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SECY-98-185

July 30, 1998

PURPOSE:

To obtain Commission approval to publish a proposed rule amending 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material."

SUMMARY:

A near-criticality incident at a low enriched fuel fabrication facility in May of 1991 prompted the U.S. Nuclear Regulatory Commission (NRC) to evaluate its safety regulations for licensees that possess and process large quantities of special nuclear material (SNM)⁽¹⁾. As a result of this review, the Commission (SRM dated January 15, 1993) and the staff recognized the need for revision of its regulatory base for these licensees and, specifically, for those possessing a critical mass of SNM. Further, the 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 integrated safety analysis (ISA). An ISA is a systematic analysis that identifies: 1) plant and external hazards and their potential for initiating accident sequences; 2) the potential accident sequences and their likelihood and consequences; and 3) the structures, systems, equipment, components, and activities of personnel relied on for safety to prevent or mitigate potential accidents at a facility.

NRC held public meetings with the nuclear industry on this issue during May and November of 1995. Industry's position on the need for revision of NRC regulations in Part 70 was articulated to the Commission by the Nuclear Energy Institute (NEI) at a July 2, 1996, meeting, and in the subsequent filing, in September 1996, of a Petition for Rulemaking (PRM 70-7) by NEI with NRC. The PRM 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 circumstances, within Part 70.

In SECY-97-137, dated June 30, 1997, the staff proposed a resolution to the NEI PRM and recommended that the Commission direct the staff to proceed with rulemaking. The staff's recommended approach to rulemaking included the basic elements of the petition, with some modification. In brief, the staff proposed to revise Part 70 to include the following major elements:

In a Staff Requirements Memorandum (SRM) dated August 22, 1997, the Commission "...approved the staff's proposal to revise Part 70" and directed staff to "...submit a draft proposed rule...by July 31, 1998." In addition, the SRM requested that the supporting guidance documents and regulatory analysis be submitted with the proposed rule.

A public meeting on modifications to Part 70 was held on May 28, 1998, with a follow-up public meeting on July 15, 1998, to share information on NRC efforts to revise Part 70 and to provide an opportunity for the presentation of public and industry views on this matter. The presentation made by NRC provided information on the proposed modifications to Part 70 and the Standard Review Plan that has been developed to support the draft proposed rule. The industry presentation noted convergence of some of the past differences between industry and NRC, and concluded that the rulemaking can and should move forward, and industry can effectively implement the rule, if the rule is modified to clarify that:

Based on information provided by NRC staff, the industry representatives expressed the view that the differences concerning the second item are now less of a concern. The remaining two items are issues that the staff addressed in SECY-97-137. Regarding the first item, the industry view is that the ISA results should not be incorporated in the license, but become a part of the documented safety basis that would be available at the site for NRC review. The ISA results would not be provided, as part of an application, to NRC for review. With respect to the third item, backfit, the main industry concern appears to be that NRC staff may want to impose safety controls that the licensee would consider unnecessary to satisfy the performance criteria established in the rule. These remaining issues were previously reviewed in SECY-97-137 in which the staff recommended that these items be rejected and the Commission endorsed the staff's position.

The NRC staff remains in favor of the approach recommended in SECY-97-137. In particular, the staff continues to believe that the safety basis for a facility, including the ISA results, should be submitted, as part of an application, to NRC for review and incorporated in the license. Also, the staff believes 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. However, given the views expressed by industry at the public meeting, the staff has included in the Federal Register notice a request for public comment on the intent to defer the implementation of a backfit provision in Part 70.

DISCUSSION:

The staff's proposed revisions to Part 70 are intended to provide a risk-informed, performance-based approach for increasing confidence in the margin of safety for licensees authorized to possess a critical mass of SNM. With two exceptions, the staff's approach is in accordance with the Commission's SRM of August 22, 1997. First, as the result of its experience in supporting the Department of Energy Regulatory Unit at Hanford with the Tank Waste Remediation System Privatization Project, the staff is recommending the inclusion of certain baseline design criteria in the proposed revisions to assure the safe design of new facilities and new processes at existing facilities. Second, the staff is recommending that the proposed revisions should not apply to all licensees authorized to possess a critical mass of SNM. Instead, the staff has identified a subset of these licensees that, based on their risk of operations, should be subject to the new requirements. A more detailed discussion of the major rule elements is presented in Attachment 1, "Discussion of the Proposed Rule's Major Elements." Most of the proposed modifications to Part 70 are found in a new Subpart H -- "Additional Requirements for Licensees Authorized to Possess a Critical Mass of Special Nuclear Material." Three draft guidance documents support the rulemaking - a Standard Review Plan, an ISA Guidance Document, and "Example Elements of an ISA Submittal -- Process Descriptions and Accident Analysis Summary."

COORDINATION:

The Office of the General Counsel has no legal objection to this paper. The Office of the Chief Financial Officer has reviewed this Commission Paper for resource implications and has no objections. The Office of the Chief Information Officer has reviewed the proposed rule for information technology and information management implications and concurs in it. However, the rule amends information collection requirements that must be submitted to and received by the Office of Management and Budget no later than the date the rule is published in the

Federal Register.

RESOURCES:

Resources to complete and implement the rule are included in the current budget.

RECOMMENDATIONS:

That the Commission:

ATTACHMENT 1

DISCUSSION OF PROPOSED RULE'S MAJOR ELEMENTS

Consequences of concern. An important element in the proposed rule is the identification of specific consequences against which licensees must provide adequate protection [10 CFR 70.60(b)]. These consequences, which are applicable to workers and members of the public, are categorized according to their level of severity (high and intermediate). Because accidents at fuel cycle facilities could result in human exposure to both radiological and chemical hazards, the proposed rule has adopted criteria that address both types of consequences. This approach satisfies the U.S. Nuclear Regulatory Commission's (NRC's) primary responsibility for radiation protection, in addition to its responsibility to protect workers and the public from the chemical hazards resulting from the processing of licensed nuclear material.

Graded Level of Protection. To ensure an acceptable level of risk at facilities that possess a critical mass of special nuclear material, the proposed rule [10 CFR 70.60(c)] calls for licensees to provide a graded level of protection against potential accidents. That level of protection must be sufficient to reduce the likelihood of such accidents to levels commensurate with their consequences. Thus, according to the proposed rule, the occurrence of any high-consequence event should be "highly unlikely," while the occurrence of any intermediate-consequence event should be "unlikely." Although the rule does not define the terms "highly unlikely" and "unlikely," the draft Standard Review Plan provides criteria for judging the likelihood of potential accidents. This guidance is based on a combination of qualitative and quantitative indicators, but does not require a probabilistic risk assessment.

Integrated Safety Analysis (ISA). According to the proposed rule [10 CFR 70.60(d)], licensees must demonstrate, based on the performance of an ISA, their ability to provide an adequate level of protection against potential accidents. An ISA is a systematic analysis to identify plant and external hazards and their potential for initiating accident sequences; the potential accident sequences and their likelihood and consequences; and the items (i.e., site, structures, systems, equipment, components, and activities of personnel) that are relied on for safety.

Measures to ensure continuous availability and reliability. Although the ISA plays a critical role in identifying potential accidents and the items relied on for safety, the performance of an ISA will not, by itself, ensure adequate protection. Instead, as required by the proposed rule [10 CFR 70.60(d)], an effective management system is needed to ensure that, when called upon, the items relied on for safety are in place and operating properly. Maintenance measures must be in place to ensure the continuous availability and reliability of all hardware relied on for safety. Training measures must be established to ensure that all personnel whose actions are relied on for safety are appropriately trained to perform their safety functions. Human-system interfaces and safety-related procedures must be developed and implemented to enable personnel relied on for safety to effectively carry out their duties. 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. In addition, quality assurance measures need to be established to ensure that the items relied on for safety, and the measures used to ensure their continuous availability and reliability, are of sufficient quality. 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 public health and safety. When operational events occur, investigations of those events must be carried out 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.

Inclusion of safety bases in the application and changes to the safety bases. The performance of the ISA to identify the items relied on for safety and the measures established to ensure the continuous availability and reliability of such items are important elements in increasing confidence in the margin of safety. Nevertheless, without formal commitments to implement these items and measures, and to keep NRC informed of any changes in such commitments, the safety bases could become uncertain over time. Thus, the proposed rule calls for the incorporation of licensee commitments to these items and measures in the license application. In addition, all changes in such commitments shall be submitted to NRC as part of a revised license application, including any changes in the ISA results (10 CFR 70.72). The rule does, however, allow for certain changes to be made, based on the results of the ISA, without prior NRC approval, as long as such changes result in, at most, a minimal increase in the risk of accidents at the facility.

ATTACHMENT 2

[7590-01-P]

NUCLEAR REGULATORY COMMISSION
10 CFR Part 70
RIN 3150 - AF22
Revised Requirements for the Domestic Licensing of Special Nuclear Material

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend its safety regulations in the provisions 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; uranium fuel fabrication; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any other activity involving a critical mass of SNM that the Commission determines could significantly affect public health and safety. 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; require the implementation of measures to ensure that the items relied on for safety are continuously available and reliable; require the inclusion of the safety bases, including the results of the ISA, in the license application; and allow for licensees to make certain changes to their facilities without prior NRC approval.

DATES: The comment period expires (insert 75 days after publication in the Federal Register.) 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: The 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. The interactive rulemaking website can then be accessed by selecting "New Rulemaking Website." 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, (301) 415-5905; e-mail cag@nrc.gov.

FOR FURTHER INFORMATION, CONTACT: Richard I. Milstein, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone (301) 415-8149; e-mail rim@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 of 1991 prompted NRC to review its safety regulations for licensees that possess and process large quantities of SNM. [See "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 its regulatory base for these licensees and, specifically, for those 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:

(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 of 1995. Industry's position on the need for revision of NRC regulations in Part 70 was articulated to the Commission by the Nuclear Energy Institute (NEI) at a July 2, 1996, meeting, and in the subsequent filing of a Petition for Rulemaking (PRM-70-7) by NEI with NRC in September 1996. NRC published in the Federal Register a notice of receipt of the PRM and requested public comments on August 21, 1996 (61 FR 60057). The PRM 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 circumstances, within Part 70.

In SECY-97-137, dated June 30, 1997, the NRC staff proposed a resolution to the NEI PRM and recommended that the Commission direct the staff to proceed with rulemaking. The NRC staff's recommended approach to rulemaking included the basic elements of the PRM, with some modification. In brief, NRC staff proposed to revise Part 70 to include the following major elements:

(1) Performance of a formal ISA, which would form the basis for a licensee's safety program. This requirement would apply to all licensed facilities (except reactors and the gaseous diffusion plants regulated under 10 CFR Part 76) or activities, subject to NRC regulation, that are authorized to possess SNM in quantities sufficient to constitute a potential for nuclear criticality;

(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 or mitigate these accidents, and the measures needed to ensure the continuous availability and reliability of these items). (This is in contrast to the PRM's approach, where the ISA results would not be included in the license application);

(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 initial conduct and implementation of the ISA by the licensees, of a qualitative backfitting mechanism to enhance regulatory stability.

In a Staff Requirements Memorandum (SRM) dated August 22, 1997, the Commission "... approved the staff's proposal to revise Part 70" and directed the NRC staff to "... submit a draft proposed rule...by July 31, 1998."

II. Description of Proposed Action

The Commission has decided to grant, in part, the NEI PRM by initiating this rulemaking. Further, the proposed rule adopts the petitioner's proposal 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 consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed such criteria;

(2) The 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 continuously available and reliable;

(4) The inclusion of the safety bases, including the ISA results, in the license application; and

(5) The allowance for licensees to make certain changes to their 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 ISA requirement be limited to "... uranium processing, fuel fabrication, and uranium enrichment plant licensees," the Commission would require the performance of an ISA for a broad range of Part 70 licensees that are authorized to possess a critical mass of SNM. The Part 70 licensees that would be affected include licensees engaged in one of the following activities: enriched uranium processing; uranium fuel fabrication; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any other activity involving a critical mass of SNM that the Commission determines could significantly affect public health and safety. The proposed rule would not apply to regulatees authorized to possess SNM under 10 CFR Parts 50, 60, 72, and 76.

Furthermore, the Commission is not currently proposing, as suggested in the NEI petition, to include a backfit provision in Part 70. Based on the discussions at a public meeting held on May 28, 1998, the purpose of the proposed backfit provision 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 bases, including the results of the ISA, are incorporated in the license application, and the NRC staff has gained sufficient experience with implementation of the ISA requirements, a qualitative backfit mechanism could be considered. Without a baseline determination of risk, as provided by the initial ISA process, it is not clear how a determination of incremental risk, as needed for a backfit analysis, would be accomplished. Furthermore, although NEI believes that a quantitative backfit approach is currently feasible, it would appear that a quantitative determination of incremental risk would require a Probabilistic Risk Assessment, to which the industry has been strongly opposed. Given the differences of opinion on this subject, the Commission requests public comment on its intent to defer consideration of a qualitative backfit provision in Part 70.

The majority of the proposed modifications to Part 70 are found in a new subpart, "Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material," that consists of Ё70.60 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 general accordance with the approach approved by the Commission in its August 22, 1997, SRM. However, the Commission has decided that the new requirements should not apply to all licensees authorized to possess a critical mass of SNM. Instead, the Commission has identified a subset of these licensees that, based on the relatively high level of risk associated 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 requirements, because of the relatively low level of risk at these facilities. This issue is further addressed in the discussion of �.62.

Section 70.4, "Definitions."

The following fourteen definitions would be added to this section to provide a clear understanding of the meaning of the new subpart H, "Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material:" Acute exposure, Acute exposure guideline levels, Controlled site boundary, Critical mass of SNM, Deviation from safe operating conditions, Double contingency, Emergency response planning guidelines, Hazardous chemicals, Integrated safety analysis, Items relied on for safety, New process, Results of the ISA, Unacceptable vulnerabilities, and Worker.

Section 70.15, "Nuclear reactors."

A new section would be added to subpart B, "Exemptions," that exempts nuclear reactors licensed under Part 50 from the new subpart H, "Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material."

Section 70.22, "Contents of applications."

Paragraph (f) would be removed. Paragraph (f) currently requires that, for plutonium processing and fuel fabrication facilities, certain additional safety-related information be submitted with an application. The new subpart H, "Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material," would contain requirements for the submittal of information called for in paragraph (f) and is sufficient to allow the Commission to make a determination of adequacy.

Section 70.23, "Requirements for the approval of applications."

Paragraphs (a)(8), and (b) would be removed. These paragraphs currently require that the Commission, to approve an application, determine that the construction of a plutonium processing and fabrication facility meet certain conditions. These conditions would be covered in the new subpart H, "Additional Requirements for Certain Applicants Authorized to Possess a Critical Mass of Special Nuclear Material."

Section 70.60, "Safety performance requirements."

These requirements would establish the purpose of the new requirements, identify the potential adverse consequences that need to be protected against, establish the level of protection that is needed to ensure that the consequences of concern do not occur, and identify the safety program elements that allow licensees to demonstrate their ability to provide an adequate level of protection.

Section 70.60(a), "Purpose."

This paragraph would address the following questions: Why are the new requirements needed? What hazards need to be considered? Who are the intended beneficiaries? In general, the new requirements are intended to ensure that workers⁽²⁾, the general public, and the environment are protected from radiological and certain chemical hazards associated with plant operations. All hazards, including fire, chemical, electrical, industrial, etc., that can potentially affect radiological safety, must be considered and addressed by licensees. In addition, chemical hazards that result from the processing of licensed nuclear material must also be considered.

The question of NRC's authority to regulate chemical hazards at its fuel cycle facilities was raised after an accident in 1986 at a Part 40 licensed facility, in which a cylinder of uranium hexafluoride ruptured and killed a worker. 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). As a result of that incident, NRC and the Occupational Safety and Health Administration (OSHA) established a memorandum of understanding (MOU) (1988) that identified the respective responsibilities of both agencies for the regulation of chemical hazards at nuclear facilities. The MOU identified the following four areas of responsibility. The NRC has responsibility for the first three areas, whereas OSHA has responsibility for 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 materials; and

(4)Plant conditions that result in an occupational risk, but do not affect the safety of licensed radioactive materials.

The purpose of the "Safety Performance Requirements," as defined in �.60(a), is consistent with the NRC/OSHA MOU.

Section 70.60(b), "Consequences of concern."

The NRC is responsible for ensuring that workers and the general public are protected from the hazards involved in the handling, processing, and storage of SNM. All hazards (including fire and chemical) that could result in radiological consequences are a subject of NRC concern. In addition, all chemical hazards resulting from the processing of licensed SNM that could directly affect a worker or member of the public are also a matter of NRC concern. Thus, NRC regulations need to address both radiological and chemical consequences. The following discussion provides information, on the consequences of human exposure to radiation and hazardous chemicals, that is relevant to the choice of appropriate consequence criteria. The actual choice of these criteria is discussed in Ё70.60(b)(1)(ii)(A) and (B); 70.60(b)(1)(iii)(A) and (C); and 70.60(b)(2)(i)(A) and (B).

Radiological Consequences. In the past, the regulation of licensees authorized to possess SNM, under 10 CFR Parts 70 and 20, has concentrated on radiation protection for persons involved in nuclear activities conducted under normal operations. The proposed amendments to Part 70 would explicitly address the potential exposure of workers or members of the public to radiation as a result of accidents. Because accidents are unanticipated events that usually occur over a relatively short period of time, a regulation that seeks to assure adequate protection of workers and members of the public must limit the risk of such accidents. This can be accomplished by identifying appropriate consequence criteria and by limiting the likelihood of occurrence of the identified consequences. In selecting the radiological consequence criteria for use in the proposed rule, the Commission has examined the radiological criteria and design basis accident scenarios used in existing NRC regulations to ensure that the proposed consequence criteria are consistent with criteria used in other Commission rules.

Chemical Consequences. The processing of SNM may involve the use or production of hazardous chemicals. For example, low enriched uranium fuel fabrication facilities convert uranium hexafluoride to uranium oxide by reaction with water (hydrolysis) to form uranyl fluoride and hydrogen fluoride. Uranyl fluoride, in addition to being radioactive, is a toxic uranium compound that can cause damage to the kidney. Hydrogen fluoride is highly toxic and poses a hazard to both workers and the general public. Other hazardous chemicals, including ammonia, nitric acid, and sulphuric acid, are also used at uranium fuel fabrication facilities. The effort to limit exposure of workers and the general public to hazardous chemicals is based on two concerns: acute exposures that could result from accidental releases, and chronic exposures (i.e., multiple and repeated exposures occurring over a long period of time -- days, months, or years), resulting from releases during normal operations.

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 hours 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 the U.S. Environmental Protection Agency (EPA) and the National Academy of Sciences (NAS), has identified a priority list of approximately 85 chemicals. Consequence criteria for 12 of these have currently been developed and criteria for approximately 30 additional chemicals per year are expected.

Another set of chemical consequence criteria, the Emergency Response Planning Guidelines (ERPGs), has been developed by the American Industrial Hygiene Association (AIHA) 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 chemical 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.

Federal regulations and internal U.S. Department of Energy (DOE) guidance require the use of ERPGs for emergency planning. Recognizing that ERPGs exist for a limited number of chemicals, DOE's Subcommittee on Consequence Assessment and Protective Actions developed Temporary Emergency Exposure Limits (TEELs) so that DOE facilities could perform complete hazard analysis and consequence assessments, even for chemicals lacking ERPGs. TEELs are not equivalent to ERPGs, but are approximations to ERPGs. They exist only until an ERPG is developed for a chemical. As of July 1997, 400 TEELs had been developed according to a methodology published in the American Industrial Hygiene Journal (1995). That methodology is not based directly on toxicological studies of the chemicals involved, but on a derived relationship between alternative exposure-limit parameters and the existing ERPG criteria. The use of the methodology results in a significant underestimation of the TEEL-2⁽³⁾ level (0.6 mg/m³) for soluble uranium and would be inconsistent with the criterion on soluble uranium intake (i.e., 30 mg) proposed in this rule.

A fourth set of chemical consequence criteria that was considered potentially applicable for acute exposure to hazardous chemicals is the Immediately Dangerous to Life and Health (IDLH) criteria established by the National Institute for Occupational Safety and Health (NIOSH). However, according to NIOSH, the IDLH criteria are defined "... only for the purpose of respirator selection." In addition, unlike the previously mentioned sets of criteria, there is only one IDLH level that has been defined. This would not facilitate the definition of multiple consequence levels for workers and the public, as intended in the proposed rule.

For chronic exposures of workers to hazardous chemicals during normal and off-normal operations, the permissible exposure limits (PELs) established by OSHA in 29 CFR 1910 are applicable. However, these limits are not relevant for acute exposures to hazardous chemicals.

Given the status of these various sets of consequence criteria, the Commission has chosen AEGLs and ERPGs, in that order, as criteria to be used for acute short-term exposure to hazardous chemicals. If a given chemical has an AEGL associated with it, that criterion should be used. If not, the ERPG criterion, if available, should be used. Appendix A contains the available AEGL values, and Appendix B contains the available ERPG values. If both AEGLs and ERPGs are available for a particular chemical, only the AEGL values will be presented. Although the TEELs cover a wide range of additional hazardous chemicals, the Commission has decided not to require their use at this time, because the methodology used to derive these values is not based on the toxicology of the chemicals involved and may, at least in certain cases, underestimate the limits. However, the use of the TEELs may be justified on a case-by-case basis in the absence of other applicable standards.

As a result of further study, new AEGL or ERPG values are expected to be established by the issuing organizations (EPA for AEGLs; AIHA for ERPGs). The Commission does not propose to engage in full, formal rulemaking with respect to these future changes, but will incorporate them in the codified appendices in final form by issuing an immediately effective final rule. The Commission believes that these purely technical changes or additions do not require comment and are, in addition, subject to the categorical exclusion in 10 CFR 51.22(c)(2).

General Approach

The consequences of concern, identified in Ё70.60(b)(1) and (b)(2), describe those consequences that licensees must protect against⁽⁴⁾. The level of protection to be provided is discussed in �.60(c) and depends on the severity of the consequences. The goal is to ensure an acceptable level of risk by limiting the likelihood of occurrence of the identified consequences. The consequences identified in �.60(b)(1) of the proposed rule are considered to be high consequences and include the occurrence of a criticality, and accidental exposure of a worker or member of the public to high levels of radiation or hazardous chemicals. The consequences identified in �.60(b)(2) are considered to be intermediate consequences and include accidental exposure of a worker or member of the public to moderate levels of radiation or hazardous chemicals, and significant releases of radioactive material to the environment. The proposed consequence criteria that are applicable to a member of the public are more restrictive than those that are applicable to a worker. Also, within each category (worker and public), NRC recognizes that the proposed radiological criteria are more restrictive (in terms of acute health effects) than the chemical criteria for a given level of severity (high or intermediate) and that this is consistent with current regulatory practice.

In some cases, a qualitative description of the consequence is used (e.g., a nuclear criticality); in other cases, a numerical criterion is used. For cases where numerical criteria have been used, NRC has based the criteria on values that have been developed previously by NRC or other government agencies or professional societies. Table 1 illustrates the radiological and chemical consequence criteria used in the proposed rule.

TABLE 1 Radiological and Chemical Consequence Criteria

Certain events that could occur at licensees' facilities are considered high-consequence events. They include the occurrence of an inadvertent criticality, the exposure of a worker or member of the public to levels of radiation at which clinically observable biological damage could occur, or concentrations of hazardous chemicals at which death or life threatening injury could occur.

Section 70.60(b)(1)(i). This paragraph deals with a nuclear criticality.

The occurrence of an inadvertent nuclear criticality is considered to be a high-consequence event. Although detecting and mitigating the consequences of a nuclear criticality are important objectives (see 10 CFR 70.63), the prevention of a criticality is a primary NRC objective.

Section 70.60(b)(1)(ii)(A). This paragraph deals with an acute exposure of a worker to a radiation dose of 1 Sv (100 rem) or greater total effective dose equivalent (TEDE).

An acute exposure of a worker to a radiation dose of 1 Sv (100 rem) or greater 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 [linear 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 she should also have an appreciation of the substantial increase in his or her lifetime risk of cancer." The use of the 1-Sv (100-rem) criterion is not intended to imply that 1 Sv (100 rem) constitutes an acceptable criterion for an emergency dose to a worker. Rather, this dose value has been proposed in this section as a reference value, which should be used by licensees to determine the level of protection (i.e., items relied on for safety, and measures to assure their continuous availability and reliability) needed to ensure an acceptably low level of risk to workers.

Section 70.60(b)(1)(ii)(B). This paragraph deals with an acute exposure of a worker to hazardous chemicals in concentrations exceeding AEGL-3 or ERPG-3 limits.

An acute exposure of a worker to hazardous chemicals at concentrations that could cause death or life-threatening injuries is considered a high-consequence event. Two existing criteria, AEGL-3⁽⁵⁾ and ERPG-3, can be used to define such concentration levels. 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, could experience life-threatening effects or death." ERPG-3 is defined as "The maximum airborne concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or developing life-threatening health effects." If, for a particular chemical, the AEGL-3 value is available, it should be used. Otherwise, the ERPG-3 value should be used. If there is no AEGL or ERPG value available, then the applicant should adopt a criterion that is comparable in severity to those that have been established for other chemicals.

Section 70.60(b)(1)(iii)(A). This paragraph deals with an acute exposure of a member of the public to a radiation dose of 0.25 Sv (25 rem) or greater TEDE.

The exposure of a member of the public to a radiation dose of 0.25 Sv (25 rem) TEDE is considered a high-consequence event. This is based on the criterion established in 10 CFR 100.11, "Determination of exclusion area, low population zone, and population center distance," and 10 CFR 50.34, "Contents of applications; technical information," where a whole-body dose of 0.25 Sv (25 rem) is used to determine the dimensions of the exclusion area and low population zone required for siting nuclear power reactors.

Section 70.60(b)(1)(iii)(B). This paragraph deals with an intake of 30 mg or greater of uranium in a soluble form by a member of the public.

The intake of 30 mg of soluble uranium by a member of the public 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...uranium intakes of 30 milligrams...." The final rule also stated that "The Commission's intended use of chemical toxicity considerations in Part 76 is consistent 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 70.60(b)(1)(iii)(C). This paragraph deals with an acute exposure of a member of the public to hazardous chemicals in concentrations exceeding AEGL-2 or ERPG-2 criteria.

An acute exposure of a member of the public to hazardous chemicals at concentrations that could cause irreversible health effects is considered a high-consequence event. Two existing criteria, AEGL-2 and ERPG-2, can be used to define such concentration levels.

AEGL-2 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, could experience irreversible or other serious, long-lasting effects or impaired ability to escape." 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 hour without experiencing or developing irreversible or other health effects or symptoms that could impair an individual's ability to take protective action." If, for a particular chemical, the AEGL-2 value is available, it should be used. Otherwise the ERPG-2 value should be used. If there is no AEGL or ERPG value available, then the applicant should adopt a criterion that is comparable in severity to those that have been established for other chemicals.

Section 70.60(b)(2)(i)(A). This paragraph deals with an acute exposure of a worker to a radiation dose of between 0.25 Sv (25 rem) and 1 Sv (100 rem) TEDE.

The exposure of a worker to a radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem) TEDE is considered an intermediate-consequence event. The basis for this choice is the use of 0.25 Sv (25 rem) as an exposure criterion in existing NRC regulations. For example, in 10 CFR 20.2202, "Notification of incidents," immediate notification is required of a licensee if an individual receives "... a total 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 receive a dose in excess of normal occupational exposure limits if a dose of this magnitude does not exceed 5 times the annual dose limits [i.e., 0.25 Sv (25 rem)] during an individual's lifetime. In addition, the EPA's Protective Action Guides (U.S. Environmental Protection Agency, 1992) and NRC's regulatory guidance (Regulatory Guide 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. NCRP 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. However, its use is not intended to imply that 0.25 Sv (25 rem) constitutes an acceptable criterion for an emergency dose to a worker. Rather, this dose value has been proposed in this section as a reference value, which should be used by licensees to determine the level of protection (i.e., items relied on for safety, and measures to assure their continuous availability and reliability) needed to ensure an acceptably low level of risk to workers.

Section 70.60(b)(2)(i)(B). This paragraph deals with an acute exposure of a worker to hazardous chemicals in concentrations between AEGL-2 (ERPG-2) and AEGL-3 (ERPG-3) criteria.

An acute exposure of a worker to hazardous chemicals at concentrations that could cause irreversible health effects (but below concentrations that could cause death or life-threatening effects) is considered an intermediate-consequence event. Two existing standards, AEGL-2 and ERPG-2, can be used to define the concentration level for irreversible health effects [see definitions in �.60(b)(1)(iii)(C), above]. Two additional standards, AEGL-3 and ERPG-3, can be used to define the concentration level for death or life-threatening effects [see definitions in �.60(b)(1)(ii)(B), above] . If, for a particular chemical, the AEGL values are available, they should be used. Otherwise the ERPG values should be used. If there are no AEGL or ERPG values available, then the applicant should adopt criteria that are comparable in severity to those that have been established for other chemicals.

Section 70.60(b)(2)(ii)(A). This paragraph deals with an acute exposure of a member of the public to a radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) TEDE.

The exposure of a member of the public to a radiation dose 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 individual 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 accident." In addition, in the regulation of geologic repository operations, 10 CFR 60.136, states that "...for Category 2 design basis events, no individual located on or beyond any point on the boundary of the preclosure controlled area will receive...a total effective dose equivalent of 5 rem...." A TEDE of 0.05 Sv (5 rem) is also the upper limit of EPA's Protective Action Guides of between 0.01 to 0.05 Sv (1 to 5 rem) for emergency evacuation of members of the public in the event of an accidental release that could result in inhalation, ingestion, or absorption of radioactive materials.

Section 70.60(b)(2)(ii)(B). This paragraph deals with an acute exposure of a member of the public to hazardous chemicals in concentrations between AEGL-1 (ERPG-1) and AEGL-2 (ERPG-2) criteria.

An acute exposure of a member of the public to hazardous chemicals at concentrations that could cause notable discomfort (but below concentrations that could cause irreversible effects) is considered an intermediate-consequence event. Two existing standards, AEGL-1 and ERPG-1, can be used to define the concentration level for notable discomfort. AEGL-1 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, could experience notable discomfort." ERPG-1 is defined as "The maximum airborne concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing other than mild transient adverse effects or perceiving a clearly defined, objectionable odor." Two additional standards, AEGL-2 and ERPG-2, can be used to define the concentration level for irreversible health effects [see definitions in �.60(b)(1)(iii)(C), above]. If, for a particular chemical, the AEGL values are available, they should be used. Otherwise the ERPG values should be used. If there are no AEGL or ERPG values available, then the applicant should adopt criteria that are comparable in severity to those that have been established for other chemicals.

Section 70.60(b)(2)(iii). This paragraph deals with a release of radioactive material to the environment.

The release of radioactive material to the environment outside the restricted area in concentrations that, if averaged over a period of 24 hours, 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 seq., and the National Environmental Policy Act of 1969, et seq.

The value established for the environmental consequence criterion is identical to the NRC Abnormal 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 on an annual basis.) 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 concentrations which, if averaged over a period of 24 hours, 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 B 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 would be about 6.8 mSv (680 mrem), or 5000 x 0.5 mSv/yr x 1 day x 1 yr/365 day. 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 contamination caused by such a release, the contamination could require considerable licensee resources to remediate. For these reasons, NRC considered the existing AO reporting criterion for discharge or dispersal of radioactive material as an appropriate consequence criterion in this rulemaking.

Several existing fuel fabrication licensees have chosen to demonstrate compliance with the public dose limit in 10 CFR 20.1301, using 10 CFR 20.1302(b)(1). However, in these cases, routine operations at the facilities do not release effluents that come anywhere close to approaching the Table 2 values in Appendix B to Part 20. Indeed, routine discharge of heavy metals such as uranium in concentrations that substantially exceed the Table 2 values in water or air effluents would be expected to cause extensive environmental contamination that would be difficult and expensive to remediate. This has been demonstrated by the extensive and expensive decommissioning actions that have been required at former fuel fabrication facilities in the United States (see NRC's "Site Decommissioning Management Plan," NUREG-1444). In addition, SNM-processing licensees would not be expected to use the compliance method in 10 CFR 20.1302(b)(2)(ii) because this is primarily directed at external radiation hazards, whereas the materials released from SNM processing facilities primarily represent internal radiation and chemical hazards. Consequently, there is no need to retain the caveat regarding alternative means of demonstrating compliance with the public dose limit, as found in the AO reporting criterion.

Section 70.60(c). This paragraph deals with the graded level of protection.

This section addresses the level of protection a licensee must provide to ensure an acceptable degree of risk at its facility. That protection must be sufficient to reduce the likelihood of potential accidents to levels commensurate with their consequences. In determining the appropriate level of protection that the licensee must provide, consideration may be given to the inherent likelihood of the accident. By inherent, we mean the likelihood of the accident, assuming no controls are in place. Thus, an accident that is initiated by an unlikely external event may require less protection (provided by the licensee) than an accident, with identical consequence, that is initiated by a more frequent event. For example, suppose a serious fire, with high consequences, could be started as the result of a process deviation that is estimated to occur once per year. The level of protection needed to prevent or mitigate this accident would be greater than that needed to protect against a similar fire resulting from an unlikely external event, such as an earthquake that might occur once in 500 years. Thus, licensees may take credit for inherent "unlikeliness" of an accident in determining the level of protection that needs to be applied.

The goal of applying a graded level of protection is to reduce the likelihood or consequences of accidents⁽⁶⁾ to ensure an acceptable level of risk at the licensee's facility. For each of the high-consequence events identified in the proposed �.60(b), the Commission believes that the occurrence of such an event should be highly unlikely to occur during any given year of plant operation. For each of the intermediate-consequence events identified in the proposed �.60(b), the Commission believes that the occurrence of such an event should be unlikely to occur during any given year of plant operation.

The Commission has decided not to include a quantitative definition of "unlikely" and "highly unlikely" in the proposed rule, because a single definition for each term may not be appropriate. Depending on the type of facility and its complexity, the number of potential accidents and their consequences, which are identified in the ISA, could differ markedly. Thus, even if the permitted likelihood for each event were quantitatively defined, the integrated risk for a given facility would depend on the number of such events that could occur and the consequences of those events. For example, some facilities may have few potential accidents in the "high-consequence" range while others may have many potential accidents in this range. Therefore, to ensure that the overall facility risk is acceptable for different types of facilities, guidelines for interpreting "likely" and "highly unlikely" may need to be adjusted accordingly. To accommodate the potential variation in these guidelines, the Commission believes that the standard review plan is the appropriate document to address these terms. The "Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility," which is being made available with the proposed rule, provides guidelines that can be applied to existing fuel cycle facilities. These guidelines have been selected so as to be consistent with the safety performance goals in the NRC Strategic Plan (NUREG-1614, Vol. 1). The Commission intends to publish standard review plans for different types of facilities licensed by NRC, as the need arises. Appropriate guidelines for such facilities can be addressed in the standard review plans at that time.

Section 70.60(d). This paragraph deals with the safety program.

ISA. The performance of an ISA, and the establishment of measures to ensure the continuous availability and reliability of items relied on for safety, are the means by which licensees are able to demonstrate their ability to provide 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 management system is needed to ensure that, when called on, the items relied on for safety are continuously in place and operating properly.

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 intermediate level, as defined in 10 CFR 70.60(b), the likelihood of such an accident must be shown to be commensurate with the consequences, as required in the proposed 10 CFR 70.60(c). Protection against accidents with consequences below the intermediate level threshold is assumed to be provided by adherence to existing NRC, OSHA, and EPA regulations.

(4) Identify the items relied on for safety (i.e., those items that are relied on to prevent or to mitigate the accidents identified in the ISA). Such items are needed to reduce the likelihood or consequences 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 the proposed 10 CFR 70.60(b).

Management control. Although the ISA plays a critical role in identifying potential accidents and the items relied on for safety, the performance of an ISA will not, by itself, ensure adequate protection. Instead, according to the proposed 10 CFR 70.60(d), an effective management system is needed to ensure that, when called on, the items relied on for safety are continuously available and reliable (i.e., in place and operating properly). Maintenance measures must be in place to ensure the continuous availability and reliability of all hardware relied on for safety. Training measures must be established to ensure that all personnel relied on for safety are appropriately trained to perform their safety functions. Human-system interfaces and safety-related procedures must be developed and implemented to enable personnel relied on for safety to effectively carry out their duties. 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. In addition, quality assurance measures need to be established to ensure that the items relied on for safety and the measures used to ensure their continuous availability and reliability are of sufficient quality. 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. When abnormal events occur, investigations of those events must be carried out 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.

Section 70.62. This section deals with requirements for the performance of ISAs and the filing of ISA results and license applications. These requirements address the question of who should perform ISAs, when they should be performed, and what ISA information should be provided to NRC.

The performance of an ISA would be required of all licensees authorized to possess a critical mass of SNM, that are engaged in one of the following activities: enriched uranium processing; uranium fuel fabrication; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; or any other activity that the Commission determines could significantly affect public health and safety. The Commission believes that possession and processing of SNM in amounts sufficient to constitute a potential for criticality is a reasonable criterion for requiring the performance of an ISA. Licensees meeting this criterion are already subject to criticality monitoring and alarm requirements that ensure an adequate response to a criticality event after it occurs. The performance of an ISA provides the means for licensees to ensure adequate measures are taken to prevent a criticality event (or other high-consequence event) before it occurs. By limiting the requirement for performance of an ISA to licensees engaged in specific activities that involve major chemical or mechanical processing of SNM, the Commission recognizes that these activities involve a higher degree of risk than the activities of licensees who are authorized to possess critical quantities of SNM, but do not perform any mechanical or chemical processing of critical or near-critical quantities of the SNM.

These types of facilities include sub-critical assemblies, where the critical mass of material is fixed in place in such a manner that an inadvertent criticality is not credible; research facilities that are authorized to possess a critical quantity of material, but do not process more than a small fraction of that material at any one laboratory; facilities that are authorized only to store the material; and facilities no longer operating, for which the material is dispersed throughout the facility as residue in walls, floors, or other fixed structures. However, potentially hazardous activities involving cleanup and decommissioning at non-operating facilities would be subject to the ISA requirement.

The proposed rule would require current Part 70 licensees, for whom the rule would be applicable to develop compliance plans and submit them to NRC within 6 months of the effective date of the rule. Each compliance plan would identify the processes that would be subject to an ISA, the ISA approach that would be implemented for each process, and the schedule for completing the analysis of each process. Licensees would be expected to complete their ISAs within 4 years of the effective date of the rule, correct any unacceptable vulnerabilities identified, and submit to NRC the results for evaluation, approval, and incorporation in the license. Pending the correction of any unacceptable vulnerabilities, licensees would be expected to implement appropriate compensatory measures to ensure adequate protection. The process description in the ISA submittal should contain information that demonstrates the licensee's compliance with the criticality monitoring and alarm requirements in 10 CFR 70.24.

Applicants operating existing facilities that could become newly subject to the Commission's authority, such as DOE facilities, would be expected to perform ISAs and submit the results as part of their applications for licenses. The ISA submittals should contain information that demonstrates the licensees' compliance with the criticality monitoring and alarm requirements in 10 CFR 70.24.

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 the proposed 10 CFR 70.60(b). In addition, the initial designs are expected to comply with the criticality monitoring and alarm requirements in 10 CFR 70.24 and the baseline design criteria in the proposed 10 CFR 70.64.

Based on these initial designs, the applicants are expected to perform preliminary ISAs before construction of facilities. If the ISA results show deficiencies in the design, the design should be modified to assure that the items and measures planned to protect against identified accidents are adequate. On the other hand, if the ISA results show that a given item at a given facility is not relied on for safety, or that it does not require full adherence to the baseline criteria, then the facility design may be modified accordingly. The applicant is expected to submit the results of the preliminary ISA, based on the modified design of the facility, to NRC before construction. However, NRC approval is not necessary for the applicant to proceed with construction. The submittal should include the identification of all cases where a deviation from the baseline criteria is proposed, along with a justification for that decision. The submittal of the preliminary ISA for review by NRC provides an opportunity for applicants to get early feedback on the design of their facilities or processes. It is much more cost-effective to correct problems identified at the design stage than after the facility has been constructed.

After construction, but before operation, applicants would be expected to update their ISAs, based on as-built conditions, taking into account the results of the preliminary ISAs, and submit the results to NRC for approval. Any inconsistencies between the results of the updated ISAs and the preliminary ISAs should be identified in the submittals.

Section 70.64. 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 licensees and applicants for a license perform an ISA. The ISA process is applied to existing designs to identify high risks that could warrant additional preventive or mitigative measures. For new facilities or new processes at existing facilities, the proposed rule calls for the performance of the ISA before construction, and the updating of the ISA before beginning operations. However, for new processes and facilities, the Commission recognizes that good engineering practice dictates that certain minimum requirements be applied as design and safety considerations for any new nuclear process or facility. Therefore, the Commission has specified baseline design criteria in �.64 that are similar 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: quality standards and records; natural phenomena hazards; fire protection; environmental and dynamic effects⁽⁷⁾; chemical protection; emergency capability; utility services; inspection, testing, and maintenance; criticality control; and instrumentation and controls. The baseline design criteria do not provide relief from compliance with the safety performance requirements of �.60. The baseline design criteria are generally an acceptable set of initial design safety considerations, which may not be sufficient to assure 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 updated ISA. For such processes and facilities, any design features that are inconsistent with the baseline design criteria should be identified and justified.

Section 70.65. This section deals with the additional content of applications.

There is additional information that would need to be submitted to NRC as part of a license application to demonstrate compliance with the additional requirements that would be established in the proposed new subpart. This information is necessary to determine whether the applicant has provided an adequate level of protection at the facility. In particular, additional information would be needed to demonstrate how the applicant's safety program complies with 10 CFR 70.60(d). This information would include a description of the plant site and structures; the processes analyzed in the ISA; an appropriate summary of the results of the ISA, including the accident sequences, the consequences and likelihoods of such sequences; and the items relied on for safety; and the measures established to ensure the continuous availability and reliability of such items. The plant and process descriptions are needed to fully understand the results of the ISA, including the rationale for choosing the items relied on for safety. The evaluation of the applicant's safety program is a critical element in determining whether the facility is safe and should be issued a license. Finally, the license application, for an operating facility, should include a description of operational events that have occurred during the past 10 years that had a significant impact on the safety of the facility. These events should be addressed in the applicant's ISA to ensure that the range of accident sequences considered in the ISA encompasses actual events that have occurred at the facility.

The license application demonstrates how the applicant intends to meet the requirements of Part 70. The application provides information about the applicant's facility and processes and commitments that ensure the health and safety of workers, the general public and the environment. To ensure confidence that these commitments will be adhered to, and will not be changed without NRC knowledge or approval, the following condition will be inserted in the license: "Authorized use: For use in accordance with the statements, representations, and conditions in the application dated ______, and supplements dated_______. The application may be revised in accordance with the provisions of 10 CFR 70.72." This condition is similar to the ones currently in use. However, it would apply to the entire license application (not just a portion of the application, as was done previously), and would allow changes to be made without prior NRC approval, in accordance with 10 CFR 70.72.

Section 70.66. This section deals with records.

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, accurate, and complete records available for NRC inspection. These records serve two major purposes. First, they can supplement information that has been submitted as part of the license application. For example, applicants would be required to submit the results of their ISAs to NRC for review. However, there may be substantial amounts of supporting material, at the licensed facility, relevant to that submittal, that NRC may wish to review. 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.

Section 70.68. This section deals with additional requirements for the approval of license applications.

In addition to the requirements found in the existing rule (i.e., 10 CFR 70.23 ), the Commission must determine that the requirements in the proposed new subpart, 10 CFR 70.60 through 70.66, will be satisfied.

Section 70.72. This section deals with changes to site, structures, systems, equipment, components, and activities of personnel.

Past incidents at fuel cycle facilities have often resulted from changes not fully analyzed, not authorized by management, or not adequately 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 confidence in the margin of safety at that facility. Any such change needs to be considered and evaluated by the licensee before the change is made. If the licensee evaluates the change, based on its ISA, and finds that it, at most, increases the risk at the facility to a minimal extent, then the licensee may make the change and then notify NRC within 60 days. Otherwise, the licensee would need to request a license amendment and get NRC approval before making the change. In either case, the change should be controlled by the licensee's configuration management system, and appropriate modifications to the license application (including, if applicable, the results of the ISA) should be submitted to NRC. Aside from providing increased confidence in the margin of safety, maintaining the license so that it reflects the current configuration of the facility would facilitate a relatively simple, cost-effective license renewal process. The ability of licensees to make certain changes to their facility without prior NRC approval, as allowed in this proposed requirement, is analogous to existing requirements in 10 CFR 70.32.

Section 70.73. This section deals with the renewal of licenses.

Under the proposed amendments to Part 70, changes to site, structures, systems, equipment, components, and activities of personnel, made by a licensee, would be reflected in the license application, which would be submitted to NRC and incorporated as a condition of the license. This process would establish a "living" license that would be maintained on a current basis. As a result, the license renewal process is expected to be a pro forma activity in which NRC, based on its current knowledge of licensee activities, as reflected in the "living license," would approve the renewal 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. This section deals with additional reporting requirements.

The new requirements that would be incorporated in the proposed changes to Part 70 suggest a revised approach for reporting of events to NRC. This new approach, based on consideration of the consequences of concern established in 10 CFR 70.60(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 scaled according to risk. The new reporting requirements are intended to supplement the requirements in the existing Part 70. A more detailed discussion of the new requirements is found in the discussion of Appendix C to Part 70.

Appendix A. "Acute Exposure Guideline Levels (AEGLs)" This appendix contains the AEGL values, for 1-hour exposures, that have been established by EPA. These values are referenced in 10 CFR 70.60(b).

Appendix B. "ERPG" This appendix contains the ERPG values that have been established by AIHA. These values are referenced in 10 CFR 70.60(b).

Appendix C. "Reportable Events"

To effectively fulfill its responsibilities, NRC needs to be aware of conditions that could result in an imminent danger to the worker or to public health and safety. In the event of an accident, NRC must be able to respond accurately to requests for information by the public and the media. In addition, to the extent possible, NRC needs to be able to provide appropriate assistance to licensees in their efforts to address potential emergencies. 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. Finally, NRC must track 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.

NRC intends to take a graded approach for reporting licensee events, as illustrated in Table 2. According to this approach, licensees would report events based on whether actual consequences have occurred or whether a potential for such consequences exists. The most serious events, and those that must be reported within the shortest timeframe (1 hour) are high-consequence events that have actually occurred. Intermediate-consequence events that have actually occurred should be reported within 4 hours.

Events that could potentially lead to a consequence of concern should also be reported. External conditions, such as a hurricane, tornado, or flood, that could pose a threat to safety at a facility, should be reported within 4 hours. Deviations from safe operating conditions should be reported within a time period that depends on the severity of the potential consequence and whether or not the licensee is able to correct the deviation within the specified period. A deviation from safe operating conditions means that a parameter that is controlled to ensure adequate protection is outside its established safety limits, or that an item relied on for safety is no longer operational or has been degraded so that it cannot perform its intended function. The reporting requirements for deviations from safe operating conditions are intended to be generally consistent with the reporting scheme established under Bulletin 91-01. For example, if a criticality control identified in the ISA is no longer operational, or degraded so that it cannot perform its intended function, that situation should be reported to NRC. If the control cannot be reestablished within 4 hours of discovery, the report should be made before expiration of the 4-hour time period. If the control has been reestablished within 4 hours of discovery, the report should be made within 24 hours. The term "reestablish" is intended to mean that the control identified in the ISA is made operative. Therefore, if a control fails and an ad-hoc control, not identified in the ISA, is established within 4 hours of discovery, a report to NRC would still have to be made before expiration of the 4-hour time period.

Another category of potential events that should 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, or by an unanticipated and unanalyzed set of circumstances. The timeframe for reporting this type of event would depend on how long it takes the licensee to remove the unsafe condition, and restore normal operations. If the licensee were unable to restore normal operating conditions within 4 hours, the report would need to be made before expiration of the 4-hour period. If the licensee were able to remove the unsafe condition and restore normal operations within 4 hours, the report would need to be made within 24 hours.

TABLE 2 Graded Reporting Requirements

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, "Proposed Methods for Regulating Major Materials Licensees," NUREG-1324, Washington, DC, February 1992.

U.S. Nuclear Regulatory Commission/ Occupational Safety and Health Administration (OSHA), "Memorandum of Understanding Between NRC and OSHA; Worker Protection at NRC-Licensed Facilities" (53 FR 43950; October 31, 1988).

U.S. Nuclear Regulatory Commission, "Certification of Gaseous 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.

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 Guide 8.29, 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 Technical Information Service, 5285 Port Royal Road, Springfield VA 22161.

Regulatory Guide 8.29 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, P.O. 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 are available by request from the NRC Public Document Room: "Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility" (Draft NUREG-1520); "Integrated Safety Analysis Guidance Document" (Draft NUREG-1513); and "Example Elements of an ISA Submittal -- Process Descriptions and Accident Analysis Summary."

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 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 continuously available and reliable; and (4) require the inclusion of the safety bases, including the results of the ISA, in the license application. The language, in the proposed rule, that defines an environmental consequence of concern, is relevant to the question of environmental impact. 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 hours, 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. 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 to the public from this action. However, 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 justice, may be submitted to NRC, as indicated under the ADDRESSES heading.

NRC has sent a copy of the environmental assessment and this proposed rule to all State Liaison Officers and requested their comments on the Environmental Assessment. The Environmental Assessment is available for inspection at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, D.C. Single copies of the environmental assessment are available from Richard I. Milstein, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone (301) 415-8149; e-mail: rim@nrc.gov.

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 (OMB) for review and approval of the paperwork requirements.

The public reporting burden for this information collection is estimated to average 70 hours per response, and the recordkeeping burden is estimated to average 500 hours per licensee, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information collection. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collections contained in the proposed rule and on the following issues:

0.Is the proposed information collection necessary for the proper performance of NRC's function? Will the information have practical utility?

1. Is the burden estimate accurate?
2. Is there a way to enhance the quality, utility, and clarity of the information to be collected?
3. 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 information collection, including suggestions for reducing the burden, to the Records Management Branch (T-6-F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail at bjs1@nrc.gov; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0009), Office of Management and Budget, Washington, DC 20503.

Comments to OMB on the information collections or on the above issues should be submitted by (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 an information collection does not display a currently valid OMB control number, 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 costs and benefits of the alternatives considered by the Commission. The draft analysis is available for inspection in the NRC Public Document Room, 2120 L Street N.W. (Lower Level), Washington, D.C. Single copies of the analysis may be obtained from Barry T. Mendelsohn, Office of 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 major nuclear fuel fabrication facilities that are authorized to possess a critical mass of SNM. 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).

Backfit Analysis

NRC has determined that the backfit rule does not apply to this proposed rule; therefore, a backfit analysis 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

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 10 CFR 70.4, the definitions of Acute exposure, Acute exposure guideline levels (AEGLs), Controlled site boundary, Critical mass of SNM, Deviation from safe operating conditions, Double contingency, Emergency response planning guidelines (ERPGs), Hazardous chemicals, Integrated safety analysis (ISA), Items relied on for safety, New process, Results of the ISA, Unacceptable vulnerabilities, and Worker are added, in alphabetical order, as follows:

� 70.4 Definitions.

Acute exposure means a single exposure or multiple exposures occurring within a short time (24 hours or less).

Acute exposure guideline levels (AEGLs) mean chemical concentration levels, established by the National Advisory Committee for Acute Guideline Levels for Hazardous Substances, that, for a defined exposure, would result in anticipated adverse health effects to humans. The following three levels have been established:

(1) AEGL-1 means 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, could experience notable discomfort.

(2) AEGL-2 means 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, could experience irreversible or other serious, long-lasting effects or impaired ability to escape.

(3) AEGL-3 means 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, could experience life-threatening effects or death.

Controlled site boundary means the physical barrier surrounding the facility that is used by the licensee to control access. It may or may not coincide with the property boundary.

Critical mass of 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.

Deviation from safe operating conditions means that a parameter that is controlled to ensure adequate protection is outside its established safety limits, or that an item relied on for safety has been lost or has been degraded so that it cannot perform its intended function.

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.

Emergency response planning guidelines (ERPGs) mean chemical concentration levels, established by the American Industrial Hygiene Association, that, for a defined exposure, would result in anticipated adverse health effects on humans. The following three levels have been established:

(1) ERPG-1 means the maximum airborne concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing other than mild transient adverse effects or perceiving a clearly defined, objectionable odor.

(2) ERPG-2 means the maximum airborne concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or developing irreversible or other health effects or symptoms which could impair an individual's ability to take protective action.

(3) ERPG-3 means the maximum airborne concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or developing life-threatening health effects.

Hazardous chemicals mean substances that are toxic, explosive, flammable, corrosive, or reactive to the extent that they can cause significant damage to property or endanger life if not adequately controlled.

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 site, structures, systems, equipment, components, and activities of personnel that are relied on for safety. As used here, integrated means joint consideration of, and protection from, all relevant hazards, including radiological, criticality, fire, and chemical.

Items relied on for safety means structures, systems, equipment, components, and activities of personnel that are relied on to prevent or to mitigate potential accidents at a facility.

New process means, for a particular licensee, a change in the basic method for processing special nuclear material, where the new method is not currently specifically authorized by the NRC license.

Results of the ISA means the information obtained as a result of performing an ISA. It includes the identification of: (1) the radiological and non-radiological hazards at the facility; (2) the accident sequences that could result from such hazards; (3) the consequence and likelihood of occurrence of each accident sequence; and (4) the items relied on for safety.

Unacceptable vulnerabilities mean deficiencies in the items relied on for safety or the measures used to assure the continuous availability and reliability of such items that need to be corrected to ensure an adequate level of protection as defined in 10 CFR 70.60(c).

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

Ё 70.13a and 70.14 [Redesignated]

5. Sections 70.13a and 70.14 are redesignated as Ё 70.14 and 70.17, respectively.

6. Section 70.15 is added to read as follows:

� 70.15 Nuclear reactors.

The regulations in Subpart H do not apply to nuclear reactors licensed under 10 CFR Part 50.

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."

� 70.22 [amended]

9. In 10 CFR 70.22, paragraph (f) is removed and paragraphs (g) through (n) are redesignated as (f) through (m).

� 70.23 [amended]

10. In 10 CFR 70.23, paragraph (a)(8) is removed, paragraph (b) is removed and reserved, and paragraphs (a)(9) through (a)(12) are redesignated as (a)(8) through (a)(11), respectively.

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

12. 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."

13. 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."

14. The undesignated center heading "MODIFICATION AND REVOCATION OF LICENSES" is redesignated as "Subpart I -- Modification and Revocation of Licenses."

Ё 70.61 and 70.62 [redesignated]

15. Sections 70.61 and 70.62 are redesignated as Ё70.81 and 70.82, respectively.

16. The undesignated center heading "ENFORCEMENT" is redesignated as "Subpart J -- Enforcement."

Ё 70.71 and 70.72 [redesignated]

17. Sections 70.71 and 70.72 are redesignated as Ё70.91 and 70.92, respectively.

18. In Part 70, a new "SUBPART H" (Ё 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

�.60 Safety performance requirements.

(a) Purpose. Each licensee engaged in enriched uranium processing, uranium fuel fabrication, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, mixed-oxide fuel fabrication, scrap recovery, or any other activity that the Commission determines could significantly affect public health and safety, shall provide protection to its workers, the general public, and the environment against radiological (including criticality), chemical, and fire hazards that could result in the adverse consequences identified in paragraph (b) of this section. Consideration must be given to radiological consequences from all causes (including those resulting from fires and hazardous chemicals), and those chemical and environmental consequences that could result from the processing of special nuclear material.

(b) Consequences of concern. Each licensee shall protect against the occurrence of the following high and intermediate adverse consequences that could result from accidents involving the handling, storage, or processing of licensed special nuclear material:

(1) High consequences.

(i) A nuclear criticality;

(ii) Acute exposure of a worker to --

(A) A radiation dose of 1 Sv (100 rem) or greater total effective dose equivalent; or

(B) Hazardous chemicals in concentrations exceeding AEGL-3 (Appendix A) or ERPG-3 (Appendix B) criteria; or

(iii) Acute exposure of a member of the public outside the controlled site boundary to:

(A) A radiation dose of 0.25 Sv (25 rem) or greater total effective dose equivalent;

(B) An intake of 30 mg or greater of uranium in a soluble form; or

(C) Hazardous chemicals in concentrations exceeding AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria.

(2) Intermediate consequences.

(i) Acute exposure of a worker to --

(A) A radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem) total effective dose equivalent; or

(B) Hazardous chemicals in concentrations between AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria and AEGL-3 (Appendix A) or ERPG-3 (Appendix B) criteria; or

(ii) Acute exposure of a member of the public outside the controlled site boundary to --

(A) A radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) total effective dose equivalent; or

(B) Hazardous chemicals in concentrations between AEGL-1 (Appendix A) or ERPG-1 (Appendix B) criteria and AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria; or

(iii) Release of radioactive material to the environment outside the restricted area in concentrations that, if averaged over a period of 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20.

(c) Graded level of protection. Each licensee shall provide a level of protection that is commensurate with the severity of the consequences resulting from credible accidents and the likelihood of any external events (e.g., natural phenomena) assumed to initiate or propagate such accidents. This graded level must apply to the items relied on for safety, identified in paragraph (d)(2)(iv) of this section, and to the measures used to assure their continuous availability and reliability, identified in paragraph (d)(3) of this section. The application of a graded level of protection must assure that --

(1) The occurrence of any of the high consequences identified in paragraph (b)(1) of this section is highly unlikely; and

(2) The occurrence of any of the intermediate consequences identified in paragraph (b)(2) of this section, is unlikely.

(d) Safety program. Each licensee shall establish and maintain a safety program that provides reasonable assurance that the accident consequences identified in paragraph (b) of this section are adequately protected against in accordance with paragraph (c).

(1) Each licensee shall compile and maintain a set of process safety information to enable the performance of an integrated safety analysis (ISA). 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.

(2) Each licensee shall perform an ISA to identify --

(i) All radiological and non-radiological hazards (e.g., chemical, fire, electrical, and mechanical);

(ii) Potential accident sequences caused by process deviations or other events internal to the plant (e.g., fires, explosions, or chemical releases) and credible external events, including natural phenomena (e.g., hurricanes, floods, tornadoes, earthquakes, tsunami, and seiches), fires, explosions, or chemical releases occurring offsite;

(iii) The consequence and likelihood of occurrence of each accident sequence identified pursuant to paragraph (d)(2)(ii) of this section; and

(iv) Items relied on for safety (i.e., structures, systems, equipment, components, and activities of personnel), that are relied on to prevent or mitigate those accidents identified under paragraph (d)(2)(ii) of this section, that exceed the consequences of concern stated in paragraph (b) of this section.

(3) To ensure the continuous availability and reliability of items relied on for safety identified under paragraph (d)(2)(iv) of this section, each licensee shall demonstrate that --

(i) Structures, systems, equipment, and components relied on for safety are designed, constructed, inspected, calibrated, tested, and maintained, as necessary, to ensure the continuous ability to perform their safety functions to satisfy paragraph (c) of this section. Items subject to this requirement include but are not limited to: principal structures of the plant; passive barriers relied on for safety (e.g., piping, glove boxes, containers, tanks, columns, vessels); active systems, equipment, and components relied on for safety; sampling and measurement systems used to convey information about the safety of plant operations; instrumentation and control systems used to monitor and control the behavior of systems relied on for safety; and utility service systems relied on for safety.

(ii) Personnel are trained, tested, and retested, as necessary, to ensure that they understand, recognize the importance of, and are qualified to perform their safety duties to satisfy paragraph (c) of this section;

(iii) Procedures relied on for safety are developed, reviewed, approved, and distributed to ensure that personnel are able to perform their safety duties to satisfy paragraph (c) of this section.

(iv) Human-system interfaces are designed and implemented to ensure that personnel relied on for safety are able to perform their safety duties to satisfy paragraph (c) of this section.

(v) Configuration changes to site, structures, process, systems, equipment, components, computer programs, personnel, procedures, and documentation are managed so that such modifications are reviewed, documented, communicated, and implemented in a systematic, controlled manner to satisfy paragraph (c) of this section.

(vi) All items relied on for safety identified under paragraph (d)(2)(iv) of this section and measures established under paragraphs (d)(3)(i) through (d)(3)(v) of this section must meet quality standards that are commensurate with the importance of the safety functions performed. Management shall establish appropriate quality assurance policies and procedures to ensure that all items relied on for safety perform their safety functions and are continuously available and reliable.

(4) Each licensee shall conduct audits and assessments of its safety program to ensure that an adequate level of protection is maintained at the facility.

(5) Each licensee shall investigate abnormal events and take corrective action to minimize the recurrence of these events.

(6) Each licensee shall establish records that will demonstrate that the requirements of paragraphs (d)(1), (d)(2), (d)(3), (d)(4), and (d)(5) of this section have been met. Each licensee shall maintain these records for the lifetime of the plant.

�.62 Requirements for the performance of ISAs and the filing of ISA results and license applications.

(a) Each applicant for a license under this subpart and each current licensee subject to this subpart shall perform an ISA as described in �.60(d)(2).

(1) Each current licensee shall --

(i) Within 6 months of the effective date of this rule, submit, for NRC approval, a compliance plan that describes the ISA approach that will be used, the processes that will be analyzed, and the schedule for completing the analysis of each process; and

(ii) Within 4 years of the effective date of this rule, perform an ISA in accordance with the compliance plan submitted under paragraph (a)(1)(i) of this section, correct any unacceptable vulnerabilities identified in the ISA, and submit the results of the ISA as part of the license application contents identified in �.65 to NRC, for approval. Pending the correction of any unacceptable vulnerabilities identified in the ISA, the licensee shall implement appropriate compensatory measures to ensure adequate protection. The process description in the ISA submittal must include information that demonstrates the licensee's compliance with the design requirements for criticality monitoring and alarms in �.24.

(2) Each applicant operating a facility that is newly subject to the Commission's authority shall perform an ISA, correct any unacceptable vulnerabilities identified in the ISA, and submit the results of the ISA as part of the license application contents identified in Ё70.22 and 70.65 to NRC, for approval. The process description in the ISA submittal must include information that demonstrates the applicant's compliance with the design requirements for criticality monitoring and alarms in �.24.

(3) Each applicant for a license to operate a new facility or a new process at an existing facility shall --

(i) Initially design the facility or process to protect against the occurrence of the adverse consequences identified in �.60(b), meet the criticality monitoring and alarm requirements of �.24, and meet the baseline design criteria in �.64;

(ii) Perform a preliminary ISA and submit the results to NRC before construction of the facility or process. The results of the preliminary ISA must demonstrate an adequate level of protection, as defined in �.60(c), against occurrence of the adverse consequences in �.60(b). The preliminary ISA submittal shall include facility and process description and design information that demonstrates the applicant's incorporation of the criticality monitoring and alarm requirements in �.24, and the baseline design criteria in �.64. Any proposed relaxation in the application of the baseline design criteria, pursuant to �.64(a), must be identified and justified in the preliminary ISA submittal; and

(iii) Before beginning operations, update the preliminary ISA and correct any unacceptable vulnerabilities identified in the ISA. The updated ISA must be based on as-built conditions and must take into account the results of the preliminary ISA. Any inconsistencies between the results of the updated ISA and the preliminary ISA must be identified.

(A) For new facilities submit the results of the ISA, as part of the license application contents identified in Ё70.22 and 70.65, to NRC for approval.

(B) For new processes submit the results of the ISA and any revisions of the approved license application as part of an application for amendment of the license under �.34.

(b) If the decommissioning of a facility involves potentially hazardous activities such as chemical treatment of wastes, each licensee shall perform an ISA of the decommissioning process, correct any unacceptable vulnerabilities identified in the ISA, and submit the results to NRC for approval before beginning such decommissioning activities.

�.64 Baseline design criteria for new facilities or new processes at existing facilities.

(a) Applicants shall address the following baseline design criteria in the design of new facilities or design of new processes at existing facilities, before performing the preliminary ISA, in accordance with �.62(a)(3)(ii). Applicants shall address these baseline design criteria in establishing minimum requirements for all items in their process design and description, which is provided in the application for a license or license amendment. Licensees shall maintain the application of these criteria unless the preliminary ISA, submitted before construction, pursuant to �.62(b)(3)(iii), 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 established and implemented in accordance with a quality assurance program, to provide adequate assurance that items relied on for safety will satisfactorily perform their safety functions. 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 hazards related to the storage, handling, and processing of licensed nuclear 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, including reliable and timely emergency power to items relied on for safety.

(8) Inspection, testing, and maintenance. The design of items relied on for safety must provide for periodic inspection, testing, and maintenance, to ensure their continued function and readiness.

(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. Features must be incorporated that enhance safety by reducing challenges to items relied on for safety. Where practicable, passive systems and features must be selected over active systems and features, to increase overall system reliability.

�.65 Additional content of applications.

In addition to the contents required by �.22, each application for a license to possess a critical mass of special nuclear material for use in the activities described in �.60(a), must contain --

(a) A description of the applicant's site, structures, and the processes analyzed in the ISA;

(b) A description of the applicant's safety program established under �.60(d), including the results of the ISA and the measures established to ensure the continuous availability and reliability of items relied on for safety; and

(c) For currently operating facilities, a description of operational events, within the past 10 years, that had a significant impact on the safety of the facility.

�.66 Records.

The applicant or licensee shall establish and maintain onsite, readily available for Commission inspection, a system of legible, current, accurate, complete, and easily retrievable records to document application-related and license-related information required by applicable parts of this chapter, Commission action, license condition, and commitments by the applicant or licensee. Records must be retained for the period specified by the applicable parts of this chapter, Commission action, license condition, and commitments made by applicant or licensee. If a retention period is not otherwise specified, these records must be retained until the Commission terminates the license or determines that they are no longer required.

�.68 Additional requirements for approval of license application.

An application for a license to possess a critical mass of SNM will be approved if the Commission determines that the applicant has complied with the requirements of �.23 and Ё70.60 through 70.66.

�.72 Changes to site, structures, systems, equipment, components, and activities of personnel.

(a) Except for a new process, subject to the requirements of �.62(a)(3), any change to site, structures, systems, equipment, components, and activities of personnel must be evaluated by the licensee before the change, to determine whether the change increases the likelihood or consequences of an accident at the facility. The evaluation must be based on the licensee's ISA results, developed in accordance with �.60(d)(2), and other safety program information, developed in accordance with �.60(d)(3), which are part of the license application contents identified in �.65.

(b) A licensee may make a change to site, structures, systems, equipment, components, and activities of personnel, without prior Commission approval, if the change --

(1) Results in, at most, a minimal increase in the likelihood or consequences of an accident previously evaluated in the ISA;

(2) Would not create the potential for an accident different from any previously evaluated in the ISA; and

(3) Is not inconsistent with NRC requirements and license conditions.

(c) For any change authorized under paragraph (b) of this section, the licensee shall submit revised pages to the license application, including any changes in the results of the ISA, to NRC within 60 days of initiation of the change.

(d) For any change that is not authorized under paragraph (b) of this section, the licensee shall file an application for an amendment of its license, as specified in �.34, that authorizes the change. As part of the application for the amendment, the licensee shall perform an ISA of the change and submit any revisions of the ISA and the license application to NRC for approval. The licensee shall also provide, as required by Part 51 of this chapter, any necessary revisions to its environmental report.

(e) The licensee shall maintain records of changes to its facility carried out under paragraph (a) of 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 (b) of this section. These records must be maintained until termination of the license.

�.73 Renewal of licenses.

Applications for renewal of a license must be filed in accordance with Ё 2.109, 70.21, 70.22, 70.33, 70.38, and 70.65. Information provided in applications, including the results of the ISA, must be current, complete, and accurate in all material respects. 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.

�.74 Additional reporting requirements.

(a) Reports to NRC Operations Center.

(1) Each licensee shall report to the NRC Operations Center the events described in paragraphs I, II, and III of Appendix C 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 (1, 4, or 24 hours) for reporting.

(3) The information provided must include a description of the event and other related information as described in paragraph V of Appendix C to Part 70.

(4) Followup information to the reports must be provided until all information required to be reported in paragraph (a)(3) of this section is complete.

(5) Duplicate reports to the Commission are not required for events when the reports are made in compliance with other parts of this chapter, provided that the reports comply with the requirements of this section concerning addressees, information content, and timeliness of filing.

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

(b) Written reports.

(1) Each licensee shall provide a written report to NRC, of the events described in paragraph IV of Appendix C to Part 70, within 30 days of discovery. The written report must contain the information described in paragraph VI of Appendix C to Part 70.

(2) Each licensee who makes a report required by paragraph (a) of this section shall submit a written followup report within 30 days of the initial report. The written report shall contain the information as described in paragraph VI of Appendix C to Part 70.

19. Appendix A to Part 70 is added to read as follows:

Appendix A to Part 70 - Acute Exposure Guideline Level Values* for 1-Hour Exposure Periods

*The values in this appendix are taken from EPA's proposed AEGL values for these chemicals (62 FR 58840; October 30, 1997).

20. Appendix B to Part 70 is added to read as follows:

Appendix B to Part 70 - Emergency Response Planning Guidelines Concentration Levels

The values in this appendix are taken from The AIHA Emergency Response Planning Guidelines and Workplace Environmental Exposure Level Guides Handbook, copyright 1998 by the American Industrial Hygiene Association (AIHA). AIHA recommends use of these values with the full documentation provided in the Emergency Response Planning Guidelines (ERPGs) published annually by AIHA. For further information, contact AIHA at (703) 849-8888.

21. Appendix C to Part 70 is added to read as follows:

Appendix C to Part 70 -- Reportable Safety Events

As required by 10 CFR 70.74, licensees who are authorized to possess a critical mass of special nuclear material shall report the following safety events (see table A-1 of this appendix):

I. Events to be reported within 1 hour of discovery, followed by a written report within 30 days.

(a) An accident from the processing of licensed material that resulted in any of the following consequences:

(1) A nuclear criticality.

(2) Acute exposure of a worker to --

(i) A radiation dose of 1 Sv (100 rem) or greater total effective dose equivalent, or

(ii) Hazardous chemicals in concentrations exceeding AEGL-3 (Appendix A) or ERPG-3 (Appendix B) criteria.

(3) Acute exposure of a member of the public outside the controlled site boundary to --

(i) A radiation dose of 0.25 Sv (25 rem) or greater total effective dose equivalent,

(ii) An intake of 30 mg or greater of uranium in a soluble form, or

(iii) Hazardous chemicals in concentrations exceeding AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria.

II. Events to be reported within 4 hours of discovery, followed by a written report within 30 days.

(a) An accident from the processing of licensed material that resulted in any of the following consequences:

(1) Acute exposure of a worker to --

(i) A radiation dose between 0.25 Sv (25 rem) and 1 Sv (100 rem) total effective dose equivalent, or

(ii) Hazardous chemicals in concentrations between AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria and AEGL-3 (Appendix A) or ERPG-3 (Appendix B) criteria.

(2) Acute exposure of a member of the public outside the controlled site boundary to --

(i) A radiation dose between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) total effective dose equivalent, or

(ii) Hazardous chemicals in concentrations between AEGL-1 (Appendix A) or ERPG-1 (Appendix B) criteria and AEGL-2 (Appendix A) or ERPG-2 (Appendix B) criteria.

(3) Release of radioactive material to the environment outside the restricted area in concentrations that, if averaged over a period of 24 hours, exceed 5000 times the values specified in Table 2 of Appendix B to 10 CFR Part 20.

(b) A deviation from safe operating conditions that has not been corrected within 4 hours and has the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph I(a) of this appendix.

(c) An external condition that poses a threat to the performance of items that are relied on for safety (e.g., site, structures, systems, equipment, components, or activities of personnel). These conditions would include natural phenomena (e.g., hurricanes, floods, tornados, earthquakes), fires, or chemical releases.

(d) A potentially unsafe condition that has not been corrected within 4 hours and that has not been identified or analyzed in the integrated safety analysis (ISA).

III. Events to be reported within 24 hours of discovery, followed by a written report within 30 days.

(a) A deviation from safe operating conditions that was corrected within 4 hours and had the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph I(a) of this appendix.

(b) A deviation from safe operating conditions that has not been corrected within 24 hours and has the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph II(a) of this appendix.

(c) A potentially unsafe condition that was corrected within 4 hours and was not identified or analyzed in the ISA.

IV. Events to be reported in writing, to NRC, within 30 days of discovery.

(a) A deviation from safe operating conditions that was corrected within 24 hours and had the potential, as identified in the ISA, for causing an accident with one or more of the consequences specified in paragraph II(a) of this appendix.

V. Licensee reports to the NRC Operations Center, as required by 10 CFR 70.74(a), shall include, to the extent that the information is applicable and available at the time the report is made, the following:

(a) Caller's name and position title.

(b) Date, time, and location of the event.

(c) Description of the event, including --

(1) Sequence of occurrences leading to the event, including degradation or failure of items relied on for safety.

(2) 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 public, and the environment, including relevant chemical and radiation data for actual personnel exposures (e.g., level of radiation exposure, concentration of chemicals, and duration of exposure).

(4) Items that are relied on to prevent or to mitigate the health and safety consequences, and whether the ability of those items to function has been affected by the event.

(5) For events involving deviations from safe operating conditions, the process parameters that are deviant, the normal operating and safety limits on these parameters, and the current values of these parameters.

(d) External conditions affecting the event.

(e) Additional actions taken by the licensee in response to the event.

(f) Status of the event (e.g., whether the event is on-going or was terminated).

(g) Current and planned site status, including any declared emergency class.

(h) Notifications related to the event that were made or are planned to any local, State, or other Federal agencies.

(i) Issue of a press release by the licensee related to the event that was made or is planned.

VI. Licensee written reports required by 10 CFR 70.74(b) shall consist of a completed NRC Form 366 and shall be forwarded to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001. Each written report must include the following information:

(1) Complete applicable information required by paragraph V of this appendix.

(2) Whether the event was identified in the ISA.

(3) Cause of the event, including all factors that contributed to the event.

(4) Corrective actions taken to prevent occurrence of similar or identical events in the future.

TABLE A-1 GRADING OF REPORTING REQUIREMENTS

TABLE A-1 FOOTNOTES

ATTACHMENT 3

PART 70 AMENDMENT
DRAFT REGULATORY ANALYSIS

July 24, 1998

Table of Contents

• 1.0 Introduction
• 2.0. Statement of the Problem
• 3.0. Objectives
• 4.0. Background
• 5.0 Alternatives
• 5.1 Option 1 Description
• 5.1.1 Option 1a
• 5.1.2 Option 1b
• 5.2 Option 2 Description
• 5.3 Option 3 Description
• 6.0. Value-Impact Analysis
• 6.1 Benefits
• 6.1.1 Increased Confidence in the Margin of Safety
• 6.1.2 Reduction in Frequency and Severity of Accidents
• 6.1.2.1 Onsite Consequences
• 6.1.2.2 Offsite Consequences
• 6.1.3 Reduction in Frequency of Incidents
• 6.2 Cost Impacts
• 6.2.1 Option 1 Costs
• 6.2.1.1 Option 1 Licensee Cost Impacts
• - Licensee Incremental Requirements of Option 1b vs Option 1a
• - Implementation Costs of Option 1b Compared to Option 1a
• - Licensee Operational/Recurring Costs of Option 1b Compared to Option 1a
• 6.2.1.2 Option 1 NRC Cost Impacts
• - NRC Option 1b Implementation Costs
• - NRC Option 1b Operational/Recurring Costs
• 6.2.2 Option 2 Costs
• 6.2.2.1 Option 2 Licensee Cost Impacts
• - Incremental Requirements of Option 2 vs Option 1b
• - Implementation Costs of Option 2 Compared to Option 1b
• - Incremental Operational Cost Impacts Compared to Option 1b
• 6.2.2.2 Option 2 NRC Cost Impacts
• - NRC Option 2 Implementation Costs
• - NRC Option 2 Operational/Recurring Costs
• 6.2.3 Option 3 Costs
• 6.2.3.1 Option 3 Licensee Cost Impacts
• - Incremental Requirements of Option 3 vs Option 1b
• - Implementation Costs of Option 3 Compared to Option 1b
• - Operational/Recurring Costs of Option 3 Compared to Option 1b
• 6.2.3.2 NRC Cost Impacts
• 6.2.4 Summary of Cost Impacts
• 7.0. Decision Rationale

Appendix

• Cost Assumptions and Averaging Approach
• A1 Estimating Cost of Performing an ISA
• A1.1 Labor Hours
• A1.2 Costs per Hour
• A1.3 Number of Systems
• A1.4 Error Sources in Estimates of Performing an ISA
• A2 Estimating Costs of Related Measures
• A3 Estimating Annual Cost of Operations
• A4 New Reporting Costs
• A5 PRA Cost Analysis
• A6 Cost Summaries

PART 70 AMENDMENT
DRAFT REGULATORY ANALYSIS

1.0 Introduction

The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend 10 CFR Part 70, "Domestic Licensing of Special Nuclear Materials," to obtain increased confidence in the margin of safety at major special nuclear material (SNM) facilities. The Commission believes that this objective can be best accomplished through a risk-informed, performance-based regulatory structure that includes: (1) the identification of appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed such criteria; (2) the performance of a comprehensive, structured, integrated safety analysis (ISA), to identify potential accidents at the facility and the items relied on for safety; and (3) the implementation of measures to ensure that the items relied on for safety are continuously available and reliable. In addition, to ensure confidence in the margin of safety, the Commission believes that the safety basis for the facility should be incorporated in the license application.

The proposed rule is, in part, NRC's response in resolution of a Petition for Rulemaking (PRM-70-7) submitted by the Nuclear Energy Institute (NEI). The scope of the proposed rule is limited to licensees (and applicants for a license), authorized to possess a critical mass of SNM, that are engaged in the following activities: enriched uranium fuel fabrication; uranium enrichment; enriched uranium hexafluoride conversion; plutonium processing; mixed-oxide fuel fabrication; scrap recovery; and any other activity that the Commission determines poses a significant potential threat to public health and safety.

The purpose of this Regulatory Analysis is to help ensure that:

• NRC's decision to issue the proposed rule is based on adequate information concerning the need for and consequences of the proposal.
• Appropriate alternatives to regulatory objectives are identified and analyzed.
• No clearly preferable alternative is available to the proposed action.
• The direct and any indirect costs of implementation are justified by its effect on overall protection of the public health and safety.

2.0. Statement of the Problem

Investigation of a potential criticality incident in May of 1991 determined that 10 CFR Part 70 does not address facility changes nor does it address changes of procedures and methods that could affect the safe operation of the facility. Change reviews were found to be handled on a case-by-case basis during the development of license conditions, with some license conditions stated in a manner that promoted the exercise of discretion on the part of the licensee in establishing the need for change reviews.⁽⁸⁾ The investigation found that the licensee's system of criticality safety controls was originally extensive and afforded true defense-in-depth. However, this system of controls deteriorated as operations proceeded and changes accumulated.⁽⁹⁾

This incident prompted the NRC staff to evaluate its safety regulations for licensees that possess and process large quantities of SNM. This evaluation concluded that NRC's existing safety regulations for materials licensees "... focus almost exclusively on radiological safety concerns, practically to the exclusion of process safety and managerial controls."⁽¹⁰⁾ Furthermore, the review found that "... each licensee needs a strong managerial program of controls and hazard assessments to ensure and maintain the level of safety that existed when it received its initial license."⁽¹¹⁾ The evaluation also found that "... hazards analyses or engineering safety analyses of plant systems and components are not routinely performed"⁽¹²⁾ by licensees.

There are a number of weaknesses with the current 10 CFR Part 70:

• It provides neither general design criteria nor performance objectives. Unlike 10 CFR Parts 50 and 72, 10 CFR Part 70 contains no "general design criteria."⁽¹³⁾ This would not be a problem if it contained detailed performance requirements in the manner of 10 CFR Part 61 or of 10 CFR 74.51. Unfortunately, the only safety performance objective mentioned in the current 10 CFR Part 70 is the overly general "protect health and minimize danger to life and property."
  
  
• It does not address clearly which facility changes require a license amendment;⁽¹⁴⁾ does not require management review or audits of changes of procedures and methods; and does not mention managerial controls, including quality assurance. Repeatedly, serious events at licensees' facilities can be traced to: lack of procedures or to failure to follow procedures; poor or no training of staff to conduct assigned duties; insufficient retraining of staff; the staff's conduct of activities without management's knowledge or approval; poor sampling and measurement of health-related, safety-related or environmentally-related media; in some cases, poor sampling and measurement of process streams where the information was not required for material control and accounting purposes, i.e., was not subject to the requirements of 10 CFR 70.57; poor maintenance; a failure by management to follow up on safety-related commitments due to a lack of a safety culture within management, to poor tracking systems and to poor commitment reporting systems; a failure by management to control changes; and a failure to properly audit for management effectiveness and to implement corrective actions when audits did occur.
  
  
• 10 CFR Part 70 contains no explicit requirements for chemical safety, fire safety, and prevention of criticality accidents.
  
  
• 10 CFR Part 70 allows a licensee to continue operating indefinitely past its license expiration date if a renewal application has been received in time. This is referred to as being in "timely-renewal." A licensee in timely-renewal may have little incentive to come to closure on contentious safety issues holding up the license renewal.
  
  
• 10 CFR Part 70 does not emphasize commitments to a safety basis. Section 70.22(a)(7) and (8) require the application to contain descriptions of equipment. facilities and procedures that will be used to protect health and safety. It does not specify that applications contain enforceable commitments. In practice, licensees and applicants for a license or for a license renewal do propose license conditions in Part 1 of their applications. Regulatory Guide 3.52, the Standard Format and Content Guide, specifies a two-part application, with only the first part containing proposed license conditions and the second part containing descriptive material. Licensees frequently have placed important safety information into the non-binding Part 2 of the application. This problem is compounded by the timely-renewal problem.
  
  
• 10 CFR Part 70 does not explicitly address licensee safety assessment. In 70.22(f), it does require plutonium processing and fuel fabrication applicants to include a "description and safety assessment of the design bases of the principal structures, systems and components of the plant," but no similar requirements apply to other SNM applicants. In practice, applicants do include safety analyses, as called for in Regulatory Guide 3.52; however, these do not comprehensively and systematically examine all hazards that could result in accidents of concern to the NRC. NUREG-1324 recommended that the regulation be revised to "require that a hazards analysis be performed for each system and component within each process that contains radioactive material or that serves as a barrier to the release of radioactive materials to an unauthorized location."

3.0. Objectives

The primary objective is to regulate major SNM licensees in an efficient, fair, and effective way, and in a manner that provides NRC with appropriate confidence in the margin of safety at these facilities. A secondary objective is to implement the resolution of a petition for rulemaking (Docket No. PRM-70-7) from NEI, as proposed in SECY-97-137.⁽¹⁵⁾

4.0. Background

On January 4, 1986, a worker lost his life during an accidental release of uranium hexafluoride (UF₆) at a facility regulated under 10 CFR Part 40. A Congressional inquiry⁽¹⁶⁾ into this accident criticized NRC's oversight of chemical hazards at NRC-regulated facilities. As a result of this accident, NRC also established an independent group, the Materials Safety Regulation Study Group (MSRSG), to evaluate regulatory practices at all fuel cycle facilities, including those regulated under Parts 40 and 70. The MSRSG concluded that there was a regulatory implementation gap over hazardous chemicals at NRC-regulated facilities.

As a result of the UF₆ release and the Study Group conclusions, an interagency Memorandum of Understanding (MOU) between NRC and the Occupational Safety and Health Administration was issued on October 31, 1988 (53 FR 433950). This MOU clarified NRC responsibility for chemical hazards resulting from processing of licensed radioactive materials. Although a branch technical position on chemical safety was published in 1989 (54 FR 11590), regulation of chemical hazards associated with processing licensed material has not been incorporated specifically into the licensing requirements of Part 70. The same is true of branch technical positions on fire safety,⁽¹⁷⁾ management controls,⁽¹⁸⁾ and requirements for operation.⁽¹⁹⁾

After a near-criticality incident on May 29, 1991, the NRC formed a Materials Regulatory Review Task Force to identify and clarify regulatory issues that need correction. The Task Force published NUREG-1324, which identified a number of weaknesses in the regulation of fuel cycle facility licensees in such areas as: quality assurance; maintenance; training and qualification; management controls and oversight; configuration management; chemical and criticality safety; and fire protection.

To determine whether the above weaknesses are still a problem today, the NRC reviewed the causes of a number of what it considers serious incidents and precursor events at fuel cycle facilities reported after 1991.⁽²⁰⁾ Serious incidents are those involving harm or serious risk of harm to persons, while precursors are events which place a facility at increased risk of a serious incident. Serious incidents examined included:

There continues to be a set of systemic program deficiencies at fuel cycle licensees that are determined to be consistent causes of serious incidents and precursors. These deficiencies are neither rare nor isolated in the industry.

An action plan for remedying deficiencies identified by NUREG-1324, approved by the Commission,⁽²¹⁾ in addition to calling for improvements in the regulatory base, fostered an approach to license renewals that encouraged inclusion of a commitment to perform an ISA as a condition of the license.

On September 30, 1996, the NRC docketed a petition for rulemaking (Docket No. PRM-70-7) from NEI. The petitioner wrote:

Over the past decade, while the formal requirements of Part 70 have not changed significantly, its application has. Licensees' documentation requirements have evolved significantly and additional requirements on the facilities have been imposed through the inspection and licensing processes. Regulatory predictability and stability associated with licensing and oversight of Part 70 facilities [have] suffered as a result. The industry believes that the ISA⁽²²⁾ requirement to evaluate risks (consequences and frequency) and the graded approach to safety (implementation and assurance), coupled with a backfit provision, would help to promote a stable and effective regulatory environment.

Staff submitted a proposed resolution to PRM-70-7 to the Commission (SECY-97-137) on June 30, 1997. That proposed resolution was endorsed by the Commission in an SRM dated August 26, 1997. Staff's recommended approach to rulemaking included the basic elements of the PRM-70-7, with some modifications. In brief, staff proposed to revise Part 70 to include the following major elements:

The objective of the proposed rulemaking is to establish a risk-informed framework, for regulating major⁽²³⁾ SNM licensees, that provides NRC with increased confidence in the margin of safety. The intent is to establish requirements that strengthen regulatory oversight while minimizing the accompanying regulatory burden.

5.0 Alternatives

The alternatives considered are:

• Option 1 -- no action;
• Option 2 -- the proposed rule and standard review plan (SRP), which are consistent with SECY-97-137 and the SRM of August 22, 1997; and
• Option 3 -- a probabilistic risk analyses (PRA) version of the proposed rule.

These alternatives are described more fully in the following paragraphs.

5.1 Option 1 Description

Two alternatives, resulting in the establishment of two different baselines, are discussed under this option. The first baseline (1a) represents the Part 70 program as required by regulation and prior to imposition of license conditions resulting from the 1993 action plan (no ISAs). The second baseline (1b) reflects the required program under Part 70 with license conditions resulting from the action plan included in most license renewals. Thus, while both alternatives are considered to be "no action," the frame of reference for each is different. This is necessary to accurately reflect the incremental cost/benefit impact of the proposed rule.

5.1.1 Option 1a

Option 1a is a so-called "no-action" alternative that corresponds to the status quo that existed before initial implementation of the 1993 action plan for fuel cycle facilities. This alternative, which ignores the fact that most licensees are now required by license condition to prepare an ISA, is needed because the existing regulations in Part 70 do not explicitly require performance of an ISA. In the timeframe of Option 1a, NRC was criticized in House Report 100 -167 for concentrating on radiological hazards and largely ignoring other hazards. An ISA addresses all hazards, not just radiological hazards.

There are several requirements in the current Part 70 that specifically address public health and safety. Section 70.23, Requirements for the approval of applications, requires, among other things, a determination that the applicant's proposed equipment, facilities, and procedures be adequate to protect health and minimize danger to life or property. Similarly, 10 CFR 70.22 requires the applicant to provide a description of equipment, facilities, and procedures to protect health and minimize danger to life or property. Section 70.22 includes such examples of equipment and facilities as "... handling devices, working areas, shields, measuring and monitoring instruments, devices for the disposal of radioactive effluents and wastes, storage facilities, criticality accident alarm systems, etc." It includes "... procedures to avoid accidental criticality, procedures for personnel monitoring and waste disposal, post-criticality accident emergency procedures, etc." as examples of procedures. However, the descriptions were not necessarily comprehensive nor enforceable license commitments because they were not proposed as, nor incorporated into, the conditions of the licenses. In addition, the existing

Part 70 does not explicitly require fire safety or chemical safety, except that fires and "... any associated chemical hazards directly incident"⁽²⁴⁾ to an accidental release of SNM are required to be considered in emergency planning for responding to accidents. Although "... procedures to avoid accidental criticality" are included as examples of proposed procedures to be contained in the license application, engineered means of preventing accidental criticality, which generally are more reliable than procedural means, and are preferred for criticality safety, are not addressed in the regulation.

For plants involved with plutonium, in addition to the above requirements, 10 CFR 70.22(f) specifically requires:

Each application for a license to possess and use special nuclear material in a plutonium processing and fuel fabrication plant shall contain, in addition to the other information required by this section, a description of the [plant site], a description and safety assessment of the design bases of the principal structure, systems, and components of the plant, including provisions for protection against natural phenomena, and a description of the quality assurance program to be applied to the design, fabrication, construction, testing and operation of the structures, systems, and components of the plant.

A footnote to 10 CFR 70.23(b) notes that for plutonium facilities, "The criteria in appendix B of part 50 of this chapter will be used by the Commission in determining the adequacy of the quality assurance program."

Regulatory Guide 3.52, Standard Format and Content for the Health and Safety Sections of License Renewal Applications for Uranium Processing and Fuel Fabrication, provides the staff position on information that should be included in the application. Because this is a guidance document rather than a regulation, compliance with it is not mandatory. Regulatory Guide 3.52 identifies a two-part license renewal application, i.e., proposed license conditions in Part I and descriptive information (demonstration and performance record) in Part II. The information in Part I is noted to be of major importance to the NRC inspection and enforcement staff and, the Regulatory Guide states that Part I should be written to be inspectable and verifiable. The information in Part II, on the other hand, is stated to be of major importance to the NRC licensing staff, during the review of the license renewal application, and should be written to provide the basis for licensing decisions.⁽²⁵⁾

According to Regulatory Guide 3.52:

In the renewal application, the applicant should analyze the plant in terms of potential hazards and the means, including appropriate margins of safety, employed to protect against these hazards. Sufficient information should be included in Part II to allow the NRC licensing staff to perform independent analyses to confirm conclusions reached by the applicant. These analyses should include but are not limited to (1) the site and its relationship to accidents from natural phenomena, (2) operations involving radiation exposures, releases to the environment, and the application of the principle of as low as is reasonably achievable (ALARA), (3) nuclear criticality safety, (4) operations involving hazardous chemicals, (5) confinement and control of radioactive materials, (6) projected effluent quantities and concentrations and effluent treatment, (7) reliability of the systems essential to safety, (8) prevention and control of fire and explosion, (9) radiological contingency planning, and (10) environmental impact associated with normal operations, abnormal conditions, and accidents.⁽²⁶⁾

The application should contain a safety analysis, including radiation safety and nuclear criticality safety, for each step of the process. The analysis should show how the commitments specified in Part I [of the application] will be met.⁽²⁷⁾

The types of accidents considered and their potential impact on occupational safety and the environment should be summarized.⁽²⁸⁾

However, these analyses did not typically include identification of all the items relied on for safety nor did they comprehensively and systematically address all the hazards, such as chemical and fire hazards, that could cause a release of licensed material.

There is nothing in the current Part 70 that explicitly requires a licensee to notify NRC of changes it makes to its facility and procedures that could make the description in Part II of the application in need of update. As noted by an NRC Incident Investigation Team:

The regulations in 10 CFR [Part] 70 do not address facility changes and changes of procedures and methods; i.e., there is no regulation comparable to that specified in 10 CFR 50.59, 'Changes, tests, and experiments.' Although the regulations in Part 70 do not explicitly address change reviews, they are handled on a case-by-case basis during the development of license conditions.⁽²⁹⁾

5.1.2 Option 1b

Under Option 1b, the actual status quo no-action alternative, NRC would retain the current Part 70 as it is. Licensees required by license condition to perform an ISA would continue to do so. An SRP would be developed, under this alternative, to promote licensing consistency and uniformity and provide standards for the quality and completeness of the ISA. NRC uses SRPs to provide guidance, to the staff, for review and evaluation of license applications. In addition to promoting uniformity and consistency in licensing reviews, SRPs help make information about regulatory reviews widely available and improve communication and understanding of the staff review process. An SRP provides guidance and compliance is not mandatory. The SRP acceptance criteria are not considered the only acceptable positions or approaches. Other positions or approaches that are consistent with the regulations may be proposed by an applicant. Under Option 1b, however, the current regulations are very broad and general (see the discussion in Option 1a, above). This allows licensees and applicants to challenge the need for performing a comprehensive and systematic ISA, for committing to use the ISA to evaluate changes, and for committing to ensure the continuous availability and reliability of the items relied on for safety, as identified in the ISA. The guidance provided in the SRP could be challenged by the absence of explicit regulatory requirements for protection against criticality, and chemical and fire hazards, as well as the absence of explicit requirements for an ISA. Furthermore, there would be no explicit regulatory requirement for configuration management and other management controls necessary to ensure that the licensee makes no changes, deliberate or inadvertent, that would decrease the continuous availability and reliability of items relied on for safety. (The regulatory basis could be said to exist currently in 10 CFR 70.32(b), which states that the Commission may incorporate in any license additional conditions and requirements necessary to protect the public health and safety. However, invoking that provision of the regulation for a generic requirement applicable to all of a class of applicants and licensees should be done through rulemaking.)

Option 1b also includes continuation of reporting criticality events under NRC Bulletin 91-01, Reporting Loss of Criticality Safety Controls, without making this reporting a regulatory requirement or expanding it to include reporting the loss of safety controls other than criticality safety controls.

5.2 Option 2 Description

Option 2 is the NRC's proposal to modify 10 CFR Part 70 by adding a new subpart that addresses the features described in SECY-97-137. This new subpart includes requirements aimed at increasing NRC's confidence in the margin of safety at certain licensed facilities authorized to possess a critical mass of special nuclear material. Option 2 is a risk-informed, performance-based regulatory approach that includes: (1) the identification of appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed such criteria; (2) the performance of an ISA to identify potential accidents at the facility and the items relied on for safety; and (3) the implementation of measures to ensure that the items relied on for safety are continuously available and reliable. In addition, in order to ensure confidence in the margin of safety, the safety bases for the facility, including the results of the ISA, would be incorporated in the license application, and revisions to the safety bases (including revisions to the ISA results) would be required to be provided to NRC.

Also included in Option 2 are new reporting requirements, which are based on consideration of the consequences and likelihood of the risk involved, and are intended to replace and expand on the approach licensees have currently been using for reporting criticality events under Bulletin 91-01. The new approach is generic, i.e., it covers all types of potential incidents (not just criticality incidents) and items relied on for safety identified and described in the ISA, and establishes a time frame for reporting that is scaled according to the risk. The new reporting requirements would supplement the reporting requirements currently in the existing 10 CFR Part 70.

An SRP, which has been developed for the proposed rule and is being made available in conjunction with this rulemaking, would be issued to provide guidance to the staff for the review and evaluation of license applications, renewals, and amendments. The SRP acceptance criteria describe ways of complying, with the new requirements, that are acceptable to NRC. The SRP also serves as regulatory guidance for applicants who need to determine what information should be presented in an application.

To assist license reviewers in determining that the applicant's proposed protection is sufficient to reduce the likelihood of potential accidents to levels commensurate with their consequences, the draft SRP includes a risk matrix of consequence categories and likelihood categories. This matrix shows which combinations the staff would find acceptable.

5.3 Option 3 Description

Option 3 is similar to Option 2, except that licensees would be required to perform the ISA using quantitative risk analyses methodology (e.g. PRAs).

Component or "basic-element" reliability data, however, do not appear to be currently available to perform quantitative ISAs on fuel cycle facilities. These facilities may employ unique equipment for which failure data may not have been kept. In addition to mechanical failures, many activities at fuel cycle facilities have considerable human interaction, the failure of which, considering both acts of commission and acts of omission, is difficult to model. Also, because of the competitive nature of the fuel cycle industry, there is no shared reliability database as there is for the nuclear power industry. Accordingly, the reliability data needed to perform a PRA, one of the methods that may be used when conducting the ISA, would at best be difficult to assemble.

6.0. Value-Impact Analysis

This section of the Regulatory Analysis discusses the benefits and costs of each alternative. Ideally, the benefits would be converted into monetary values, as would any non-cost impacts, such as radiation exposure that could be involved in a rule that required entries into a radiation area for its implementation. The total of benefits and costs would then be algebraically summed to determine for which alternative the difference between the values and impacts was greatest.

However, for this rulemaking, the assignment of monetary values to benefits is not possible because:

• No model exists for assigning a monetary value to the benefit of increased NRC confidence in the margins of safety at the affected facilities.
• Available guidance for Regulatory Analyses provides a monetary conversion for stochastic exposure to radioactivity, but not for injuries and fatalities due to exposure to hazardous chemicals, which are a primary concern at these essentially chemical processing facilities.
• There also are no monetary criteria to use for injuries or fatalities due to high radiation doses from criticality accidents, because the Regulatory Analysis guidelines of $2000 per person-rem "...is not applicable to deterministic health effects, including early fatalities."⁽³⁰⁾
• Furthermore, available estimates of the likelihood and consequences of an accident at any of these facilities are subject to large uncertainties.

While better estimates may be available after the completion of the ISAs being performed by most fuel fabrication facilities as a condition of their last license renewal, non-quantifiable attributes will remain the primary benefits. Subjective judgement still would be required as to which of the alternatives best solves the problems identified in section 2 of this report. Thus in section 6.1 we discuss the benefits of each alternative in a qualitative manner only. In section 6.2 we present estimates of the cost to an average licensee and to the NRC for implementing each alternative. The costs in section 6.2 do not include potential savings in terms of averted worker lives lost, averted injuries, averted offsite contamination and cleanup, and averted incident investigation.

6.1 Benefits

6.1.1 Increased Confidence in the Margin of Safety

The performance, by fuel fabrication and enrichment applicants and licensees, of a comprehensive and systematic hazards analysis, as part of an ISA, together with implementation of any corrective actions identified by the ISA, and associated licensee commitments to maintain the items relied on for safety, are key elements for increasing NRC's confidence in the margin of safety at these facilities. Safety analyses that consider chemical, fire, criticality, and radiation safety separately, as opposed to in an integrated manner, can result in measures that enhance safety in one area but degrade it in another. As an obvious example, water may not be an acceptable fire-suppression medium in a moderator-controlled area. But other examples may not be so obvious. For instance, installation of a drip pan under a valve, to confine radioactive contamination, could constitute a criticality safety concern if its shape was not a safe geometry. The performance of ISAs will significantly improve licensee and NRC knowledge, regarding potential accidents and the items relied on for safety, to prevent or mitigate the consequences of these accidents. Only Options 2 and 3 ensure that: (a) ISAs will be performed by all affected licensees in an acceptable manner; (b) items relied on for safety will be identified and reviewed; (c) continuous reliability and availability of those items will be maintained; and, (d) future changes will not decrease safety at the facilities without NRC review.

Options 2 and 3 would correct the weaknesses identified with the current 10 CFR Part 70 (see section 2 of this Regulatory Analysis). The new section 70.60 would provide explicit safety performance requirements as well as baseline design criteria for new facilities. The risk-informed regulation specifies the consequences of concern and the graded levels of protection that must be provided. Proposed section 70.72 clarifies what changes the facility may make without submitting an amendment application, and ensures that all changes, whether or not an amendment is required, are subjected by the licensee to an appropriate safety review. In section 70.60(d)(3) the rule would require a safety program that includes configuration management controls and quality assurance. It also requires personnel to be trained and tested to ensure they understand the safety features that are relied on to prevent accidents. The required ISA would have to address chemical, fire, and criticality hazards, as well as radiological hazards.

In addition, Options 2 and 3 would mitigate the timely-renewal issue, because the safety features of the license would be kept up to date making it a living license. In addition, the issue of what commitments of the license application are binding in the license will be resolved because the items identified by the ISA as relied on for safety as well as the measures to ensure the continuous reliability and availability of those items will be enforceable license commitments.

The PRA approach (Option 3) would provide additional numerical values associated with the likelihood of accident sequences and would provide a basis for more refined grading of protection, if the data were available to allow the quantitative approach without excessive uncertainty bounds. In addition, with the availability of PRAs, it may be possible, for NRC to quantify the benefits of proposed changes to requirements on these facilities. Thus, any backfit analysis, which the Commission may wish to impose on itself in the future before new staff positions or regulatory requirements could be adopted, could be based on the results of a PRA. Otherwise, backfit analyses would have to be primarily qualitative in nature, which makes implementation difficult. However, on balance, NRC believes that Option 3 would provide only a relatively small benefit compared with Option 2, and Option 3 is beset with problems associated with the unavailability of data and relatively little experience in the chemical industry with quantitative models.

6.1.2 Reduction in Frequency and Severity of Accidents

The processing of SNM at facilities licensed to possess a critical mass of SNM could result in a number of potential accidents with varying consequences. These accidents could include an inadvertent criticality; public or worker intake of uranium or plutonium; public or worker exposure to radiation; and public or worker exposure to hazardous chemicals that are used or generated during the processing of SNM.

6.1.2.1 Onsite Consequences

Deaths of two workers are directly attributable to accidents involving licensed nuclear material.⁽³¹⁾ (In contrast, there have been no deaths, because of licensed radioactive material usages, from accidents at U.S.-licensed reactors.) Additional worker injuries and health concerns have resulted from radiation and chemical exposures resulting from NRC-licensed SNM processing operations.

Options 1b, 2 and 3 have the potential to prevent and mitigate the consequences and reduce the likelihood of accidents, compared with Option 1a, through the correction of any vulnerabilities discovered by licensees in their performance of ISAs. To the extent that they enhance plant personnel awareness of their plant's safety features and measures relied on to ensure the continuous reliability and availability of those features, these options have additional potential to reduce the likelihood of accidents.

Options 2 and 3 would be expected to be more effective than Option 1b in reducing the consequences and likelihood of accidents because they would apply uniformly to all major SNM licensees. Under Option 1b not all licensees have license conditions that require performance of ISAs and there is considerable variability in the license conditions regarding maintenance of the safety features. Furthermore, Option 1b is considerably more limited than Options 2 or 3 in maintaining ISAs as a tool for evaluating facility changes.

6.1.2.2 Offsite Consequences

Accidents at licensed fuel fabrication facilities have resulted in offsite releases of uranium compounds and contamination of offsite property. At least one has involved significant government and licensee effort to track, measure, and account for the material released. The types of accidents that would be of most concern to offsite population are a release of UF₆ to the atmosphere, a major fire resulting in loss of confinement of SNM, or accidents sending SNM or toxic chemicals through the ventilation stacks. As in the case of onsite accidents, Options 2 and 3 offer the greatest potential for reducing opportunities for accidents with significant offsite consequences. Only Options 2 and 3 provide the offsite consequence criteria against which to judge the adequacy of protection.

6.1.3 Reduction in Frequency of Incidents

There have been and continue to be several incidents annually of safety significance. Reporting of these incidents to NRC causes both licensee and NRC resource expenditures to investigate and resolve such incidents. This reporting has value in that it provides the NRC with information needed for it to perform its oversight responsibility and requires a licensee to consider what went wrong and what steps might be needed to prevent a recurrence of this safety degradation, but the trend should be toward fewer incidents happening so that they do not require reporting. Under Option 1b, Bulletin 91-01 requests licensees to report loss of one or more criticality safety controls, but does not mandate those reports and does not address loss of other safety controls. Under Option 1b the NRC's confidence in the margin of safety would remain the same, and the annual number of incidents would also be unchanged.

Reversion to Option 1a, which does not include Bulletin 91-01, would cause a decrease in NRC confidence in the margin of safety. Option 1a would also not require any ISAs, and, therefore:

As a result, more accident precursor incidents could be expected by a reversion to Option 1a.

Options 2 and 3 include a requirement that expands the reporting required by the current Part 70 to include reporting criticality incidents (Bulletin 91-01 incidents) as well as loss of other safety controls. The reporting requirements in these options have been written with consideration of risks associated with the full range of incidents of concern, to ensure that safety incidents in addition to criticality are included, but at the same time, to minimize the burden on licensees of reporting inconsequential or low-risk events. Options 2 and 3 would increase NRC confidence in the margin of safety. They should also lead to a reduction in accident precursor incidents due to the requirement that all major licensees perform ISAs, maintain them and use them to evaluate changes.

6.2 Cost Impacts

This section presents the incremental costs of transition from the baseline (Option 1b) to the proposed rule (Option 2) and from Option 2 to the PRA option (Option 3). It also discusses the sunk cost that was involved in the transition from the pre-1993 action plan (Option 1a) to Option 1b. Details on supporting cost assumptions are discussed in the Appendix.

Most existing licenses for facilities within the scope of the proposed rule (Option 2) contain license conditions that require the performance of an ISA, although not necessarily to the standards that would be established by the proposed rule and the guidance provided by the SRP. To a varying degree, some of the other provisions of the proposed rule and SRP are required by license condition in existing licenses. Following the usual practice for NRC Regulatory Analyses, no credit is given as sunk costs for licensee practices that can be discontinued by the licensee without a license amendment. On the other hand, licensee practices that are commitments included in a license application, provisions of a safety evaluation report (SER), provisions of a license condition, or provisions of a regulation, are considered to be part of the cost baseline (i.e., sunk costs).

The details of the costs are provided below and in the Appendix. A summary of the cost impacts is shown in Table 6.2-1. For licensees that have already implemented a set of license conditions that most nearly approaches the requirements of the proposed rule (Option 2), the range of estimated average incremental costs to implement the proposed rule are about $140,000 to $400,000 one-time costs and $20,000 to $40,000 per year. For those licensees with fewest changes in their license conditions under Option 1b, the per licensee range of estimated average incremental costs to implement Option 2 are about $700,000 to $2,200,000 one-time costs and $150,000 to $230,000 per year.

6.2.1 Option 1 Costs

6.2.1.1 Option 1 Licensee Cost Impacts

- Licensee Incremental Requirements of Option 1b vs Option 1a

Option 1a assumes a reversion to the licensing basis before the action plan was adopted. Incremental changes in requirements due to the action plan (i.e, Option 1b) varies by licensee, but for most licensees (5 of the current 7), included a license condition requiring the performance of an ISA. The standards for the ISA are not defined, and neither are the consequences of concern. Those licensees required by license condition to perform an ISA were all assumed to have to update their design basis documents to as built conditions before beginning the ISA. To varying degrees, Option 1b required establishing or upgrading existing configuration control, quality assurance, training and other measures for ensuring continuous reliability and availability of safety items identified by the ISA. There is considerable nonuniformity in these measures from one licensee to another under Option 1b. Option 1b also includes a license condition requiring 4 of the 7 current licensees to periodically update the demonstration part of their license applications. To account for these individual variations, weighted averages were used for the average costs of licensees already required to perform much of the proposed rule under Option 1b and those licensees currently required to perform little of the proposed rule.

- Implementation Costs of Option 1b Compared to Option 1a

Most of the cost involved in going from Option 1a to the Option 1b baseline has already been expended or is in the process of being expended, and is considered sunk cost. Costs that licensees have already expended or will spend in complying with license conditions on establishing configuration management programs, in updating piping and instrumentation drawings to match as-built and as-modified equipment, including the performance of ISAs, are considered as the licensee sunk implementation costs for no-action baseline Option 1b. They are part of the baseline for this Regulatory Analysis. The licensees who are required to perform an ISA under Option 1b, implement measures to ensure the reliability and availability of items relied on for safety, are estimated to have license conditions costing on average⁽³²⁾ about $700,000 to $2,200,000 per licensee, with variations depending on several factors.

Table 6.2-1 Summary of Incremental Cost Impacts

Another factor affecting average costs is whether or not the license conditions for a licensee required to perform an ISA include associated requirements for implementation of new measures or upgrading of existing measures to assure the reliability and availability of items relied upon for safety. For example, only 2 of the 7 licensees are required to update their quality assurance of items relied on for safety, 3 have additional record keeping requirements, and 4 have new configuration management requirements. Furthermore, additional Option 1b requirements pertaining to staff training and to self-inspection and maintenance of items relied on for safety were imposed on 6 of the 7 licensees, not just the 5 required to perform an ISA.

Those licensees not performing an ISA under Option 1b are assumed to have incurred some incremental costs compared to Option 1a as a result of their last license renewal. These costs are associated with required enhancements or improvements to staff training, configuration management, quality assurance, and similar measures intended to better ensure safe operations. Average implementation costs for such actions for these licensees are estimated to be in the range of $80,000 to $110,000 per licensee.

- Licensee Operational/Recurring Costs of Option 1b Compared to Option 1a

For a licensee with appropriate conditions in its license, the annual operational (recurring) sunk costs of Option 1b include the costs associated with maintaining configuration control, quality assurance, training and other measures for ensuring reliability and availability of safety items identified by the ISA. There are also recurring costs associated with facility changes which will require updating the ISA. In total, these recurring costs are estimated to average about $170,000 to $240,000 per licensee per year for those licensees required by license conditions to perform periodic updates of their ISAs and the demonstration sections of their license applications. Other licensees, with minimal requirements for improving Option 1a measures, are also assumed to expend, on average, about $40,000 per licensee-year more under their existing Option 1b requirements than under Option 1a.

6.2.1.2 Option 1 NRC Cost Impacts

- NRC Option 1b Implementation Costs

Additional NRC implementation costs are assumed to be required to develop an SRP for Option 1b, because the SRP draft that has been developed assumes the proposal of Option 2 is adopted as a regulation. Not having to expend those funds would be a cost savings in Options 2 and 3 relative to the baseline. These savings for Options 2 and 3 compared to Option 1b are estimated to be approximately 1 FTE (full-time equivalent), or about $110,000⁽³⁶⁾.

Under Option 1b, the NRC would incur implementation costs in reviewing ISAs for the five licensees required to performing an ISA and in evaluating the actions taken to better assure the availability and reliability of items relied on for safety. These NRC reviews and evaluations are estimated to require, on average, about 900 to 2000 staff-hours per licensee, or incremental NRC expenditures on the order of $50,000 to $110,000 per licensee for the five licensees performing ISAs under Option 1b.

- NRC Option 1b Operational/Recurring Costs

As discussed below, it is estimated that the NRC will have recurring net savings averaging about $19,000 per year per licensee over the long term under Option 1b compared to Option 1a.

The NRC incurs operational costs with Option 1b compared to Option 1a in reviewing periodic updates to the demonstration sections of the license applications. Four fuel cycle facility licensees are required to provide these periodic updates to the NRC. The review costs are estimated to be about $6,000 per licensee per year.

The NRC also expends additional time reviewing the increased number of event reports submitted by licensees as a result of the Bulletin 91-01 requests (and which are assumed to be part of the overall changes from Option 1a to Option 1b). These additional event report reviews are estimated to cost the agency between $3,600 and $9,000 per year, or between $500 and $1,300 per year per licensee.

On the other hand, the NRC's costs associated with performing license renewal reviews are expected to be reduced for those licensees submitting periodic updates to the demonstration sections of their license applications. With Option 1b, four licensees are required to provided these updates. The estimated savings to the agency from reduced license renewal review expenditures is estimated to be about $25,000 per year per each of the four licensees.

6.2.2 Option 2 Costs

6.2.2.1 Option 2 Licensee Cost Impacts

- Incremental Requirements of Option 2 vs Option 1b

If a licensee were not required to do so for the Option 1b baseline alternative (two of the seven current licensees are not presently required to perform ISAs), Option 2 would include developing and documenting the required ISAs, including the identification of items relied on for safety and measures to ensure their availability and reliability. Those licensees performing an ISA under Option 1b would likely have to upgrade their existing analyses to meet the standards required by Option 2.

The safety of all existing operating licensees is considered to be adequate, and the licensees are considered competent to safely perform operations with SNM. Accordingly, it is expected that the changes in the current safety basis will not be dramatic, but rather a matter of refinement. It is assumed that for some licensees Option 2 would involve merely a review of their existing measures that ensure the reliability and availability of their safety items, while other licensees may have to establish some new, or upgrade existing, measures. Required actions would include:

• Establish or upgrade measures to ensure that items relied upon for safety meet quality standards commensurate with their importance, and establish corresponding policies and procedures.
  
  
• Establish and maintain configuration control to assure that changes to processes and systems are reviewed, documented, communicated and implemented in a manner which satisfies safety requirements.
  
  
• Establish or upgrade any additional measures needed to ensure that items relied upon for safety are designed, constructed, inspected, calibrated, tested and maintained as necessary.
  
  
• Establish or upgrade training programs to ensure that personnel are trained, tested, and retested to assure they recognize and understand safety concerns.
  
  
• Establish records that demonstrate adherence to the foregoing requirements.
  
  
• New reporting requirements. (Option 2 also includes strengthening the event reporting requirements for affected licensees.)

Table 6.2-2 indicates the number of current Part 70 licensees judged likely to incur cost impacts by the foregoing provisions of the proposed rule with Option 2. Also shown are estimates of the relative efforts needed to establish measures or bring existing measures into compliance with the Option 2 requirements. The "relative effort needed to achieve compliance" is indicated as a fraction. A low value indicates that licensees in that group already have measures which are expected to largely satisfy the proposed rule requirements, and that the remaining effort to achieve full compliance is relatively small. A high value (1.0 is the maximum) indicates that existing measures are expected to need substantial improvement to comply with the proposed rule. A value of 1.0 assumes that affected licensees would be given essentially no credit for existing measures, and that an entirely new program would have to be established. The judgments of the relative effort needed to achieve compliance are based on NRC fuel cycle licensing staff suggestions and on comparisons of existing license conditions with the requirements of the proposed rule and with acceptance criteria of the draft Standard Review Plan.

Table 6.2-2 Relative Impact of Proposed Rule Reliability
and Availability Requirements on Affected Part 70 Licensees

- Implementation Costs of Option 2 Compared to Option 1b

Each affected applicant or licensee would incur some implementation costs under Option 2, even if the licensee already had conducted an ISA under Option 1b. One time implementation costs that licensees already required to perform an ISA would expend to go from Option 1b to Option 2 could include upgrading of the ISA to Option 2 standards (e.g., to review the ISA and update it where necessary based on the consequences of concern and other rule and SRP provisions). Weighted average incremental costs for upgrading existing ISAs to Option 2 standards and for measures to ensure reliability and availability of items relied on for safety are estimated at $140,000 to $400,000 per licensee for licensees already required to perform ISAs under Option 1b.

The licensees who have not committed to perform an ISA under Option 1b would have to do so under Option 2. Weighted average costs to perform an ISA and for measures to ensure reliability and availability of items relied on for safety are estimated to range from $700,000 to $2,200,000 per licensee for licensees who had minimal license conditions imposed under Option 1b.

- Incremental Operational Cost Impacts Compared to Option 1b

Once these measures were implemented, the licensees would incur recurring operational costs for maintenance and for periodic updates associated with changes to systems and processes. These costs include updates to ISAs to reflect changes to systems and processes, and recurring costs associated with additional personnel training, maintenance of configuration management, enhanced maintenance, testing, inspection activities, enhanced quality assurance, maintaining design basis information, and similar ongoing activities. In addition, Option 2 includes strengthening the event reporting requirements for affected licensees.

The incremental annual recurring or operational costs per licensee are estimated at $20,000 to $40,000 for an average licensee already required by Option 1b to do much of Option 2 requirements. The average annual cost for other licensees is estimated at $150,000 to $230,000.

6.2.2.2 Option 2 NRC Cost Impacts

- NRC Option 2 Implementation Costs

The NRC's incremental implementation activities under Option 2 would consist of initial evaluations of ISA submittals for those licensees who did not commit to perform an ISA under Option 1b, as well as reviews of revised ISAs for the other licensees. The costs of ISA reviews will depend on the type of ISA results documentation submitted by licensees. Option 2 would require licensees to submit sufficiently detailed ISA summaries, in contrast to the very brief or no submittals that are expected under Option 1b. The sufficiently detailed summaries are expected to reduce NRC staff expenditures of time and effort associated with reviewing ISAs, and reduce the need to spend time at licensee sites reviewing ISAs. For each of the two licensees performing an ISA for the first time, the NRC review and onsite evaluation costs with the ISA summaries are estimated at from $10,000 to $30,000 less than the comparable costs would have been under Option 1b, or an average cost of $40,000 to $80,000 per licensee.⁽³⁸⁾ For the five licensees whose initial ISAs were reviewed under Option 1b, the NRC's review of the revised ISAs under Option 2 is estimated to average about 120 to 360 staff hours, or about $6,000 to $20,000 per licensee.

Associated with the ISA evaluations would be reviews to assess the adequacy of licensee measures to ensure the reliability and availability of items relied upon for safety. These incremental implementation costs are assumed to require about 80 to 120 staff hours, or about $4,000 to $6,500 per licensee for licensees required to perform an ISA under Option 1b and on the order of $20,000 to$30,000 per licensee for the two licensees who did not perform an ISA under Option 1b.

- NRC Option 2 Operational/Recurring Costs

Incremental recurring NRC activities with Option 2 include reviews of ISA updates and reviews of additional licensee event reports expected under Option 2. Costs associated with license renewals are expected to be different with Option 2 compared to Option 1b.

Licensees would be required to submit updates to their ISA results as their ISAs are modified to reflect changes to systems and processes. NRC review of ISA updates for the three licensees not required to provide updates to the demonstration part of their license applications under Option 1b is estimated to cost the NRC about $3700 per licensee per year under Option 2. On the other hand, NRC review of the ISA updates provided by the other four licensees is expected to require less labor effort per review than the update reviews under Option 1b, because the licensee submittals under Option 2 are expected to be more comprehensive, and hence easier to review, than under Option 1b. This is estimated to be a savings of about $2,000 per licensee per year. With a savings of $8000 per year for four licensees and an additional cost of $11,200 per year for three licensees, the net cost to the NRC is $3,200 per year, or a weighted average of $490 per year per licensee.

The increased number of licensee event reports expected with Option 2 are estimated to increase NRC operational costs by $10,000 to $25,000 annually compared to the cost of reviews under Option 1b, or $1,400 to $3,600 per licensee, NRC costs associated with Option 2 license renewal efforts are expected to be reduced compared to those experienced with Option 1b, because all licensees will be required to submit periodic ISA updates and updates to other licensing information. These updates will enable the NRC to better keep abreast of changes made to licensee processes, systems, and facilities on an ongoing basis, which will reduce the review burden for license renewal applications. These savings are estimated to amount to about $18,000 per licensee per year.

6.2.3 Option 3 Costs

6.2.3.1 Option 3 Licensee Cost Impacts

- Incremental Requirements of Option 3 vs Option 1b

Option 3 is identical to Option 2 except that it would require PRA methodology to be used for performance of ISAs. In Option 2, PRA methodology is an option that licensees may elect to use for the performance of ISAs, but are not required to use. In general, NRC would not expect any licensees to elect to use PRA methodology under Option 2.

- Implementation Costs of Option 3 Compared to Option 1b

Option 3 is estimated have many of the same implementation costs as Option 2, but to be considerably more costly than Option 2 because of the PRA requirement. According to the Center for Chemical Process Safety:

Although elements of the CPQRA⁽³⁹⁾ are being practiced today in the [chemical process industry], only a few organizations have integrated this process into their risk management program. 匱he reason that these methods are not in more widespread use is that detailed CPQRA techniques are complex and cost-intensive, and require special resources and trained personnel.⁽⁴⁰⁾

Based on the assumptions discussed in section A5 of the Appendix, the cost increase for implementation of Option 3 compared to Option 1b ranges from $350,000 to $1,400,000 for the average licensee required to perform an ISA under Option 1b and from $500,000 to $3,200,000 for the average licensee not required to perform an ISA under Option 1b.

- Operational/Recurring Costs of Option 3 Compared to Option 1b

Option 3 would have similar incremental operational costs as Option 2, but also additional costs, both because of the requirement to use quantitative ISAs (PRAs) to evaluate changes and additions to facilities and processes and because of the continued need to collect and update reliability data.

6.2.3.2 NRC Cost Impacts

No additional NRC costs or savings are attributed to the incremental requirement from Option 2 to Option 3.

6.2.4 Summary of Cost Impacts

Incremental implementation and operational costs for each alternative are shown in Table 6.2-1 for two "average" licensees, one that was required under Option 1b to perform an ISA and one that was not. The differences in high and low costs for each situation reflect, among other things, differences between AIChE estimates and licensee estimates of the cost of performing an ISA.

For licensees that have already implemented a set of license conditions that most nearly approaches the requirements of the proposed rule (Option 2), the range of estimated average incremental costs to implement the proposed rule are about $140,000 to $400,000 one-time costs and $20,000 to $40,000 per year. For those licensees with fewest changes in their license conditions under Option 1b, the per licensee range of estimated average incremental costs to implement Option 2 are about $700,000 to $2,200,000 one-time costs and $150,000 to $230,000 per year. Option 3 implementation costs are estimated to be considerably higher.

7.0. Decision Rationale

Based on the above analysis, NRC believes that the proposed rule, if adopted, would provide the needed increase in the confidence in the margin of safety, at affected facilities, in the least costly manner.

Regulatory Analysis - Appendix

Cost Assumptions and Averaging Approach

A1 Estimating Cost of Performing an ISA

The cost of performing an ISA was estimated on the basis of three factors, namely, the labor hours to analyze a single complex system, the cost per hour of that labor loaded with overhead factors, and the equivalent number of complex systems to be analyzed. A simple system is estimated to require about one-fourth the effort of a complex system.

A1.1 Labor Hours

With regard to the factor of labor hours per system, the information obtained from licensees implies that most of their ISA efforts to date consisted of HAZOP⁽⁴¹⁾ analyses, and What-If was used to a lesser extent. An evaluation of the total projected ISA effort of one licensee indicated that a split of 2/3 HAZOP, 1/3 What-If may be a reasonable assumption. The labor required to accomplish these analyses can vary widely, depending on the type of analysis performed, the complexity of the target systems, and the number of people making up the evaluation team.

Guidance in the AIChE document on qualitative hazards analysis⁽⁴²⁾ was used to estimate the range in the labor requirements for HAZOP and What-if analyses. The estimate is based on the following assumptions:

• the minimum team size would be 5 people, and the maximum size would be 8 people.
  
  
• the documentation efforts would be performed by only two members of the team.
  
  
• the estimates apply to complex systems.

The results are shown in Table A.

Using the above HAZOP/What-if split with the foregoing "mean" efforts, and noting that not all team members are needed to perform certain of the activities, gives an estimate of 800 labor-hours for analysis of one complex system. This value was used as one basis for estimating ISA efforts.

In addition to the labor effort included above for documenting the ISA, an additional effort by licensees was assumed to be needed for those options requiring the submittal of comprehensive summaries of ISA results to the NRC. The effort to prepare these ISA submittals was estimated to require about two person-weeks (80 hrs) per system to prepare.

A1.2 Costs per Hour

Average industry labor rates for skill categories assumed to be representative of the work required were estimated based partially on information obtained from fuel fabrication licensees. Licensee actions and activities involved in performing work that might be required by alternatives under consideration in this Regulatory Analysis were assumed to be accomplished by two types of work groups. Group 1 could be used to perform analytical efforts which were not overly complex, and could include activities such as creating or revising procedures.

Group 2 would be needed to perform more complex evaluations such as performing ISAs and determining measures needed to assure the reliability and availability of items relied on for safety. Each group was assumed to include management, engineers, and clerical staff. Somewhat different mixes were assumed for each group. For example, Composite Group 1 was assumed to require 15% management, 70% engineering staff, and 15% clerical support, while Composite Group 2 had 15 % management, 75% senior engineering staff, and 10% clerical support. The resulting composite labor rate as generated accounted for basic wages, applicable overheads, fringe benefits, and profit. The resulting loaded labor rates for licensees were $50.50 for Composite Group 1 and $57.00 for Composite Group 2. (These labor rate estimates may be somewhat overestimated, because chemical industry experience applying HAZOP and What-if is that teams need someone trained in the hazards analysis methodology but usually need no management member, only a single engineer, and the balance are typically process operators and maintenance personnel.)

NRC labor rates were derived from NUREG/CR-4627⁽⁴³⁾. These rates were escalated to 1997 dollars using trends established by the Gross National Product Price Deflator index. The resulting NRC labor rate was taken to be $53.40 per hour. Following standard practice for NRC Regulatory Analyses, this rate is not fully burdened, but represents base wages for staff plus an allowance for management efforts and for efforts by support staff.

A1.3 Number of Systems

The third factor in determining the cost of performing an ISA is the number of complex and simple systems at an average facility. A major fuel fabrication facility generally includes the process steps listed in Table B. Following AIChE guidelines, this type of facility can be considered to consist of four complex and six simple systems.

Of the current seven major fuel cycle licensees that would be subject to this rulemaking, four can be characterized as equivalent to the above plant description. One only loads pellets into fuel rods and assembles rods into fuel bundles, so has no complex process systems, and therefore its ISA should require much less effort. Another facility is also primarily involved in mechanical rather than chemical processes, except for wet scrap recovery operations. It is estimated to have about three complex and a dozen simple systems. The seventh current major licensee is estimated to have about 12 complex systems.

The AIChE guidelines indicate that an ISA for a simple system, using HAZOP and What-If analysis, can be performed for about one-quarter of the effort required for a complex system. On average then, it could be assumed that a typical major fuel cycle licensee has the equivalent of about 6 complex systems.

However, information from one major fuel fabrication licensee is that it has 28 systems in its ISA (complexity not specified), which implies a different breakdown than indicated in Table B. Using 28 systems, 18 of which conservatively are assumed to be complex, for the four licensees whose operations may be roughly characterized by Table B, it was estimated that the seven major fuel cycle licensees averaged the equivalent of 15.5 complex systems per licensee. The same licensee provided an estimate of its cost to perform an ISA, from which it was estimated that they used about 1780 labor hours per system. Licensee opinion that the AIChE estimates may be too low was also stated by a second major licensee, who could not provide cost of performing an ISA but did claim that the AIChE labor estimates per system were a factor of three too low.

A1.4 Error Sources in Estimates of Performing an ISA

The AIChE estimates may be somewhat low because they neither include criticality as a hazard nor include any accident analyses that might be necessary. The possibility is also recognized that information provided by licensees could include costs that may not be solely attributable to the performance of an ISA, such as the cost of criticality analyses that would be done even if an ISA was not performed, and the cost of bringing plant diagrams up to date, which we are considering as a cost separate from the ISA. The true costs of performing an ISA probably lies somewhere between these two extremes.

A2 Estimating Costs of Related Measures

In addition to the costs of preparing for, performing, and documenting the ISA, there are several related activities that may have cost impacts. Licensees that expended resources in upgrading measures (e.g., training) under Option 1b requirements, but that were considered not to fully meet the standards to be imposed by Option 2, were assumed to expend the balance of the resources under Option 2 needed to achieve a complete program (i.e., to meet acceptance criteria in the SRP). For example, if a licensee expended 70% of the resources under Option 1b needed to establish a suitable employee training program, that licensee was assumed to expend 30% under Option 2 to achieve a fully compliant program (0.7 + 0.3 = 1.0). The only exception to this approach was that the five licensees performing an ISA under Option 1b were assumed on average to expend 15% of the full ISA development costs under Option 2 to bring their ISAs up to Option 2 standards.

Table C indicates the level of effort estimated for these upgrade or implementation actions. Estimated implementation costs for these activities are also shown.

Most of the activities listed in Table C had their implementation efforts estimated on a per-system basis. The exception is the staff training/retraining. The training efforts assumed that training manuals would be upgraded based on ISA results and that affected staff members would be required to take enhanced training. The number of affected staff members per facility was based on the number of individuals at fuel facilities with measurable doses (see NUREG-1272). Record keeping expenditures assumed that new storage space and new storage equipment (I.e., new filing cabinets, new computer data storage systems) would have to be provided, and were assumed to be dependent on the number of systems characterizing the facility.

The implementation costs to establish or upgrade the measures needed to assure the reliability and availability of items relied on for safety were assumed to affect all licensees to some degree under Option 2, depending on the quality and comprehensiveness of their existing measures. The relative impacts for various licensee groups were noted in Table 6.22. Table D indicates the associated cost ranges for upgrading these existing measures or establishing needed measures.

A3 Estimating Annual Cost of Operations

Operational costs for each alternative were estimated using incremental annual operational costs associated with the alternative. Costs that occur less frequently than annually were prorated to an annual basis, using the assumption of a 20 year remaining plant life. To convert to present value, a discount rate of 7% was used. The 7% discount rate is suggested in NUREG/BR-0058, Rev. 2.

Incremental licensee operational costs associated with alternatives may include maintaining system and process safety information current, retraining and testing personnel, maintaining configuration control records, and updating process safety information. Table E shows the estimated licensee efforts and costs associated with these activities.

In addition, past history indicates that changes are frequently made to systems and facilities licensed under Part 70, or new processes are added to existing facilities. The data accumulated by the NRC over the past several years indicated that, on average, fuel fabrication licensees had roughly five minor modifications per year, and also had the equivalent of two substantial modifications or additions every three years, or about two thirds of a major modification per year. Major modifications require license amendments. The cost of demonstrating the safety of a proposed amendment will possibly be less with an ISA available to help provide a basis for demonstrating safety, but no credit for such savings was taken in this Regulatory Analysis. Table E includes the annual estimated hours for updating ISAs for minor process modifications. The effort needed to update an ISA for these types of modifications was estimated to be about 20% of the effort needed to evaluate a complex system. Thus, the annual ISA updating effort was assumed to be the equivalent of each licensee performing an ISA of slightly more than one complex system.

The estimates provided in Table E do not give credit for existing measures that could partially or completely satisfy the specified requirements. Such existing measures and measures already required by current license conditions could reduce the actual cost impacts to licensees. Accordingly, the estimates in Table E were multiplied by indicated factors to arrive at the cost estimates reported in section 6.2.

The maintenance of ISAs and the requirement to keep licensing basis information current are expected to reduce considerably the effort expended by licensees in preparing license renewal submittals. The NRC currently expends in excess of three staff years in renewing the license of a typical fuel cycle facility. The assumption was made that licensees probably expend about three times this amount in preparing their renewal applications. The assumption was also made that licensee efforts associated with license renewals would be reduced by about a factor of three under the proposed rule conditions compared to the situation that exists today. The value of these savings over the remaining plant life (assumed to be 20 years) is estimated to average a present value of $580,000 to $860,000 per licensee, or about $55,000 to $80,000 (savings) per licensee per year.

Table F summarizes the estimated recurring cost impacts of Option 2 compared to Option 1b. The licensee groups with the lower costs are those that, under Option 1b, are already performing some or all of the required actions called for with Option 2; the converse is true for the licensee groups with the higher cost ranges.

A4 New Reporting Costs

The assumption was made that the issuance of new reporting requirements under Options 2 and 3 would result in event reporting trends analogous to what was experienced with the issuance of Bulletin 91-01. That trend showed a several-fold increase in the number of event reports per year for the first 3-4 years after issuance of the bulletin, and then subsequently decreasing to a level about two and one-half times the number of event reports experienced prior to issuance of the bulletin. The current average number of these reports in recent years has been about 2.1 per licensee-year for major licensees. The estimate of incremental reporting costs assumed that this historical trend will be repeated, starting from the current level of event reports. The number of such events was assumed to be proportional to the number of equivalent complex systems characterizing fuel cycle facilities. To estimate costs, it was further assumed that licensees would expend about one person-week in preparing each event report and responding to NRC inquiries. The resulting average incremental reporting cost is estimated to be in the range of $4,000 to $11,000 per licensee per year (averaged over remaining facility lifetime).⁽⁴⁴⁾

A5 PRA Cost Analysis

It is estimated that implementation of a quantitative ISA based on PRA methodology would be at least 1.5 times more expensive than a qualitative ISA. In addition, the quantitative ISA is assumed to require a reliability data collection effort to support the analysis. The qualitative ISAs already committed to by licensees could be helpful for the PRAs, and credit was given for these commitments. This basis resulted in estimated incremental quantitative ISA costs of $185,000 to $1.1 million per licensee, on average, for licensees performing a qualitative ISA under Option 1b. Licensees not performing an ISA under Option 1b would incur costs, on average, of between $400,000 and $2.4 million per licensee to perform the quantitative analysis. (This is the incremental cost from Option 1b, rather than the incremental cost from Option 2.) In addition, the initial data collection efforts (e.g., failure rates) necessary for PRAs are estimated to cost an additional $60,000 to $160,000 per licensee. Other implementation costs for Option 3 would be the same as those noted for Option 2.

Operational costs would also be higher. The annual data collection efforts are estimated to cost between $2,000 and $6,000 per licensee. For licensees with ISA commitments under Option 1b, the efforts associated with performing quantitative ISA updates are in the range of $50,000 to $120,000 more per licensee annually than those for qualitative ISAs. Licensees without ISA commitments under Option 1b would be expected to expend about $75,000 to $160,000 annually per licensee to update quantitative ISAs.

A6 Cost Summaries

Table 6.2-1 itemizes estimated cost impacts to licensees in transitioning from one option to another. Costs are shown for the transitions from Option 1a to Option 1b (considered to be sunk costs), from Option 1b to Option 2, and from Option 1b to Option 3. Estimates are provided for both implementation and operational/recurring activities. All costs are on a per-licensee basis. Table 6.2-1 provides estimates for two categories of licensees: those which, in the context of the transition being considered, have already been required to implement a license condition that encompasses the proposed requirement to some significant degree, and those which have either not previously had such a license condition or whose implementation of the license condition is expected to need substantial improvement to satisfy the proposed alternative.

As shown in Table 6.2-1, there are large variations in the costs to each licensee, because of variations in licensees processes, variations in the current licensing basis for the licensees, and uncertainties in the cost estimates.

To summarize these cost estimates, the low and high average costs for each cost element were added. In addition, Table G shows total costs to the seven current licenses and "average costs." The values in Table G were rounded off in Table 6.2-1, so as not to imply a high degree of certainty in the estimates.

Table A. AIChE Labor Estimates for Performing a Complex System ISA

d=day, w=week

*Activity typically performed by 2 team members

Table B. Systems Characterizing Typical Full Scope Fuel Fabrication Facilities

Table C. ISA-Related Implementation Activities

Table D. Cost Impacts of Proposed Rule Reliability
and Availability Requirements on Affected Part 70 Licensees

Table E Estimated Incremental Operational Activities Burden Per Licensee Per Year

Table F. Licensee Recurring Cost Impacts of Option 2 Relative to Option 1b

Table G - Incremental Cost Impacts
(In thousands of 1997 dollars)

G-1 Incremental Cost of Option 1.b Compared to 1a

ATTACHMENT 7

NRC CONSIDERS CHANGES TO REGULATIONS
FOR SPECIAL NUCLEAR MATERIAL LICENSEES

The Nuclear Regulatory Commission is considering amending its regulations to provide increased confidence in the safety margin for some licensed facilities that possess and process large quantities of certain types of uranium and plutonium.

The proposed amendments would require affected licensees to analyze their facilities carefully to identify potential accidents. The licensees would have to take actions to reduce the likelihood and effects of the postulated accidents if their consequences could exceed specified limits.

The proposed changes result from an NRC review of its regulations after a fuel fabrication facility nearly had an unintended criticality (i.e., a nuclear chain reaction) in May of 1991. As a result of this review, the NRC concluded that, in order to increase confidence in the safety margin, similar licensees should perform an integrated safety analysis. Such an analysis would identify:

(1) Plant and external hazards and their potential for causing accidents;

(2) Potential accident sequences and their likelihood and consequences;

(3) Structures, systems, equipment, components and activities of personnel relied on to prevent or mitigate potential accidents at the facility.

The regulations would apply to licensees that are authorized to possess a "critical mass" of "special nuclear material" and that are engaged in one of the following activities: enriched uranium processing, uranium fuel fabrication, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, mixed-oxide fuel fabrication, scrap recovery, or any other activity involving a critical mass of special nuclear material that the Commission determines could significantly affect public health and safety.

"Special nuclear material" refers to plutonium, uranium-233, uranium enriched in the isotope 233 or in the isotope 235, and any other material that the Commission determines to be special nuclear material, but does not include natural uranium. The term also refers to any material artificially enriched by any of these materials.

A "critical mass" of special nuclear material contains more than: 700 grams of uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1,500 grams of 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.

Currently the NRC's regulation of licensees authorized to possess special nuclear material concentrates on protecting public health and safety during nuclear activities conducted under normal operations. The proposed amendments would extend NRC's regulatory framework to address explicitly the potential exposure of workers or members of the public to radiation and hazardous chemicals as a result of accidents.

The NRC held public meetings on these issues in May and November of 1995 and again on May 28 and July 15 of this year.

The proposed rule amendments would partly grant and partly deny a petition filed in September 1996 by the Nuclear Energy Institute, which sought changes to Part 70, "Domestic Licensing of Special Nuclear Material," of the NRC's regulations.

The proposed new regulations would require licensees to perform an integrated safety analysis, as described above, which would include identification of the radiological and chemical consequences of credible potential accidents at their facilities. A plan for performing the analysis would have to be submitted within six months of the effective date of the amendments to the regulations, and the analysis would have to be conducted within four years. Results of the analysis would be incorporated into the license.

Licensees further would have to establish a safety program that provides reasonable assurance of protection against accidents that could result in releases of radioactive materials or hazardous chemicals in excess of NRC criteria.

Licensees also would have to ensure that structures, systems, equipment and components relied on for safety are designed, constructed and maintained so they will perform their safety function. Licensee personnel would have to be trained and tested to confirm their qualifications to perform their safety duties. Management would have to establish appropriate quality assurance procedures to ensure that items relied on for safety perform their safety functions and are continuously available and reliable.

Applicants for a license to operate a new facility would have to design that facility to provide an adequate level of protection against accidents. They would also have to perform a preliminary integrated safety analysis before constructing the facility and update the analysis before beginning operations.

Other provisions of the proposed revisions to the regulation are discussed in a Federal Register notice to be issued shortly.

Interested persons are invited to submit comments, within 75 days of the Federal Register notice, to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff. Comments may also be submitted electronically through the NRC web site at http://ruleforum.llnl.gov/cgi-bin/rulemake.

###

ATTACHMENT 8

The Honorable Dan Schaefer, Chairman
Subcommittee on Energy and Power
Committee on Commerce
United States House of Representatives
Washington, DC 20515

Dear Mr. Chairman:

The U.S. Nuclear Regulatory Commission (NRC) has sent to the Office of the Federal Register for publication, the enclosed proposed rule to amend 10 CFR Part 70, concerning domestic licensing of special nuclear material.

The objective of the proposed rule is to increase confidence in the margin of safety at facilities authorized to possess special nuclear material in sufficient quantities to be of criticality concern. The proposed rule would: 1) identify appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed these criteria; 2) require affected licensees to perform an integrated safety analysis (ISA) to identify potential accidents at the facility and the items relied on for safety; 3) require licensees to implement measures to ensure that the items relied on for safety are continuously available and reliable; 4) require the inclusion of the safety bases, including the results of the ISA, in a facility's license application; and 5) allow for licensees to make certain changes to their facilities without prior NRC approval.

The proposed amendments will be subject to a 75-day public comment period.

Enclosure: As stated

cc: Representative Ralph Hall

Dear Mr. Chairman:

The U.S. Nuclear Regulatory Commission (NRC) has sent to the Office of the Federal Register for publication, the enclosed proposed rule to amend 10 CFR Part 70, concerning domestic licensing of special nuclear material.

The objective of the proposed rule is to increase confidence in the margin of safety at facilities authorized to possess special nuclear material in sufficient quantities to be of criticality concern. The proposed rule would: 1) identify appropriate consequence criteria and the level of protection needed to prevent or mitigate accidents that exceed these criteria; 2) require affected licensees to perform an integrated safety analysis (ISA) to identify potential accidents at the facility and the items relied on for safety; 3) require licensees to implement measures to ensure that the items relied on for safety are continuously available and reliable; 4) require the inclusion of the safety bases, including the results of the ISA, in a facility's license application; and 5) allow for licensees to make certain changes to their facilities without prior NRC approval.

The proposed amendments will be subject to a 75-day public comment period.

Enclosure: As stated

cc: Senator Bob Grahams

ATTACHMENT 9

UNITED STATES NUCLEAR REGULATORY COMMISSION

Environmental Assessment and Finding of No Significant Impact
(PRE-DECISIONAL DRAFT)
For
Proposed Amendments to 10 CFR Part 70

July 21, 1998

Division of Fuel Cycle Safety and Safeguards
Office of Nuclear Material Safety and Safeguards
Environmental Assessment and Finding of No Significant Impact

TABLE OF CONTENTS

• Description of the Proposed Action
• Need for the Proposed Action
• Environmental Impacts of Proposed Alternatives
• Environmental Justice
• Finding of No Significant Impact
• List of Agencies and Persons Contacted
• References
• Principal Contributors:

Environmental Assessment and Finding of No Significant Impact
(PRE-DECISIONAL DRAFT)
For
Proposed Amendments to 10 CFR Part 70

Description of the Proposed Action

The Nuclear Regulatory Commission (NRC) is proposing to amend its regulations in 10 CFR Part 70 (Part 70), "Domestic Licensing of Special Nuclear Material," to establish a risk-informed, performance-based framework for regulating special nuclear material (SNM) licensees engaged in uranium processing, uranium fuel fabrication, scrap recovery, or related activities. This action is being taken in response to a Petition for Rulemaking (PRM 70-7) filed by the Nuclear Energy Institute (NEI) in September 1996, as documented in SECY-97-137, "Proposed Resolution to Petition for Rulemaking Filed by the Nuclear Energy Institute." The amendments are intended to provide for increased confidence in the margin of safety at fuel cycle facilities by ensuring that licensees systematically identify items (i.e., structures, systems, equipment, components and personnel activities) necessary for protection of health and environmental safety and ensure that these items remain continuously available and reliable. The revised Part 70 would apply to certain facilities that are authorized to process SNM in quantities sufficient to constitute a critical mass (except reactors and gaseous diffusion plants).

NRC is proposing to add safety performance requirements with the following major elements:

1. Identification of appropriate consequence criteria and items relied on for safety to prevent or mitigate accidents that exceed the established criteria;
   
   
2. Performance of an Integrated Safety Analysis (ISA) to identify potential accidents at the facility and the items relied upon for safety;
   
   
3. Measures to ensure that items relied on for safety are continuously available and reliable;
   
   
4. Inclusion of the safety bases, including the results of the ISA, in the license application; and
   
   
5. Flexibility for licensees to make certain changes to their facilities, based on the results of the ISA, that do not increase risk without prior NRC approval.

The Commission's approach, outlined above, agrees in principle with the NEI's petition, with the modifications described in SECY-97-137. These new requirements would apply to licensees engaged in various activities , listed above, including seven currently operating commercial nuclear fuel cycle facilities in the United States. These facilities are already licensed by NRC and subject to the existing requirements in 10 CFR Part 70.

Need for the Proposed Action

The proposed amendments to Part 70 are necessary to provide for increased confidence in the margin of safety at SNM facilities that possess more than a critical mass of SNM. In general the new requirements are intended to ensure that workers, the general public, and the environment are protected from radiological and certain chemical hazards associated with plant operations. A near-criticality incident at a low enriched fuel fabrication facility in May of 1991 prompted NRC staff to evaluate its safety regulations for large materials licensees. (See NUREG-1324 and NUREG-1450 for additional details.) As a result of this review, the Commission and the staff recognized the need for revision of its regulatory basis for these facilities and, specifically, those possessing a critical mass of special nuclear material. Although licensee programs at existing SNM processing facilities are adequate to protect the public, more than three decades of experience with fuel fabrication and SNM processing in the U.S. has surfaced systemic deficiencies in licensee safety programs, especially in the areas of configuration management, maintenance, quality assurance, and safety analysis. The weaknesses identified with the current Part 70 regulatory framework parallel these deficiencies. That is, the current Part 70 does not require the identifications of items relied on for safety; does not require licensees to address fire and chemical process safety; does not require the prevention of an inadvertent criticality; does not require the reporting of all significant facility changes to NRC; and does not require implementation of most managerial controls, including maintenance and quality assurance. It is not a risk-informed regulation in that no specific performance objectives are established and no systematic safety analysis is required to demonstrate compliance with such objectives.

In summary, the existing regulations do not explicitly require a comprehensive, systematic and integrated analysis to identify hazards, such as criticality, fire, chemical releases, and their potential for causing accidents that could affect workers, the public and the environment. Nor do the existing regulations require the identification of items relied on for safety and the measures to assure their continuous availability and reliability. There is a need, therefore, to revise the existing regulations to include these features so as to provide increased confidence in the margin of safety and in the continuous availability and reliability of the items relied on for safety at licensed facilities. The Commission believes such revisions to Part 70 constitute a risk-informed, performance-based approach in which the items relied on for safety and the measures to assure their continuous availability and reliability are selected commensurate with the risk.

The two primary alternatives to be considered are: 1) Option 1-no-action, and 2) Option 2- the proposed rule revision and development of a standard review plan (SRP). Option 2 is consistent with SECY-97-137 which was approved by the Commission in SRM of August 22, 1997.

Option 1

Option 1 is the status-quo, no action alternative that reflects the current Part 70 requirements with added varying license conditions requiring ISAs in most, but not all, license renewals. In the time frame prior to Option 1, NRC was criticized in House Report 100-167 for concentrating on radiological hazards and largely for ignoring other hazards. Under Option 1, those licensees required to perform an ISA would continue to do so. An SRP could be developed to promote some consistency and uniformity and provide standards for the quality and completeness of the ISA. However, in addition to current inconsistencies among licensees under Option 1, there are other licensees that are not performing ISAs at all. Therefore, an SRP would not bridge this regulatory gap, since requirements are set in the regulations and the current Part 70 lacks such needed requirements.

Option 2

Option 2 is the Commission's proposal to modify 10 CFR Part 70 by adding a new subpart, "Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear Material," that consists of 10 CFR 70.60 to 70.74. This new subpart includes requirements aimed at increasing NRC's confidence in the margin of safety. It will also establish consistency in the manner that affected licensees are regulated. These new requirements, although briefly discussed above, are discussed in detail in the Statement of Consideration and Regulatory Analysis to the proposed Part 70.

Environmental Impacts of Proposed Alternatives

Option 2

The potential environmental impacts of Option 2, the proposed action, are those which arise from the additional effort licensees may require to perform an ISA and implement the safety-related performance requirements⁽⁴⁷⁾, and the benefits to the public health and safety and the environment.Using a risk-informed regulatory framework, the proposed action establishes specific performance objectives and requires licensees to conduct an integrated safety analysis (ISA) to demonstrate compliance with these objectives. Adherence to the new performance objectives, which include the establishment of consequence criteria and corresponding likelihood goals, is expected to lessen potential impacts on workers, members of the public, and the environment from accidents at the SNM processing facilities.

Option 2, the proposed action, has positive effects on environmental protection, i.e., it would decrease the likelihood of worker, public, and environmental exposure to radioactive and hazardous materials as a result of an accident. Specifically, the proposed action would require that licensees do the following:

The benefits of the proposed action in reducing the likelihood of potential accidents and mitigating environmental impacts are real although not readily quantifiable. As discussed in the Regulatory Analysis, the implementation of the proposed action is expected to reduce the frequency and severity of accidents at affected licensed facilities. The reduction should translate into fewer accident-related injuries, fewer exposures to workers, reduced cleanup, and less environmental contamination. Quantification of these benefits was not performed because of the lack of risk information, i.e., baseline data relating to the number, impact, severity, and consequence of accidents, that was available. Therefore, negative and positive impacts in this environmental assessment are assessed qualitatively.

Option 1

The first alternative, Option 1-no action or status quo, does not provide increased confidence in the margin of safety because it fails to provide a risk-informed performance-based regulatory framework. There are no specific performance objectives in the existing rule, and there is no requirement for licensees to perform a safety analysis to identify potential accidents and the items relied on for safety. Further, without such a risk-informed, performance-based regulatory framework and the consistency fostered by the proposed action, a large amount of licensee and NRC resources could be consumed by continuing to implement the existing requirements. The impact of the first alternative is a likelihood of more incidents of environmental significance which could have been anticipated and prevented had proper requirements been in place. Although it is possible that licensees would have already identified the possibility of such accidents and have effective controls in place, this outcome cannot be reliably expected because the regulatory framework is not in place to require such outcomes. Under this option, licensees would have considerable freedom in deciding which accidents are significant and should be protected against, the method of determining which items would be relied on for safety, and which measures would assure the continuous availability and reliability of these items.

Under this no action alternative, the result would be a potentially higher risk of accidents with significant consequences, with additional NRC staff and licensee resources expended for subsequent investigations and enforcement.

Summary

The potential environmental impacts of the proposed action are expected to be positive and are preferable to the no action, status-quo alternative because the proposed action accomplishes the greatest gain in protecting the environment for the administrative resources expended. This conclusion may be summarized from Table 1 below.

TABLE 1: ENVIRONMENTAL IMPACTS OF PROPOSED ALTERNATIVES

Environmental Justice

NRC is committed to complying with Executive Order 12898 -- Federal Actions to Address Environmental Justice in Minority Populations and Low-Income Populations

(EO 12898), dated February 11, 1994, in all its actions. As no significant environmental impacts have been identified, NRC staff has determined that there can be no disproportionately high and adverse effects or impacts on minority or low-income populations. Consequently, further evaluation of environmental justice concerns, as outlined in Executive Order 12898, is not warranted.

Finding of No Significant Impact

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 these proposed amendments, 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 determination of this environmental assessment is that there will be no significant environmental impact from this action. 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 these proposed amendments but somehow were not addressed. NRC uses the following working definition of "environmental justice:" the fair treatment and meaningful involvement of all people, regardless of race, ethnicity, culture, income, or educational level with respect to the development, implementation, and enforcement of environmental laws, regulations, and policies.

List of Agencies and Persons Contacted

Nuclear Energy Institute
General Electric Company
Westinghouse Electric Company
U.S. Department of Energy

References

NUREG-1324, Proposed Method for Regulating Major Material Licensees, USNRC, February 1992.

NUREG-1450, Potential Criticality Accident at the General Electric Nuclear Fuel and Component Manufacturing Facility, May 29, 1991, USNRC, August 1991.

Draft Regulatory Analysis for Proposed Revisions to 10 CFR Part 70, USNRC, 1998

Draft Statement of Consideration for Proposed Revisions to 10 CFR Part 70, 1998

Principal Contributors:

FCSS staff

1. The results of this review were documented in NUREG 1324, "Proposed Method for Regulating Major Materials Licensees."

2. A worker, in the context of this rulemaking, 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 individual who is subject to an occupational dose as in 10 CFR 20.1003).

3. TEEL-2 is defined as the maximum airborne concentration below which it is believed that nearly all individuals could be exposed for up to 1 hour without experiencing or developing irreversible or other health effects or symptoms which could impair an individual's ability to take protective action.

4. The proposed rule does not address chemical and radiological consequences to workers and members of the public resulting from routine operations. These consequences are covered in other regulations (i.e., 10 CFR Part 20 and 29 CFR Part 1910).

5. Three levels of consequences are defined for each chemical (AEGL-1, AEGL-2, and AEGL-3) for four different exposure times: 30 minutes; 1 hour; 4 hours; and 8 hours. The AEGL value for a 1-hour exposure is chosen for consistency with the definition of ERPG.

6. For exposures of workers or members of the public to radioactive or hazardous chemical materials during normal operations, adherence to the existing requirements of 10 CFR 20 and 29 CFR 1910 should be sufficient to protect the public health and safety.

7. 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., under normal operating conditions, expected accident conditions, etc.). These conditions could include high temperatures, or a corrosive environment. It could also include dynamic changes in surrounding conditions caused by an accident (e.g., the bursting of a high-pressure pipe).

8. Discussed on page 12-4; NUREG-1450, Potential Criticality Accident ..., May 29, 1991; published August, 1991.

9. Ibid., page 7-16.

10. NUREG-1324, Proposed Method for Regulating Major Materials Licensees; published February, 1992; page 17.

11. Ibid., page 18.

12. Ibid., page 27.

13. Ibid., pages 17 and 30.

14. NUREG-1450, page 7-17.

15. Staff Requirements Memorandum, SECY-97-137 - Proposed Resolution to Petition for Rulemaking Filed by the Nuclear Energy Institute, August 22, 1997.

16. NRC's Regulation of Fuel Cycle Facilities: A Paper Tiger, Eighth Report by the Committee on Government Operations, June 18, 1987.

17. Branch Technical Position on Fire Protection for Fuel Cycle Facilities, published in the Federal Register (54 FR 11595-98) dated March 21, 1989. See also NRC Information Notice 92-014, U Oxide Fires at Fuel Cycle Facilities, and draft Regulatory Guide DG-3006, Standard Format & Content For Fire Protection Sections of License Applications for Fuel Cycle Facilities, issued for comment April 30, 1993.

18. Branch Technical Position on Management Controls/Quality Assurance for Fuel Cycle Facilities, published in the Federal Register (54 FR 11591-92) March 21, 1989.

19. Branch Technical Position on Requirements for Operation for Fuel Cycle Facilities, published in the Federal Register (54 FR 11591-92) March 21, 1989.

20. Updated from Attachment 3 (Regulatory Concerns from Precursor Events at Fuel Cycle Facilities) to Improving the regulation of Fuel Cycle Facilities: Overview, distributed at the NRC Public Workshop on Improving NRC's Regulation of Fuel Cycle Facilities, November, 30, 1995.

21. Staff Requirements Memorandum (SRM) on action plan for fuel cycle facilities (SECY-93-128), dated June 7, 1993.

22. The Petition uses ISA to stand for integrated safety assessment. NRC prefers the term integrated safety analysis.

23. Major SNM licensee, in the context of this rulemaking, means, in general, a licensee whose approved activity involves mechanical or chemical processing of critical quantities of SNM. See the scope of the proposed rule for more detail.

24. 10 Section CFR 70.22(I)(1)(ii).

25. Regulatory Guide 3.52, Revision 1, November, 1986, page vii.

26. Ibid., page viii.

27. Ibid., page 29 (Section 15.2).

28. Ibid., page 30 (Chapter 16).

29. NUREG-1450

30. NUREG-1530, Reassessment of NRC's Dollar per Person-Rem Conversion Factor Policy, December, 1995. NUREG-1530 explains that applying cost to non-stochastic fatalities is inconsistent with the Commission's Safety Goal Policy wherein the Commission made clear that no death will ever be "acceptable" in the sense that the Commission would regard it as a routine or permissible event.

31. One death from a criticality at a licensed SNM scrap recovery plant, July 24, 1964, and one from the hydrogen fluoride vapor cloud resulting from release of UF₆ at Sequoyah Fuels Gore, Oklahoma, conversion plant, January 4, 1986.

32. In addition to variation in the average cost per licensee, individual licensees can expect to have cost variations about an average.

33. 1997 dollars.

34. Savings are indicated as negative values, shown in parentheses.

35. No difference in NRC cost was estimated for Option 3 versus Option 2.

36. The NRC labor rates used in this Regulatory Analysis are discussed in the "Costs per Hour" portion of the Appendix to this Regulatory Analysis.

37. Replacement components are required to be of the correct design and materials.

38. It is assumed that reviews of ISAs prepared under Option 1b are completed prior to implementation of Option 2. Otherwise, the NRC cost of reviews would show a savings of $4,000 to $10,000 for each of the licensees preparing ISAs under Option 1b.

39. The Center for Chemical Process Safety states, "The term 'Chemical Process Quantitative Risk Analysis' (CPQRA) is used to emphasize the unique character of this methodology as applied to the [chemical process industry]." For the purposes of this Regulatory Analysis, the more familiar term PRA has been used for chemical process quantitative risk analysis.

40. Guidelines for Chemical Process Quantitative Risk Analysis, Center for Chemical Process Safety of the American Institute of Chemical Engineers, 1989, page xvii.

41. A description of HAZOP and What-If analysis methodologies may be found in the draft NUREG-1513, Integrated Safety Analysis Guidance Document, which is included in this rulemaking package and is available at the NRC Public Document Room. A more detailed description is available in Guidelines for Hazard Evaluation Procedures, Second Edition with Worked Examples, Center for Chemical Process Safety of the American Institute of Chemical Engineers (AIChE), 1992. This is one of the chemical industry references cited by the Occupational Safety and Health Administration in its rulemaking on Process Safety Management rulemaking (10 FR 6356, February 24, 1991.)

42. Ibid.

43. NUREG/CR-4627, Rev.2, Generic Cost Estimates; Science and Engineering Associates, February 1992.

44. Event reporting is assumed to increase by a factor of about 5 over baseline values for the first 3-4 years after the new requirements are issued, and then to about 2.5 times the pre-change level for the balance of the facility life. Thus, the reporting expenditures are not constant over the remaining life of a facility. Averaging over remaining facility life is a way of presenting the equivalent annual costs without getting into the complexity of the early year costs versus the later year costs.

45. The estimated per licensee costs in this table account for cost differences due to differences in the number of systems assumed for affected facilities. The range does not account for uncertainties in the individual estimates. The labor efforts and costs shown do not give credit for existing measures to which licensees may already be committed. Adjustments for sunk costs for existing commitments are discussed in section 6.2.1.

46. Replacement components are required to be of the correct design and materials.

47. Administrative burdens associated with the proposed revisions to Part 70 are discussed in detail in the Regulatory Analysis of the rule.