instances in which an item relied on for safety or management measure has failed or been discovered to be non-operational. The NRC will, therefore, already possess all of the information sought in the "log of §70.62(a)(3). Tabulating data that the NRC already possesses and has presumably internally analyzed, seems to be a wasteful and inefficient use of licensee and NRC resources that should be focused exclusively on safety-significant issues. This is an unnecessarily prescriptive requirement.

NEI, therefore, recommends that §70.62(a)(3) be deleted from the rule.

4

Integrated Safety Analysis (§70.62)

NEI has two comments with the timing requirements specified in §70.62(c)(3) for completion of an ISA by existing licensees:

(i) for consistency the phrase " ... [the date of publication of the final rule]'

in the first sentence should be replaced by " ... the effective date of the rule ..." as has been done in subsections (i), (ii) and (iii), and (ii) the 4-year period for conducting the ISA and for modifying the plant to address any identified unacceptable performance deficiencies may be too short. Also, we recommend that the period should start on the date on which the NRC approves the plan required in subsection 3(i). If the clock starts on the effective date of the rule and the NRC takes one year to approve the ISA plan, the licensee will be unduly hampered.

There should be some incentive for the NRC to complete its approval process in a timely manner. NEI is also concerned over the limited time available for a licensee to not only conduct the ISA, but also to implement any modifications to the facility as is required by

§70.62(c)(3)(iii). Based on the fact that licensees who have already committed to perform ISAs were generally given five years to complete them, NEI recommends that an existing licensee be granted 5 years to complete the ISA. We also recommend that appropriate and sufficient time be allowed for the licensee to present to the NRC and to implement a plan to correct any identified unacceptable performance deficiencies. Finally, we recommend of imposition of a 90-day time frame on the NRC to issue a decision on the acceptability of a licensee's ISA approach. NEI recommends that subsection (ii) be re-written to read: " ... (ii) Within 5 years of the date of NRC approval of the licensee's plan, complete an ... "

• For consistency with the language in §70.62(a) ( ... "the safety program may be graded such that management measures applied are commensurate with the reduction of risk attributable to that item .. .'), NEI recommends that the second sentence in §70.62(d) be revised to include the term "graded." This sentence would then read: "... The measures applied to a particular engineered or administrative control or control system may be graded commensurate with the reduction of the risk attributable to that control or control system ... "

Baseline Design Criteria (§70.64) 10 CFR70.22(i)(l)(ii) and 70.22(i)(3) require a license applicant to design an emergency plan to respond to the radiological hazards of an accidental release of special nuclear material. This plan must address how on-site workers will be protected (§70.22(i)(3)(v)). The 'Emergency Capability' design criterion presented in

§70.64(a)(6) requires a license applicant to plan for the ' ... evacuation of personnel ... "

5

This criterion should be made specific to on-site personnel to be consistent with the requirements of 10 CFR 70.22. A licensee should not be required to demonstrate that the design provides for evacuation of off-site personnel, for this implies the involvement of FEMA and constitutes emergency preparedness measures that have never before been placed on fuel cycle facility licensees. NEI recommends that

§70.64(a)(6)(ii) be written as: " ... Evacuation of on-site personnel; and ... "

The proposed definition of 'items relied on for safety' (§70.4) encompasses both physical devises and activities of personnel. §70.64(a)(8) requires that the design of items relied on for safety provides for inspection, testing and maintenance of items relied on for safety. However, such activities (testing and maintenance) can not be applied to activities of personnel. The breadth of the term items relied on for safety req_uir_es_addition_of language that is appropriate -to administ!!ative controls and--to personnel activities. Either the definition of 'items relied on for safety' should be rewritten to separate engineered and administrative safety controls (e.g. 'items relied on for safety' versus '[personnel] activities relied on for safety'),or additional qualifying text should be inserted. NEI recommends that baseline design criterion (8) be rewritten as follows:

" ... the design of items relied on for safety must provide for adequate inspection, testing, and maintenance, or adequate training, testing and qualification for personnel whose activities relied on for safety. to ensure their availability and reliability to perform their function when needed ... "

§70.64(b) directs a license applicant to apply 'defense-in-depth practices' to the facility design and then indicates that engineered controls should be used in preference to administrative safety controls. Consistent with the ability granted a licensee to grade all aspects of its safety program (cf. §70.62(a)), grading of the defense-in-depth safety concepts in the design of the facility should also be permitted. Safety design criterion (b)(l) appears unnecessarily prescriptive by discouraging a licensee from using anything but an engineered safety control. So long as the licensee can satisfactorily demonstrate that an Rdministrative safety control or a system of administrative and engineered controls will enable the performance criteria to be satisfied, the choice of items relied on for safety and the nature of 'defense-in-depth' practices that is applied should be flexible.

Technical Edits:

(i) §70.64(a): the last sentence refers to a paragraph (c). There is no paragraph (c) in §70.64. This sentence should be deleted.

(ii) §70.64(a)(5): the words "produced from" in the first sentence should be deleted. The sentence should read: " ... The design must provide for adequate protection against chemical risks of licensed material, plant ... "

6

(iii) §70.64(a)(9): for consistency with correct terminology, NEI recommends that the word 'nuclear' be added before 'criticality.'

ISA Summary (§70.65)

NEI's comments on §70.65(b), which outlines the content of an ISA Summary, relate to the level of detail that will be expected in this document. The rule should not prescribing an acceptable level of detail, but should defer this issue to be developed in the SRP. Use of terms such as " ... types of accident sequences ... " rather than detailed description of each accident sequence in §70.65(b)(3) is commended.

However, in §70.65(b)(6) the required level of descriptive detail for items relied on for safety (" ... sufficient detail ... ") remains vague. NEI recommends that information at the 'systems level' should be required, rather than at the 'component' or 'sub-component' level.

The ISA will examine the proposed locations, quantities and risks of chemicals at the facility and determine whether any could serve as an initiating event for a credible accident sequence or potentially affect the safety of licensed materials. For any chemicals falling into either category and posing a high or intermediate risk, their locations and characteristics would need to be specified in the ISA Summary.

A licensee should not be required to keep the NRC appraised of the contents of bulk storage tanks kept in the back 40" if the ISA has shown that they pose no credible risk to the facility, its operations, the public and the environment. Thus, NEI believes the §70.65(b)(7) requirement for information on the locations of on-site chemicals is unnecessary.

§70.65(b)(3) seeks information on each process analyzed in the ISA including the hazards identified for each. This information should not extend to include process safety information that is specifically excluded from the requirements of §70.65. As the ISA Summary only requires specification of the items relied on for safety for high- and intermediate-consequence events (cf. §70.65(b)(6)), there should be no need for an applicant to include in the ISA Summary information on processes and hazards for accident sequences and processes that are determined in the ISA not to produce consequences that exceed the performance criteria of §70.61. Such information will, however, be maintained at the facility site for review by NRC staff.

The ISA Summary should, consequently, only address those processes for which accident sequences have been identified that would produce consequences that exceed the performance criteria of §70.61.

§70.65(b)(6) requires the applicant to list all items relied on for safety for high- and intermediate-consequence events and any other accident sequences for which the licensee has defined items relied on for safety. This is far too broad a requirement.

The items should only need to be described at the systems level, rather than at the component or sub-component level. While this list will include " ... activities of 7

personnel relied on for safety ... " it should not include procedures that the personnel must follow. As procedures are constantly being adjusted, revised and improved, their inclusion in the list of items relied on for safety would necessitate frequent revisions to the ISA Summary that may have little if any safety significance.

The ISA Summary should, therefore, provide a concise summary of pertinent information on technology, equipment and hazardous materials used in each process, but not include detailed process safety information that is maintained at the facility as 'ISA documentation.' As stated before, the on-site ISA documentation is available for review by the NRC Staff if such detailed information need be examined.

Facility Change Mechanism (§70. 72)

The intent of the new Facility Change Mechanism is to permit, based on specific criteria, the licensee to make certain changes to the facility and its operations that maintain or improve safety without seeking NRC pre-approval. The merits of this new mechanism are threefold: (i) the NRC need only assess safety-significant changes, (ii) the licensees' regulatory burden (and commitment of resources) to filing license amendment requests for even the most safety benign changes is reduced, and (iii) protection of public health and safety and the environment will be enhanced by directing regulatory attention to potentially higher-risk conditions.

However, as currently worded, §70. 72 will not achieve the intended goals.

§70. 72(a) requires that any change to the facility be formally evaluated by means of the configuration management (CM) system to evaluate, among other things, its potential impact on safety and the need to modify the ISA and ISA Summary. This requirement is too broad and all-encompassing and would require CM evaluation of changes having no or absolutely minimal effect on health and safety (e.g. office remodeling, planting of shrubbery, changing paint colors). Rather than to first evaluate every change by means of CM, the licensee should first rely on internal procedures to initially screen any proposed changes for their potential adverse safety impacts. If this preliminary screening indicates that implementing the change could place the licensee at risk of not meeting the performance requirements of 10 CFR 70.61, then the change would be evaluated by the ISA methodology and the CM system. What appears to be outlined in §70. 72(a) are steps in the ISA methodology that will be used in evaluating a proposed change. The CM system function will be applied in evaluating the change and in recording it in the facility on-site documentation so as to ensure consistency among design requirements, physical configuration and facility documentation. §70.72(b), like §70.72(a),

contains unsatisfactory language by requiring that any change to the facility be first evaluated by CM to establish if there is need for a license amendment.

8

§70. 72(c) prescribes what facility changes can be made without NRC pre-approval.

This section is patterned after a similar provision for nuclear power reactors described in 10 CFR 50.59(a) and (b). In contrast to 10 CFR 50.59, which addresses changes to the facility's safety analysis report (the equivalent document for Part 70 licensees being the ISA Summary), §70. 72(c) would again apply to any change to the facility, its operating procedures or items relied on for safety. The NRC pre-approval exclusion provision should be patterned after 10 CFR 50.59, whereby only changes to contents of the ISA Summary would be required to be reviewed under

§70.72(c). Incorporating corrections to terminology (discussed below) this section should be revised to read: "(c) the licensee may make changes to the site, processes or items relied on for safety as described in the ISA Su'!',,mary, without prior ... "

§70. 72(c)(l)(i) parallels the language of 10 CFR 60.69(a)(2) whereby NRC pre-approval would be required for " ... an accident ... of a different type than any evaluated previously in the safety analysis report ... " NEI concurs with this criterion.

However, the footnote appended to "new types of accidents" is contrary to the stated goal of limiting requests for license amendments to those that are safety significant.

The footnote's reference to accident initiators, changes in consequences and changes in the safety function of a control could be literally interpreted to require essentially any change to the facility to require NRC pre-approval and a license amendment.

NEI strongly recommends that the footnote be deleted for consistency with the intent of 10 CFR 70. 72.

§70. 72(d) requires notification to the NRC within 90 days of any change that does not require NRC pre-approval, but for which changes to the ISA Summary were necessary. The corresponding reporting period for nuclear power licensees for such changes can be as long as 24 months (cf. 10 CFR 50. 71(e)(3)(i)). Information pertaining to the change (e.g. ISA analysis and supporting documentation, CM information) will be available at any time at the facility for NRC inspection and review. For consistency with the Facility Change Mechanism reporting requirements (§70. 72 (d)(3)), NEI recommends that all changes be reported annually to NRC headquarters consistent with the recommendations of the Commissioners.

NEI understands the intent of §70. 72(d)(3) to be a requirement for licensees to submit annually a brief summary of facility changes that are implemented without NRC pre-approval, whether or not they affect the ISA Summary. However, the wording of this section f' ... records required by §70.62(a)(2) of this part ...") will inadvertently and significantly expand the information that would have to be reported. §70.62(a)(2) requires records not only pertaining to the ISA (and ISA Summary), but also pertaining to process safety information and management measures. §70. 72(d)(3) will, therefore, require the licensee to submit voluminous information that could include updates of process safety information including drawings, process flow diagrams, piping and instrument diagrams, information on process chemicals, technology, equipment and process conditions (temperatures, 9

intermediates, pressures, etc.), updates of material safety data sheets (MSDS), etc.

The licensee has made a binding license commitment to maintain this detailed information at the facility as one of the components of the facility safety program (cf. 10 CFR 70.62(a)). Information that would be reported annually under

§70. 72(d)(3) pertains to changes having a low risk or safety significance. Annual submission to NRC Headquarters of such detailed information of low safety significance seems unnecessary and as such, this section should be reworded to read:

" ... a brief summary of all changes to the integrated safety analysis and ISA Summary, that are made without prior Commission approval, must be submitted to the NRC every 12 months ..."

To summarize, the Facility Change Mechanism (§70.72) should be revised to:

(i) incorporate consideration of risk in deciding which changes should be controlled by the CM system and clearly distinguish between ISA functions and CM (ii) demonstrate consistency in the use of terminology (e.g. 'items relied on for safety' rather than the formerly used 'structures, systems and components')

(iii) delete the footnote for §70.62(c)(l)(i)

(iv) permit a licensee to improve or enhance an item relied on for safety without seeking NRC pre-approval (v) lengthen the reporting time frame of §70.72(d)(l) to one year in accordance with the Commissioners' in the July 1999 SRM (vi) clarify the annual reporting requirements of §70. 72(d)(3) to encompass descriptions of changes made to the facility without NRC pre-approval and to exclude submission of up-dated data that should remain at the facility (referred to as ISA documentation).

Backfit Provision:

NEI has documented in letters to former Chairman Jackson (May 26, 1999) and to Dr. Carl Paperiello (February 12, 1999) and in its September 1996 Petition for Rulemaking (PRM-70-7) why an immediately effective backfit provision should be included in 10 CFR 70. We continue to believe that application of the backfit provision upon the effective date of the revised rule is justified and appropriate.

Our concern with the timing of the backfit provision is accentuated by the refusal of the Staff to implement an immediately-effective backfit provision in 10 CFR 76 that the Commission had approved and directed.

NEI clearly believes that the safety bases of Part 70 facilities are sufficiently well understood to permit a backfit provision now. The fuel cycle facilities' exemplary operating history must provides demonstrable evidence that there is current 10

understanding of "safety bases" (even in the absence of an ISA). The ISA's primary advantage is that it will better and more efficiently direct the NRC and licensees' attention to what has been existing practice of licensees - focusing attention and resources on safety significant issues. By conducting systematic safety analyses of Part 70 facilities, a licensee will be able to qualitatively assess the improvement in public health and safety that a change can afford. NEI has consistently advocated implementation of a qualitative methodology to derive the safety benefit of a back.fit modification. This approach obviates the need to establish the incremental risk of a proposed facility modification and acknowledges the inappropriateness of applying quantitative methodologies to Part 70 facilities.

We are particularly concerned with the open-ended time frame to implement the back.fit provision. The NRC has approved license renewals for fuel fabricators that incorporated an ISA into their license renewal application. For those licensees that have prepared an ISA and had their license renewed, NEI believes that an immediately effective back.fit provision should be implemented.

In summary, NEI recommends that:

(i) back.fit language be included as part of the proposed 10 CFR 70 revisions, and (ii) the back.fit provision be immediately effective to those processes or parts of an existing facility for which the ISA has been completed.

NEI provided the NRC with streamlined language for an immediately effective back.fit provision in its letter to former Commissioner Shirley Jackson on May 26, 1999. NEI recommends that this language be incorporated into the Part 70 reV1s1ons .

• Implementation Provision:

The proposed revisions to 10 CFR 70 should have an implementation provision similar to that presented in 10 CFR 20.1008. NEI believes that such an implementation provision should be included in the Part 70 revisions to address potential conflicts between existing license conditions and the new Part 70 requirements. We believe this additional provision is necessary, especially in light of license conditions modeled after proposed Part 70 revisions that have added to licenses recently renewed by the NRC.

Topics for Which Comment Are Solicited:

The Federal Register notice solicited public and stakeholder comments on the following four topics:

11

(1) Backfit Provision~- Reg. P.41340]: NEI has commented above that the back.fit provision should be immediately effective upon approval of a licensee's license renewal, and in any case, upon submittal of an ISA Summary to the NRC (See comments and rationale above)

(2) NRC-OSHA Preemption [Egg. Reg. P.41342 re §70.61]: The NRC/OSHA MOU is, in NEI's view, consistent with the statutory allocation of jurisdiction between the NRC and OSHA, and serves as a useful frame of reference for discussing these issues. The proposed treatment of chemical hazards in Part 70 revisions will not encroach in any way on OSHA's traditional authority over non-radiological chemical hazards at NRC licensed facilities. Had the NRC retained authority over purely non-radiological hazardous chemicals that could have no impact on radiological safety, OSHA would have had a legitimate concern that the rule would unintentionally "preempt" OSHA regulations. However, the draft rule makes clear that the ruJe encompasses: (1) the hazards of NRC-licensed materials; and (2) non-radiological hazards that may affect the safety of NRC-licensed materials; but not (3) purely chemical hazards or other potentially hazardous working conditions that are within OSHA's province. By clarifying these parameters of the rule, the NRC has appropriately limited its role so as not to intrude on OSHA's traditional authority. This would include OSHA's Process Safety Management rules, its Permissible Exposure Limits (PELs), and other OSHA requirements.

The §70.62 requirement to perform an ISA will not preempt OSHA requirements. While licensees will need to determine whether any non-radiological chemicals present at their sites could affect the safety of NRC-licensed materials, the NRC would not impose any restrictions on the use or handling of such chemicals unless, through the ISA, it was determined that they could have such an effect. (In doing so, the NRC would actually be regulating the safety of the licensed material itself, rather than regulating the "direct" hazards of the non-radiological chemicals). Finally, NEI does not believe that OSHA would be precluded from addressing workplace hazards arising out of the decommissioning process -- so long as it does not attempt to regulate the hazards of licensed material subject to NRC jurisdiction.

(3) ISA Methodology: ~ - Reg. P.41346 re §72.62(c)] NEI believes the rule offers a license applicant or licensee sufficient flexibility in selecting an appropriate ISA methodology. We have no comments to add.

(4) ISA Summary Update Frequency~- Reg. P.41348 re §72.72(c)]: NEI wholeheartedly supports the Commissioners' recommendation that the timeframe for reporting changes to the ISA Summary to the NRC be 12

r lengthened from 90 days to 12 months. As discussed above, we believe the reporting period should be consistent with that imposed on reactor licensees (12 to 24 months). The analyses of any changes made to the ISA (and which may have to be included in the ISA Summary) will be available at the facility for review and inspection. We do not foresee an adverse safety impact by retaining the updated information at the facility and submitting it to the NRC on an annual basis.

Miscellaneous Comments: NEI urges the NRC to correct inaccuracies that are present in Part II 'Description of Proposed Action' of the Federal Register notice.

We are concerned that such inaccuracies could contribute to public misunderstanding of the intent of the rule's provisions. Examples of such inaccuracies include the following:

(1) Terminology: several terms that we understood were included in earlier drafts, but since were deleted from the rule remain. For example: 'facility vulnerability' (instead of 'unacceptable performance deficiency'),

'structures, systems and components' (instead of 'items relied on for safety')

(2) SRM (December 1998): statements of approach are included with no comment that they were subsequently deleted, and therefore give the reader an incorrect impression of what is contained in the revisions. For example, on p. 41339 it is stated that the ISA results are part of the application and license and that a preliminary ISA and decommissioning ISA are needed.

(3) Safety Grading: the text continues to erroneously indicate that safety grading is required (p. 41341) even though the rule language was changed.

(4) ISA: the text (p. 41347) states that the ISA can be used to supplement the final design of the facility. This suggests that a second ISA (or at least an update of the initial ISA) may be required, even if no changes are made to the facility design during construction Ref \Ftleo\Part 70\Federal Reei.ster Comment Letter.m.sw 13

DOCKET NUMBERPR PROPOSED RULE. '11J October 12, 1999 8

'99 OCT 14 P3 :49 L-99-004 l I Secretary of the Commission AU.

Attn. Rulemakings and Adjudications Staff U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

USEC Comments on Proposed 10 CFR 70, "Domestic Licensing of Special Nuclear Material; possession of a Critical Mass of Special Nuclear Material", dated 7/30/99

Dear Sirs:

The following comments are submitted in response to the July 30, 1999, Federal Register notice (64FR41338) proposing to amend 10 CFR Part 70, Domestic Licensing of Special Nuclear Material.

The new 10 CFR 70, and its precursor 10 CFR 76, are important steps in the evolution of nuclear materials licensing to prepare for the submittal of the next major fuel facility license application. The next application may be submitted by USEC for an enrichment facility. These amendments should develop an improved basis for regulation of future 10 CFR 70 facilities by achieving better consistency of regulatory practice with the content of the regulations. Further, they should provide for a more predictable and stable licensing process to increase the likelihood of success for the next enrichment plant license application.

USEC has been an active participant in activities aimed at improving regulation under 10 CFR 70 since 1993. During this time many 10 CFR 70 proposals, comments and suggestions have been exchanged between NRC and the industry. The proposed rule represents the culmination of an arduous process to craft a regulation that will be acceptable by many competing interests. USEC would like to recognize NRC's perseverance and offer the attached constructive comments to assist with completion of this task. The explanation of the comments is followed with a copy of 10 CFR 70 revised to reflect the comments in redline/strikeout and as-revised formats.

Sincerely, J. William Bennett Vice President, Advanced Technology ~CT f 9 1999

~edged by cara ...... .,..,,,....,.~.......

6903 Rockledge Drive, Bethesda, MD 20817-1818 Telephone 301-564-3200 Fax 301-564-3201 http://www.usec.com Offices in Livermore, CA Paducah, KY Portsmouth, OH Washington, DC

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USEC Comments on Proposed 10 CFR 70, "Domestic Licensing of Special Nuclear Material; possession of a Critical Mass of Special Nuclear Material", dated 7/30/99 The following comments are in the order in which they appear in 10 CFR 70.

1) Performance Requirements: The Federal Register Notice requests comments with respect to-the clarity-and effectiveness of the language used per the-June 1, 1998 Presidential Memorandum. We find the language in §70.61(b) and §70.61(c) could be substantially clearer and have offered a plain language version of this section in the attachment.

2) Safety Program: The safety program is broader than the three elements identified in 70.62(a)(l) as: 1) process safety information, 2) integrated safety analysis, and 3) management measures. Fuel cycle facility safety programs encompass the three elements identified plus all of the other topics addressed in the license application.

This includes, for example, radiation safety, criticality safety, chemical safety and frre protection in addition to the three elements directly associated with the integrated safety analysis.

This comment can be readily addressed by deleting the reference to the three elements in 70.62(a)(l) and clarifying the application requirements in 70.65(a) as provided in the attachment.

3) Submittal of changes: §70.72(d)(l) requires submittal of affected pages ofthe integrated safety analysis summary within 90 days of making a change pursuant to

§70.72. §70.72(d)(3) requires a brief summary of changes made that do not affect the integrated safety analysis summary every 12 months. This frequency is to allow NRC staff to review the changes being made to the facility in enough time to ensure that the licensee's evaluations of potential impacts to health and safety are accurate and to maintain facility and safety information on the docket current. The Statements of Consideration indicate that the Commission is particularly interested in comments concerning the 90 day time period for submitting updated ISA summary pages.

10 CFR 50.59 allows changes to be made to production and utilization facilities without prior NRC approval under certain conditions. While different in content than

§70.72, the general concept is the same. That is, changes to the facility may be made without prior NRC approval. Both §50.59 and §70.72 also require that NRC be notified of these changes.

§50.59 requires that a brief description, including a summary safety evaluation; be submitted to NRC for each change either annually or with the update of the FSAR which may not exceed 24 months. This frequency has been adequate for many years for production and utilization facilities. The frequency for updating facility and safety information on the docket and informing the NRC staff of what changes have been made ihould be the same for fuel cycle facilities as for pro~uction and

utilization facilities. As there has been over 25 years experience with 10 CPR 50.59, the frequency established there should be sufficient.

As noted in the Statements of Consideration, the content proposed for §70.72 is deliberately different than §50.59. The 70.72 proposal is expected to result in fewer license amendments than one patterned after §50.59. The amount and type of license amendments required by §70.72 is expected to be consistent with past practice at fuel cycle facilities. USEC believes ,that the content of §70. 72 is appropriate for .fuel cycle-facilities and should be different than §50.59.

However, the difference in content of §50.59 and §70.72 does not justify a difference in the frequency for submitting information regarding the changes to NRC, as suggested in the Statements of Consideration. For the past 30 years or so, fuel cycle licensees have been able to make many changes without any notification ofNRC.

The addition of §70.72 will give NRC additional information on changes at the facilities. This additional information can be acquired without more frequent submittals than those at production and utiliz.ation facilities.

We would note one other difference between the submittal requirements of §70.72 and §50.59. §50.59 requires a brief description and summary safety evaluation be submitted for each change. §70. 72 requires a brief summary of changes that do not affect the ISA Summary and revised ISA Summary pages without explanation for changes affecting the ISA Summary. USEC believes that NRC will benefit from a description of changes made to the ISA Summary. Accordingly, §70.72 should require brief descriptions and summary safety evaluations of each change made pursuant to §70.72 and require that an updated ISA Summary be provided on a biennial basis. More current information will be maintained available for NRC inspection at the site as required by §70.72(f). The text of proposed 10 CPR 70 has been modified accordingly in the attachment.

4) Facility changes and change process: §70.72 adds requirements for a configuration management system and for making changes. As written, the requirements apply to all site, structures, processes, systems, equipment, components, computer programs, and activities of personnel regardless of safety significance. Compliance with these requirements would appear to require configuration management and change control applied to everything on the site of the licensed facility. This could include the wastewater treatment facility, a laser facility, the administration building, maintenance of the shrubbery, etc. Every change would require an evaluation and a summary submitted to the NRC.

Inclusion of items on the site that make no contribution to the NRC regulated safe operation of the facility would place an undue burden on the licensee. To remedy this, we propose that the configuration and change process be limited to any " ....

changes to the site, processes or items relied on for safety as described in the ISA Summary, without prior. .. ". The text of proposed 10 CPR 70 has been modified accordingly in the attachment.

5) Operation: The Statements of Consideration at 64 FR 41346 indicates that applicants for licenses to operate new facilities or new processes at existing facilities would be expected to update their ISAs based on as-built conditions and submit the results to NRC before operation. The process for uranium enrichment facilities that must comply with §70.23a would differ from this description. Uranium enrichment facilities would submit a complete license application, including an ISA summary, for constru_ctiou and operation.. This. application would.be the basis for NRC review and culminate in issuance of a license for construction and operation. Following issuance of the license, the licensee would institute change control pursuant to §70.72. The licensee would then be required to submit summaries of changes and ISA summary updates as required by §70.72. An inspection would verify that the facility has been constructed in accordance with the license prior to operation as required by §70.32(k).

No pre-operational submittal and review of an updated ISA summary is anticipated for uranium enrichment facilities as their configuration would be controlled since issuance of the construction and operation license. No changes to 10 CFR 70 are needed to resolve this comment.

6) Back:fit: The Statements of Consideration request specific comments on the Commission's intent to defer consideration of a qualitative backfit provision. It further solicits suggestions for backfit provisions specifically applicable to fuel cycle backfit needs; requests identification of information available to conduct the analysis associated with backfits; and asks what period of time is reasonable before a backfit provision should be implemented.

USEC firmly believes that deferring consideration of a backfit provision would be evading an extremely important issue. The revisions to 10 CFR 70 will result in a dramatic change in the regulations applicable to fuel cycle facilities. Regardless of intentions to make the new regulations clear and explicit, there are many opportunities for interpretations of the new regulations. Differing interpretations of the language in the rule are predictable. A key difference in interpretations could result in the need to modify or add to plant systems, structures, components, procedures, or organization. It is certain that some key differences in understanding, interpretation or position will lead to justifiable differences of opinion between members of the staff and the licensee.

USEC's view ofbackfitting aligns almost exactly with NRC's "Backfitting Guidelines" (NUREG-1409) dated July 1990. The following two paragraphs describe this view.

Backfits are expected to occur as part of the regulatory process to ensure safety. It is important for sound and effective regulation, however, that backfitting be conducted by a controlled and defined process. The backfitting process is intended to provide for a formal, systematic, and disciplined review of new or changed positions before imposing them.

The backfit process enhances regulatory stability by ensuring that changes in regulatory staff positions are justified and suitably defined. For example, even if not needed to meet the standard for adequate protection or to ensure compliance, back:fitting is proper if a substantial safety benefit is realized and the costs are justified by the safety benefit.

The proposed 10 CFR 70 changes many things. It adds substantive new performance requirements, new design basis criteria, new reporting requirements,- new safety -

analysis requirements, new requirements for management measures and a new change control process. All of these new provisions can add uncertainty to the regulation of fuel cycle facilities. It is vital that a formal, systematic, and disciplined review of new, changed or differing positions that could backfit existing facilities be applied to increase regulatory certainty. The back:fit provision provides for this systematic review.

10 CFR 76, Certification of Gaseous Diffusion Plants contains a back:fit provision

(§76. 76). §76. 76 is very similar to 10 CFR 50.109 and should serve as a model for a provision to be included in 10 CFR 70. Many of the same arguments that have been raised in opposition to the inclusion of a back:fit provision in 10 CFR 70 were raised in opposition to §76.76.

Former Commissioner Remick, commenting on SECY 93-285, addressed similar concerns registered by the staff regarding the incorporation of a back:fit provision in 10 CFR 76. He wrote:

"I believe that the proposed regulations should contain a back:fit provision which is as much like §50.109 as possible. I would think, for instance, that all of

§50.109(a)(2)-(7) and (c) could apply in the new context. We should make use of the experience embodied in the back:fit rule. Doing so will add some consistency

• to our regulatory practices. The only flexibility it will deprive us of is the flexibility to impose ill-considered backfits."

No change to the back:fit language in 10 CFR 50.109 is needed to allow for qualitative analysis. There has been considerable discussion of a qualitative versus a quantitative backfit provision. NEI proposed and USEC endorses the use of the tried and true backfit language used successfully in 10 CFR 50.109. This is neither a quantitative nor a qualitative backfit provision. The standard incorporated in the rule is that backfitting will be required if there is a "substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection." NRC's own guidance in NUREG/BR-0058, Rev 2, "Regulatory Analysis Guidance of the U.S. Nuclear Regulatory Commission" states the Commission's preference that quantitative analyses are much preferred over qualitative ones.

The staff contends that a quantitative determination of incremental risk would require a Probabilistic Risk Assessment. This is clearly not the case. While the existence of Probabilistic Risk Assessments may aid the staff in quantifying the increase in the overall protection of the public, it is by no means essential that Probabilistic Risk Assessments exist as a basis for backfit analyses.

The Commission revised the reactor backfit rule (10 CFR 50.109) in 1985 "to establish standards and an agency discipline for future management ofbackfitting for power reactors." This was well before Probabilistic Risk Assessments were available for many reactors. Indeed, it wasn't until late 1991, as required by Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR

§50.54(£)," that risk analysis information became widely available for reactors. This was years after the revision of 10 CFR 50.109.

NRC guidance recognizes the need for flexibility in quantification and offers substantial information available for fuel cycle risk quantification. "Backfitting Guidelines" (NUREG-1409) gives examples of situations in which the backfit rule does not require a strict quantitative showing that benefits exceed costs, but rather "that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived.

"Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission" (NUREG/BR-0058, Rev.2) anticipates the need for flexibility in quantification. It states:

"Estimated values and impacts should be expressed in monetary terms whenever possible; many regulatory actions, such as those affecting ... materials licensees, may not be supported by available PRA analysis ... the staff needs to make every reasonable effort to apply alternative tools that can provide a quantitative perspective ... concerning the value of the proposed action;

• [Where PRAs or other statistics-based analyses are not available] the generally recommended approach is to utilize whatever data may be available within a simplified model to provide some quantitative perspective;

[Where quantification is not possible] reliance on the qualitative approach should be a last resort, to be used only after efforts to develop pertinent data or factual information have proved unsuccessful; "The Regulatory Analysis Technical Evaluation Handbook" (NUREG/BR-0184) provides guidance to the analyst on how to prepare regulatory analysis and implements the policy in NUREG/BR-0058. Appendix C ofNUREG/BR-0184 provides information for performing regulatory analysis for non-reactor facilities.

Appendix C discusses the need for quantification as follows:

" ... the analyst should strive to use quantitative attributes when performing a regulatory analysis for non-reactor licensees. The Commission has determined, for example, that PRA should be used for analyses involving materials licensees when the potential safety consequences warrant its use, sufficient data are

available, and the licensees can reasonably b e expected to be capable of performing such analyses (NRC 1996c). However, it should be recognized that there are many benefits of improved regulation of non-reactor facilities that do not lend themselves to quantification. For example, increased confidence in the margin of safety may be a non-quantifiable benefit of a particular proposed regulatory requirement. As noted in Section 4.5, non-quantifiable benefits and costs can be significant elements of a regulatory analysis and need to be considered.by the analyst,and decision-maker as appropriate:" -

NUREG/BR-0184, Appendix C contains estimated accident frequencies and other information and references to assist the analyst in quantifying regulatory analyses for fuel cycle facilities.

PRA was not a prerequisite to §50.109, nor to 10 CFR 76.76, nor is it required to prepare a quantitative determination of incremental risk for fuel cycle facilities. NRC guidance recognizes the need for qualitative as well as quantitative arguments. Just as a regulatory analysis was prepared for proposed 10 CFR 70, responsible regulatory analyses can be performed on potential back.fits of fuel cycle facilities.

This new regulation will be applied to facilities that have been operating for over 30 years. Changes will likely be required at the facilities, most of which will be voluntarily undertaken by the licensee. There will also likely be differences between the licensee and some members of the NRC staff regarding what, and the extent of, changes that should be made. Adoption of a back.fit provision allows these differences to be examined on a cost/benefit basis through a disciplined process.

The Statements of Consideration state "Without a baseline determination of risk, as provided by the initial ISA process, it is not clear how a determination of incremental risk, as needed for a back.fit analysis, would be accomplished." USEC does not believe that a comprehensive risk baseline is necessary before reasoned judgements can be made on the benefits and risks of a proposed back:fit. USEC agrees with the staff that conducting an ISA is beneficial and will enhance our mutual ability to understand the integrated risk of operation of these facilities. However, fuel cycle facilites, like our gaseous diffusion plants, have operated for many years. The risks associated with the facilities are largely known from years of operational experience and from numerous analyses that have been performed. NUREG/BR-0184 Appendix C provides a comprehensive summary of the information that is available. There is plenty of basis on which to evaluate the relevant benefits and costs of potential back:fits and this will be added to with the performance of ISAs.

Fuel cycle back:fit needs are not dissimilar to production and utilization back.fit needs.

§50. l 09 was the product of a concerted effort by the industry to stem the flow of new staff requirements and positions that started shortly after the Three Mile Island incident in March 1979. This incident prompted the issuance of numerous bulletins, orders and other NRC direction that resulted in modifications or additions to plant systems, structures, components, procedures and organization. The Commission saw

the need to formalize and achieve a disciplined process for review of new or changed NRC staff positions before imposing them. The Sequoyah Fuels incident in 1986 and the General Electric incident in 1991 were the Three Mile Islands of the fuel cycle industry. It is appropriate and needed to enhance regulatory certainty in the fuel cycle industry by ensuring that changes in regulatory staff positions are justified and suitably defined by inclusion of an immediately effective backfit provision in 10 CFR 70.

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, 124.6"(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).

Sections 70.1(c) and 70.20a(b) also issued under secs. 135,141, Pub. L.97-425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161). Section 70.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Section 70.21(g) also issued under sec. 122, 68 Stat.

939 (42 U.S.C. 2152). Section 70.31 also issued under sec. 57d, Pub. L.93-377, 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36 and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Section 70.61 also issued under secs. 186, 187, 68 Stat. 955 (42 U.S.C. 2236, 2237). Section 70.62 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).

2. The undesignated center heading "GENERAL PROVISIONS" is redesignated as "Subpart A--General Provisions."

3. In Sec. 70.4, the definitions of Acute, Available and reliable to perform their function when needed, Configuration management, Critical mass of special nuclear material, Double contingency, Hazardous chemicals produced from licensed material, Integrated safety analysis (ISA), Integrated safety analysis summary, Items relied on for safety, Management measures, Unacceptable performance deficiencies, and Worker are added, in alphabetical order, as follows:

Sec. 70.4 Definitions.

Acute as used in this part means a single radiation dose* or chemical exposure event or multiple radiation dose or chemical exposure events occurring.within a short time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less).

Available and reliable to perform their function when needed as used in subpart H of this part means that, based upon the analyzed, credible conditions in the integrated safety analysis, items relied on for safety will perform their intended safety function when needed and management measures will be implemented that ensure continuous compliance with the performance requirements of Sec. 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.

Configuration management (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 when needed.

Critical mass of special nuclear material (SNM) means special nuclear material in a quantity exceeding 700 grams of contained uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1500 grams of contained uranium-235, if no uranium enriched to more than 4 percent by weight of uranium-235 is present; 450 grams of any combination thereof; or one-half such quantities if massive moderators or reflectors made of graphite, heavy water, or beryllium may be present.

Double contingency means *a process design that incorporates sufficient factors*of -*

safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.

Hazardous chemicals produced from licensed materials means substances having licensed material as precursor compound(s) or substances that physically or chemically interact with licensed materials; that are toxic, explosive, flammable, corrosive, or reactive to the extent that they can endanger life or health if not adequately controlled. These include substances commingled with licensed material, and include substances such as hydrogen fluoride that is produced by the reaction of uranium hexafluoride and water, but do not include substances prior to process addition to licensed material or after process separation from licensed material.

Integrated safety analysis (ISA) means a systematic analysis to identify plant and i external hazards and their potential for initiating accident sequences, the potential accident sequences, their likelihood and consequences, and the items relied on for safety. As used here,

• integrated means joint consideration of, and protection from, all relevant hazards, including radiological, nuclear criticality, fire, and chemical. However, with respect to compliance with the regulations of this part, the NRC requirement is 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.

Integrated safety analysis summary means the document submitted with the license

-application, license amendment application, or license renewal application that provides a synopsis of the results of the integrated safety analysis and contains the information specified in Sec. 70.65(b).

Items relied on for safety means structures, systems, equipment, components, and activities of personnel that are relied on to prevent potential accidents at a facility that could exceed the performance requirements in Sec. 70.61 or to mitigate their potential consequences.

This does not limit the licensee from identifying additional structures, systems, equipment, components, or activities of personnel (i.e., beyond those in the minimum set necessary for compliance with the performance requirements) as items relied on for safety.

Management measures mean the functions performed by the licensee, generally on a continuing basis, that are applied to items relied upon for safety, to ensure the items are available and reliable to perform their functions when needed. Management measures include configuration management, maintenance, training and qualifications, procedures, audits and assessments, incident investigations, records management, and other quality assurance elements.

Unacceptable performance deficiencies mean deficiencies in the items relied on for safety or the management measures that need to be corrected to ensure an adequate level of protection as defined in 10 CFR 70.61(b), (c), or (d). .J

Worker means an individual whose assigned duties in the course of employment involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation (i.e., an individual who is subject to an occupational dose as in 20 CFR 20.1003).

4. In Sec. 70.8 paragraph (b) is revised to read as follows.

Sec. 70.8 Information collection requirements: 0MB approval.

(b) The approved information collection requirements contained in this part appear in Secs. 70.9, 70.17, 70.19, 70.20a, 70.20b, 70.21, 70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, 70.39, 70.42, 70.50, 70.51, 70.52, 70.53, 70.57, 70.58, 70.59, 70.61, 70.62, 70.64, 70.65, 70.72, 70.73, 70.74 and Appendix A.

5. The undesignated center heading "EXEMPTIONS" is redesignated as "Subpart 8--Exemptions."

• Secs. 70.13a and 70.14 [Redesignated]

6. Sections 70.13a and 70.14 are redesignated as Secs. 70.14 and 70.17, respectively.

7. The undesignated center heading "GENERAL LICENSES" is redesignated as "Subpart C-General Licenses." ,

8. The undesignated center heading "LICENSE APPLICATIONS" is redesignated as

Subpart D-License Applications."

9. The.undesignated center heading "LICENSES" is redesignated as "Subpart E--Licenses."

10. The undesignated center heading "ACQUISITION, USE AND TRANSFER OF SPECIAL NUCLEAR MATERIAL, CREDITORS' RIGHTS," is redesignated as Subpart F-Acquisition, Use, and Transfer of Special Nuclear Material, Creditors' Rights."

11. The undesignated center heading SPECIAL NUCLEAR MATERIAL CONTROL RECORDS, REPORTS AND INSPECTIONS" is redesignated as "Subpart G--Special Nuclear Material Control Records, Reports, and Inspections."

12. In Sec. 70.50 paragraph (c) is revised and paragraph (d) is added to read as follows.

Sec. 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 Sec. 70.74 and appendix A of this part if applicable, by_telephone to the NRC Operations .

Center.<SUP>3&It;/SUP> To the extent that the information is available at the time of notification, the information provided in these reports must include:

\3\ The commercial telephone number for the NRC Operations Center is (301) 816-5100.

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(i) Caller's name, position title and call back telephone number; (ii) Date, time, and exact location 9.f the event; (iii) Description of the event, including; (A) Radiological or chemical hazards involved including isotopes, quantities, and chemical and physical form of any material released; (B) Actual or potential health and safety consequences to the workers, the public, and the environment, including relevant chemical and radiation data for actual personnel exposures to radiation or radioactive materials or chemicals (e.g., level of radiation exposure, concentration of chemicals, and duration of exposure);

(C) The sequence of occurrences leading to the event, including degradation or failure of structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences; and (D) Whether the remaining structures, systems, equipment, components, and activities of personnel relied on to prevent potential acqidents or mitigate their consequences are available and reliable to perform their function.

(iv) External conditions affecting the event; (v) Additional actions taken by the licensee in response to the event; (vi) Status of the event (e.g., whether the event is on-going or was terminated);

(vii) Current and planned site status, including any declared emergency class; (viii) Notifications related to the event that were made or are planned to any local, State, or other Federal agencies; -

(ix) Status of any press releases related to the event that were made or are planned.

(2) Written report. Each licensee who makes a report required by paragraph (a) or (b) of this section, or by Sec. 70. 74 and appendix 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:

(i) Complete applicable information required by Sec. 70.50(c)(1);

(ii) The probable cause of the event, including all factors that contributed to the event and the manufacturer and model number (if applicable) of any equipment that failed or malfunctioned; (iii) Corrective actions taken or planned to prevent occurrence of similar or identical events in the future and the results of any evaluations or assessments; and (iv) For licensees subject to subpart H of this part, whether the event was identified and evaluated in the Integrated Safety Analysis.

(d) The provisions of Sec. 70.50 do not apply to licensees subject to Sec. 50.72. They do apply to those part 50 licensees possessing material licensed under part 70 who are not subject to the notification requirements in Sec. 50. 72.

13. The undesignated center heading "MODIFICATION AND REVOCATION OF LICENSES" is redesignated as "Subpart I--Modification and Revocation of Licenses."

Secs. 70.61 and 70.62 [Redesignated]

14. Sections 70.61 and 70._62 are redesignated as Secs. 70.81 and 70.82, respectively.

15. The undesignated center heading "ENFORCEMENT' is redesignated as "Subpart J-Enforcement."

Secs. 70.71 and 70.72 [Redesignated],

16.-Sections 70.71' and-70:72 are redesignated as Secs. 70.91 and 70.92, respectively.

17. In part 70, a new subpart H (Secs. 70.60-70.74) is added to read as follows:

Subpart H-Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear Material Sec.

70.60 Applicability.

70.61 Performance requirements .

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.

70.72 Facility changes and change process.

7d. 73 Renewal of licenses.

70. 74 Additional reporting requirements.

Sec. 70.60 Applicability.

The regulations in Sec. 70.61 through Sec. 70.74 apply, in addition to other applicable

,Commission regulations, to each applicant or licensee that is or plans to be: authorized to possess greater than a critical mass of special nuclear material, and engaged in enriched uranium processing, fabrication of uranium fuel or fuel assemblies, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap recovery of special nuclear material,' or any other activity that the Commission determines could significantly affect public health and safety. The regulations in Sec. 70.61 through Sec. 70.74 do not apply to decommissioning activities performed pursuant to other applicable Commission regulations including Sec. 70.25 and Sec. 70.38 of this Part.

Also, the regulations in Sec. 70.61 through Sec. 70.74 do not apply to activities that are certified by the Commission pursuant to Part 76 of this chapter or licensed by the Commission pursuant to other parts of this chapter.

Sec. 70.61 Performance requirements.

(a) Each applicant or licensee shall evaluate, in the integrated safety analysis performed in accordance with Sec. 70.62, its compliance with the performance requirements in paragraphs (b), (c), and (d) of this section.

(b) The rislc of each credible high consequence event must be limited, unless the event is highly unlikely, through the application of engineered controls, administrati*;e controls, or both, that reduce the likelihood of occurrence of the event or its consequeneeeach

~ -~--

c~1bie

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accident sequence whose unmitigated consequences could exceed those below must be reduced to be highly unlikely through the application of engineered controls, administrative controls, or both. Applieetief'I ef edditief'lel eeF1tfels is Aet required fef these high eeF1sequeF1ee e't'eF1ts deffieF1stFeted te be highly Uf'llil(ely.(Comment 1)

(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 located outside the controlled area identified pursuant to paragraph (f) of this section; or (4) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that:

(i) Could endanger the life of a worker, or (ii) Could lead to irreversible or other serious, long-lasting health effects to any individual located outside the controlled area identified pursuant to paragraph (f) of this section.

High eeF1sequeF1ee e't'ef'lts ere these iF1teFF1elly ef exteFFlelly if'litieted e't'ef'lts tt'tet result if'I:

(2) AFI acute dese ef 0.25 S*, (25 Feffi) ef gFeetef tetel effeeti't'e dese equi't'elef'lt te eF1y if'ldi't'iduel leceted eutside the eeF1tfelled eree ideF1tified puraueF1t te peFegFept, (f) ef this seetief'I; (3) AFI if'lt8I(e ef 30 ffig ef gfeetef ef Uf8FliUffi if'I seluble fefffi by BFIY iFldi't'iduel leceted eutside the eeF1trnlled eFee ideF1tified puFsuef'lt te peFBgFept't (f) ef this seetief'I; ef (4) AFI eeute et'teft'lieel expesure telf 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 Sec. If BFI epplieef'lt pessesses ef plef'ls te pessess queAtities ef ffieteFiel eepeble ef suet'! et'teffiieel expesures, tt'teFI the epplieef'lt shell prepese eppFepFiete queAtiteti't'e steF1defds fer these t'teeltt't effects, es70.65 of this part ef the iFlfefffietieFI subffiitted pursuef'lt te See.

70.85 ef this peft(c) The likelihood of each credible accident sequence whose unmitigated consequences could exceed those below must be reduced to unlikely through the application of engineered controls, administrative controls, or both.

- - - (e) The rislt(Comment 1)

(1) An acute worker dose of 0.25 Sv (25 rem) or greater total effective dose equivalent; (2) An acute dose of eect't eredible if'ltefffiediete eeF1sequeF1ce e't'ef'lt ft'lust be lift'lited, uF1less the e't'ef'lt is uAlil(ely, tt'!Feugt'I0.05 Sv (5 rem) or greater total effective dose equivalent to any individual located outside the epplieetief'I ef eF1giF1eered eeF1tFels, edffiif'listFeti't'e eeF1tfels, er beth, tt'tet reduee the lil(elit'teed ef eceuFFeF1ce ef the e't'ef'ltcontrolled area identified pursuant to paragraph (f) of this section; (3) A 24-hour averaged release of radioactive material outside the restricted area in concentrations exceeding 5000 times the values in table 2 of appendix B to 10 CFR part 20; or its eeF1sequeF1ee (4) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that:

(i) Could lead to irreversible or other serious, long-lasting health effects to a worker, or (ii) Could cause mild transient health effects to any individual located outside the controlled area as specified in paragraph (f) of this section. Applieetief'I ef edditieF1el ceF1tfels is Aet required fef these iF1teFffiediete eeF1sequeF1ce e1,eF1ts deffieF1stFeted te be Uf'llil(ely.

IAterft'lediete ceF1sequeF1ee e't'eF1ts ere these iF1teFF1elly ef exteFFlelly if'litieted e't'ef'lts, tt'tet ern F1et high eeF1sequeF1ee e't'eF1ts, that result if'I:

(2) AFI aeute dese ef 0.05 S't' (5 reffi) ef grnatef tetel effeeti*,e dese equi't'alef'lt te &Fly if'ldi*,idual leeated eutside the Cef'ltFelled BfeB ideFltified f'UfSUBFlt te paFegFeph (f) ef this seetief'I;

(4) Afl eeute el-lemieel mcpesur:e te Bfl ifldi¥iduel fFem lieeflsed meteFiel eF 1-lezefdeus el-lemieels pFedueed ffem lieeflsed meteFiel ti-let:

(i) Geuld lead te iFfe¥eraible eF eU~eF seFieus, leflg lestiflg 1-leeltl-l effects te a **YeFIEeF, eF (ii) Geuld eeuse mild tFSAsieflt 1-leeltl-l effects te efly ifldi¥iduel leeeted eutside the eefltFelled area es specified ifl peFegFepl-l (f) ef tl-lis sectiefl. 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 Sec. 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 margin of subcriticality for safety. Preventive controls and measures must be the primary means of protection against nuclear criticality accidents.

(e) Each engineered or administrative control or control system necessary to comply with paragraphs (b), (c), or (d) of this section shall be designated as an item relied on for safety .

The safety program, established and maintained pursuant to Sec. 70.62 of this part, shall ensure that each item relied on for safety will be available and reliable to perform its intended function when needed and in the context of the performance requirements of this section.

(f) Each licensee must establish a controlled area, as defined in Sec. 20.1003, in which the licensee retains the authority to determine all activities, including exclusion or removal of personnel and property from the area. For the purpose of complying with the performance requirements of this section, individuals who are not workers, as defined in Sec. 70.4, may be permitted to perform ongoing activities (e.g. , at a facility not related to the licensed activities) in the controlled area, if the licensee:

(1) Demonstrates and documents, in the integrated safety analysis, that the risk for those individuals at the location of their activities does not exceed the performance requirements of paragraphs (b)(2) , (b)(3), (b)(4)(ii), (c)(2), and (c)(4)(ii) of this section; or (2) Provides: training in accordance with 10 CFR 19.12(a)(1 )-(5) to these individuals to ensure that they are aware of the risks associated with accidents involving the licensed activities as determined by the integrated safety analysis, and conspicuously posts and maintains notices stating where the information in 10 CFR 19.11 (a) may be examined by these individuals. Under these conditions, the performance requirements for workers specified in paragraphs (b) and (c) of this section may be applied to these individuals.

Sec. 70.62 Safety program and integrated safety analysis.

(a) Safety program. (1) Each licensee shall establish and maintain a safety program that demonstrates compliance with the performance requirements of Sec. 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. Tl-le tl-!Fee elemeflts ef ti-le safety pFegFem; Aemely, prneess safety iflfeFmetiefl, ifltegFeted safety eAelysis,(Comment 2)

(2) Each licensee shall establish and meAegemeflt measures, ere deseFibed iflmaintain records that demonstrate compliance with the requirements of paragraphs (b) through (d) of this section.

Page 7

(2) Eaeh lieeRsee shall establish aRd maiRtaiR reeer=ds that demeRstrate eem~liaRee 't't'ith the re~uiremeRts ef ~aragra~hs (b) threugh (d) ef this seetieR.(3) Each licensee shall establish and maintain a log, available for NRC inspection, documenting each discovery that an item relied on for safety or management measure has failed to perform its function either in the context of the performance requirements of Sec. 70.61 or upon demand. This log must identify the item relied on for safety or management measure that has failed and the safety function affected, the date of discovery, date (or estimated date) of the failure, duration (or estimated duration) of the time that the item was unable to perform. its function, any other affected items relied on for safety or management measures and their safety function, affected processes, cause of the failure, whether the failure was in the context of the performance requirements or upon demand or both, and any corrective or compensatory action that was taken. The log must be initiated at the time of discovery and updated promptly upon the conclusion of each investigation of a failure of an item relied on for safety or management measure.

(b) Process safety information. Each licensee or applicant shall maintain process safety information to enable the performance of an integrated safety analysis. This process safety information must include information pertaining to the hazards of the materials used or produced in the process, information pertaining to the technology of the process, and information pertaining to the equipment in the process.

(c) Integrated safety analysis. (1) Each licensee or applicant shall conduct an integrated safety analysis, that is of appropriate detail for the complexity of the process, that identifies:

(i) Radiological hazards related to possessing or processing licensed material at its facility; (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 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 Sec. 70.61 (e) of this part, the characteristics of its preventive, mitigative, or other safety function, and the assumptions and conditions under which the item is relied upon to support compliance with the performance requirements of Sec. 70.61 .

(2) Integrated safety analysis team qualifications. In order to assure the adequacy of the integrated safety analysis, the analysis must be performed by a team with expertise in engineering and process operations. The team shall include at least one person who has experience and knowledge specific to each process being evaluated, and persons who have experience in nuclear criticality safety, radiation safety, fire safety, and chemical process safety.

One member of the team must be knowledgeable in the specific integrated safety analysis methodology being used.

(3) Requirements for existing licensees. Notwithstanding other provisions regarding the effective date for part 70, subpart H, requirements, licensees shall comply with the provisions in paragraphs (c)(3)(i), (ii), and (iii) of this section beginning on [the date of publication of the final rule]. Individuals holding an NRC license on [the date of publication of the final rule] shall, with regard to existing licensed activities:

Page 8

(i) Within 6 months of the effective date of the rule, submit for NRC approval, a plan that describes the integrated safety analysis approach that will be used, the processes that will be analyzed, and the schedule for completing the analysis of each process.

(ii) Within 4 years of the effective date of the rule, complete an integrated safety analysis, correct all unacceptable performance deficiencies, and submit an integrated safety analysis summary in accordance with Sec. 70.65 or the approved plan submitted under paragraph (c)(3)(i) of this section.

(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 measures to provide continuing assurance of compliance with the performance requirements of Sec. 70.61 . 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 administrative controls and control systems that are identified as items relied on for safety pursuant to Sec.

70.61 (e) of this part are designed, implemented, and maintained, as necessary, to ensure they are available and reliable to perform their function when needed, in the context of compliance with the performance requirements of Sec. 70.61 of this part.

Sec. 70.64 Requirements for new facilities or new processes at existing facilities.

(a) Baseline design criteria. Each prospective applicant or licensee shall address the following baseline design criteria in the design of new facilities. Each existing licensee shall address the following baseline design criteria in the design of new processes at existing facilities that require a license amendment under Sec. 70. 72. The baseline design criteria must be applied to the design of new facilities and new processes, but do not require retrofits to existing facilities or existing processes (e.g. , those housing or adjacent to the new process);

however, all facilities and processes must comply with the performance requirements in Sec.

70.61 . Licensees shall maintain the application of these criteria unless the evaluation performed pursuant to paragraph (c) of this section demonstrates that a given item is not relied on for safety or does not require adherence to the specified criteria.

(1) Quality standards and records. The design must be developed and implemented in accordance with management measures, to provide adequate assurance that items relied on for safety will be available and reliable to perform their function when needed. Appropriate records of these items must be maintained by or under the control of the licensee throughout the life of the facility.

(2) Natural phenomena hazards. The design must provide for adequate protection against natural phenomena with consideration of the most severe documented historical events for the site.

(3) Fire protection. The design must provide for adequate protection against fires and explosions.

(4) Environmental and 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.

Page 9

(5) Chemical protection. 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) Emergency capability. The design must provide for emergency capability to maintain control of:

(i) Licensed material; (ii) Evacuation of personnel; and (iii) Onsite emergency facilities and services that facilitate the use of available offsite services.

(7) Utility services. The design must provide for continued operation of essential utility services.

(8) Inspection, testing, and maintenance. The design of items relied on for safety must provide for adequate inspection, testing, and maintenance, to ensure their availability and reliability to perform their function when needed.

(9) Criticality control. The design must provide for criticality control including adherence to the double contingency principle .

(10) Instrumentation and controls. The design must provide for inclusion of instrumentation and control systems to monitor and control the behavior of items relied on for safety.

(b) Facility and system design and plant layout must be based on defense-in-depth practices.<SUP>4&It;/SUP> The design process must incorporate, to the extent practicable:

\4\ As used in Sec. 70.64, defense-in-depth practices means a design philosophy, applied from the outset and through completion of the design, that is based on providing successive levels of protection such that health and safety will not be wholly dependent upon any single element of the design, construction, maintenance, or operation of the facility. The net effect of incorporating defense-in-depth practices is a conservatively designed facility and system that will exhibit greater tolerance to failures and external challenges. The risk insights obtained through performance of the integrated safety analysis can be then used to supplement the final design by focusing attention on the prevention and mitigation of the higher-risk potential accidents.

(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.

Sec. 70.65 Additional content of applications.

(a) In addition to the contents required by Sec. 70.22, each application must include a description of the applieaflt's safety pFegFaffimanagement measures established under Sec.

70.62, iReludiRg U:ie ifltegFated safety aRalysis sufflffiary70.62(d) and a deseFiptiefl ef the fflaRageffleflt ffleasuresan integrated safety analysis summary.

- - - (Comment 2)

Page 10

(b) The integrated safety analysis summary must be submitted with the license or renewal application (and amendment application as necessary), but shall not be incorporated in the license. However, changes to the integrated safety analysis summary shall meet the conditions of Sec. 70. 72. The integrated safety analysis summary must eentein:

(1) A gener-el dcseFiptien ef the site with emphasis en these feeteFS that eeuld effect safety (ishall be updated biennially to reflect changes made to the facility a maximum of 6 months prior to the date of filing. (COMMENT 3) The integrated safety analysis summary must contain:

(1) A general description of the site with emphasis on those factors that could affect safety (i.e. , meteorology, seismology);

(2) A general description of the facility with emphasis on those areas that could affect safety, including an identification of the controlled area boundaries; (3) A description of each process (defined as a single reasonably simple integrated unit operation within an overall production line) analyzed in the integrated safety analysis in sufficient detail to understand the theory of operation; and, for each process, the hazards that were identified in the integrated safety analysis pursuant to Sec. 70.62(c)(1)(i)-(iii) and a general description of the types of accident sequences; (4) Information that demonstrates the licensee's compliance with the performance requirements of Sec. 70.61 ; the requirements for criticality monitoring and alarms in Sec. 70.24; and, if applicable, the requirements of Sec. 70.64; (5) A description of the team, qualifications, and the methods used to perform the integrated safety analysis; (6) A list briefly describing all items relied on for safety which are identified pursuant to Sec. 70.61 (e) in sufficient detail to understand their functions in relation to the performance requirements of Sec. 70.61 ;

(7) A description of the proposed quantitative standards used to assess the consequences from acute chemical exposure to licensed material or chemicals produced from licensed materials which are on-site, or expected to be on-site as described in Sec. 70.61 (b)(4) and (c)(4);

(8) A descriptive list that identifies all items relied on for safety that are the sole item preventing or mitigating an accident sequence that exceeds the performance requirements of Sec. 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.

Sec. 70.66 Additional requirements for approval of license application.

An application for a license from an applicant subject to subpart H will be approved if the Commission determines that the applicant has complied with the requirements of Sec. 70.21 ,

Sec. 70.22, Sec. 70.23 and Sec. 70.60 through Sec. 70.65.

Sec. 70. 72 Facility changes and change process.

(a) The licensee shall establish a configuration management system to evaluate, implement, and track each change to the site, structures, processes, systems, equipment, Page 11

components, computer programs, and activities of personnel deseFibed iA the ISA Summarythat affect records required by §70.62(a)(2) (COMMENT 3).-4t This system must be documented in written procedures and must assure that the following are addressed prior to implementing any change:

(1) The technical basis for the change; (2) Impact of the change on safety and health or control of licensed material; (3) Modifications to existing operating procedures including any necessary training or retraining before operation; (4) Authorization requirements for the change; (5) For temporary changes, the approved duration (e.g. , expiration date) of the change; and (6) The impacts or modifications to the integrated safety analysis, integrated safety analysis summary, or other safety program information, developed in accordance with Sec.

70.62.

(b) Any change to site, structures, processes, systems, equipment, components, computer programs, and activities of personnel deseFibed iA ISA Summarythat affect records required by §70.62(a)(2) (COMMENT 4) must be evaluated by the licensee as specified in paragraph (a) of this section, before the change is implemented. The evaluation of the change must determine, before the change is implemented, if an amendment to the license is required to be submitted in accordance with Sec. 70.34.

(c) The licensee may make changes to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel deseFibed iA the ISA Summarythat affect records required by §70.62(a)(2) (COMMENT 4}3), without prior Commission approval, if the change:

( 1) Does not:

(i) Create new types \5\ of accident sequences that, unless mitigated or prevented, would exceed the performance requirements of Sec. 70.61 and that have not previously been described in the integrated safety analysis summary; or

\5\ Any change in the defining characteristics of the elements of an accident sequence may change the * *type" of the accident sequence for a given process. For example, a new type of accident could involve a different initiator, significant changes in the consequence, or a change in the safety function of a control (e.g., temperature limiting device versus a flow limiting device).

(ii) Use new processes, technologies, or control systems for which the licensee has no prior experience; (2) Does not remove, without at least an equivalent replacement of the safety function, an item relied on for safety that is listed in the integrated safety analysis summary; (3) Does not alter any item relied on for safety, listed in the integrated safety analysis summary, that is the sole item preventing or mitigating an accident sequence that exceeds the performance requirements of Sec. 70.61 ; and (4) Is not otherwise prohibited by this section, license condition, or order.

(d)(1) For et=ty-ehanges that affect the iAtegFated safety aAalysis summary, as submitted iA aeeeFdaAee 't't1ithrequire pre-approval under Sec. 70.65, but de Aet reeiuiFe NRG pFe apprn*t1al, Page 12

the lieeAsee shall submit re*tised pages to the iAtegr-ated safety aAalysis summary, to NRG,

• • 't'ithiA 90 days of the ehaAge.

(2) For ehaAges that re~uire pre approval uAder See. 70.72, the licensee shall submit an amendment request to the NRC in accordance with Sec. 70.34 and Sec. 70.65.

(3)(2) A brief summary of all changes made in accordance with 70.72(c), that are made without prior Commission approval, must be submitted to NRC with the reeordsintegrated safety analysis summary update required by See70.65(b).- 70.62(a)(2) of this part, that are made vtithout prior GommissioA appro\*al, must be submitted to NRG e\*ery 12 moAths(COMMENT 3)

(e) If a change covered by Sec.

(e) If a ehaAge covered by See. 70.72 is made, the affected on-site documentation must be updated promptly.

(f) The licensee shall maintain records of changes to its facility carried out under this section. These records must include a written evaluation that provides the bases for the 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.

Sec. 70. 73 Renewal of licenses.

Applications for renewal of a license must be filed in accordance with Secs. 2.109, 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 references are clear and specific.

Sec. 70. 74 Additional reporting requirements.

(a) Reports to NRC Operations Center. (1) Each licensee shall report to the NRC Operations Center the events described in appendix A to part 70.

(2) Reports must be made by a knowledgeable licensee representative and by any method that will ensure compliance with the required time period for reporting.

(3) The information provided must include a description of the event and other related information as described in Sec. 70.50(c)(1).

(4) Follow-up information to the reports must be provided until all information required to be reported in Sec. 70.50(c)(1) of this part is complete.

(5) Each licensee shall provide reasonable assurance that reliable communication with the NRC Operations Center is available during each event.

(b) Written reports. Each licensee who makes a report required by paragraph (a)(1) of this section shall submit a written follow-up report within 30 days of the initial report. The written report must contain the information as described in Sec. 70.50(c)(2).

18. Appendix A to part 70 is added to read as follows:

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:

Page 13

(a) One hour reports. Events to be reported to the NRC Operations Center within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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:

(1) An inadvertent nuclear criticality.

(2) An acute intake by an individual of 30 mg or greater of uranium in a soluble form.

(3) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that exceeds the quantitative standards established to satisfy*the requirements in Sec. 70.61 (p)(4).

(4) An event or condition such that no items relied on for safety, as documented in the Integrated Safety Analysis summary, remain available and reliable, in an accident sequence evaluated in the Integrated Safety Analysis, to perform their function:

(i) In the context of the performance requirements in Sec. 70.61 (b) and Sec. 70.61 (c), or (ii) Prevent a nuclear criticality accident (i.e., loss of all controls in a particular sequence).

(5) Loss of controls such that only one item relied on for safety, as documented in the Integrated Safety Analysis summary, remains available and reliable to prevent a nuclear criticality accident, and has ~en in this state for greater than eight hours.

(b) Twenty-four hour reports. Events to be reported to the NRC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of dis~v~ry, supplemented with the information in 10 CFR 70.50(c)(1) as it becomes available, followed by a written report within 30 days:

(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 Sec. 70.61.

(2) Loss or degradation of items relied on for safety th~t results in failure to meet the performance requirement of Sec. 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 satisfy

• the requirements of Sec. 70.61 (c)(4).

(4) Any natural phenomenon or other external event, including fires internal and external to the facility, that has affected or may have affected the intended safety function or availability or reliability of one or more items re.lied on for safety.

(5) An occurrence of an event or process deviation that was considered in the Integrated Safety Analysis and:

(i) Was dismissed due to its likelihood; or *

(ii) Was categorized as unlikely and whose associated unmitigated consequences would have exceeded those in Sec. 70.61 (b) had the item(s) relied on for safety not performed their safety function(s). .

(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.

For the Nuclear Regulatory Commission.

Dated at Rockville, Maryland, this 23rd day of July, 1999.

Annette Vietti-Cook, Secretary of the Commission.

Page 14 ,

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).

Sections 70.1(c) and 70.20a(b) also issued undersecs.135, 141, Pub. L.97-425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161). Section 70.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851 ). Section 70.21 (g) also issued under sec. 122, 68 Stat.

939 (42 U.S.C. 2152). Section 70.31 also issued under sec. 57d, Pub. L.93-377, 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36 and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Section 70.61 also issued under secs. 186, 187, 68 Stat. 955 (42 U.S.C. 2236, 2237). Section 70.62 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).

2. The undesignated center heading "GENERAL PROVISIONS" is redesignated as "Subpart A-General Provisions." ,

3. In Sec. 70 4, the definitions of Acute, Available and reliable 'to perform their function when needed, Configuration management, Critical mass of special nuclear material, Double contingency, Hazardous chemicals produced from licensed material, Integrated safety analysis (ISA), Integrated safety analysis summary, Items relied on for safety, Management measures, Unacceptable performance deficiencies, and Worker are added, in alphabetical order, as follows:

Sec. 70.4 Definitions.

Acute as used in this part means a single radiation dose or chemical exposure event or multiple radiation dose or chemical exposure events occurring within a short time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less).

Available and reliable to perform their function when needed as used in subpart H of this part means that, based upon the analyzed, credible conditions in the integrated safety analysis, items relied on for safety will perform their intended safety function when needed and management measures will be implemented that ensure continuous compliance with the performance requirements of Sec. 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.

Configuration management (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 when needed.

Critical mass of special nuclear material (SNM) means special nuclear material in a quantity exceeding 700 grams of contained uranium-235; 520 grams of uranium-233; 450

• grams of plutonium; 1500 grams of contained uranium-235, if no uranium enriched to more than 4 percent by weight of uranium-235 is present; 450 grams of any combination thereof; or one-half such quantities if ma$sive moderators or reflectors made of graphite, heavy water, or beryllium may be present.

Double contingency means a process design that incorporates sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.

Hazardous chemicals produced from licensed materials means substances having licensed material as precursor compound(s) or substances that physically or chemically interact with licensed materials; that are toxic, explosive, flammable, corrosive, or reactive to the extent that they can endanger life or health if not adequately controlled. These include substances commingled with licensed material, and include substances such as hydrogen fluoride that is produced by the reaction of uranium hexafluoride and water, but do not include substances prior to process addition to licensed material or after process separation from licensed material.

Integrated safety analysis (ISA) means a systematic analysis to identify plant and external hazards and their potential for initiating accident sequences, the potential accident sequences, their likelihood and consequences, and the items relied on for safety. As used here, integrated means joint consideration of, and protection from, all relevant hazards, including radiological, nuclear criticality, fire, and chemical. However, with respect to compliance with the regulations of this part, the NRG requirement is limited to consideration of the effects of all relevant hazards on radiological safety, prevention of nuclear criticality accidents, or chemical hazards directly associated with NRG licensed radioactive material.

• Integrated safety analysis summary means the document submitted with the license application, license amendment application, or license renewal application that provides a synopsis of the results of the integrated safety analysis and contains the information specified in Sec. 70.65(b).

Items relied on for safety means structures, systems, equipment, components, and activities of personnel that are relied on to prevent potential accidents at a facility that could exceed the performance requirements in Sec. 70.61 or to mitigate their potential consequences.

This does not limit the licensee from identifying additional structures, systems, equipment, components, or activities of personnel (i.e., beyond those in the minimum set necessary for compliance with the performance requirements) as items relied on for safety.

Management measures mean the functions performed by the licensee, generally on a continuing basis, that are applied to items relied upon for safety, to ensure the items are available and reliable to perform their functions when needed. Management measures include configuration management, maintenance, training and.qualifications, procedures, audits and assessments, incident investigations, records management, and other quality assurance elements.

Unacceptable performance deficiencies mean deficiencies in the items relied on for safety or the management measures that need to be corrected to ensure an adequate level of protection as defined in 10 CFR 70.61 (b), (c), or (d).

Worker means an individual whose assigned duties in the course* of employment involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation (i.e., an individual who is subject to an occupational dose as in 20 CFR 20.1003).

4. In Sec. 70.8 paragraph (b) is revised to read as follows.

Sec. 70.8 Information collection requirements: 0MB approval.

(b) The approved information collection requirements contained in this part appear in Secs. 70.9, 70.17, 70.19, 70.20a, 70.20b, 70.21, 70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, 70.39, 70.42, 70.50, 70.51, 70.52, 70.53, 70.57, 70.58, 70.59, 70.61, 70.62, 70,64, 70.65, 70.72, 70.73, 70.74 and Appendix A.

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

. Secs. 70.13a and 70.14 [Redesignated]

6. Sections 70.13a and 70.14 are redesignated as Secs. 70.14 and 70.17, respectively.

7. The undesignated center heading "GENERAL LICENSES" is redesignated as.

• • Subpart C-General Licenses."

8. The undesignated center heading "LICENSE APPLICATIONS" is redesignated as "Subpart D-License Applications."

9. The undesignated center heading "LICENSES" is redesignated as "Subpart E--Licenses."

10. The undesignated center heading "ACQUISITION, USE AND TRANSFER OF SPECIAL NUCLEAR MATERIAL, CREDITORS' RIGHTS," is redesignated as .. Subpart F-Acquisition, Use, and Transfer of Special Nuclear Material, Creditors' Rights."

11. The undesignated center heading SPECIAL NUCLEAR MATERIAL CONTROL RECORDS, REPORTS AND INSPECTIONS" is redesignated as "Subpart G-Special Nuclear Material Control Records, Reports, and Inspections."

12. In Sec. 70.50 paragraph (c) is revised and paragraph (d) is added to read as follows:

Sec. 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 Sec. 70. 74 and appendix A of this part if applicable, by telephone to the NRC Operations Center.<SUP>3&It;/SUP> To the extent that the information is available at the time of notification, the information provided in these reports must include:

.\3\ The commercial telephone number for the NRC Operations Center is (301) 816-5100.

(i) Caller's name, position title and call back telephone number; (ii) Date, time, and exact location of the event; (iii) Description of the event, including; _

(A) Radiological or chemical hazards involved including isotopes, quantities, and chemical and physical form of any material released;

• (!3) Actual or potential health and safety consequences~to the-workers, the publi_c, and the environment, including relevant chemical and radiation data for actual personnel exposures to radiation or radioactive materials or chemicals (e.g., level of radiation exposure, concentration of chemicals, and duration of exposure);

(C) The sequence of occurrences leading to the event, including degradation or failure of structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences; and '-

(D) Whether the remaining structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences are available and reliable to perform their function. *

(iv) External conditions affecting the event; (v) Additional actions taken by the licensee in response to the event; (vi) Status of the event (e.g., whether the event is on-going or was terminated);

(vii) Current and planned site status, including any declared emergency class; (viii) Notifications related to the event that were made or are planned to any local, State, or other Federal agencies; (ix) Status of any press releases related to the event that were made or are planned.

(2) Written report. Each licensee who makes a report required by paragraph (a) or (b) of this section, or by Sec. 70. 74 and appendix 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:

(i) Complete applicable information required by Sec. 70.50(c)(1 );

(ii) The probable cause of the event, including all factors that contributed to the event and the manufacturer and model number (if applicable) of any equipment that failed or malfunctioned; '

(iii) Corrective actions taken or planned to prevent occurrence of similar or identical events in the Mure and the results of any evaluations or assessments; and (iv) For licensees subject to subpart H of this part, whether 'the event was identified. and evaluated in the Integrated Safety Analysis.

(d) The provisions of Sec. 70.50 do not apply to licensees subject to Sec. 50.72. They do apply to those part 50 licensees possessir\g material licensed-under: part 70 who are not subject to the notification requirements in Sec. 50. 72.

13. The undesignated center heading "MODIFICATION AND REVOCATION OF LICENSES" is redesignated as "Subpart !--Modification and Revocation of Licenses."

Secs. 70.61 and 70.62 [Redesignated]

14. Sections 70.61 and 70.62 are redesignated as Secs. 70.81 and 70.82, respectively.

15. The undesignated center heading "ENFORCEMENT' is redesignated as "Subpart J-Enforcement."

Secs. 70.71 and 70.72 [Redesignated]

_16. Sections 70.71 and 70.72 are redesignated as Secs. 70.91 and 70.92; respectively.

17. In part 70, a new subpart H (Secs. 70.60-70.74) is added to read as follows:

Subpart H-Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special t:Juclear Material Sec.

70.60 Applicability.

70.61 Performance requirements.

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.

70. 72 Facility changes and change process.

70. 73 Renewal of licenses.

?:O. 74 Additional reporting requirements.

Sec. 70.60 Applicability.

The regulations in Sec. 70.61 through Sec. 70.74 apply, in ad~ition to other applicable Commission regulations, to each applicant or licensee that is or plans to be: authorized to possess greater than a critical mass of special nuclear material, and engaged in enriched uranium processing, fabrication of uranium fuel or fuel assemblies, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap recovery of special nuclear material, or any other activity that the Commission determines could significantly affect public health and safety. The regulations in Sec. 70.61 through Sec. 70.74 do not apply to decommissioning activities performed pursuant to other applicable Commission regulations including Sec. 70.25 and Sec. 70.38 of this.Part.

Also, the regulations in Sec. 70.61 through Sec. 70.74 do not apply to activities that are certified by the Commission pursuant to Part 76 of this chapter or licensed by the Commission pursuant to other parts of this chapter.

Sec. 70.61 Performance requirements.

(a) Each applicant or licensee shall evaluate, in the integrated safety analysis performed in accordance with Sec. 70.62, its compliance with the performance requirements in paragraphs (b), (c), and (d) of this section. '

(b) The likelihood of each credible accident sequence whose unmitigated consequences could exceed those below must be reduced to be highly unlikely through the application of engineered controls, administrative controls, or both. (Comment 1)

(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 fonn by any individual located outside1 the controlled area identified pursuant to paragraph (f) of this section; or (4) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that:

(i) Could endang_er the life_ of a worker, or (ii) Could lead to irreversible or other serious, long-lasting health effects to any individual located outside the controlled area identified pursuant to 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 infonnation submitted pursuant to Sec. 70.65 of this part.

(c) The likelihood of each credible accident sequence whose unmitigated consequences could exceed those below must be reduced to unlikely through the application of engineered controls, administrative controls, or both. (Comment 1)

(1) An acute worker dose of 0.25 Sv (25 rem) or greater total effective dose equivalent; (2) An acute dose of 0.05 Sv (5 rem) or greater total effective dose equivalent to any individual located outside the controlled area identified pursuant to paragraph (f) of this section; (3) A 24-hour averaged release of radioactive material outside the restricted area in concentrations exceeding 5000 times the values in table 2 of appendix 8 to 10 CFR part 20; or (4) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that:

(i) Could lead to irreversible or other serious, long-lasting health effects to a worker, or (ii) Could cause mild transient health effects to any individual located 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 infonnation_

submitted pursuant to Sec. 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 nonnal and credible abnonnal conditions, all nuclear processes are subcritical, including use of an approved margin of subcriticality for safety. Preventive controls and measures must be the primary means of protection against nuclear criticality accidents.

(e) Each engineered or administrative control or control system necessary to comply with paragraphs (b), (c), or (d) of this section shall be designated as an item relied on for safety.

The safety program, established and maintained pursuant to Sec. 70.62 of this part, shall ensure that each item relied on for safety will be available and reliable to perfonn its intended function when needed and in the context of the perfonnance requirements of this section.

(f) Each licensee must establish a controlled area, as defined in Sec. 20.1003, in which the licensee retains the authority to detennine all activities, including exclusion or removal of personnel and property from the area. For the purpose of complying with 'the perfonnance requirements of this section, individuals who are not workers, as defined in Sec. 70.4, may be permitted to perform ongoing activities (e.g., at a facility not related to the licensed activities) in the controlled area, if the licensee:

(1) Demonstrates and documents, in the integrated safety analysis, that the risk for those individuals at the location of their activities does not exceed the perfonnance requirements of paragraphs (b)(2), (b)(3), (b)(4)(ii), (c)(2), and (c)(4)(ii) of this section; or

(2) Provides: training in accordance with 10 CFR 19.12(a)(1 )-(5) to these individuals to ensure that they are aware of the risks associated with accidents involving the licensed activities as determined by the integrated safety analysis, and conspicuously posts and maintains notices stating where the information in 10 CFR 19.11 (a) may be examined by these individuals. Under these conditions, the performance requirements for workers specified in paragraphs (b) and (c) of this section may be applied to these individuals.

Sec. 70.62 Safety program and integrated safety analysis.

(a) Safety program. (1) Each licensee shall establish and maintain a safety program that demonstrates compliance with the performance requirements of Sec. 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. (Comment 2)

(2) Each licensee shall establish and maintain records that demonstrate compliance with the requirements of paragraphs (b) through (d) of this section.

(3) Each licensee shall establish and maintain a log, available for NRC inspection, documenting each discovery that an item relied on for safety or management measure has failed to perform its function either in the context of the performance requirements of Sec. 70.61 or upon demand. This log must identify the item relied on for safety or management measure that has failed and the safety function affected, the date of disc,overy, date (or estimated date) of the failure, duration (or estimated duration) of the time that the item was unable to perform its function, any other affected items relied on for safety or management measures and their safety function, affected processes, cause of the failure, whether the failure was in the context of the performance requirements or upon demand or both, and any corrective or compensatory action that was taken. The log must be initiated at the time of discovery and updated promptly upon the conclusion of each investigation of a failure of an item relied on for safety or management measure.

(b) Process safety information. Each licensee or applicant shall maintain process safety information to enable the performance of an integrated safety analysis. This process safety information must include information pertaining to the hazards of the materials used or produced in the process, information pertaining to the technology of the process, and information pertaining to the equipment in the process.

(c) Integrated safety analysis. (1) Each licensee or applicant shall conduct an integrated safety analysis, that is of appropriate detail for the complexity of the process, that identifies:

(i) Radiological hazards related to possessing or processing licensed material at its facility; (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 identified pursuant to paragraph (c)(1 )(iv) of this section, and the methods used to determine the consequences and likelihoods; and Page 7

(vi) Each item relied on for safety identified pursuant to Sec. 70.61 (e) of this part, the characteristics of its preventive, mitigative, or other safety function, and the assumptions and conditions under which the item is relied upon to support compliance with the performance requirements of Sec. 70.61.

(2) Integrated safety .analysis team qualifications. In order to assure the adequacy of the integrated safety analysis, the analysis must be performed by a team with expertise in engineering and process operations. The team shall include at least one person who has experience and knowledge specific to each process being evaluated, and persons who have _

experience in nuclear criticality safety, radiation safety, fire safety, and chemical process safety.

One member of the team must be knowledgeable in the specific integrated safety analysis methodology being used.

(3) Requirements for existing licensees. Notwithstanding other provisions regarding the effective date for part 70, subpart H, requirements, licensees shall comply with the provisions in paragraphs (c)(3)(i), (ii), and (iii) of this section beginning on [the date of publication of the final rule]. Individuals holding an NRG license on [the date of publication of the final rule] shall, with regard to existing licensed activities:

(i) Wrthin 6 months of the effective date of the rule, submit for NRG approval, a plan that describes the integrated safety analysis approach that will be used, the processes that will be analyzed, and the schedule for completing the analysis of each process.

(ii) Wrthin 4 years of the effective date of the rule, complete an integrated safety analysis, correct all unacceptable performance deficiencies, and submit an integrated safety analysis summary in accordance with Sec. 70.65 or the approved plan submitted under paragraph (c)(3)(i) of this section.

(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 measures to provide continuing assurance of compliance with the performance requirements of Sec. 70.61. 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 administrative controls and control systems that are identified as items relied on for safety pursuant to Sec.

70.61 (e) of this part are designed, implemented, and maintained, as necessary, to ensure they are available and reliable to perform their function when needed, in the context of compliance with the performance requirements of Sec. 70.61 of this part.

Sec. 70.64 Requirements for new facilities or new processes at existing facilities.

(a) Baseline design criteria. Each prospective applicant or licensee shall address the following baseline design criteria in the design of new facilities. Each existing licensee shall address the following baseline design criteria in the design of new processes at existing facilities that require a license amendment under Sec. 70. 72. The baseline design criteria must be applied to the design of new facilities and new processes, but do not require retrofits to existing facilities or existing processes (e.g., those housing or adjacent to the new process);

however, all facilities and processes must comply with the performance requirements in Sec.

70.61. Licensees shall maintain the application of these criteria unless the evaluation performed Page 8

pursuant to paragraph (c) of this section demonstrates that a given item is not relied on for safety or does not require adherence to the specified criteria.

(1) Quality standards and records. The design must be developed and implemented in accordance with management measures, to provide adequate assurance that items relied on for safety will be available and reliable to perform their function when needed. Appropriate records of these items must be maintained by or under the control of the licensee throughout the life of the facility.

(2) Natural_ phenomena hazards. The design must provide for adequate protection against natural phenomena with consideration of the most severe documented historical events for the site.

(3) Fire protection. The design must provide for adequate protection against fires and explosions.

(4) Environmental and 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) Chemical protection. 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) Emergency capability. The design must provide for emergency capability to maintain control of:

(i) Licensed material; (ii) Evacuation of personnel; and (iii) Onsite emergency facilities and services that facilitate the use of available offsite services.

(7) Utility services. The design must provide for continued operation of essential utility services.

(8) Inspection, testing, and maintenance. The design of items relied on for safety must provide for adequate inspection, testing, and maintenance, to ensure their availability and reliability to perform their function when needed.

(9) Criticality control. The design must provide-for criticality control including adherence to the double contingency principle.

_(10) Instrumentation and controls. The design must provide for inclusion of instrumentati9n -and control systems to monitor and control the behavior of items relied on for safety.

(b) Facility and system design and plant layout must be based on defense-in-depth practices.<SUP>4&It;/SUP> The design process must incorporate; to the extent practicable:

\4\ As used in Sec. 70.64, defense-in-depth practices means a design philosophy, applied from the outset and through completion of the design, that is based on providing successive levels of protection such that health and safety will not be wholly dependent upon any single element of the design, construction, maintenance, or operation of the facility. The net effect of incorporating defense-in-depth practices is a conservatively designed facility and system that will exhibit greater tolerance to failures and external challenges. The risk insights obtained through performance of the integrated safety analysis can be then used to supplement Page 9

the final design by focusing attention on the prevention and mitigation_of the higher-risk potential accidents.

(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.

Sec. 70.65 Additional content of applications.

(a) In addition to the contents required by Sec. 70.22, each application must include a description of the management measures established under Sec. 70.62(d) and an integrated 1 safety analysis summary. (Comment 2)

(b) The integrated safety analysis summary must be submitted'with the license or renewal application (and amendment application as necessary), but shall not be incorporated in the license. However, changes to the integrated safety analysis summary shall meet the conditions of Sec. 70. 72. The integrated safety analysis summary shall be updated biennially* to reflect changes made to the facility a maximum of 6 months prior to the date of filing.

(COMMENT 3) The integrated safety analysis summary must contain:

(1) A general description of the site with emphasis on those factors that could affect safety (i.e., meteorology, seismology);

(2) A general description of the facility with emphasis on those areas that could affect safety, including an identification of the controlled area boundaries; (3) A description of each process (defined as a single reasonably simple integrated unit operation within an overall production line) analyzed in the integrated safety analysis in sufficient detail to understand the theory of operation; and, for each process, the hazards that were identified in the integrated safety analysis pursuant to Sec. 70.62(c)(1)(i)-(iii) and a general description of the types of accident sequences; (4) Information that demonstrates the licensee's compliance with the performance requirements of Sec. 70.61; the requirements for criticality monitoring and alarms in Sec. 70.24; and, if applicable, the requirements of Sec. 70.64; (5) A description of the team, qualifications, and the methods used to perform the integrated safety analysis; (6) A list briefly describing all items relied on for safety which are identified pursuant to Sec. 70.61 (e) in sufficient detail to understand their functions in relation to the performance requirements of Sec. 70.61; (7) A description of the proposed quantitative standards used to assess the consequences from acute chemical exposure to licensed material or chemicals produced from licensed materials which are on-site, or expected to be on-site as described in Sec. 70.61 (b)(4) and (c)(4);

(8) A descriptive list that identifies all items relied on for safety that are the sole item preventing or mitigating an accident sequence that exceeds the performance requirements of Sec. 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.

Page 10

Sec. 70.66 Additional requirements for approval of license application.

An application for a license from an applicant subject to subpart H will be approved if the Commission determines that the applicant has complied With the requirements of Sec. 70.21, Sec. 70.22, Sec. 70.23 and Sec. 70.60 through Sec: 70.65.

Sec. 70. 72 Facility_ changes and change process.

(a) The licensee shall establish a configuration management system to evaluate, implement, and track each change to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel that affect records required by

§70.62(a)(2) (COMMENT 3). This system must be documented in written procedures and must assure that the following are addressed prior to implementing any change:

(1) The technical basis for the change; (2) Impact of the change on safety and health or control of licensed material; (3) Modifications to existing operating procedures including any necessary training or retraining before operation; (4) Authorization requirements for the change; (5) For temporary changes, the approved duration (e.g., expiration date) of the change; and (6) The impacts or modifications to the integrated safety analysis, integrated safe~y analysis summary, or other safety program information, developed in accordance with Sec.

70.62.

(b) Any change to site, structures, processes, systems, equipment, components, computer programs, and activities of personnel that affect records required by §70.62(a)(2)

(COMMENT 4) must b"e evaluated by the licensee as specified in paragraph (a) of this section, before the change is implemented. The evaluation of the change must determine, before the change is implemented, if an amendment to the license is required to be submitted in accordance with Sec. 70.34.

(c) The licensee may make changes to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel that affect records required by §70.62(a)(2) (COMMENT 3), without prior Commission approval, if the change:

(1) Does not:

(i) Create new types \5\ of accident sequences that, unless mitigated or prevented, would exceed the performance requirements of Sec. 70.61 and that have not previously been described in the integrated safety analysis summary; or

\5\ Any change in the defining characteristics of the elements of an accident sequence may change the "type" of the accident sequence for a given process. For example, a new type of accident could involve a different initiator, significant changes in the consequence, or a change in the safety function of a control (e.g., temperature-limiting device versus a flow limiting device).

Page 11

/

(ii) Use new processes, technologies, or control systems for which the licensee has no prior experience; (2) Does not remove, without at least an equivalent replacement of the safety function, an item relied on for safety that is listed in the integrated safety analysis summary; (3) Does not alter any item relied on for safety, listed in the integrated safety analysis summary, that is the sole item preventing or mitigating an accident sequence that exceeds the performance requirements of Sec. 70.61; and (4) Is not otherwise prohib~ed by this section, license condition, or orde_r.

(d)(1) For changes that require pre-approval under Sec. 70.72, the licensee shall submit an amendment request to the NRC in accordance with Sec. 70.34 and Sec. 70.65.

(2) A brief summary of all changes made in accordance with 70.72(c), that are made without prior Commission approval, must be submitted to NRC with the integrated safety analysis summary update required by 70.65(b). (COMMENT 3)

(e) If a change covered by Sec. 70.72 is made, the affected on-site documentation must be updated promptly.

(f) The licensee shall maintain records of changes to its facility carried out under this section. These records must include a written evaluation that provides the bases for the 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.

Sec. 70. 73 Renewal of licenses.

Applications for renewal of a license must be filed in accordance with Secs. 2.109, 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 references are c:lear and specific.

Sec. 70. 74 Additional reporting requirements.

(a) Reports to NRC Operations Center. (1) Each licensee shall report to the NRC Operations Center the events described in appendix A to part 70.

(2) Reports must be made by a knowledgeable licensee representative and by any method that will ensure compliance with the required time period for reporting.

(3) The information provided must include a description of the event and other related information as described in Sec. 70.50(c)(1).

(4) Follow-up information to the reports must be provided until all information required to be reported in Sec. 70.50(c)(1) of this part is complete.

(5) Each licensee shall provide reasonable assurance that reliable communication with the NRC Operations Center is available during each event.

_ (b) Written reports. Each licensee who makes a report required. by paragraph (a)( 1) of this section shall submit a written follow-up report within 30 days of the initial report. The written report must contain the information as described in Sec. 70.50(c)(2).

18. Appendix A to part 70 is added to read as follows:

Page 12

Appendix A to Part 70-Reportable Safety Events As required by 10 CFR 70.74, licensees subject to the requirements in subpart Hof part 70, shall report:

(a) One hour reports. Events to be reported to the NRC Operations Center within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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:

(1) --An inadvertent nuclear

. criticality.

(2) An acute intake by an individual of 30 mg or greater of uranium in a soluble form.

(3) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that exceeds the quantitative standards established to satisfy the requirements in Sec. 70.61 (b)(4).

(4) An event or condition such that no items relied on for safety, as documented in the Integrated Safety Analysis summary, remain available and reliable, in an accident sequence evaluated in the Integrated Safety Analysis, to perform their function:

(i) In the context of the performance requirements in Sec. 70.61 (b) and Sec. 70.61 (c), or (ii) Prevent a nuclear criticality accident (i.e., loss of all controls in a particular sequence)'.

(5) Loss of controls such that only one item relied on for safety, as documented in the Integrated Safety Analysis summary, remains available and reliable to prevent a nuclear criticality accident, and has been in this state for greater than eight hours.

(b) Twenty-four hour reports. Events to be reported to the NRC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery, supplemented with the infor'mation in 10 CFR 70.50(c)(1) as it becomes available, followed by a written report within 30 days:

(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 Sec. 70.61.

(2) Loss or degradation of items relied on for safety that results in failure to meet the performance requirement of Sec. 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 satisfy the requirements of Sec. 70.61 (c)(4).

(4) Any natural phenomenon or other external event, including fires internal and external to the facility, that has affected or may have affected the intended safety function or availability or reliability of one or more items relied on for safety.

(5) An occurrence of an event or process deviation that was considered in the Integrated Safety Analysis and:

(i) Was dismissed due to its likelihood; or (ii) Was categorized as unlikely and whose associated unmitigated consequences would have exceeded those in Sec. 70.61 (b) had the item(s) relied on for safety not performed their safety function(s).

(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.

For the Nuclear Regulatory Commission.

Page 13

Dated at Rockville, Maryland, this 23rd day of July, 1999.

Annette Vietti-Cook, Secretary of the Commission .

Page 14

BWX Technologies, Inc.

Babcock & Wilcox, a McDermott company ooc.::=TED 11'" ,, (' Naval Nuclear Fuel Division P.O. Box 785 Lynchburg, VA 24505-0785 (804) 522-6000

'99 OCT 14 P3 :48 99-099 October 11, 1999 Secretary of the Commission U.S. Nuclear Regulatory Commission DOCKET NU BER ATTN: Rulemakings and Adjudication Staff PROPOSED AUL 7O Washington, DC 20555-0001 (fl'IFl<c/1338)

Reference:

Comments on Proposed Rule 10CFR70: Domestic Licensing of Special Nuclear Material; Possession of Critical Mass of Special Nuclear Material (Federal Register Vol. 64, No. 146, pp. 41338-41357, dated July 30, 1999)

Dear Sir:

In response to a request for public input in the referenced Federal Register notice, BWXT's Naval Nuclear Fuel Division submits the attached comments.

BWXT generally endorses the proposed revision to 10CFR70 as a positive step toward achieving a risk-informed/performance-based regulatory framework. The attached comments reflect areas in the rule language where BWXT believes improvements are necessary.

BWXT would also like to endorse the comments submitted by the Nuclear Energy Institute on behalf of the fuel cycle industry .

Sincerely, Arne F. Olsen Attachment OCT 19 1999

-'eknowfedged by card ................"".'"""" ... ,. .

U.S.

RU qq

>>- 1!~,JI *' ' ~

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Section Comment 70.4 The definition of "available and reliable to perform their function when needed" appears throughout the proposed revisions to the rule. This definition requires measures be implemented that "ensure continuous compliance". BWXT believes this language indicates a level of certainty that is not realistic. A better choice of terminology would be "provides reasonable assurance".

70.62(a) This section implies the Safety Program has only three elements. This may be true when discussing an Integrated Safety Analysis, which will identify Items Relied on for Safety and their associated Management measures. This is not true, however, in relation to the requirements of 10CFR70.22 for contents of a license application. BWXT believes the Safety Program is much more comprehensive and includes occupational safety (e.g., Radiation Protection Program required by 10CFR20) as well as accident safety, which is the focus of Subpart H. BWXT suggests that attempts to define the Safety Program be deleted from 70.62(a). The requirements contained in 70.62 a-d can be retained without creating a less than comprehensive definition of the Safety Program.

70.62(a)(3) This section is very prescriptive in requiring a "log" to be available which documents failures of items relied on for safety. BWXT believes the requirement should be rewritten to be performance based rather than prescriptive. A performance-based requirement could state "each licensee shall maintain records of failures ... which are retrievable and available for inspection". Most licensees have an incident reporting and corrective action system, which is used for all activities at the facility. As long as these systems meet the performance objective it seems unnecessary for the rule language to be prescriptive in how it is met.

70.62(c)(3)(i) This section requires a plan to be submitted within 6 months of the effective date of the rule. This requirement should pertain only if a licensee has not already completed the actions outlined in 70.62(c)(3)(ii) 70.62(c)(3) There is no mention of timeframe for a licensee to come into compliance with the revisions to the rule that are not related to completion of the ISA and submittal of a summary. When 10CFR20 was revised, licensees were given one year until the requirements become effective in which to implement programmatic changes. 10CFR20 .1008 specifically addressed potential contradictions between license applications and regulations. It seems probable that conforming license amendments will be required to correct inconsistencies in areas not related to the ISA (e.g., reporting requirements) and to achieve compliance with 70.65(a). BWXT recommends an effective date sufficiently far into the future that programmatic changes can be implemented at the operating facilities and that any necessary conforming license amendments can be completed.

Section Comment 70.65(a) The concept of establishing a safety program under 70.62 is confusing. As stated in the previous comment on 70.62(a), the requirements for including the additional information as part of a license application can be included without creating a narrowly focused definition of the safety program.

70.72(c)(l)(i) This section seems clear until the reader tries to understand the footnote, which attempts to explain new types of accident sequences. Taken literally, which we must be able to do with regulations, this footnote will require nearly all process changes to require a license amendment. This outcome is in direct conflict with commission directives issued during the development of the rule. BWXT recommends the footnote be deleted. The language in 70.72(c)(l)(i) is completely adequate in the absence of the footnote.

70.72(d)(l) BWXT believes the 90-day update requirement is unnecessary and is inconsistent with the requirements in 10CFR50.71 for reactor licensees whose potential consequences are significantly greater than those at fuel facilities. BWXT supports an annual update of the ISA Summary.

70. 72(d)(3) This section requires annual submittal summarizing all changes to records required by 70.62(a)(2). The requirements for records in 70.62(a)(2) apply to all records described in 70.62(b) through (d). These records include Process Safety Information (70.62(b)) which enables the performance of the Integrated Safety Analysis. This would include procedures, drawings, detailed equipment lists, etc. BWXT does not believe NRC requires a summary of changes to this type information.

70.73 NRC should consider including a maximum timeframe for license renewal that is substantially longer than the current practice of 10 years. If a "living license" is truly the outcome, as described in the Supplementary Information, it seems renewal periods as long as 20 years would be appropriate.

Appendix A The terminology in (b)(l) clearly ties the failure to the performance (b) requirements. The phrase, "and which results in failure to meet the performance requirements of Sec. 70.61", is very clear. This phrase should be consistently included in (b)(2)-(5) using the exact same wording.

Specific comments were also solicited in the following topics:

1. Backfit Provision BWXT believes the Backfit Provision should be immediately effective. This view has been clearly articulated in past meetings and in the NEI comments on this rule.

If the backfit provision is not immediately effective, an alternative would be to make it effective for facilities or systems for which the ISA has been completed and take ISA Summary submitted to NRC.

In either case, backfit language should be included in the rule now with dates or circumstances under which it is effective.

2. NRC-OSHA Preemption BWXT has no comment

3. ISA Methodology BWXT believes the proposed rule offers sufficient flexibility .

4. ISA Summary Update Frequency As stated in comments to 72.72, BWXT believes an annual update to the ISA Summary is adequate and consistent with 10CFR50.

DOCKET NUMBER PROPOSED RULE PR 0 CP'IFR1l 33~

DOC _7::0 DUKE COGEMA STONE & WEBSTER O!- .

J ADJ!

Annette Vietti-Cook, Secretary of the Commission 13 October 1999 US Nuclear Regulatory Commission PSH-99-001 Attention: Rulemakings and Adjudications Staff Washington, DC 20555-0001

SUBJECT:

Comments on Proposed Rule 10 CFR Part 70, Domestic Licensing ofSpecial Nuclear Material; Possession of a Critical Mass ofSpecial Nuclear Material

[64 FR 41338, 30 July 1999]

Dear Ms. Vietti-Cook:

Duke Cogema Stone & Webster, LLC (DCS), the contractor and prospective licensee responsible for the design, construction, and operation of the US Department of Energy's Mixed-Oxide (MOX) Fuel Fabrication Facility, offers the enclosed comments to the proposed revision to 10 CFR 70, in response to the NRC's request for public comment.

• We note that the Nuclear Energy Institute (NEI), on behalf of the nuclear fuel cycle industry, is submitting comments separately. DCS is a member of NEI, and acknowledges and endorses NEI's comments. The enclosed recommendations are in addition to, and not in lieu of, NEI's comments.

While DCS is a new participant in these proceedings, the interactions we have ob$erved to date in the development of this proposed rule have been encouraging, ancl we ting. ~h ~ a&t f,Ilajority of the proposed changes to be positive and productive. Our comments primarily propose clarifications or minor changes to ensure consistency throughout the rule. -We. *al$o propose elimination of the plutonium-specific provisions deleted in an earlier draft of the proposed change, as we believe they are redundant to - and in some ways contrary to - otner* proposed changes.

PO Box 3147 400 South Tryon Street, WC-32G Charlotte, NC 28231 -1847 Charlotte, NC 28202 G:V..icensing\Letter to NRG - Proposed Revision to 10CFR70.doc

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Secretary of the Commission PSH-99-001 13 October 1999 Page 2 of2 We look forward to continued interaction in the development of this rule change, the ongoing work on a generic Standard Review Plan (i.e., NUREG-1520), and a MOX-specific SRP. If I can provide any additional information, please do not hesitate to contact me.

Sincerely,

~S-~ngs Licensing Manager

• xc: Edward J. Brabazon, DCS Donald J. Chamberlain, DCS Marvin S. Fertel, NEI Robert H. Ihde, DCS James V. Johnson, USDOE/MD Felix M. Killar, NEI John E. Matheson, DCS Toney A. Mathews, DCS Mark A. Michelsen, DCS Andrew Persinko, USNRC Charles F. Sanders, DCS Theodore S. Sherr, USNRC

DCS Comments on Proposed Rule 10 CFR Part 70 Page 1 of 5 Items Relied on For Safety (§70.4, §70.64)

DCS notes that NEI has commented, with regard to §70.64(a)(8), that the use of "item relied on for safety" (IROFS) is problematic in the context of design, inspection, and maintenance, owing to the definition of IROFS including "activities of personnel" (§70.4). DCS shares this concern and proposes that changing the definition in §70.4 to limit IROFS to " structures, systems, equipment, and components" would ameliorate this concern. It is reasonably straightforward to classify physical items as being relied upon for safety, and to apply graded QA controls, including management measures, to design, construction, operation, maintenance, etc., of those physical items, based on their respective safety functions. It can be confusing to try and classify and grade items when they include "personnel activities," since an activity has little importance absent the context of its influence on a physical item' s safety function. Removing "personnel activities" from the definition of IROFS would not limit their importance, but rather would put activities in context with the structures, systems, equipment, or components to which they are related, without necessitating a change in the balance of the proposed rule.

Doing so will also help address the concern raised by NEI with regard to §70.65(b)(6), where they recommend that IROFS listed in the ISA Summary be limited to the systems level and that they not include personnel activities such as the use of procedures, which change constantly.

Definition of Worker, Definition of Controlled Area, and Consideration of Collocated Workers (§70.4, §70.61)

DCS notes that NEI has addressed this issue in part in its comments associated with §70.61 ,

noting further that they selected the MOX facility as an example of the problem associated with the current proposed language. DCS fully endorses and reiterates NEI's comments in this regard.

As indicated in the definition of "worker" in §70.4, it is apparent that the NRC intends to consider individuals outside of the controlled area boundary as workers if they are subject to 10 CFR 20 requirements. We expect that the US Department of Energy (DOE) will also comment on this matter, and concur with their position that 10 CFR 835 provides an equivalent level of protection, such that collocated workers - inside or outside the controlled area - who are subject to the requirements of either 10 CFR 20 or 10 CFR 835 (or other equivalent control) should be considered "workers," provided the licensee can demonstrate the ability to provide management measures (e.g., notification, evacuation, etc., as appropriate) in the event of an emergency.

Plutonium-Specific Provisions (§70.22, §70.23)

An earlier draft revision to 10 CFR 70 (as documented in SECY 98-185 and subsequent discussions) included the deletion of plutonium-specific provisions contained in §§ 70.22(f),

70.23(a)(8), and 70.23(b). The latest draft retains these provisions, indicating that NRC Staff

DCS Comments on Proposed Rule 10 CFR Part 70 Page 2 of 5 now intend that they remain in 10 CFR 70. DCS notes some problems with this approach, and proposes reconsideration of the original plans to delete these sections. To wit:

§70.22(t) states that plutonium-related applicants shall provide information on the plant site, design basis of principal structures, systems, and components (SSCs), etc., as part of the license application. This section requires information that is also required in other sections of the revised rule, and is at best redundant in this regard (and therefore unnecessary).

More importantly, this section has not been revised to reflect the provisions of §70.65, which calls for an ISA Summary (containing the results of the safety assessment, also required in §70.22(f)) to be submitted with the license application, but not to be included as part ofthe license. As written, §70.22(f) seems to contradict §70.65 in this regard.

§70.23(a)(8) states that the Commission will approve a plutonium facility's license application only after construction of principal SSCs has been completed in accordance with the application. Certainly this is not a requirement unique to plutonium facilities. The NRC already has the authority to grant licenses conditional upon successful completion of certain actions (such as successful startup testing, training, etc.). Completion of construction in accordance with the license application seems such an obvious condition that this specific provision seems redundant and therefore unnecessary.

§70.23(b) states that the Commission will approve construction only after determination that the design bases of those SSCs, and the attendant quality assurance program, are adequate to protect against natural phenomena and the consequences of potential accidents. Our concerns are:

( 1) This provision as written seems contrary to other changes being proposed under the draft rule, as it addresses consequences of potential accidents, as opposed to the risk associated with credible accidents. Further, if this provision were amended to address risk as opposed to consequences - i.e., for consistency with the proposed §70.61 - it would be redundant to those proposed changes.

(2) The standard set in §70.23(a)(7) for other 10CFR70 licensees is that construction can commence based on a conclusion by the Director ofNMSS that environmental impacts have been appropriately addressed. Even in the absence of a mandated PHA submittal (a provision of the earlier draft also struck from the latest version), the discretion afforded the NRC under §70.23(a)(7) - i.e., NRC's authority over construction associated with "any .. .activity which the Commission determines will significantly affect the quality of the environment" - is adequate to ensure the sufficiency of information provided to NRC to authorize or disallow construction.

In consideration of these issues, and in the interest of statutory efficiency, DCS proposes that

§70.23(a)(7) be clarified for applicability to plutonium facilities, and §§ 70.22(f), 70.23(a)(7),

DCS Comments on Proposed Rule IO CFR Part 70 Page 3 of 5 and 70.23(b) be eliminated as previously proposed. Doing so would avoid the preconception that, irrespective of design features and material composition, plutonium is "more special" than other special nuclear materials.

DCS understands and shares the NRC's sensitivity to issues associated with licensing a plutonium facility. Any particular requirements associated with a MOX facility (such as the requirement for specific content in support of a request for construction authorization) could then be applied via guidance in the MOX-specific Standard Review Plan, or via Commission Order, taking advantage of existing discretion under other provisions of the rule without unnecessarily prescriptive language in the rule that conflicts with other provisions intended to support risk-informed, performance-based regulation. DCS is committed to working with the NRC staff and the Commission to ensure that our license application is complete and sufficient to enable you to conclude that its design basis adequately protects the health and safety of the public, our workers, and the environment. We believe that we can develop such a basis within the proposed rule, without undue risk, and that the other, thoughtfully considered changes proposed to 10 CFR 70 provide for adequate demonstration our facility's safety basis.

Decommissioning Requirements (§70.22, §70.38)

While not a part of the proposed change under consideration, DCS anticipates DOE will submit a comment requesting consideration of modifying the current rule to account for DOE-owned facilities. DCS shares this concern, which would presumably affect §§ 70.22(a)(9) and 70.38.

In SECY-99-177, NRC Staff proposed that this issue could be resolved without a change to current regulations, but DCS is unaware as to a final Commission position in this regard. DCS intends to engage the NRC in this issue soon, to understand whether the decommissioning requirements for the MOX fuel fabrication facility will require a rulemaking. If it is apparent to the NRC that such a rulemaking will be required, DCS suggests, in the interest of efficiency, that it be addressed in this revision to IO CFR 70.

Worker Dose (§70.61)

DCS notes that the application of occupational dose limits for accident conditions is virtually unprecedented in facilities and activities regulated by the NRC. In fact, in a recent change to I 0 CFR 60, NMSS provided a rationale for why accident dose limits for workers were unnecessary:

Although it is the Commission 's intent that the regulations in part 20 also be observed to the extent practicable [emphasis added] during emergencies, the Commission also recognizes that, in an actual emergency, operations that do not conform to the regulations may be necessary to protect public health and safety. Notwithstanding the general applicability of these regulations to all operational situations, it is not the Commission 's intent that these requirements apply to Category 2 events [i.e., 10 CFR

DCS Comments on Proposed Rule 10 CFR Part 70 Page 4 of 5 60 's "unlikely " and "highly unlikely " events) as a design basis for the facility. [60 FR 15181]

The Commission notes that dose limits are not proposed for protection of workers during Category 2 design basis events, consistent with the policy in practice for facilities regulated by the Commission under parts 50 and 72. The Commission has determined that specific standards for the protection of workers during Category 2 events are not needed for part 60. First for some design basis events, the repository design and quality assurance enhancements employed to satisfy the proposed requirements, for protection of members of the public, during Category 2 events, will also provide a measure of protection for onsite workers. Second, onsite workers would have access to protective equipment (e.g., respirators) and clothing, should the need ever arise. Third, onsite workers would be trained in emergency response and procedures to deal with operational problems related to these kinds off events. Fourth, part 20 should provide adequate worker protection standards. [60 FR 15183]

DCS agrees with this rationale, and proposes the elimination of explicit worker accident doses from the proposed rule. While it is true that there are differences between 10 CFR 70 facilities (specifically, in DCS' case, a MOX fuel fabrication facility), commercial reactors regulated under 10 CFR 50, and a potential preclosure repository operations area regulated under 10 CFR

60. It is not apparent, however, that 10 CFR 70 facilities are inherently more haz.ardous to workers. Commercial reactors involve considerably more energetic systems and potentially volatile source terms than do fuel fabrication facilities, and proposed repository operations involve analogous material handling operations within hot cells (for transfer of - notably irradiated - spent fuel). DCS proposes, therefore, that the NRC maintain consistency with past precedent in this regard, and eliminate the specific worker dose limits in 10 CFR 70.

Log of Failures (§70.62), Facility Changes (§70.72)

DCS notes NEI has commented extensively on these very important issues, and reiterates NEI's comments in this regard, especially with regard to the redundancy of a failure log (§70.62),

requirements for notification of "any change" to the facility §§ 70.72(a) and 70.72(b), and the unnecessarily onerous 90-day notification requirements of§ 70. 72(d).

Additional Content of Application and ISA Summary (§70.65)

DCS notes in §70.65(b) that, consistent with discussions to date, the NRC anticipates the ISA Summary will be submitted with the license application, but not incorporated in the license.

The wording in §70.65(a), however, seems to contradict this position, given the general heading Additional Content of Applications. DCS proposes removing "including the integrated safety analysis summary and a description of the management measures" to clarify the issue.

DCS Comments on Proposed Rule IO CFR Part 70 Page 5 of 5 Absent this or some other clarification, DCS is concerned that §70.65(a), as written, leaves the impression that the ISA Summary is part of the application (and by reference in the material license certificate, part of the license). The requirement to include the ISA Summary is adequately covered in §70.65(b). If necessary (i.e., if not sufficiently implicit in the ISA Summary requirements), additional discussion of the inclusion of management measures as part of the ISA Summary could be included in §70.65(b).

DCS also notes that NEI has expressed a related concern (i.e., under the heading Safety Program Definition [§70.65]. DCS shares this concern as well, and suggests that the NRC clarify the relationship of the ISA Summary to the license and the safety basis to ensure consistency throughout the rule with the intent expressed in §70.65(b).

DCS specifically notes NEI's comment with regard to §70.65(b)(3), regarding the extent of information required in the ISA Summary for processes which have been evaluated but have no safety implications, and reiterates NEI' s comments .

CD Department of Energy OUI **r - If

• ' I\

Germantown , MD 20874-1290

• 99 October 13 , 1999 0

AD._

Secretary of the Commission U.S. Nuclear Regulatory Commission Attention: Rulemakings and Adjudications Staff Washington, D .C. 20555-0001 DOCKET U BER PROPOSED RU '10

Dear Sir/Madam:

( lP'-1Ff<'!I:33~

In reference to the Comments on Proposed Rule 10 CFR 70: Domestic Licensing of Special Nuclear Material ; Possession of a Critical mass of Special uclear Material (Federal Register

• Vol. 64, No. 146, pp. 41338-41357 dated July 30, 1999). The Department of Energy submits the attached comments on the proposed revisions to IO CFR 70 in response to a request for public input in the July 30, 1999 Federal Register notice.

Please contact Dr. Jacques Read, of my staff, if you have any quesitons on this. He can be reached at (301) 903-2535, e-mail: jacques.read@eh.doe.gov.

s~(~

Richard L. Black, Director Office of Nuclear Safety Policy and Standards Attachments cc:

The Honorable Greta Joy Dicus, Chairman, NRC The Honorable its J. Diaz, Commissioner, NRC The Honorable Edward McGaffigan, Jr., Commissioner, NRC The Honorable Jeffrey S. Merrifield, Commissioner, NRC Dr. William D. Travers, EDO/NRC Dr. Carl A. Paperiello, NMSS, NRC Mr. Frank Miraglia, EDO, NRC OCT I 9 1999 l\cknowfedged by card ......................-.,.. ,

@ Printed with soy ink on recycled paper

JS.

AU

DOE Comments on Proposed 10CFR 70 Rule Published in the Federal Register July 30, 1999

1. Section 70.4, "Worker" "individual whose assigned duties in the course of employment involve exposure to radiation and/or radioactive material from licensed and unlicensed sources ofradiatiori (i.e., an individual who is subject to an occupational dose as in 10 CFR 20 .1003 )".

The following change should be made to the definition provided: " .... from licensed sources of radiation, and radiation from man-made non-regulated sources ( e.g., an individual ..... ). As originally defined, persons who are subject to occupational doses from natural sources of radiation, for example airline pilots and astronauts subject to high cosmic background might be included, whereas workers involved with the manipulations of unlicensed radioactive materials might not be. The proposed change removes this source of confusion.

2. Section. 70.11 should be revised to reflect the applicability of the NRC authority over a MOX fuel fabrication facility owned by DOE, pursuant to changes in law last year.

3. Section 70.22 (f) should be coordinated with 70.65. As written, it is not clear whether the requirements are collateral, complementary, or redundant.

I

4. Section 70.23 (b) should be e,camined to clarify the need for this requirement in light of similar information being submitted pursuant to 70.65. Irrespective of 70.65, 70.23 (b) appears to be an unnecessary step and should be considered for deletion by NRC. IfNRC chooses to retain 70.23 (b), NRC should clarify how the authorization process would be conducted, given that the procedural step has never been exercised to the knowledge of DOE. Furthermore, NRC should identify how the "design basis" authorization is defined, why it is necessary, and how it relates to the ISA.

--\

5. Section 70.61, Performance Requirements:

This section of the rule sets the dose limits only for high-consequence and intermediate-consequence events with the likelihood of highly unlikely and unlikely and does not set the limits for anticipated occurrences similar to that in 10CFR72, parts 104 and 106. The dose limit for anticipated occurrences is much less than the limits for high-consequence and intermediate-consequence events and the anticipated occurrences, when analyzed unmitigated, could result in doses that potentially exceed the limits for high-consequence and intermediate-consequence events. The NRC should specify the dose limits for po:tential anticipated occurrences at tlie nuclear fuel cycle facilities. This part of the rule then will cover the range of likelihood (anticipated, likely, unlikely, and highly unlikely) of potential accidents that could occur at nuclear cycle facilities. This could result in an increase in the number of

\..

structures, systems, and components relied on for safety and will impact the design, operation, and licensing of the MOX facility.

a) Section 70.61 (d) is not related to 70.61 (b) or 70.61 (c) yet the three conditionals are all linked together. Subpart (d) should be segregated from (b) and ( c) if (d) is preserved as an independent entry (as would seem preferable).

Otherwise, (d) should be subsumed under (b) and/or (c), and the regulatory basis for criticality prevention should be predicated on the risks and/or consequences of the accidents, rather than the presence of initiator precursor per se. ( editorial) b) Section 70.61(f), Each licensee must establish a controlled area, as defined in section 10 CPR 10.1003, in which the licensee retains the authority to determine all activities, including exclusion or removal of personnel and property from the area. For the purpose of complying with performance requirements of this section, individuals who are not workers, as defined in sec. 70.4 may be permitted to perform ongoing activities (e.g., at a facility not related to the licensed activities) in the controlled area, if the licensee demonstrates compliance with 70.61(:t) (1) or (2).

These requirements consider the individuals working in the nearby facilities as public when performing an accident analysis to determine the consequences of the accidents that may occur at the facility. 1bis would result in a more stringent application of safety requirements for the protection of workers ( e.g., additional items relied on for safety) at the Mixed Oxide (MOX) Fuel Fabrication Facility (FFF), Pit Disassembly, Conversion Facility, Immobiliz.ation Facility, and any other nearby DOE facilities. This also would have a substantial impact on the cost of the MOX facility. The workers in the nearby DOE facilities are protected under DOE Code of Federal Regulations 10 CFR 835, "Occupational Radiation Protection" and DOE Order 5400.5, "Radiation Protection of the Public and the Environment," and potentially by draft 10 CPR 834, "Radiation Protection of Public and the Environment," which are comparable to the protection afforded the workers under NRC 10 CFR 20.

Therefore, the NRC should consider changing Section 70.60(f) (1) to read as follows: Demonstrates and documents, in the integrated safety analysis, that those individuals at the location of their activities do not exceed the performance requirements of paragraphs (b) (1), (b) (3), (b) (4)(ii), (c)(l) and (c) (4)(i) of this section, including the Section 70.60(f)(2) requirement in Section 70.22 (h)(2)(ii)(3). Accordingly, the paragraph could be rewritten as follows: "Each licensee must ensure that a controlled area can be established as defined in Sec 20.1003 in which the licensee has the authority to enable control over all activities.

6. Section 70.62 (d) Management Measures. Second sentence: "The measures applied to a particular engineered or administrative control or control system may be

commensurate with the reduction of risk attributable to that control and control system."

The management measures.are to be applied to items relied on for safety b ~ on their contribution to a reduction in risk. The failure data for most fuel facility equipment are not well documented. The frequency of failure of equipment is a major

', , factor in determining the reduction of risk. Therefore, the NRC should consider the graded approach to mapagement measures, using risk as one of the factors in applying the management measures to items relied on for safety. 01:Qer factors should include consequences, life cycle, a:nd magnitude of hazard involved. Balanced and integrated criteria for determining th~ appropriate management measures can ensure the safety ,

and integrity of the facility.

7. Section 70.64 a(4) Environmental and dynamic effects. The design must provide for adequate protection from environmental conditions and dynamic effects associated with normal operation, maintenance, testing and postulated accidents that could lead to loss of safety functions. ,

/

This requirement is unclear. What does it mean? Is formal Equipment Environmental Qualification Program required similar to that required under 10 CFR 50.49 and Regulatory Guide 1.89? The NRC should clarify this requirement and should not impose requirements that may not be appropriate or necessary because of the nature of the processes at non-reactor nuclear facilities. *

8. Section 70.64(b). Facility and systems design and layout must be based on defense-in-depth practices. The defense-in-depth definition as used in Section 70.64 does not reflect the defense-in-depth design philosophy as defined in W ASH-1250, The J Safety of Power Reactor and Related facilities," which outlined three levels of safety concepts in the design of a nuclear Facility. The three levels concern different design considerations in the facility; however, these design considerations intermesh and

'J overlap so that distinctions as to whether certain design features belong to one or the other of these levels are somewhat arbitrary. -

, The definition in the rule oversimplifies the concept of defense in depth, to w~ere-it loses its basic purpose. For,example, Sections 70.64-(b)(l) and (2) do not adequately represent the implementation

-, of defense-in-depth philosophy in the design, The selection of engineered controls over administrative controls and features that reduce challenges to items relied on for safety are partially implemented in the concept.

\

For non-reactor nuclear facilities, one level of safety by itseIBmay not be sufficient to protect against the release of radioactive materials. However, a combination of any of these levels should provide a sufficient level of protection to the public, workers, ap.d '

environment. The NRC should reexamine the definition and the application of the defense-in-depth philosophy to be commensurate with the level of hazard and associated consequences and risk. NRC should clarify how defense- in- depth philosophy applies to the regulation of facility types stated in section 70. 60.

9. Section 70.65 (9). A description of the definitions of likely, unlikely, highly unlikely, and credibl~ as used in the evaluations in the integrated safety analysis.

The NRC should define the terms likely, unlikely, highly unlikely, and credible in the rule so that there will be one set of definitions applied to all nuclear fuel facilities.

This will minimiz.e the interpretation and application of these terms in the integrated safety analysis.

10. Section 70.73 states that a description of changes made to structures, systems, components, etc., should be sent periodically by the licensee to the NRC. the term "periodically should be defined.

11. On the ISA update summary, the 90 day period appears to be too cumbersome. An annual update (similar to the annual FSAR updates for reactors per 10CFR50.71 (e))

should suffice. If the spirit of the regulation is not being met based on experience, the licensee should face enforcement action.

12. A backfit process similar to that in 10 CFR 50.109 or 10 CFR 76.76 should be incorporated into the revisions to Part 70 and should apply to the current proposed changes to the extent they apply to existing facilities.

13. Because DOE facilities do not have the uncertainty of continued corporate sponsorship inherent in commercial facilities, the timeliness and schedule requirements in the decommissioning requirements of § 70.38 should be revised to include separate requirements for DOE facilities.

14. The criticality requirements of§ 70.24 should be revised to permit alternate criticality control provisions to be accepted for DOE facilities without requiring an exemption.

15. As additional DOE facilities are licensed by the NRC under the provisions of Part 70,,.

NRC should ensure that the requirements address the full range of fissionable and 1

fissile materials at these facilities. -

Comments on NUREG 1513, Integrated Safety Analysis Guidance Document Guidance on the quality assurance of the ISA process itself should be supplied.

Comments on NUREG 1520, Standard Review Plan (SRP)

This review focuses on the Integrated Safety Analysis OSA) Chapter 3 of the SRP, since most of remaining SRP Chapters are dependent on the ISA results. The comments only address Chapter 3, ISA, and Appendix A.

Quantitative and non-q_uantitative determination of likelihood of accidents:

The quantitative determination of event likelihood is dependent on several factors, such as the equipment failure rate; operator error rate; and surveillance, inspection,

_ maintenance, and testing intervals. These factors must be known to determine the frequency of occurrence of an event. The non-quantitative determination of the likelihood imposes design criteria such as redlllldancy, independence, concurrency, and assurance measures for reliability and availability of items relied on for safety. The likelihood index, which is a summation of preventive and mitigation controls failures, does not consider the interdependency of these controls, nor does it reflect the actual performance of these controls under the expected operating conditions. For example, the integration of failure rates over a range of potential failures for controls that are independent and the --

summation of failure rates for dependent controls would be more likely to represent the actual performance and likelihood of failure of these controls.

The criteria are subjective and open to arbitrary interpretation by a reviewer. The risk for potential disagreement on appropriate assigned accident likelihood, duration index, and failure rates is extremely high and could render the results of the integrated safety r-analysis (ISA) unacceptable. (This could jeopardize the chances of obtaining a facility

_ license and impact the cost and schedule of the project) See the comment on the SRP risk matrix in our response to Section 70.65 aboye.

Failure rate of components credited/or prevention and mitigation of accidents and reductiQn of risk and/or likelihood:

The fuel fabrication facilities in the US do not maintain a failure-rate database on their equipment, and failu,re-rate data for structures, systems, and components for a MOX fuel fabrication facility do not exist. The available failure-rate data in the US are geared more toward the commercial nuclear power industry. In addition, the type of equipment used in

'-1 reactors is significantly different from that used in the fuel fabrication facilities. The Europ~ may maintain failure-rate data on their equipment. However;the use of this data will depend oil the basis and quality of the data, e.g., collection and control of the data. The European data may have to be validated under US quality assurance practices.

The use of failure data for specific equipment without consideration of the total systems integration failure (i.e., system interactions, support system failures, etc.) may not reflect the effectiveness of these engineered features in mitigating the risk from the potential b.a:zards. The ISA attempts to implement the performance-based, risk-informed approach but fails to recognize that witholl;t comprehensive and valid equipment failure data, the approach cannot be implemented in a meaningful fashion. '-

Summation offrequencies of all accident sequences:

The two performance ~ety measures established as part of the Nuclear Regulatory Commission (NRC) Strategic Plan are (1) no inadvertent criticality and (2) no increase in reportable radiation releases. The argument used to justify using the summation of frequencies of all potential accident sequences that could occur at the facility during its

service life is not supported by any technical justifications. It is unrealistic to take the 5-year-average of reportable radiation exposures, allocate 10% of the average, and divide by the number of currently operating fuel facilities to establish a safety performance goal for the facility. There will be only one mixed oxide (MOX) fuel facility; therefore, the performance safety goal for MOX could be set at a much higher level. For example, the performance safety goal for accidents with immediate consequences at the MOX Fuel Fabrication Facility (FFF) could be 4E-2 and for high consequences could be lE-2, whereas for low-enriched uranium (LEU) facilities, the goal would be set at 4E-3 and lE-3, respectively, a factor of 10 lower.

The summation of frequencies of all accidents and comparison of the result to a set of quantitative goals may or may not reflect the actual risk from the facility because these goals are set without sufficient basis or adequate data. The data used to set the safety performance goal numbers are insufficient and statistically insignificant In addition, the number of operating facilities should not be considered as a significant factor in determining the safety performance goal. The 10% of the average of reportable increase in radiation exposure is an arbitrary number. Why not 15% or 20%? Clarification should be requested to ensure that MOX and LEU facilities are judged on the same scale of risk to the health and safety of the public, workers, and environment.

Risk index evaluation:

The risk index evaluation includes factors such as frequency of the initiatir{g event, duration of vulnerability, and frequency of the preceding system/control failure. In Table 5 of the ISA, a duration index is assigned to the duration of the vulnerable state.

For example, a duration index of -2 is assigned for an average failure of a few days and a duration in years of0.001, ahd a-5 duration index is assigned for a 5-minutes-average failure and a lE-5 quration in years. What does a few days mean? Does it mean 2, 4, or 6 days? How does a 5-minute-average failure result in an index of-5?

The bases for duration index numbers appear to be selected arbitrarily. The duration of a control/system failure is important in determining the overall risk index; however, these numbers should be based on credible data and properly factored in the index. The data and the methodology for assigning a duration index number should be referenced, and bases for the assigning of index numbers also should be provided.

ISA process:

-The ISA process includes the use of several tables to assess the risk from potential accidents and the acceptability of an item relied on for safety to prevent or mitigate the consequences of an accident. The process steps are complex and very hard to follow. A

• complex process may not produce results that reflect the conditions analyzed. The I~A process should be tailored to accommodate the complexity and uniqueness of the operation to be analyzed and simple enough that it can be easily understood and applied.

A logic diagram or procedure should be included to describe the process better.

Determination of assurance measures for safety controls:

The ISA calls for every item relied on for safety in accident sequence categories '2 or 3 (high and intermediate consequences) to be assigned at least a minimal set of measures to defend against a coIIlilion mode failure of all controls. In additio!!, it specifies this minimal set as configuration management, regulk auditing, adequate labeling and training, written procedures, surveillance, and corrective and preventive maintenance.

The minimal set of assurance measures for items relied on for safety appears to be selected arbitrarily, and there is no logic or basis to support it The rule calls for tq.e assurance measures to be selected based on the importance of the item to safety and the level of risk associated with its failure. This could have a substantial impact on the design, construction, testing, operation, and maintenance of the facility.

In summary, the ISA process is too prescriptive. This has made the process very complex and confusing. Continued attention should be given to the process as it evolves. (A clear and well-defined ISA process will minimiz.e the risk to the MOX project.)

Appendix A - Example Procedure The discussion appearing in this section contains virtually nq firm guidance as to how to quantitatively justify category assignments. It does, however, contain logical flaws, and must be rewritten.

~ the early days of nuclear power plant regulation, a British authority named Farmer proposed a method of judging the acceptability of risk from accidents. This method involved estimatmg a consequen.ce measure, such as calculated dose at the site boundary, and a likelihood for the accident yielding that consequence, given~ the units of per year.

If all identified accidents were given as points on *a log-log plot of consequence versus likelihood, then Farmer postulated that,there would be a line or curve on that plot such that if all points had consequences and likelihoods less than the curve, then the overall risk could be accepted. He was shown to be wrong; while providing a possible method of graphically presenting comparative risks, the plot cannot be used to assign acceptable areas of risk. ,

This "Farmer's curve" cannot be a straight line, because the acceptable risk of an accident does not remain constant as the accident consequences increase - i.e., higher consequence accidents must be more than proportionally less likely than lower consequence accidents. This is because no measure of consequence is appli~ble over a broad spectrum of accidents. For example, worker doses ofless than 1 Sievert can be

• - acceptable at very much higher probabilities than worker doses over 10 Sievert, since the first consequence is unlikely to be fatal, while the second -is likely to be fatal. The risk of death is an additional risk that 10 Sv doses have that is above and beyond ten times the health risk of 1 Sv doses. Similarly, the risk to people off-site from a catastronhe I is entirely different from the risk associated with a precautionary evacuation. Mortality and morbidity are two entirely separate hazards, and.the acceptable likelihood often deaths is not just simply one tenth that of one death.

)

The proposed guidance for judging the acceptability of consequences replaces the fallacy of Farmer's curve with that of a histogram.

Tlns view of risk is not new, it was first described by D. Bernoulli in the eighteenth

~ century (the "St Petersburg Paradox"), and was extensively investigated by L.J. Savage in the decade following the second World War (the "Sure Thing" principle). These principles state that risk management cannot be the acceptance of a likelihood, but of a consequence. If a consequence is too large to be accepted, then the design must reduce its likelihood such that its occurrence can be viewed as virtually impossible. If a consequence is acceptable, then its risk can be managed by~control of likelihood.

In practice, this means that unacceptably high consequence accidents must be both prevented and mitigated, i.e., in addition to design features to interrupt accident scenarios leading to the unacceptable consequences, there must be design features to mitigate the consequences to humans sufficient to render them acceptable.

I

LAW OF FICES GARVEY, SCHUBERT S BARER A P ARTNERS HIP OF PROFESSIONAL CORPORATIONS SEATTLE PORTLAND FIFTH FLOOR EIGHTEENTH F"LOOR ELEVENTH F"LOOR 1000 POTOMAC STREET N .W.

1191 SECOND AVENUE 121 S .W . MORRISON STREET SEATT L E, WASHINGTON 96101-2939 WASH IN GTON, O .C. 20007 PORTLAND, OREGON 97204-314 1 (206) 464-3939 (202) 965 -7880 (503) 226 -3939 F"AX : (202) 965 - 1729 PLEASE REPLY TO WASHINGTON, D .C . O F"F" IC E October 12, 1999

)> 0 0-r r DOCKET U ER r -

By Hand Delivery PROPOSED AU Ms. Annette Vietti-Cook

• Secretary of the Commission U. S . Nuclear Regulatory Commission One White Fl in t North Building 11555 Rockville Pike Rockville , MD 20852 0

w

-0 U1 Attn : Rulemakings and Adjudications Staff Notice of Proposed Rulernaking: Domestic Licensing of Special Nuclear Material (10 C.F.R. Part 70)

Dear Ms . Vietti - Cook :

In light of the recent criticality accident at the Tokai-Mura fuel fabrication facility in Japan and the Commi ssion ' s announcement that it intends to make its own study of the Japanese accident in cooperation with Japanese authorities (see NRC Press Release No . 99 - 214, October 8 , 1999) , my client , the Nuclear Control Institute , 1 respectfully requests that the commen t per iod on the Commission's proposed amendments to i ts regulations governing the domestic licensi ng of special nuclear material, as published in the Federal Register on J uly 30 , 1999 (64 Fed . Reg . 41338) , be extended for an additional period of at least sixty (60) days. Such an extension is important in order to help ensure that the implications of this accident may be fully considered by the Commission in its rulemaking process .

1 The Nuclear Control Institute is a non-profit membersh i p corporation , organized and existing under the laws of the District of Co lumb i a , which is dedicated to halting the pro li feration of nuclear weapons. Its address and te l ephone number are: 1000 Connecticut Avenue , N.W., Suite 804, Washington ,

D. C . 20036 ; Tel.: (202) 822-8444.

DOClrn p lf-nI A ., f _

[' -/ ~ 4?;125_

Ms. Anne t te Vietti-Cook October 12, 1999 Page 2 Indeed, the Institut e b el ieves it is imperative that the Commission inform itself and inte rested stakeholders on the signi fi cance of the accident for t he Part 70 regulatory structure before any final decisions a re made concerning the proposed revisions to Part 70.

Thank you fo r your consider a tion of our views .

Sincere l y,

• Eldon V. C .

Counsel to e Nuclear Control Institute cc : Paul L . Leventhal

DOCK~Tl:O l rc:- ,

[7590-01-P]

"99 JUL 26 P.? :37 NUCLEAR REGULATORY COMMISSION 10 CFR Part 70 01 r"

RIN 3150-AF22 Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

SUMMARY

The U.S. Nuclear Regulatory Commission {NRC) is proposing to amend its regulations governing the domestic licensing of special nuclear material {SNM) for licensees authorized to possess a critical mass of SNM, that are engaged in one of the following activities: enriched uranium processing; fabrication of uranium fuel or fuel assemblies; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; fabrication of mixed-oxide fuel or fuel assemblies; scrap recovery of special nuclear material; or any other activity invoMng a critical mass of SNM that the Commission determines could significantly affect public health and safety or the environment. The proposed amendments would identify appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed these criteria; require affected licensees to perform an integrated safety analysis {ISA) to identify potential accidents at the facility and the items relied on for safety necessary to prevent these potential accidents and/or mitigate their consequences; require the implementation of measures to ensure that the items relied on for safety are available and reliable to perform their function when need~d; require the inclusion of the safety bases, including a summary of the ISA, with the license application; and allow for licensees to make certain changes to their safety program and facilities without prior NRC approval.

~ l31 tC/99 DATES: The comment period expires {i,.sert 15 days after f!'UBlieetien i,. the l=eder:al Reeietert 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 rulemaking website through the NRC home page (http://www.nrc.gov). From the home page, select "Rulemaking" from the tool bar at the bottom of the page. The interactive rulemaking website can then be accessed by selecting "Rulemaking Forum." This site provides the ability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Ms. Carol Gallagher by telephone at (301) 415-5905 or e-mail cag@nrc.gov.

FOR FURTHER INFORMATION, CONTACT: Theodore S. Sherr, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone (301) 415-7218; e-mail tss@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Background II. Description of Proposed Action I. Background A near-criticality incident at a low enriched fuel fabrication facility in May 1991 prompted NRC to review its safety regulations for licensees that possess and process large quantities of SNM. [See NUREG-1324, "Proposed Method for Regulating Major Materials Licensees" (U.S.

Nuclear Regulatory Commission, 1992) for additional details on the review.] As a result of this review, the Commission and the staff recognized the need for revision of the regulatory base for these licensees, especially for tho~e possessing a critical mass of SNM. Further, the NRC staff 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:

2

.(1) Plant and external hazards and their potential for initiating accident sequences; (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.

NRC held public meetings with the nuclear industry on this issue during May and November 1995. The Nuclear Energy Institute (NEI) explained, to the Commission, industry's position on the need for revision of NRC regulations, in 10 CFR Part 70, at a July 2, 1996, meeting, and in a subsequent filing of a Petition for Rulemaking (PRM-70-7) in September 1996. NRC published in the Federal Register a notice of receipt of the PAM and requested public comme~ on August 21, 1996 (61 FR 60057). The PAM requested that NRC amend Part 70 to:

(1) Add a definition for a uranium processing and fuel fabrication plant; (2} Require the performance of an ISA, or acceptable alternative, at uranium processing, fuel fabrication, and enrichment plants; and (3) Include a requirement for backfit analysis, under certain circum~tances, within Part 70.

In SECY-97-13?, dated June 30, 1997, the staff proposed a resolution to the NEI PAM and recommended that the Commission direct the staff to proceed with rulemaking. The staffs recommended approach to rulemaking included the basic elements of the PAM, with some modification. In brief, staffs proposed resolution was to revise Part 70 to include the following major elements:

(1) Performance of a formal ISA, that would form the basis for a licensee's safety program. This requirement would apply to all licensed facilities or activities, subject to NRC regulation, that are authorized to possess SNM in quantities sufficient to constitute a potential for nuclear criticality (except power reactors and ,the gaseous diffusion plants regulated under 10 CFR Part 76);

, (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 3

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 NRC prior approval; and (5) Consideration by the Commission, after licensees' initial conduct and implementation of the ISA, of a qualitative backfitting mechanism to enhance regulatory stability.

In an SRM dated August 22, 1997, the Commission"... approved the staffs proposal to revise Part 70" and directed the NRC staff to " ... submit a draft proposed rule ... by July 31, 1998."

A draft proposed rule was provided to the Commission in SECY-98-185, "Proposed Rulemaking - Revised Requirements for the Domestic Licensing of Special Nuclear Material,"

dated July 30, 1998. The draft proposed rule reflected the approach recommended in SECY-97-137. In particular, the safety basis ior a facility, including the ISA results, would be submitted as part of an application to NRC, for review, and Incorporated in the license. Also in SECY 98-185, the staff recommended that a qualitative backfit mechanism should be considered for implementation only after the safety basis, including the results of the ISA, is established and incorporated in the license, and after licensees and staff have gained experience with the implementation of the ISA requirement.

In response to SECY-98-185, the Commission issued an SRM .dated December 1, 1998, which directed the staff not to publish the draft proposed rule for public comment Instead, the Commission directed the staff to obtain stakeholder input and revise the draft proposed rule. In that SRM, the Commission also directed the staff to:

(1) Decide what is fundamental for NRC's regulatory purposes for inclusion as part of the license or docket and what can be justified from a public health and safety and cost-benefit 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; 4

(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 ~substantiar increase In safety test; (5) Require the reporting of certaln significant events because of their potential to impact worker or public health and safety; (6) Clarify the basis for use of chemical safety and chemical consequence criteria, particularly within the context of the Memoranda of Understanding with the Occupational Safety and Health Administration (OSHA) and other government agencies; (7) Critically review the Standard Review* Plan (SRP) to ensure that by providing specific acceptance ~riteria, it does not inadvertently prevent licensees or applicants from suggesting alternate means of demonstrating compliance with the rule; and (8) Request Input on how applicable ISA methodologies should be employed in the licensing of new technologies for use within new or existing facilities.

As directed in the SRM, stakeholder input was solicited and obtained at public meetings held in December 1998 and -

January and March 1999. . A website was established to facilitate communication with stakeholders and to so!icit further input. The nuclear industry submitted comments by letters and postings on the website. This revised proposed rule incorporates much of the December .1, 1998 SRM direction and reflects language responsive to many of the comments received. It appears that most of the major concerns with the earlier draft proposed rule have been re~lved.

5

II. Description of_ Proposed Action The proposed rule grants the NEI September 1996 PRM in part and modifies the petitioner's proposal as indicated in the following discussion.

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 ris!< levels, considering consequence criteria and the level of protection

. needed. to prevent accidents that could exceed such criteria; (2) ljle performance of an ISA to identify potential accidents at the facility and the items relied on for safety; (3) The implementation of measures to ensure that the items relied on for safety are available and reliable to perform their function when needed; *

(4) The inclusion of the safety bases, including the ISA summary, In the license appllcation;and (5) The allowance for licensees to make certain changes to their safe~ program and facilities without prior NRC approval.

The Commission's approach agrees in principle with the NEI petition. However, in contrast to the petition's suggestion that the* 1sA requirement be limited to "*.. uranium processing, fuel fabrication, and. uranium enrichment pl~nt licensees," the Commission would require the performance of an ISA for a broader range of Part 70 lice~e that are authorized to possess a critical mass of SNM. "The Part 70 llcen$8es that would t?e-affected include licensees engaged in one of the following activities: enriched uranium processing; fabrication of uranium fuel or fuel assemblies; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; fabrication of mixed-oxide fuel or fuel assemblies; scrap recovery of special nuclear material; or any other activity lnvoMng a critical mass of SNM that the Commission. determines could significantly affect public health and safety. The proposed rule would not apply to licensees authorized to possess SNM under 1 0 CFR Parts 50, 60, 72, and 76.

Furthermore, the Commission Is not currently proposing, as suggested in the NEI petition, to include a back.fit provision in Part 70. Based on the discussions at public meetings held on May 28, 1998, and March 23, 1999, the purpose of the NEl-proposed backfit provision 6

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 Commission believes that once the safety basis, including the ISA summary, is incorporated in the license application, and the NRC staff has gajned 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 initiaJ ISA process, It Is not clear how a determination of incremental risk, as needed for a backfit analysis, would be accomplished. Furthermore, although NEI previously stated that a quantitative backfit approach is currently feasible, it would appear that a quantitative determination of incrementaJ risk would require a ProbabHistic Risk Assessment, to which the industry has been I strongly opposed. The Commission requests public comment on its Intent to defer consideration of a quaJltative backflt provision in Part 70; any specific suggestions for backfit provisions that would specifically address fuel cycle backfit needs and the information that would be available to conduct the I

associated analysis; and what would constitute a reasonable period of time, inclualng supporting rationale, before a backfit provision should be implemented.

I The majority of the proposed modifications to Part 70 are found In a new Subpart H, "AdditionaJ Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear Material," that consists of 10 CFR 70.60 through 70.74. These proposed modifications to Part 70, discussed in detail below, are required to increase confidence in the margin of safety and are in generaJ accordance with the approach approved by the Commission in its SRMs of August 22, 1997, and December 1, 1998*

• Section 70.4 Definitions.

Definitions of the following 12 terms would be added to this section to provide a clear understanding of the meaning of the new Subpart H: '"Acute", '"Available and reliable to perform their function when needed", "Configuration management", '"CriticaJ mass of SNM", '"Double contingency", "Hazardous materials produced from licensed materiaJs", '"Integrated safety analysis", "Integrated safety analysis summary", "Items relied on for safety", "Management measures", "Unacceptable performance deficiencies", and 'Worker."

7

Section 70.14 Foreign military aircraft.

This paragraph reflects an administrative change to renumber the paragraph from 70.13a.

Section 70.17 Specific exemptions.

This paragraph reflects an administrative change to renumber the paragraph from 70.14. \

Section 70.50 Reporting requirements.

Paragraph (c) would be reworded to include information to be transmitted when making verbaJ or written reports to NRC.

• The new information derives from the specifics of the new Subpart H, such as sequence of events and whether the event was evaluated in theJSA. To the extent the new information is also applicable to licensees not subject to Subpart H, the Information was added with no differentiation noted. The new information that would only apply to Subpart H licensees is noted.

Section 70.60 Applicability.

. This section lists the types of NRC licensees or applicants who would be subject to the new Part 70, Subpart H. The Commission has decided that the new requirements should not apply to all licensees authorized to possess critical masses of SNM. Instead, the Commission has identified a subset of these licensees that, based on the risk ~ated with operations at these facilities, should be subject to the new requirements. This change would exclude certain facilities (e.g., those authorized only to store SNM or use SNM In sealed form for research and educational purposes) from the new reql!irements, because of the relatively low level of risk at these facilities. In general, the new Subpart is intended to ensure that the signtticant accidents that are possible at fuel fabrication facilities (and the other listed facility types) have been analyzed in advance, and that appropriate controls or measures are established to ensure adequate protection of workers, 1 public, and the environment The requirements and provisions 1

A worker, In the context of this rulemaklng, is defined as an individual whose assigned duties In the course of employment Involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation (I.e., an lndMdual who is subject to an occupational dose as In 10 CFR 20.1003).

8

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 (radiologicaJ 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 [§70.38(g)(4)0ii)] requires an approved decommissioning plan (DP) that includes " a description of methods used to ensure protection of workers and the environment against radiation hazards during decommissioning." Because the DP is submitted for NRC approval before initiation of "... procedures and activities necessary to carry out decommissioning of the site or separate building or outdoor area," the DP will continue to be the '

vehicle for regulatory approval of the licensee's practices for protection of health and safety during decommissioning. The ISA should provide valuable information with respect to developing the DP and the use of the ISA In this manner Is encouraged.

Section 70.61 Performance Requirements.

i,, the past, the regulation of licensees authorized to-possess SNM, under 1o CFR Parts 20 and 70, has concentrated on radiation protection for persons involved in nuclear activities conducted under normal operations. The proposed amendments to Part 70 would explicitly

• address potential exposures to workers or members of the public and envii:onmental releases as a result of accidents. Part 20 continues to be NRC's standard for protection of workers and public from radiation during normal operations, anticipated upsets (e.g., minor process upsets that are likely to occur one or more times during the life of the facility), and accidents. Although it is the Commission's in~ent that the regulations in Part 20 also be observed to the extent practicable during an emergency, it is not the Commission's intent that the Part 20 requirements apply as the design standard for all possible accidents at the facility, irrespective of the likelihood of those accidents. Because accidents are unanticipated .events that usually occur over a relatively short period of time, the Part 70 changes seek to assure adequate protection of workers, members of the public, and the environment by limiting the risk (combined likelihood and consequence) of such accidents.

9

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 crltlcality be limited by assuring that all processes must remain subcrltical 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 the acceptable likelihood for high-consequence events.

A risk-infon:ned approach must consider not only the consequences of potential accidents, but also their likelihood of occu~nce. As mentioned above, the performance requirements rely on the terms "unlikely" and i,ighly unlikely" to focus on the risk of accidents.

However, the Commission has decided not to include quantitative definitions "unlikely" and "highly unlikely" in the proposed rule, because a single definition for each term, that would apply to all the facilities regulated by Part 70, may not be appropriate. Depending on the type of facility and its complexity, the number of potential accidents and their consequences could differ markedly. Therefore, to ensure that the overall facility risk from accidents is acceptable for diffe1 cnt types of facilities, the rule requires app1icants to develop, for NRC 8f.Jproval (see

§70.65), the meaning of "unlikely" and "highly unlikeJy" specific to their processes and facility.

To accommodate this development, the Commission believes that the SAP is the appropriate document to include guidelines for licensees to use. A draft "Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility" has been developed. The draft SAP provides one acceptable approach for the meaning of "unlikely" and i,1ghly unlikely" that can be applied t6 existing fuel cycle facilities.

The general approach for complying with the performance requi(ements is that, at the time of licensing, each hazard (e.g., fire, chemical, electrical, industrial) that can potentially affect radiological safety is identified and evaluated, in an ISA, by the licensee. The impact of accidents, both internal and external, associated with these hazards is compared with the three performance requirements. Any (and all) structures, systems, components, or human actions, for which credit is taken in the ISA for mitigating (reducing the consequence of) or preventing (reducing the likelihood of) the accident such that all three performance requirements are satisfied, must be identified as an "item relied on for safety." "Items relied on for safety" is a

tenn 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 actuaJ "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 perfonn Its safety function wh~n needed. The combination of the set of "items relied on for safety" and the "management measures" applied to each item will detennine 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 perfonnance requirements).

The proposed performance requirements also address certain chemical hazards that result from the processing of licensed nuclear material. The question of the extent of NRC's authority to regulate chemical hazards at its fuel ~ facilities was raised after an accident in 1986 at a Part 40 licensed facility, in which a cylinder of uranium hexafluoride ruptured and resulted in a worker fatality. The cause of the worker's death was the inhalation of hydrogen fluoride gas, which was produced from the chemical reaction of uranium hexafluoride and water (humidity in air). Partly as a result of the coordinated Federal response and resulting Congressional.investigation into that accident, NRC and the OSHA ~ntered into an MOU, in 1988, that clarified the agencies' interpretations of their respective responsibi_lities for the regulation of chemical hazards at nuclear facifities. The MOU identified tho following four areas of responslblllty. Generally, NRC covers the first three areas, whereas OSHA covers the fourth area:

(1) Radiation risk produced by radioactive materials; (2) Chemical risk produced by radioactive materials; -

(3) Plant conditions that affect the safety of radioactive mat~rials; and (4) Plant conditions that result in an occupationaJ risk, but do not affect the safety of licensed radioactive materials.

One goaJ of the perfonnance requirements in §70.61 is to be consistent with the NRG-OSHA MOU. Therefore, the perfonnance requirements in §70.61 Include explicit standards for the MOU's first two areas of responsibility. In addition, the third MOU area of responsibility is specifiqally evaJuated by licensees under the ISA requirements of §70.62(c)(1)(iii). As an example of the third MOU area, if the failure of a chemical system adjacent to a nuclear system could affect the safety of the nuclear system such that the radiation dose (and associated likelihood of that accident) exceeded a performance requirement, the chemical system failure 11

would be within the scope of the ISA and the means to prevent the chemical system failure from 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 State agencies acting under section 2021 of title 42, exercise statutory authority to prescribe or enforce standards or regulations affecting occupational safety or health.D [29 U.S.C. §653(b)(1)]

NRC staff subsequently met with OSHA officials on February 25, 1999, and some clarifications and further information were provided at that meeting. As a result of the meeting discussions, some changes were made to the rule language to more cfearty specify the scope of NRC involvement However, these changes do not fully resolve the basic preemption issue.

The problems ide11tified with the rule are not unique, i.e., ~e preemption issue is generic and may already exist for any NRG-licensed facilities where there are requirements to analyze

• hazards. At the February 25 meeting, OSHA confirmed that the rule language IS consistent with the October 21, 1988 MOU; indicated that they have no-suggested changes to the MOU; and indicated that they are not opposed to the proposed rule. The CQmmission's view is that the proposed rule is consistent with NRC responsibilities and authority under the Atomic* Energy Act, and consistent with the OSHA MOU. The only resolution of the preemption issue appears to be a legislative modification of the OSH Act. Public comments would be appreciated on any options that may have been overlooked.

Within each performance requirement, NRC recognizes that the proposed radiological standards are more restrictive, in terms of acute health effects to workers or the public, than the chemical standards for a given consequence (high or intermediate) and that this is consistent with current regulatory practice. The choice of each criterion is discussed below in a paragraph-by-paragraph discussion of §70.61.

12

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 vaJue, 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 that are not expected to occur during the operating life of the facility.

Section 70.61 (b). This section addresses performance requirements for high-consequence events.

The consequences identified in §70.61 (b) of the proposed rule are referred to as "high-consequence events" and include accidental exposure of a worker or an individual located outside of the controlled area to high levels of radiation or hazardous chemicals. These accidents, if they occurred, would represent radiation doses to a worker or an individual located outside of the controlled area at levels with clinically observable biological damage or concentrations of hazardous chemicals- produced from licensed material at which death or life-threatening injury could occur. The goal is to ensure an acceptable level of risk by limiting the combination of the likelihood of occurrence and the identified consequences. Thus, high-consequence events must be sufficiently mitigated to a lower consequence or prevented such that the event is highly unlikely (or lower). The application of "items relied on for safety"

• provides this prevention or mitigation function.

Section 70.61 (b}(1). An acute exposure of a worker to a radiation dose of 1 Sv (100 rem) or greater total effective dose equivalent (TEDE) is considered to be a high-consequence event. According to the National Council on Radiation Protection and Measurements (NCRP, 1971 ), life-saving actions - including the " ... search for and removal of injured persons, or entry to prevent conditions that would probably injure numbers of people" - should be undertaken only when the " ... planned dose to the whole body shall not exceed 100 rems." This is consistent with a later NCRP position (NCRP, 1987) on emergency occupational exposures, that states " ...when the exposure may approach or exceed 1 Gy (100 rad) of low-LET ~inear energy transfer] radiation (or an equivalent high-LET exposure) to a large portion of the body, In a short time, the worker needs to understand not only the potential for acute effects but he or 13

she should also have an appreci~tlon of the substantial increase in his or her lifetime ris~ of cancer."

I Section 70.61 (b)(2). The exposure of an individual located 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 025 Sv (25 rem) is used to determine the dimensions of the exclusion a~ and low-population zone required for siting nuclear power reactors.

Section 70.6Hb}(3). The.intake of 30 mg of soluble uranium by an individual located outside of the controlled area is considered a high- consequence event This choice, which is based on a review of the available literature [Pacific Northwest Laboratories (PNL), 1994], is consistent with the selection of 30 mg of uranium as a criterion that was discussed during the Part 76 rulemaking, "Certification of Gaseous Diffusion Plants." In particular, the final rule that established Part 76 (59-FR 48944; September 23, 1994) stated that "The NRC will consider

• whether the potential consequences of a reasonable spectrum of postulated accident scenarios exceed ... uranlum inta!(es of 30 milligrams...." The final rule also stated that "The Commission's intended use of chemical toxicity considerations in Part 76 Is consist~nt with its practice elsewhere [e.g., 10 CFR 20.1201(e)], and prevents any potential regulatory gap in public protection against toxic effects of soluble uranium."

Section 70.61 (b}(4). An acute chemical exposure to hazardous chemicals produced from licensed material at concentrations that either (1) could cause death or life-threatening injuries to a worker; or (2) could cause irreversible health effects to an individual located outside of the controlled area, is considered a high-consequence event Chemical consequence criteria corresponding to anticipated adverse health effects to humans from acute exposures (i.e., a single exposure or multiple exposures occurring within a short time - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less)' have been developed, or are under development, by a number of organizations. Of particular interest, the National Advisory Committee for Acute Guideline Levels for Hazardous Substances is developing Acute Exposure Guideline Limits (AEGLs) that will eventually cover approximately 400 industrial chemicals and pesticides. The committee, which works under the auspices of 14

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 devt1loped 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 Association 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 engineers, implementers of environmental regulations such as the Superfund Amendment and Reauthorization Act, industrial hygienists, and toxicologists, transportation safety engineers, fire protection specialists, and government agencies ...." {DOE Risk Management Quarterly. 1997).

Despite their general acceptance, there are currently only approximately 80 ERPG criteria available, and some chemicals of importance (e.g., nitric acid) are not covered.

The qualitative language in the performance requirement allows the applicant/licensee to propose and adopt an appropriate standard, which may be an AEGL or ERPG standard, or where there is no AEGL or ERPG value available, the applicant may develop or adopt a criterion that is comparable in severity to those that have been established for other chemicals.

For example, for the worker performance requirement, existing criteria that can be used by licensees to define appropriate concentration levels to satisfy the performance requirement are the AEGL-3 and ERPG-3. AEGL-3 is defined as "The airborne concentration (expressed in ppm or mg/m3) of a substance at or above which it is predicted that the general population, including susceptible, but excluding hypersusceptible, individuals, COL:1ld experience life-threatening effects or death." ERPG-3 is defined as "The maximum ai~ome concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without experiencing or developing life-threatening health effects." Similarly, for the public, AEGL-2 is defined as "The airborne concentration (expressed in ppm or mg/m 3) of a substance at or above which it is predicted that the general population, including susceptible, but excluding hypersusceptible, individuals, could experience Irreversible or other serious, long-lasting effects or impaired ability to escape," and ERPG-2 is defined as "The maximum airborne concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without experiencing or developing irreversible or other health effects or symptoms that could impair an individual's ability to take protective action."

15

Section 70.61 (c). This section addresses performance requirements for Intermediate-consequence events.

The consequences identified in §70.61 (c) of the proposed rule are referred to as "intermediate-consequence events" and include accidental exposure of a worker or an 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 sufficiently mitigated to a lower consequence or prevented such that the event is unlikely (or lower). The application of "items relied on for safety" provides this prevention or mitigation function.

Sectjon 70.61 (c)(1 ). A worker radiation dose between 0.25 Sv {25 rem) and 1 Sv (100 rem) TEDE is conside.red an intermediate-consequence event [over 1 Sv (100 rem) is a high-consequence event]. This value was chosen because of the use of 0.25 Sv {25 rem) as a criterion in existing NRC regulations. For eXB.J; *1-'ie, in 10 CFR 20.2202, '"Notification of incidents," immediate notification is required of ,a licensee If an individual receives"~-- a totaJ effective dose equivalent of 0.25 Sv {25 rem) or more." Also, in 10 CFR 20.1206, "Planned special exposures," a licensee may authorize an adult worker to reC61v'e a dose in exc85$ of normal occupational exposure limits if a dose of this magnitude does not exceed 5 times the annual dose limits p.e., 0.25 Sv {25 rem)] during an individual's lifetime. In addition, EPA's .

Protective Action Guides (U.S. Environmental Protection Agency, 1992) and NRC's regulatory .*

guidance {Regulatory Gulde 8.29, 1996) identify 0.25 Sv {25 rem) as the whole-body dose limit to workers for life-saving actions and protection of large populations. ~CRP has also stated that a TEDE of 0.25 Sv {25 rem) corresponds to the once-In-a-lifetime accidental or emergency dose for workers.

Section 70.61 (c)(2). A dose to any lndMdual located outside of the controlled area between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) is considered an intermediate-consequence event. NRC has used a 0.05-Sv {5-rem) exposure criterion in a number of Its existing regulations. For example, 10 CFR 72.106, "Controlled area of an ISFSI or MRS," states that

• Any indMdual located on or beyond the nearest boundary of the controlled area shall not receive a dose greater than 5 rem to the whole body or any organ from any design basis 16

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 co1.:1ld result in inhalation, Ingestion, or absorption of radioactive materials.

Section 70.61 (c)(3). The release of radioactive material to the environment outside the restricted area in concentrations that, ff averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 5000 times the values specified in Table 2 of Appendix B to Part 20, is considered an intermediate-consequence event. In contrast to the other, consequences criteria that directly protect workers and members of the public, the intent of this criterion is to ensure protection of the environment from the occurrence of accidents at certain facilities authorized to process greater than critical mass quantities of SNM. This implements NRC's responsibility for protecting the environment, in accordance with the Atomic Energy Act of 1954, et seg .. and the National Environmental ,

Policy Act of 1969, et seg.

The value established for the environmental consequence criterion is identical to the NRC A::,,1ormal Occurrence (AO) criterion that addresses the discharge or dispersal of radioactive material from its intended place of confinement (Section 208 of the Energy Reorganization Act of 1974, as amended, requires that AOs be reported to Congress annually) .

• In particular, AO reporting criterion 1.B.1 requires the reporting of an event that involves*~...the release of radioactive material to an unrestricted area in concentratio"!9 which, if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20, unless the licensee has demonstrated compliance with 10 CFR 20.1301 using 10 CFR 20.1302(b)(1) or 10 CFR 20.1302(b)(2)(ii)" [December 19, 1996; 61 FR 67072]. The concentrations listed in Table 2 of Appendix 8 to Part 20 apply to radioactive materials in air and water effluents to unrestricted areas. NRC established these concentrations based on an implicit effective dose equivalent limit of 0.5 mSV/yr (50 mrem/yr) for each medium, assuming an individual were continuously exposed to the listed concentrations present in an unrestricted area for a year.

If an individual were continuously exposed for 1 day to concentrations of radioactive material 5000 times greater than the values listed in Appendix B to Part 20, the projected dose 17

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 until it was properly remediated. Depending on the extent of environmental contamlnati_on 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 rulemaking.

Section 70.61 (c){4). An acute chemical exposure to hazardous chemicaJs produced from licensed material at concentrations that either; a} to a worker, could cause irreversible, health effects (but at concentrations below those which could cause death or life-threatening effects}; orb} to an individual located outside of the controlled area, could cause notable discomfort (but at concentrations below those which could caus*e Irreversible effects}, is considered an intermediate-consequence event Chemical consequence criteria corresponding to anticipated adverse health effects to humans from acute exposures (i.e., a single exposure or multiple exposures occurring within a short time - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less} have been developed, or are under developm,ent, by a number of organizations. Of particular interest, two existing stan~,mis, AEGL-2 and ERPG-2, can be used to define the *concentration leval for irreversible health effects, and two existing standards, AEGL-1 and ERPG-1, can be used to define the concentrati9n level for notable discomfort. The qualitative language in the performance requirement allows the applicant/licensee to adopt and propose an appropriate standard, which may be an AEGL or ERPG standard, or where there is no AEGL or ERPG va.lu~ available, the applicant may develop or adopt a criterion that is comparable'I in severity to ~those that have been established for other chemicals.

Section 70.61 {d). This section addresses performance requirements for an accidental nuclear criticality.

The third performance requirement states that the risk of nuclear criticaJity accidents must be limited by assuring that under normal and credible abnormal conditions, all nuclear processes are subcrltlcal, including use of an approved margin of subcriticaJity for safety. It also requires that preventive controls and measures shall be the primary means of protection against nuclear criticality accidents. Although detecting and mitigating the consequences of a 18

nuclear criticality are important objectives (e.g., for establishing aJarm 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 goaJ of "... no accidental criticality invoMng 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 Satety in Operations with Fissionable Materials Outside Reactors."

Section 70.61 {e). This section addresses items relied on for safety and management measures.

Paragraph 70.61 (e) would require that each engineered or administrative control or control system that is needed to meet the performance requirements be designated as an item relied on for safety. This means that any control or control system that is necessary to maintain the acceptable combination of consequence and likelihood for an accident is designated an item relied on for safety. The importance of this section is that, once a control is designated as an item relied on for safety, It falls into the envelope of the safety program required by section 70.62'. For example, records will be kept regarding the item, and management measures such as the configuration control program are applied to the item and to changes that affect the item, to ensure* that the item will be available and reliable to perform its function when needed.

The failure of an item relied on for safety does not necessarily mean that an accident will occur which will cause one of the consequences listed in the performance requirements to be

• exceeded. Some control systems may have parallel (redundant or diverse) control systems that would continue* to prevent the accident The need for such defense-in:-depth and single-failure I

resistance would ideally be based on the severity and likelihood of the µ9tential accident. . In other cases, the failure of an item may mean that the particular accident sequence is no longer i,ighly unlikely", or "unlikely." In these cases, the performance requirement is not met, and the expectation would be that a management measure would exist (possibly in the form of an operating procedure) that ensured that the facility would not operate In a condition that exceeds the performance requirement. For example, a facility that relies on emergency power could not*

operate for an extended time in the absence of an emergency power source even if grid power is available. In this manner, the items relied on for safety and the management measures complement each other to ensure adequate protection from accidents at any given time.

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Section 70.61 m. This section addresses 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 activrties 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. Control includes the power to exclude individuals, 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 §20.1003), access to which is controlled by the licensee for purposes of radiation safety.]

Individuals who do not receive an occupational dose (as that tefm is used in Part 20) in the controlled area will be subject to the dose limits for members of the public in 10 CFR 20.1301. However, the Commission recognizes that certain licensees may have ongoing activities at their site (i.e., within the controlled area) that are not related to the licensed activities. For example, a non-nuclear facility may be adjacent to the nuclear facility but both are within the controlled area (which may be defined similar to the site boundary). This raises a question regarding the appropriate accident standard for these individuals. Protection of the individuals at the non-nuclear facility must consider that the nature of many potential accidents at a fuel cycle facility, is such that there may not be sufficient time during which to take action to exclude individuals from the controlled area. Therefore, for purposes of the ISA accident evaluation, the rule explicitly contains two options for these individuals (as well as an implicit third option). In the first option, the licensee evaluates, in the ISA, the risk at its location (as opposed to that at any point at or beyond the controlled area boundary) and determines that it meets the performance requirements for members of the public. In the second option, performance requirements for workers may be applied to individuals in the controlled area if the provisions of Section 70.61 (f)(2) are satisfied. These conditions ensure that the individuals are aware of the risks to them from the potential accidents at the nuclear facility and have received appropriate training and access to information. This parallels and is consistent with the use of the term, "Exclusion area", by 10 CFR Parts 50 and 100, which states; "Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result." The Implied third option is to define (or redefine) a controlled area such that within it only activities associated with the licensed nuclear facility are permitted.

20

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). Use of the ISA to detem'line the risks to these individuals would need to consider second-order effects such as the probability of the indMdual being present at the time that the unlikely (or highly unlikely) accident occurred. This level of detail is unnecessary to 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 degree Of protection to the infrequently present individuaJs.

Section 70.62 Safety Program and Integrated Safety Analysis.

This paragraph addresses the safety program, that includes process safety information, ISA, and management measures. The performance of an ISA, and the establishment of measures to ensure the availability and reliability of items relied on for safety when needed, are the means by which licensees demonstrate an adequate level of protection at their facilities.

The ISA is a systematic analysis to identify plant and external hazards and their potential for initiating accident sequences; the potential accident sequences and their consequences; and the site; structures, systems, equipment, components, and activities of personnel relied on for safety. As used here, "integrated* means joint consideration of, and protection from, all relevant hazards, including radiological, criticality, fire, and chemical. The structure of the

• safety program recognizes the critical role that the ISA plays in identifying potential accidents and the items relied on for safety. However, it also recognizes that the performance of the ISA, by itself, will not ensure adequate protection. Instead, an effective man~gement system is needed to ensure that the items relied on for safety are available and reliable to perform their function when needed. Detail~ requirements for each part of the safety program are included in this section.

Section 70.62(a). Each licensee would be required to establish and maintain a safety program that demonstrates compliance with the performance requirements of §70.61. Although the ISA would be the primary tool in identifying the potential accidents requiring consequence mitigation and accident prevention, process safety information would be used to develop the ISA, and management measures would be used to ensure the availability and reliability of items 21

relied on for safety identified through the ISA. The management measures may be graded according 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.

Finally, each licensee would also be required to establish and maintain a log documenting each discovery that an item relied on for safety has failed to perform its function either i!l the context of the performance requirements of §70.61 or on demand. The phrase

" .. .in the context of the performance requirements of §70.61

• means that items relied on for safety that fail would require logging even if their failures did not result in process upsets or accidents but could have resulted in the accident conditions they are protecting against, had all conditions been optimum for the accident This would not include failures during times, such as routine maintenance on an item, when the item or measure was clearly documented to not be available. The log must contain: (a) the identity of the item that failed and the safety function affected; (b) date of discovery of the failure; (c) duration of time that the item was unable to perform its function; (d) any other affected items relied on for safety and their safety function; (e) affected processes; (f) the cause of the failure; (g) whether the failure was in the context of performance requirements, or on demand, or both; and (h) any corrective or compensatory actions taken. The log should be initiated at the time of discovery and _updated promptly at the completion of each investigation of a failure of an item relied on for safe~. The purpose of the log is to assist NRC in determining whether items relied on for safety are, in fact, available and reliable and in detecting system problems that may impact ISA evaluations.

Section 70.62(b). This paragraph would require the licensee to maintain process-safety information pertaining to the hazards of the materials used or produced in the process, the technology of the process, and the equipr,:ient in the process. NRC confidence in the margin of safety at its licensed facilities depends, in part, on the ability of licensees to maintain a set of current, accu~te, and complete records available for NRC inspection. The process-safety information should be used in support of development of an ISA.

22

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 plant, or by credible external events, including natural phenomena such as floods, earthquakes, etc. To accomplish the task of identifying potential accidents, the licensee needs to ensure that detailed and accurate information about plant processes is maintained and made available to the personnel performing the ISA. <

(3) Determine the consequences of each accident that has been identified. For an accident with consequences at a "high" or "intermecfiate level," as defined in 10 CFR 70.6l,.the likelihood of such an accident must be shown to be commensurate with the consequences, as required. in 1o CFR 70.61.

(4) Identify the items relied on for safety (i.e., those Items that are relied on to prevent accidents or to mitigate their consequences, identified in the ISA). These items are needed to reduce the consequences or likelihood of the accidents to acceptable levels. The identification of items relied on for safety is required only for accidents with consequences at a high or intermediate level, as defined In 10 CFR 70.61.

It Is expected that the licensee or applicant would perform the ISA using a "team" of Individuals with expertise in engineering and process operations related ~o the system being evaluated; the team should indude*persons with experience in nuclear criticality safety, radiation safety, fire safety, and chemical process safety, as warranted by the materials and potential hazards associated with the process being evaluated. At leaet one member of the JSA team should be an lndMdual who has experience and knowledge that is ~clflc to the process being evaluated. Finally, at least one lncftvidual in the team must be knowledgeable in the specific ISA methodology being used.

Current Part 70 licensees, for whom the rule applies, would be required to develop ,

plans and submit them to NRC within 6 months of the effective d~te of the rule. Each plan would identify the processes that would be subject to an ISA, the ISA approach that would be 23

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 contain~ 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 would be expected to design their facilities or processes to protect against the occurrence of the adverse consequences identified in 10 CFR 70.61, using the baseline design criteria 10 CFR 70.64(a). Before operation, applicants would be expected to update their ISAs, based on as-built conditions and submit the results to NRC as ISA summaries, along with the applications, following the requirements in 10 CFR 70.65(b).

The Commission believes that sufficient flexibility is permitted in the ISA methodology chosen to be able to accommodate a*wide range of technologies. However, to assure that sufficient flexibility exists, the Commission Is h::i'-luesting comments on this matter.

Section 70.62(d). Although the ISA would play a critical role in identifying potential accidents and the Items relied oh for safety, the periormance of an l.'.:,A would not, by itself, ensure adequate protection. In addition, as would be provided for in 10 CFR 70.62(d), an effective management system would be needed to ensure that the items relied on for safety are available and reliable to perform their function when needed. As stated before, management measures may be graded to better implement the results of the ISA. .

Management measures are functions performed by the licensee, in general on a continuing basis, that are applied to items relied on for safety. Management measures include:

a) configuration management; b) maintenance; c) training and qualifications; d) procedures; e) audits and assessments; f) incident Investigations; g) records management; and h) other quality assurance elements. Changes in the configuration of the facility need to be carefully controlled to ensure consistency among .

the

. facility design and operational requirements, the physical configuration, and the facility documentation. Maintenance measures must be in place to ensure the availability and reliability of all hardware, identified as items relied on for safety, to perform their function when needed. Training measures must be established to ensure that all personnel relied on for safety are appropriately trained to perform their saf~ty functions.

24

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 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 the performance requirements of this section. The utility of the two modifying requirements, "When neededt and "in the context of the performance requirements of this section," is clarified as follows:

The phrase "when needed" is used to acknowledge that a particular safety control need not be continuously functioning. For example, it may not be operational during maintenance or calibration testing, or may not be required when the process 1s not operational or when special nuclear material is not present. However, the phrase, when needed, does not relieve a licensee from compliance with the performance requirements. For example, if a particular component is out for maintenance, the licensee must consider credible event sequences in developing the ISA and identifying items relied on for safety- a high-consequence event sequence still has to be highly unlikely. Compliance with the performance requirements in these cases can be established by various means including identification of additional items relied on for safety (and

• application of safety program management measures to them), or by limiting operations or placing the plant in a different operating mode during the maintenance_ of the item relied on for safety.

To illustrate, a loss of offsite power during a one-week maintenance outage of the emergency diesel generator that is relied on for safety would still be a credible event sequence.

If the loss of power, combined with the generator's inoperable status, could result in a combination of dose and likelihood that exceeds a performance requirement, then the licensee would not be in compliance with the performance requirements of §70.61. A licensee cannot

• claim, after the maintenance, that since the power was not lost, the generator was available when needed. The concept is that the ISA is used as a risk-informed, forward-look at the credible facility hazards and their effects on plant systems and modes of operation. The rule would require that each item necessary to comply ,with the performance requirements be 25

Identified as important to safety and placed under the safety program management controls. In I

identifying each item, the ISA must consider various modes of operation and the likelihood that 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 I

in cases where one control does not result in compliance with the performance requirement or has periods when it is inoperable. Using the loss of offsite power example again, a licensee would still be required to meet the risk-informed performance requirements of the -rule when an emergency diesel generator used as an item relied on for safety is not operable or out of service for maintenance.

Section 70.64 Requirements for new facilities or new processes at existing facilities.

This section deals with baseline design criteria for new facilities or new processes at existing facilities.

A major feature of the proposed amendments to Part 70 is the requirement that licenboeS and applicants for a license perform an ISA and use the ISA proce&; to develop risk-informed decisions regarding facility safety. The ISA process Is applied to existing designs to identify risk insights on those areas that warrant additional preventive or mitigative measures.

For new facflities, the proposed rule would require the performance of the ISA before construction [see the existing §70.21 (f) and §70.23(a)(7)], and the u~ting of the ISA before beginning operations. For new processes and facilities, the Commisslo~ recognizes that good engineering practice dictates that certain minimum requirements be applied as design and safety considerations for any new nuclear process or facility. In addition, a fundamental element of NRC's safety philosophy Is that designs and operations should provide for defense-in-depth protection against accidents. -Therefore, the Commission has specified baseline design criteria in §70.64 that are similar In use to the general design criteria in Part 50 Appendix A; Part 72, Subpart F; and 10 CFR 60.131. The baseline design criteria identify 10 initial safety design considerations, including: a) quality standards and records; b) natural phenomena 26

• hazards; c) fire protection; d) environmental and dynamic effects2; 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 §70.61. The baseline design criteria are generally an acceptable set of inltial 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 these processes and facilities, any design features that are inconsistent with the baseline design criteria should be identified and justified.

Using the baseline design criteria and considering defense-in-depth practices In the design should result in a new facility design that is based on providing successive levels of protection such that health and safety will not be wholly dependent on any single element of the design, construction, maintenance, or ~ration of the facility. The net effect of incorporating defense-in-depth practices is a conservatively designed facility and system that will exhibit greater tolerance tor failures and external challenges, The risk insights obtained through .

performance of the ISA can be. then used to supplement the final design by focusing attention on the prevention and mitigation of the potential accidents having higher-risk.

. Section 70.65 Additional content of appUcations.

In addition to the information that currently must be submitted to NRC, under §70:.22, for a license application, this section requires additional information to be_ submitted to demonstrate compliance with the proposed new subpart. In particular, this additionaJ information would need to include a description of the applicant's saf~ty program established under §70.62, a description of the management measures, and an ISA summary.

The ISA summary would contain: a) a description of the site and the facility; b) a description of the team qualifications and ISA methodology; c) the processes analyzed in the ISA and the maximum consequences of each; d) a demonstration of how the licensee meets 2

Environmental and dynamic effects are effects that could be caused by ambient conditions. For example, an item relied on for safety will need to function within its expected environment (i.e., imder normal operating condltlons, expected accident conditions, etc.). These conditions could include high temperatures, or a corrosive environment It could also Include dynamic changes in surrounding concfrtions caused by an accident (e.g., the bursting of a high-pressure pipe).

27

the requirements for criticality monitoring and alanns In §70.24; e) a demonstration of how the licensee meets the performance requirements of §70.61 and, if applicable, §70.64; f) a list of Items 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

§70.22, there may be information about the facility which would be too detailed for inclusion in the general site description, but would be needed to be included here to understand the ISA and ISA results. The demonstration of how the licensee meets §§70.24, 70.61, and 70.64 is a critical element in determining whether the applicant understands and complies with the I

regulations and can operate the facility safely. Another critical element Is the applicant's identification of the items relied on for safety. Through the ISA process, the applicant should have identified potential accidents that can occur in individual processes and In the facility as a whole. As discussed earlier, these accidents are prevented or their consequences mitigated using controls that are Identified in the ISA summary as items relied on for safety.. It is important for NRC staff to review the items relied on for safety, that were identified as such by the applicant or licensee, to determine whether potential accidents are adequately prevented or mitigated. Since items relied on for safety play a key role In assuring that the performance requirements are met, and because the applicant has .great flexibility in selecting and identifying what the actual "items* are (as discussed in relation to §70.61), the-items relied on for safety would be clearly and unambiguously Identified on a list This list of ite_ms is then managed and controlled by the applicant through the management measures in §70.61 to ensure*that they continue to perform the safety function required. By evaluating the JSA methodology, and the ISA summary, supplemented by reviewing the ISA and other information, as needed, at the licensee's facility, the staff can better understand the potential hazards at the facility, how the applicant plans to address these hazards, and thereby have confidence in the safety basis on which the license will be issued.

The ISA summary would be required to be submitted on the docket in conjunction with the license application but would not be considered part of the license. The ISA, on which the ISA summary is based, would be maintained current at the licensee's facility and available for NRC review, but it would not be submitted and docketed. The information and commitments 28

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 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 ISAsummary, 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 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 ~70.72.

Section 70.66 Additional regujrements for the approval of lice~ applications.

In addition to the requirements found in the existing rule (i.e., 1o CFR 70.23), the

• Commission must determine that the requirements in the new subpart, 10 CFR 70.60 through 70.66, will be satisfied.

Section 70.72 Facility changes and change process.

This section deals with changes to site, structures, systems, equipment, components, and activities of personnel after a license application has been approved.

Past incidents at fuel PYCfe facilities have often resulted from changes norfully analyzed,

• not authorized by licensee management, or not ~deq~tely understood by facility personnel.

Therefore, effective control of changes to a facility's site, structures, systems, equipment, -

components, and activities of personnel is a key element In assuring safety at that facility. This section would require the licensee to establish and use a system to eyaluate changes and the potential impacts of those changes before implementing-them. By using this system to evaluate, implement and track changes to the facility, the licensee can make certain changes without NRC pre-approval. If the change affects information contained in the ISA summary, the licensee would be required to notify NRC within 90 days of the change by submitting updated ISA summary pages in that time. For changes that affect the on-site documentation, such as the ISA, management measures or process-safety information, the licensee would be required to notify NRC within 12 months of the change. This update frequency would allow NRC staff to review the ct,anges being made to the facility In enough time to ensure that the licensee's evaluations of potentlaJ impacts to health and safety were accurate. It also allows NRC staff to maintain relatively current facility and safety information on the docket at all times. In addition, 29

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. The Commission is particularly interested in comments concerning the 90 day time period for submitting updated ISA summary p~ges that reflect changes to a facility's site, structures, systems, equipment, components, and activities of personnel.

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 on heaJth 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. .

The advantages ,of Option 1 are that it focuses on the most signtflcant changes to the facility and is equivalent to looking at the highest risk*changes. It contains very little subjective criteria and is therefore easier to implement and inspect. It also would likely only result in a few license amendments a year which is generally consistent with the past practice at these facilities. Since Option 1 would permit more changes without NRC pre-approval, a relatively short timeframe (90 days) for submitting updated ISA summary pages is required in order for I

NRC to have information that reflects the current status of the facility and to be confident that adequate protection is still provided with the changes, as reflected in the ISA summary. The advanta.ges of Option 2 are that NRC would have more control over the changes at the

. facilities, i.e., staff expects that more charmes would be reviewed by ~e staff before being implemented; thus, it would be less likely that NRC would have a concern with a change after the fact; and it is consistent with the change control process at power reactors, where changes are reported only after 12 months.

The proposed rule language reflects Option 1.

Section 70.73 Renewal of licenses.

Under the proposed amendments to Part 70, changes to site, structures, systems, equipment, components, .and actMtles of personnel made by the licensee pursuant to §70.72 would be. documented on a continuing basis on-site. A description of those changes would also be sent to NRC periodically. This process is intended to keep the documents, which support 30

the license, current and thereby establish a "living" license. In the past, the license renewal process was burdensome to NRC and the licensee because all changes made to the facility since the last license renewal would be reviewed at one time. However, with the proposed "living license," 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 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.

Section 70.74 Additional reporting requirements.

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 consideration of the risk and consequences established in 10 CFR 70.61(b) is intended to replace and expand on the .approach licensees have currently been using for reporting criticality events under Bulletin 91-01. The new approach would cover all types of events, not just criticality events, and establish a timeframe for reporting that is .x.-a.led according to risk. The new reporting requirements are intended to supplement the requirements in the existing Parts 20 and 70 and elsewhere in the regulations. A more detailed discussion of the new requirements is found in the following discussion of Appendix A to Part 70.

Appendix A Reportable Events .

• The reporting of events supports NRC's need to be aware of conditions that could result in an imminent danger to the worker or to public health and safety or_to the environment In particular, NRC needs to be aware of licensee efforts to address potential emergencies.

Further, once safe conditions have been restored after an event, NRC has an interest in disseminating information on the event to the nuclear industry and other interested parties, to reduce the likelihood that the event will occur in the future. Also, in the event of an accident, NRC must be able to respond accurately to requests for information by the public and the media. Finally, NRC must evaluate the performance of individual licensees and the industry as a whole to fulfill its statutory mandate to protect the health and safety of the worker and the public and the environment.

Licensee reporting of events would consist of two reporting classes based on the hazard

- reports that must be made in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and those to be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. According to 31

this approach, licensees would report events based on two criteria: 1} whether actual consequences have occurred or whether a potential for such consequences exists; and 2) the seriousness of the consequences. The events that must be reported within the shortest timeframe (1 hour} are high-consequence events. These events encompass unintended criticalities and loss of criticality controls, and loss of chemical controls or the occurrence of chemical exposures that exceed the performance requirements in §70.61 (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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These include chemical exposure to licensed material or hazardous chemicals that exceed the lower threshold limits in §70.61 (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, or fire, either internal or external to the plant, that aff~cted or could have affected a facility, must be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This reporting requirement would capture, for example, a tornado that strikes a facility, an earthquake motion experienced by a facility, or any type of fire. Since these events could have affected a facility, NRC would want to know about such events to assess a licensee's conclusion of whether any detrimental effects did in fact occur, or could have occurred in the absence of controls that were present but not part of the safety basis.

Anothe; category of potential events that would be reported Is one that involves the existence of an unsafe condition that is not identified in the ISA. This condition could be caused by a deviation from established safe operating conditions, by an unanticipated and unanalyzed set of ci~mstances, or by an improper analysis. This type of event would be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The proposed rule also would require concurrent reporting of events when a news release is made or if other Government agencies are notified, as is done under 10 CFR Part 50.72, to support NRC's ability to be responsive to questions concerning the safety of NRC-licensed facilities.

32

REFERENCES Graig, D.K., et al.,, "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 Radiation Protection Criteria," NCRP Report No. 39, Washington, DC, 1971.

National Council on Radiation Protection and Measurements (NCRP),

"Recommendations on Limits for Exposure to Ionizing Radiation," NCRP Report No. 91, Washington, DC, 1987.

U.S. Nuclear Regulatory Commission, "ProposE:Jd Methods for Regulating Major Materials Licensees," NUREG-1324, Washington, DC, February 1992.

U.S. Nuclear Regulatory Commission/ Occupational Safety and Healt;1 Administration (OSHA), "Memorandum of Understanding Between NRC and OSHA; Worker Protection at NRG-Licensed Facilities" (53 FR 43950; October 31, 1988).

U.S. Nuclear Regulatory Commission, "Certification of Gaseou_s Diffusion Plants" (59 FR 48944; September 23, 1994).

U.S. Nuclear Regulatory Commission, "Abnormal Occurrence Reports: Implementation of Section 208 of Energy Reorganization Act of 1974" (61 FR 67072; December 19, 1996).

U.S. Nuclear Regulatory Commission, "Site Decommissioning Management Plan,"

NUREG-1444, Washington, DC, October 1993.

33

U.S. Nuclear Regulatory Commission, "Strategic Plan, Fiscal Year 1997 - Fiscal Year 2002," NUREG-1614, Washington, DC, September 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 Gulde 829, Rev. 1, February 1996.

Theide, L, "Emergency Information Where It's Needed," DOE Risk Management Quarterly, Vol 5, No 2, Richland, WA, May 1997.

These documents are available for inspection and copying for a fee at the NRC Public Document Room, 2120 L Street, N.W. (Lower Level), Washington DC 20555-0001.

Copies of NUREG-1324, NUREG-1614, and NUREG-1444 may also be purchased from the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington DC 20402-9328. Copies are also available from the National 7echnical

-Information Service, 5285 Port Royal Road, Springfield VA 22161.

, Regulatory Guide 829 may be purchased from the Government Printing Office (GPO) at the current GPO price. Information on current GPO prices may be obtained by contacting the Superintendent of Documents, U.S. Government Printing Office, ~ .0. Box 37082, Washington DC 20402-9328. Issued guides may also be purchased from the National Technical Information Service on a standing-order basis. Details on this service may be obtained by writing NTIS, 5285 Port Royal Road, Springfield, VA 22161.

Copies of the following draft regulatory guidance documents may be requested by writing to U.S. Nuclear Regulatory Commission, Reproduction and Distribution Services, Washington, DC 20555-0001: "Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility" (Draft NUREG-1520); and "Integrated Safety Analysis Guidance Document" (Draft NUREG-1513).

34

Plain Language The Presidential Memorandum dated June 1, 1998, entitled "Plain Language in Government Writing," directed that the Federal government's writing be in plain language. The NRC requests comments on this proposed rule specifically with respect to the clarity and effectiveness of the language used. Comments should be sent to the address listed above.

Finding of No Significant Environmental Impact Availability 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 accidents at the facility and the ite,,1s relied on for safety; (3) require the implementation of measures to ensure that the items relied on for safety are available and reliable to perform their function when needed; and (4) require the inclusion of the safety bases, as reflected in the ISA summary, in

• the license application. The language, In the proposed rule, that defines an environmental consequence of concern, is relevant to the question of environmental jmpact. Licensees would be required to provide an adequate level of protection against a " ... release of radioactive material to the environment outside the restricted area in concentrations that, if averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20."

Implementation of the new amendments, including the requirement to protect against events that could damage the environment, is expected to result in a significant improvement in licensees' (and NRC's) understanding of the risks at their facilities and their ability to ensure that those risks are acceptable. For existing licensees, any deficiencies identified in the ISA would need to be promptly addressed. For new licensees, operations would not begin unless licensees demonstrated an adequate level of protection against potential accidents identified in the ISA. As a result, the safety and environmental impact of the new amendments is positive.

35

There will be less adverse impact on the environment from operations carried_out In accordance with the proposed rule than if those operations were carried out in accordance with the existing Part 70 regulation.

The determination of this Environmental Assessment is that there will be no significant offsite impact on the public from this action. Howev.er, the general public should note that NRC welcomes public participation. NRC has also committed to complying with Executive Order (EO) 12898, "Federal Actions to Address Environmental Justice in Minority Populations and Low-Income Populations," dated February 11, 1994, in all its actions. Therefore, NRC has also determined that there are no disproportionate, high, and adverse impacts on minority and low-income populations. In the letter and spirit of EO 12898, NRC is requesting public comment on any environmental justice considerations or questions that the public thinks may be related to this proposed rule, but somehow*were not addressed. Comments on any aspect of the Environmental Assessment, including environmental justi~. 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 EnvironmentaJ Assessment The Enyironmental 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 the environmental ~ment are available from Barry Mendelsohn, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone (301) 415-7262; e-mail: btm1@nrc.gov. ,1 Paperwork Reduction Act Statement This proposed rule amends information collection .requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501. et seq.). This. rule has been submitted to the Office of Management and Budget (0MB) for review and approval of the paperwork requirements.

The public reporting burden for this information collection is estimated to average 99 hours0.00115 days <br />0.0275 hours <br />1.636905e-4 weeks <br />3.76695e-5 months <br /> per response, and the recordkeeping burden is estimated to average 560 hours0.00648 days <br />0.156 hours <br />9.259259e-4 weeks <br />2.1308e-4 months <br /> per licensee, including the time for reviewing Instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information r

36

collection. NRC is seeking public comment on the potential impact of the information collections contained in the proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper performance of NRC's function? Will the information have practical utility?

2. Is the burden estimate accurate?

3. Is there a way to enhance the quality, utility, and clarity of the information to be collected?

4. How can the burden of the information collection be minimized, including the use of automated collection techniques?

Send comments on any aspect of this proposed informatior:;i 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 0MB on the information collections or on the above issues should be submitted by (insert 30 days after publication in the Federal Register). Comments received 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.

Public Protection Notification

• If a means used to impose an information collection does not display a currently valid 0MB control number, the NRC may not conduct nor sponsor, and a person is not required to respond to, the information collection.

Regulatory Analysis The Commission has prepared a draft Regulatory Analysis on this proposed regulation.

The analysis examines the benefits and costs of the alternatives considered by the Commission. The draft Regulatory Analysis is available for inspection in the ~RC Public Document Room, 2120 L Street N.W. (Lower Level), Washington, D.C. and the Part 70 website. Single copies of the analysis may be obtained from Barry T. Mendelsohn, Office of 37

Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, telephone (301) 415- 7262, e-mail: btm1@nrc.gov.

The Commission requests public comment on the draft Regulatory Analysis. Comments on the draft analysis may be submitted to NRC as indicated under the ADDRESSES heading.

Regulatory Flexibility Certification As required by the Regulatory Flexibility Act, as amended, 5 U.S.C. 605(b), the Commission certifies that this proposed rule, if adopted, would not have a significant economic impact on a substantial number of small entities. This proposed rule would affect facilities that are authorized to possess a critical mass of SNM and who are engaged in one of the following activities: a) enriched uranium processing; b) fabrication of uranium fuel or fuel assemblies; c) uranium enrichment; 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 invoMng 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 the size standards published by NRC (10 CFR 2.810).

Voluntary Consensus Standards The National Technology Transfer Act of 1995, Pub. L. 104-113, requires that Federal Agencies use technical standards that are developed or adopted by voluntary consensus standards bodies unless the use of such a standard is inconsistent with applicable law or otherwise impractical. In this proposed rule, the NRG proposes to use the following voluntary consensus standard, ANSI/ANS Standard 8.1-1983, "Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors," developed by the American Nuclear Society.

Portions of the standard were used in the definition of double contingency and in §70.61 (d).

The NRC invites comment on the applicability and use of other standards.

Backfit Analysis

• 38

NRC has determined that the backfit rule does not apply to this proposed rule; therefore, a backflt arialysis is not required for this proposed rule because these amendments do not involve any provisions that would impose backfits as defined in 10 CFR Chapter I.

List of Subjects in 10 CFR Part 70 Criminal penalties, Hazardous materials transportation, Material control and accounting, Nuclear materials, Packaging and containers, Radiation protection, Reporting and recordkeeping requirements, Scientific equipment, Security measures, Special nuclear material.

For the reasons set out in the preamble and under the authority of the Atomic Energy 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

• 39

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).

Sections 70.1 (c) and 70.20a(b) also issued under secs. 135, 141, Pub. L.97-425, 96 Stat.

2232, 2241 (42 U.S.C. 10155, 10161). Section 70.7 also issued under Pub. L 95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Section 70.21 (g) also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Section 70.31 also issued under sec. 57d, Pub. L 93-377, 88 Stat. 475 (42 U.S.C.

2077). Sections 70.36 and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C.

2234). Section 70.61 also issued under secs. 186, 187, 68 Stat. 955 (42 U.S.C. 2236, 2237).

Section 70.62 also issued under sec. 108, 68 Stat 939, as amended (42 U.S.C. 2138}.

2. The undesignated center heading "GENERAL PROVISIONS" is redesignated as "Subpart A - General Provisions."

3. In §70.4, the definitions of Acute, Available and reliable to perform their function when needed, Configuration Management, Critical mass of special nuclear material, Double contingency, Hazardous chemicals produced from licensed material, Integrated safety analysis (ISA), Integrated safety analysis summary, Items relied on for safety, Management _measures, Unacceptable performance deficiencies, and Worker are added, in alphabetical order, as follows:

§70.4 Definitjons.

Acute as used in this Part means a single radiation dose or chemical exposure event or multiple radiation dose or chemical exposure events occurring within a short time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less}.

Available and reliable to perform their function when needed as used in Subpart H of this Part means that, based upon the analyzed, credible conditions in the integrated safety analysis, 40

items relied on for safety will perform their intended safety function when needed and management measures will be implemented that ensure continuous compliance with the performance requirements of §70.61 of this Part, considering factors such as necessary mainten~ce, operating limits, common cause failures, and the likelihood and consequences of failure or degradation of l

the items and measures.

Configuration management (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 when needed.

CriticaJ mass of special nuclear material (SNM) means special nuclear material in a quantity exceeding 700 grams of contained uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1500 grams of contained uranium-235, if no uranium enriched to more than 4 percent by weight of *uranium-235 is present,. 450 grams of any combination ther~of; or one-half such quantities if massive moderators Qr reflectors made of g1aphite, heavy water, or beryllium may be present.

• * '

• r * *

• Double contingency means a process de3ign that incorporates sufficie, if factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.

Hazardous chemicals produced from licensed materials means ~ubstances having licensed material as precursor compound(s) or substances that physically or* chemically interact with licensed materials; that are toxic, explosive, flammable, corrosive, or reactive to the extent that they can endanger life or health if not adequately controlled. These include substances commingled with licensed material, and include substances such as hydrogen fluoride that is produced by the reaction of uranium hexafluoride and water, but do not include substances prior to process addition to licensed material or after process separation from licensed material.

Integrated safety analysis (ISA) means a systematic analysis to identify plant and external hazards and their potential for initiating accident sequences, the potential accident sequences, their likelihood and consequences, and the items relied on for safety. As used here, integrated means 41

joint consideration of, and protection from, all relevant hazards, including radiological, nuclear criticality, fire, and chemical However, with respect to compliance with the regulations of this Part, the NRC requirement is 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 materiaJ.

Integrated safety analysis summary means the document submitted with the license application, license amendment application, or license renewal application that provides a synopsis of the results of the integrated safety analysis and contains the information specified in §70.65(b).

I Items relied on for safety means structures, systems, equipment, components, and activities of personnel that are relied on to prevent potential accidents at a fac{lity that could exceed the performance requirements in §70.61 or to mltigate.their'potential consequences. This does not limit the licensee from identifying additional structures, systems, equipment, components, or activities of personnel (i.e., beyond those in th~ minimum set necessary for compliance with the performance requirements) as items relied on for safety.

Management measures mean the functions performed by the licensee, generally on a continuing basis, that are applied to items relied upon for safety, to e~ure tt.-3 items are available and reliable to perform their functions when needed. Management measures include configuration management, maintenance, training and qualifications, procedures, audits *and assessments, incident investigations, records management, and other quality assurance elements.

Unacceptable performance deficiencies mean deficiencies itJ the _items relied on for safety or the management measures that need to be corrected to ensure an adequate level of protection as defined in 10 CFR 70.61(b), (c), or (d).

Worker means an indMdual whose assigned duties in the course of employment involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation (i.e., an indMdual who is subject to an occupational dose as in 20 CFR 20.1003) .

. 4. In §70.8 paragraph (b) is revised to read as follows.

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§70.8 Information collection requirements: 0MB approval.

(b) The approved information collection requirements contained in this part appear in

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§§ 70.9, 70.17, 70.19, 70.20a, 70.20b, 70.21, 70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, 70.39, 70.42, 70.50, 70.51, 70.52, 70.53, 70.57, 70.58, 70.59, 70.61, 70.62, 70.64, 70.65, 70.72, 70.73, 70.74 and Appendix A.

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

Exemptions."

§§ 70.13a and 70.14 [Redesignated}

6. Sections 70.13a and 70.14 are redesignated as§§ 70.14 and 70.17, respectively.

7. The undesignated center heading "GENERAL LICENSES- is redesignated as "Subpart C - General Licenses."

8. The undesignated center heading "LICENSE APPLICATIONS" is redesignated as "Subpart D - License Applications."

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

Licenses."

• 10. The undeslgnated center heading "ACQUISITION, USE AND TRANSFER OF SPECIAL NUCLEAR MATERIAL, CREDITORS' RIGHTS," Is redesignat~ as "Suf?p~rt F-Acquisition, Use, and Transfer of Special Nuclear Material, Creditors' Rights."

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11. The undesignated center heading "SPECIAL NUCLEAR MATERIAL CONTROL RECORDS, REPORTS AND INSPECTIONS" is redesignated as "Subpart G - Special Nuclear Material Control Records, Reports, and Inspections:

12.. In §70.50 paragraph (c) is revised to read as follows.

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§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 Center. To the extent that the information is available at the time of notification, the information provided in these reports must include:

O} Caller's name, position title and call back telephone number;

{ii) Date, time, and exact location of the event; (iii) Description of the event, including;

{A) Radiological or chemical hazards involved including isotopes, quantities, and chemical and physical form of any material released; (8) Actual or potential health and safety consequences to the workers, the public, and the environment, including relevant chemical and radiation data for actual personnel exposures to radiation or radioactive materials or chemicals (e.g., level of radiation exposure, concentration of chemicals, and duration of exposure);

(C) The sequence of occurrences leading to the event, including degradation or failure of structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences; and (D) Whether the remaining structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences are available and reliable to perform their function.

(iv) External conditions affecting the event;

{v) ~ditional actions taken by the licensee in response to the event; (vi) Status of the event (e.g., whether the event is on-going or was terminated);

(vii). Current and planned site status, including any declared emergency class; (viiQ Notifications related to the event that were made or are planned to any local, State, or other Federal agencies; (ix) Status of any press releases related to the event that were made or are planned.

3 The commercial telephone number for the NRC Operations Center is (301) 816-5100.

44

(2) Written report. Each licensee who makes a report required by paragraph (a) or (b) of this section, or by §70.74 and Appendix 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 aJI 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 10CFR Part 20. The reports must include the following:

(i) Complete applicable information required by §70.SO(c}(1};

(ii) The probable cause of the event, including aJI factors that contributed to the event and the manufacturer and model number (If applicable) of any equipment that failed or malfunctioned; (iii) Corrective actions taken or planned to prevent occurrence of similar or identical events In the future and the results of any evaluations or assessments; and (iv) For licensees subject to Subpart H of this Part, whether the event was identified and evaluated in the Integrated Safety Analysis.

(d) The provisions of §70.50 do not apply to l;censees subject to §50.72. They do apply to those Part 50 licensees possessing material licensed under Part 70 who are not subject to the notification requirements in §50. 72.

13. The undeslgnated center heading "MODIFICATION AND REVOCATION OF LICENSEf is redesignated as "Subpart I - Modifi~tion and Revocation of Licenses."

§§ 70.61 and 70.62 [Redesignated]

14. Sections 70.61 and 70.62 are redesignated as §§70.81 and _70.82, respectively.

15. The undesignated center heading aENFORCEMENT" is redesignated as asubpart J -

Enforcemene

§§ 70.71 and 70.72 [Redesignated]

16. Sections 70.71 and 70.72 are redeslgnated as §§70.91 and 70.92, respe~ely.

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17. In Part 70, a new "SUBPART H" (§§ 70.60 - 70.74) is added to read as follows:

Subpart H - Additional Requirements for Certain Licensees Authorized to Possess a CriticaJ Mass of Special Nuclear Material.

Sec.

70.60 Applicability.

70.61 Performance requirements.

70.62 Safety program and integrated safety analy~is.

70.64 Requirements for new facilities or new processes at existing facilities.

70.65 Adcfrtional content of applications.

70.66 Additional requirements for approval of license application.

70.72 Facility changes and change process.

70. 73 Renewal of licenses. ,

70.74 Additional reporting requirements.

§70.60 Applicability.

The regulations in §70.61 through §70.74 apply, in addition to other.applicable Commission regulations, to each applicant or licensee that is or plans to be: (1) authorized to possess greater than a critical mass of special nuclear material, and (2) engaged in enriched uranium processing, fabrication of uranium fuel or fuel assemblies, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap recovery of special nuclear material, or any other activity that the Com_mission determines could significantly affect public health and safety. The regulations in §70.61 through §70.74 do not apply to decommissioning activities performed pursuant to other applicable Commission regulations including §70.25 and §70.38 of this Part. Also, ~e regulations in §70.61 through §70.74 do not apply to activities that are certified by the Commission pursuant to Part 76 of this chapter or licensed by the Commission pursuant to other parts of this chapter.

§70.61 Pertonnance Requirements.

(a) Each applicant or licensee shaJI evaluate, in the integrated safety analysis performed in accordance with §70.62, its compliance with the performance requirements in paragraphs (b),

(c), and (d) of this section.

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(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 ,wants 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 located outside the controlled area identified pursuant to paragraph (f) of this section; or (4) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that (i) Could endanger the life of a worker, or (Ii) Could lead to irreversible or other serious, long-lasting health effects to any individual located outside the controlled area identified pursuant to paragraph (f) of this section. If an applicant possesses' or plans to possess q~.::;.;.tities of material capable of such chemicaJ exposures, then the applicant shall propose appropriate quantitatjve standards for these health effects, as part of the information submitted pursuant to §70.65 of this Part.

(c) The risk of each credible intermediate-consequence event must be limited, unless the event is 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 intermediate-consequence events demonstrated to be unlikely.

Intermediate-consequence events are those internally or externally initiated events, that are not high-consequen~ events, that result in:

(1) An acute worker dose of 0.25 Sv (25 rem) or greater total effective dose equivalent; (2) An acute dose of 0.05 Sv (5 rem) or greater total effective dose equivalent to any individual located outside the controlled area identified pursuant to paragraph (f) of this section; (3) A 24-hour averaged release of radioactive material outside the restricted area in concentrations exceeding 5000 times the values in Table 2 of Appendix B to 1 0 CFR Part 20; or (4) An acute chemical exposure to an individual from licensed material or hazardous chemlca.Js prcx:tuced from licensed material that (i) Could lead to Irreversible or other serious, long-lasting health effects to a worker, or 47

(ii) Could cause mild transient health effects to any Individual located outside the controlled area as specffied 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 normaJ and credible abnormal conditions, all nuclear processes are subcritical, including use of an approved margin of subcriticalityfor safety. Preventive controls and measures must be the primary means of protection against nuclear criticality accidents.

(e) Each engineered or administrative control or control system necessary to comply with paragraphs (b), (c), or {d) of this section shall be designated as an item relied on for: safety. The safety program, established and maintained pursuant to §70.62 of this Part, shall ensure that each item relied on for safety will be available and reliable to perform its intended function when needed

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and in the context of the performance requirements of this section.

{~ Each licensee must establish a controlled area, as defined in §20.1003, in which the licensee retains the authority to determine aJI activities, including exclusion or removal of personnel and property from the area. For the purpose of complying with the performance requirements of this section, individuals who are not workers, as defined in §70.4, may be permitted to perform ongoing activities (e.g., at a facility not related to the licensed activities) in the controlled area, if the licensee:

{1) Demonstrates and documents, in the integrated safety analysis, that the risk for those Individuals at the location of their activities does not exceed the performance requirements of paragraphs {b)(2), {b){3), (b)(4){ii), {c)(2), and (c)(4){ii) of this section; or (2) Provides: (i) Training in accordance with *1 O CFR 19.12{a)(1 )-(5) to these individuals to ensure that they are aware of the risks associated with accidents invoMng the licensed activities as determined by the integrated safety analysis, and (ii) Conspicuously posts and maintains notices stating where the information in 10 CFR 19.11 (a) may be examined by these individuals. Under these conditions, the performance requirements for workers specffied In paragraphs (b) and (c) of this section may be applied to these individuals.

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§70.62 Safety Program and Integrated Safety Analysis (a) Safety program. (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 J

(b) through (d) of this section.

(2) Each licensee shall establish and maintain records that demonstrate compliance with the requirements of paragraphs (b) through (d) of this section.

(3) Each licensee shall *establish and maintain a log, available for NRC inspection, documenting each discovery that an item relied on for safety or management measure has failed to perform its function either in the context of the performance requirements of §70.61 or upon demand. This log must Identify the item relied on for safety or management measure that has failed and the safety function affected, the date of discovery, date (or estimated date) of the failure, duration (or estimated duration) of the time that the item was unable to perform its function, any other affected items relied on for safety or management measures and their safety function, affe~:d processes, cause of the failure, whether the fallur~ was in the context cf the performance requirements or upon demand or both, and any corrective or compensatory action that was taken.

The log must be initiated at the time of discovery and updated promptly upon the conclusion of

• each investigation of a failure of an item relied on for safety or management measure.

(b) Process safety information. Each licensee or applicant shall maintain process safety information to enable the performance of an Integrated safety analysis. This process safety information must include information pertaining to the hazards of the materials used or produced in the process, Information pertaining to the technology of the process, and information pertaining to the equipment in the process.

(c) Integrated safety analysis. (1) Each licensee or applicant shall conduct an integrated safety analysis, that is of appropriate detail for the complexity of the process, that identifies:

0) Radiological hazards related to possessing or processing licensed material at its facility; 49

(IQ 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 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) of this Part, the characteristics of its preventive, mitigative, or other safety function, and the assumptions and conditions under which the item is relied upon to support compliance with the performance requirements of §70.61 .

(2) Integrated safety analysis team qua/ifica'tions. In order to assure the adequacy of the integrated safety analysis, the analysis must be performed by a team with expertise in engineering and process operations. The team shall include at least one person who has experience and knowledge specific to each process being evaluated, and persons who have experience in nuclear criticality safety, radiation safety,' fire safety, and chemical process safety. One member of the team must be knowledgeable in the specific integrated safety analysis metr.:xk>logy being used.

• (3) Requirements for existing licensees. Notwithstanding other provisions regarding the effective date for Part 70 Subpart H requirements, licensees shall comply with the provisions in paragraphs (c)(3)(Q, (ii), and (iii) of this section beginning on <the date of publication of the final rule>. lndMduals holding an NRC license on <the date of publication (?f the final rule> shall, with regard to existing licensed activities:

(i) Within 6 months of the effective date of the rule, submit for NRC approval, a plan that describes the integrated safety analysis approach that will be used; the processes that will be analyzed, and the schedule for completing the analysis of each process.

(ii) Within 4 years of the effective date of the rule, complete an integrated safety analysis, correct all unacceptable performance deficiencies, and submit an integrated safety analysi~

summary in accordance with §70.65 or the approved plan submitted under paragraph (c)(3)(i) of this section.

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(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 measures to provide continuing assurance of compliance with the performance requirements ~f

§70.61. 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 administrative controls and control systems that are identified as items relied on for safety pursuant to §70.61 (e) of this Part are designed, implemented, and maintained, as necessary, to ensure they are available and reliable to perform their function when needed, in the-context of compliance with the performance requirements of §70.61 of this Part.

§70.64 Reguirements for new facilities or new processes at existing facilities.

(a) Baseline design criteria. Each prospective applicant or licensee shall address the

. following baseline design criteria in the design of new facilities. Each existing licensee shall address the follow:ng baseline design criteria in the design of new processes at existing facilities that require a license amendment under §70.72. The baseline design criteria must be applied to the design of new facilities and new processes, but do not require retrofits to existing facilities or existing processes (e.g., those housing or adjacent to the new process); however, aJI facilities and processes must comply with the performance requirements in §70.61: Licensees shall maintain the application of these criteria unless the evaluation performed pursuant to paragraph (c} of this section demonstrates that a given item is not relied on for safety or does not require adherence to the specified criteria (1) Quality standards and record§. The design must be developed and implemented in accordance with management measures, to provide adequate assurance that items relied on for safety will be available and reliable to perform their function when needed. Appropriate records of these items must be maintained by or under the control of 'the licensee throughout the life of the facility.

(2) Natural phenomena hazards. The design must provide for*adequate protection against natural phenomena with consideration of the most severe documented historical events for the site.

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(3) Fire protection. The design must provide for adequate protection against fires and explosions.

(4) Environmental and 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) Chemical protection. 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) Emergency capability. The design must provide for emergency capability to maintain control of:

(i) Licensed material; (ii) Evacuation of personriel; and (iii) Onsite emergency facilities a*nd services that facilitate the use of available offsite services.

(7) Utility services. The design must provide for continued operation of essential utility services.

(8) Inspection, testing. and maintenance. The design of items relied on for safety must provide for adequate inspection, testing, and maintenance, to ensure their availability and reliability to perform their function when needed.

(9) Criticality control. Th_e design must provide for criticality control including adherence to the double contingency principle.

(1 O) lnstrumehtatiqn and controls. The design must provide for inclusion of instrumentation and control systems to monitor and control the behavior of items reli~ on for safety.

(b) Facility and system design and plant layout must be based on defense-in-depth practices4

• The design process mU$! incorporate, to the extent pra~le:

4 As used in §70.64, defense-In-depth practices means a desig~ philosophy, applied from the outset and through completion of the design, that is based on providing successive levels of protection such that health and safety will not be wholly dependent upon any single element of the design, construction, maintenance, or operation of the facility. The net effect of Incorporating defense-In-depth practices is a conservatively designed facility and system that will exhibit greater tolerance to failures and external challenges. The risk insights obtained through performance of the Integrated safety analysis can be then used to supplement the final design by focusing attention on the prevention and mitigation of the higher-risk potential accidents.

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(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.

§70.65 Additional content of applications.

(a) In addition to the contents required by §70.22, each application must include a description of the applicant's safety program established under §70.62, 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 or renewal application (and amendment application as necessary), but shall not be incorporated in the license.

However, changes to_the integrated safety analysis summary shall meet the conditions of §70.72.

The integrated safety analysis summary must contain:

(1) A general description of the site with emphasis on those factors that could affect safety (i.e., meteorology, seismology);

(2) A general description of the facility with emphasis on those areas that could affect safety, including an identification of the controlled area boundaries; (3) A description of each process (defined as a single reasonably simple integrated unit operation within an overall production line) analyzed in the integrated safety analysis in sufficient

. detail to understand the theory of operation; and, for each process, the hazards that were identified in the integrated safety analysis pursuant to §70.62(c)(1 )(i)-(iii) and a general description of the

. types of accident sequences; (4) Information that demonstrates the licensee's compliance with the performance requirements of §70.61; the requirements for criticality monitoring and alarms in §7024; and, if applicable, the requirements of §70.64; (5) A description of the team, qualifications, and the methods used to perform the integrated safety analysis; (6) A list briefly describing all items relied on for safety which are identified pursuant to

§70.61 (e) in sufficient detail to understand their functions in relation to the performance requirements of §70.61 ;

(7) A description of the proposed quantitative standards used to assess the consequences from acute chemical exposure to licensed material or chemicals produced from licensed materials which are on-site, or expected to be on-site as described in §70.61 (b)(4) and (c)(4);

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(8) A descriptive list that identifies all items relied on for safety that are the sole item 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.

§70.66 Additional requirements for approval of ficense application.

An application for a license from an applicart subject to Subpart H will be approved if the

. Commission determines that the applicant has complied with the requirements of §70.21, §70.22,

§70.23 and §70.60 through §70.65.

§ 70. 72 Facility changes and change process.

(a) The licensee shall establish a configuration management system to evaluate, implement, and track each change to the site, structures, processes, systems, equipm~nt, componer.its, computer programs, and activities of personnel. This system must be documented in written procedures and must assure that the following are addressed prior to implementing any change:

(1) The technical basis for the change; (2) Impact of the change on safety and health or control of licensed material; (3) Modifications to existing operating procedures including any necessary training or retraining before operation; (4) Autho~tion requirements for the change; (5) For temporary changes, the approved duration (e.g., expiration date) of the change; and (6) The impacts or modifications to the integrated safety analysl_s, integrated safety analysis summary, or other safety program information, *develcped In accordance with §70.62.

(b) Any change to site, structures, processes, systems, equipment, components, computer programs, and activities of personnel must be evaluated by the licensee as specified in paragraph (a) of this section, before the change is implemented. The evaluation of the change must determine, before the change is implemented, If an amendment to the license is required to be submitted in accordance with §70.34.

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(c) The licensee may make changes to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel, without prior Commission approval, If the change:

(1) does not (I) Create new types5 of accident sequences that, unless mitigated or prevented, would exceed the performance requirements .of §70.61 and that have not pr,eviously been described in the integrated !¥ifety analysis summary; or (ii) Use new processes, technologies, or control systems for which the Hcensee has no prior experience; (2) Does not remove, without at least an equivalent replacement of the safety function, an item relied on for safety that is listed in the integrated safety analysis summary; (3) Does not alter any item relied on for safety, listed in the integrated safety analysis summary, that is the sole item preventing or mitigating an accident sequence that exceeds the

• performance requirements of §70.61; and (4) Is not otherwise prohibited by this section, license condition, or order.

(d)(1) For any changes that affect the integrated safety analysis summary, as submitted in accordance with §70.65, but do not require NRC. pre-approval, the licensee shall submit revised pages to the integrated safety anaJysls summary, to NRC, within 90 c:..ys of the change.

(2) For changes that require pre-approval under §70. 72, the licensee shall submit an amendment request to the NRC In accordance with §70.34 and §70.65.

(3) A brief summary of all changes to the records required by §70.62(a)(2) of this Part, that

. *are made without prior Commission approval, must be submitted to N_RC every 12 months.

(e) If a change covered by §70.72 Is made, the affected on-site documentation must be updated promptly.

(f) The licensee shall maintain records of changes to its facility carried out under this section. These records must include a written evaluation that provides the bases for the 6

Any change in the defining characteristics of the elements of an accident sequence may change the "type" of the accident sequence for a given process. For example, a new type of accident could involve a different initiator, significant changes In the consequence, or a change in the safety function of a control (e.g., temperature limiting device versus a flow limiting device).

55

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 li~nse.

§70. 73 Renewal of licenses.

Applications for renewal of a license must be filed in accordance with §§2.109, 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 references are clear and specific.

§70.74 Additional reporting regujrements.

(a) Reports to NRG Operations Center.

(1) Each licensee shall report to the NRG Operations Center the events described in Appendix A to Part 70.

(2) Reports must be made by a knowledgeable licensee representative and by any method that will ensure compliance with the required time period for reporting.

(3) Th_e information provided must include a description of the event and other related informa~~n as described in §70.50(c)(1)

(4) Follow-up information to the reports must be provided until all information required to be reported in §70.50(c)(1) of this Part is complete.

(5) Each licensee shall provide reasonable assurance that reliable communication with the NRG Operations Center is available during each event.

(br Written Reports. Each licensee who makes a report required by paragraph (a)(1) of this section shall submit a written follow-up report within 30 days of the initial report. The written report must contain the information as described in §70.50(c)(2).

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Appendix A to Part 70 - Reportable Safety Events As required by 10 CFR 70.74, licensees subject to the requirements in Subpart Hof Part 70, shall report:

(a) One hour reports. Events to be reported to the \ NRC Operations Center within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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:

(1) An inadvertent nuclear criticality.

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(2) An acute intake by an indMdual of 30 mg or greater of uranium in *a soluble form.

(3) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that exceeds the quantitative standards established to satisfy the requirements in §70.61 (b)(4).* *

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(4) An event or condition such that no'items relied on for safety, as documented in the lnte~,* ated Safety Analysis summary, remain available and reliable, in an c.ccident sequence evaluated in the Integrated Safety Analysis, to perform their function:

(i) In the context of the performance requirements in §70.61 (b) and §70.61 (c), or 00 Prevent a nuclear criticality accident (i.e., loss of all controls in a particular sequence).

(5) Loss of controls such that only one item relied on for~ . 1,' as documenteo ::-, the Integrated Safety Analysis summary, remains -available and reliable to prevent a nucl~r crincality accident, and has been in this state for greater than eight hours.

(b) Twenty:efour hour reports. Events to be reported to the NRC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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:

57

(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 satisfy the requirements of §70.61 (c)(4).

(4) Any natural phenomenon or other external event, including fires internal and external to the facility, that has affected or may have affected the intended, safety function or availability or reliability of one or more items relied on for safety. *

(5) An occurrence of an event or process deviation that was considered in the Integrated Safety Analysis and:

(i) Was dismissed due to Its likelihood; or (ii) Was categorized as unlikely and whose associated unmitigated ..onsequences would have exceeded those in §70.61 (b) had the item(s) relied on for safety not performed their safety function(s).

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(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 ~J__ day of July, 1999.

For the Nuclear Regulatory Commission.

~~~

• Secretary of the Commission.

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