Domestic Licensing of Special Nuclear Material; Possession of a Critical Mass of Special Nuclear Material

NUCLEAR REGULATORY COMMISSION

10 CFR Part 70

RIN 3150-AF22

Domestic Licensing of Special Nuclear Material; Possession of a

Critical Mass of Special Nuclear Material

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to

amend its regulations governing the domestic licensing of special

nuclear material (SNM) for licensees authorized to possess a critical

mass of SNM, that are engaged in one of the following activities:

enriched uranium processing; fabrication of uranium fuel or fuel

assemblies; uranium enrichment; enriched uranium hexafluoride

conversion; plutonium processing; fabrication of mixed-oxide fuel or

fuel assemblies; scrap recovery of special nuclear material; or any

other activity involving a critical mass of SNM that the Commission

determines could significantly affect public health and safety or the

environment. The proposed amendments would identify appropriate

consequence criteria and the level of protection needed to prevent or

mitigate accidents that exceed these criteria; require affected

licensees to perform an integrated safety analysis (ISA) to identify

potential accidents at the facility and the items relied on for safety

necessary to prevent these potential accidents and/or mitigate their

consequences; require the implementation of measures to ensure that the

items relied on for safety are available and reliable to perform their

function when needed; require the inclusion of the safety bases,

including a summary of the ISA, with the license application; and allow

for licensees to make certain changes to their safety program and

facilities without prior NRC approval.

DATES: The comment period expires October 13, 1999. 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.

of the ISA, with the license application; and allow

for licensees to make certain changes to their safety program and

facilities without prior NRC approval.

DATES: The comment period expires October 13, 1999. Comments received

after this date will be considered if it is practical to do so, but,

the Commission is able to ensure consideration only for comments

received on or before this date.

ADDRESSES: Submit comments to: Secretary of the Commission, U.S.

Nuclear Regulatory Commission, Washington, DC, 20555-0001, Attention:

Rulemakings and Adjudications Staff.

Deliver comments to: 11555 Rockville Pike, Rockville, Maryland,

between 7:30 a.m. and 4:15 p.m. on Federal workdays.

You may also provide comments via NRC's interactive rulemaking

website through the NRC home page (http://www.nrc.gov). From the home

page, select ``Rulemaking'' from the tool bar at the bottom of the

page. The interactive rulemaking website can then be accessed by

selecting ``Rulemaking Forum.'' This site provides the ability to

upload comments as files (any format), if your web browser supports

that function. For information about the interactive rulemaking

website, contact Ms. Carol Gallagher by telephone at (301) 415-5905 or

e-mail [email protected].

FOR FURTHER INFORMATION CONTACT: Theodore S. Sherr, Office of Nuclear

Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,

Washington, DC, 20555-0001, telephone (301) 415-7218; e-mail

[email protected].

SUPPLEMENTARY INFORMATION:

I. Background

II. Description of Proposed Action

I. Background

A near-criticality incident at a low enriched fuel fabrication

facility in May 1991 prompted NRC to review its safety regulations for

licensees that possess and process large quantities of SNM. [See NUREG-

1324, ``Proposed Method

ton, DC, 20555-0001, telephone (301) 415-7218; e-mail

[email protected].

SUPPLEMENTARY INFORMATION:

I. Background

II. Description of Proposed Action

I. Background

A near-criticality incident at a low enriched fuel fabrication

facility in May 1991 prompted NRC to review its safety regulations for

licensees that possess and process large quantities of SNM. [See NUREG-

1324, ``Proposed Method

for Regulating Major Materials Licensees'' (U.S. Nuclear Regulatory

Commission, 1992) for additional details on the review.] As a result of

this review, the Commission and the staff recognized the need for

revision of the regulatory base for these licensees, especially for

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 1995. The Nuclear Energy Institute (NEI)

explained, to the Commission, industry's position on the need for

revision of NRC regulations, in 10 CFR Part 70, at a July 2, 1996,

meeting, and in a subsequent filing of a Petition for Rulemaking (PRM-

70-7) in September 1996. NRC published in the Federal Register a notice

of receipt of the 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

er 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 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 PRM, with some

modification. In brief, staff's proposed resolution was to revise Part

70 to include the following major elements:

(1) Performance of a formal ISA, that would form the basis for a

licensee's safety program. This requirement would apply to all licensed

facilities or activities, subject to NRC regulation, that are

authorized to possess SNM in quantities sufficient to constitute a

potential for nuclear criticality (except power reactors and the

gaseous diffusion plants regulated under 10 CFR Part 76);

(2) Establishment of criteria to identify the adverse consequences

that licensees must protect against;

(3) Inclusion of the safety bases in a license application (i.e.,

the identification of the potential accidents, the items relied on for

safety to prevent these accidents and/or mitigate their consequences,

and the measures needed to ensure the availability and reliability of

these items);

(4) Ability of licensees, based on the results of an ISA, to make

certain changes without NRC prior approval; and

ion of the safety bases in a license application (i.e.,

the identification of the potential accidents, the items relied on for

safety to prevent these accidents and/or mitigate their consequences,

and the measures needed to ensure the availability and reliability of

these items);

(4) Ability of licensees, based on the results of an ISA, to make

certain changes without NRC prior approval; and

(5) Consideration by the Commission, after licensees' initial

conduct and implementation of the ISA, of a qualitative backfitting

mechanism to enhance regulatory stability.

In an SRM dated August 22, 1997, the Commission ``. . . approved

the staff's proposal to revise Part 70'' and directed the NRC staff to

``. . . submit a draft proposed rule . . . by July 31, 1998.''

A draft proposed rule was provided to the Commission in SECY-98-

185, ``Proposed Rulemaking--Revised Requirements for the Domestic

Licensing of Special Nuclear Material,'' dated July 30, 1998. The draft

proposed rule reflected the approach recommended in SECY-97-137. In

particular, the safety basis for a facility, including the ISA results,

would be submitted as part of an application to NRC, for review, and

incorporated in the license. Also in SECY 98-185, the staff recommended

that a qualitative backfit mechanism should be considered for

implementation only after the safety basis, including the results of

the ISA, is established and incorporated in the license, and after

licensees and staff have gained experience with the implementation of

the ISA requirement.

In response to SECY-98-185, the Commission issued an SRM dated

December 1, 1998, which directed the staff not to publish the draft

proposed rule for public comment. Instead, the Commission directed the

staff to obtain stakeholder input and revise the draft proposed rule.

In that SRM, the Commission also directed the staff to:

gained experience with the implementation of

the ISA requirement.

In response to SECY-98-185, the Commission issued an SRM dated

December 1, 1998, which directed the staff not to publish the draft

proposed rule for public comment. Instead, the Commission directed the

staff to obtain stakeholder input and revise the draft proposed rule.

In that SRM, the Commission also directed the staff to:

(1) Decide what is fundamental for NRC's regulatory purposes for

inclusion as part of the license or docket and what can be justified

from a public health and safety and cost-benefit basis, and assure that

Part 70 captures for submittal those few significant changes that

currently would require license amendments;

(2) Require licensees/applicants to address baseline design

criteria and develop a preliminary ISA for new processes and new

facilities;

(3) Justify, on a health and safety or cost-benefit basis, any

requirement to conduct a decommissioning ISA;

(4) Require that any new backfit pass a cost-benefit test, without

the ``substantial'' increase in safety test;

(5) Require the reporting of certain significant events because of

their potential to impact worker or public health and safety;

(6) Clarify the basis for use of chemical safety and chemical

consequence criteria, particularly within the context of the Memoranda

of Understanding with the Occupational Safety and Health Administration

(OSHA) and other government agencies;

(7) Critically review the Standard Review Plan (SRP) to ensure that

by providing specific acceptance criteria, it does not inadvertently

prevent licensees or applicants from suggesting alternate means of

demonstrating compliance with the rule; and

within the context of the Memoranda

of Understanding with the Occupational Safety and Health Administration

(OSHA) and other government agencies;

(7) Critically review the Standard Review Plan (SRP) to ensure that

by providing specific acceptance criteria, it does not inadvertently

prevent licensees or applicants from suggesting alternate means of

demonstrating compliance with the rule; and

(8) Request input on how applicable ISA methodologies should be

employed in the licensing of new technologies for use within new or

existing facilities.

As directed in the SRM, stakeholder input was solicited and

obtained at public meetings held in December 1998 and January and March

1999. A website was established to facilitate communication with

stakeholders and to solicit further input. The nuclear industry

submitted comments by letters and postings on the website. This revised

proposed rule incorporates much of the December 1, 1998 SRM direction

and reflects language responsive to many of the comments received. It

appears that most of the major concerns with the earlier draft proposed

rule have been resolved.

II. Description of Proposed Action

The proposed rule grants the NEI September 1996 PRM in part and

modifies the petitioner's proposal as indicated in the following

discussion.

The Commission is proposing to modify Part 70 to provide increased

confidence in the margin of safety at certain facilities authorized to

process a critical mass of SNM. The Commission believes that this

objective can be best accomplished through a risk-informed and

performance-based regulatory approach that includes:

(1) The identification of appropriate risk levels, considering

consequence criteria and the level of protection needed to prevent

accidents that could exceed such criteria;

(2) The performance of an ISA to identify potential accidents at

the

facility and the items relied on for safety;

can be best accomplished through a risk-informed and

performance-based regulatory approach that includes:

(1) The identification of appropriate risk levels, considering

consequence criteria and the level of protection needed to prevent

accidents that could 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 available and reliable to perform their function when

needed;

(4) The inclusion of the safety bases, including the ISA summary,

in the license application; and

(5) The allowance for licensees to make certain changes to their

safety program and facilities without prior NRC approval.

The Commission's approach agrees in principle with the NEI

petition. However, in contrast to the petition's suggestion that the

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 broader 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;

fabrication of uranium fuel or fuel assemblies; uranium enrichment;

enriched uranium hexafluoride conversion; plutonium processing;

fabrication of mixed-oxide fuel or fuel assemblies; scrap recovery of

special nuclear material; or any other activity involving a critical

mass of SNM that the Commission determines could significantly affect

public health and safety. The proposed rule would not apply to

licensees authorized to possess SNM under 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

material; 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

licensees 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 public meetings held on May 28, 1998,

and March 23, 1999, the purpose of the NEI-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 basis, including the ISA

summary, is incorporated in the license application, and the NRC staff

has gained sufficient experience with implementation of the ISA

requirements, a qualitative backfit mechanism could be considered.

Without a baseline determination of risk, as provided by the initial

ISA process, it is not clear how a determination of incremental risk,

as needed for a backfit analysis, would be accomplished. Furthermore,

although NEI previously stated that a quantitative backfit approach is

currently feasible, it would appear that a quantitative determination

of incremental risk would require a Probabilistic Risk Assessment, to

which the industry has been strongly opposed. The Commission requests

public comment on its intent to defer consideration of a qualitative

backfit provision in Part 70; any specific suggestions for backfit

provisions that would specifically address fuel cycle backfit needs and

the information that would be available to conduct the associated

analysis; and what would constitute a reasonable period of time,

including supporting rationale, before a backfit provision should be

implemented

ent to defer consideration of a qualitative

backfit provision in Part 70; any specific suggestions for backfit

provisions that would specifically address fuel cycle backfit needs and

the information that would be available to conduct the associated

analysis; and what would constitute a reasonable period of time,

including supporting rationale, before a backfit provision should be

implemented.

The majority of the proposed modifications to Part 70 are found in

a new Subpart H, ``Additional Requirements for Certain Licensees

Authorized to Possess a Critical Mass of Special Nuclear Material,''

that consists of 10 CFR 70.60 through 70.74. These proposed

modifications to Part 70, discussed in detail below, are required to

increase confidence in the margin of safety and are in general

accordance with the approach approved by the Commission in its SRMs of

August 22, 1997, and December 1, 1998.

Section 70.4 Definitions

Definitions of the following 12 terms would be added to this

section to provide a clear understanding of the meaning of the new

Subpart H: ``Acute'', ``Available and reliable to perform their

function when needed'', ``Configuration management'', ``Critical mass

of SNM'', ``Double contingency'', ``Hazardous materials produced from

licensed materials'', ``Integrated safety analysis'', ``Integrated

safety analysis summary'', ``Items relied on for safety'', ``Management

measures'', ``Unacceptable performance deficiencies'', and ``Worker.''

Section 70.14 Foreign Military Aircraft

This paragraph reflects an administrative change to renumber the

paragraph from 70.13a.

Section 70.17 Specific Exemptions

This paragraph reflects an administrative change to renumber the

paragraph from 70.14.

Section 70.50 Reporting Requirements

ied on for safety'', ``Management

measures'', ``Unacceptable performance deficiencies'', and ``Worker.''

Section 70.14 Foreign Military Aircraft

This paragraph reflects an administrative change to renumber the

paragraph from 70.13a.

Section 70.17 Specific Exemptions

This paragraph reflects an administrative change to renumber the

paragraph from 70.14.

Section 70.50 Reporting Requirements

Paragraph (c) would be reworded to include information to be

transmitted when making verbal or written reports to NRC. The new

information derives from the specifics of the new Subpart H, such as

sequence of events and whether the event was evaluated in the ISA. To

the extent the new information is also applicable to licensees not

subject to Subpart H, the information was added with no differentiation

noted. The new information that would only apply to Subpart H licensees

is noted.

Section 70.60 Applicability

This section lists the types of NRC licensees or applicants who

would be subject to the new Part 70, Subpart H. The Commission has

decided that the new requirements should not apply to all licensees

authorized to possess critical masses of SNM. Instead, the Commission

has identified a subset of these licensees that, based on the risk

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. In general,

the new Subpart is intended to ensure that the significant accidents

that are possible at fuel fabrication facilities (and the other listed

facility types) have been analyzed in advance, and that appropriate

controls or measures are established to ensure adequate protection of

workers,\1\ public, and the environment

because

of the relatively low level of risk at these facilities. In general,

the new Subpart is intended to ensure that the significant accidents

that are possible at fuel fabrication facilities (and the other listed

facility types) have been analyzed in advance, and that appropriate

controls or measures are established to ensure adequate protection of

workers,\1\ public, and the environment. The requirements and

provisions in Subpart H are in addition to, and not a substitute for,

other applicable requirements, including those of the U.S.

Environmental Protection Agency (EPA) and the U.S. Department of Labor,

OSHA. The requirements being added by NRC only apply to NRC's areas of

responsibility (radiological safety and chemical safety directly

related to licensed radioactive material). In this regard, the

requirements for hazards and accident analyses that NRC is adding are

intended to complement and be consistent with the parallel OSHA and EPA

regulations.

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\1\ 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).

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The regulation states that Subpart H does not apply to

decommissioning activities. NRC notes that the existing regulation

[Sec. 70.38(g)(4)(iii)] requires an approved decommissioning plan (DP)

that includes ``a description of methods used to ensure protection of

workers and the environment against radiation hazards during

decommissioning.'' Because the DP is submitted for NRC approval before

initiation of ``. . . procedures and activities necessary to carry out

decommissioning of the site or

existing regulation

[Sec. 70.38(g)(4)(iii)] requires an approved decommissioning plan (DP)

that includes ``a description of methods used to ensure protection of

workers and the environment against radiation hazards during

decommissioning.'' Because the DP is submitted for NRC approval before

initiation of ``. . . procedures and activities necessary to carry out

decommissioning of the site or

separate building or outdoor area,'' the DP will continue to be the

vehicle for regulatory approval of the licensee's practices for

protection of health and safety during decommissioning. The ISA should

provide valuable information with respect to developing the DP and the

use of the ISA in this manner is encouraged.

Section 70.61 Performance Requirements

In the past, the regulation of licensees authorized to possess SNM,

under 10 CFR Parts 20 and 70, has concentrated on radiation protection

for persons involved in nuclear activities conducted under normal

operations. The proposed amendments to Part 70 would explicitly address

potential exposures to workers or members of the public and

environmental releases as a result of accidents. Part 20 continues to

be NRC's standard for protection of workers and public from radiation

during normal operations, anticipated upsets (e.g., minor process

upsets that are likely to occur one or more times during the life of

the facility), and accidents. Although it is the Commission's intent

that the regulations in Part 20 also be observed to the extent

practicable during an emergency, it is not the Commission's intent that

the Part 20 requirements apply as the design standard for all possible

accidents at the facility, irrespective of the likelihood of those

accidents. Because accidents are unanticipated events that usually

occur over a relatively short period of time, the Part 70 changes seek

to assure adequate protection of workers, members of the public, and

the environment by limiting the risk (combined likelihood and

consequence) of such accidents

n standard for all possible

accidents at the facility, irrespective of the likelihood of those

accidents. Because accidents are unanticipated events that usually

occur over a relatively short period of time, the Part 70 changes seek

to assure adequate protection of workers, members of the public, and

the environment by limiting the risk (combined likelihood and

consequence) of such accidents.

There are three risk-informed performance requirements for the

rule, each of which is set out in 10 CFR 70.61: (1) Section 70.61(b)

states that high-consequence events must meet a likelihood standard of

highly unlikely; (2) section 70.61(c) requires that intermediate-

consequence events must meet a likelihood standard of unlikely; and (3)

section 70.61(d) requires that risk of nuclear criticality be limited

by assuring that all processes must remain subcritical under any normal

or credible abnormal conditions. The term ``performance requirements''

thus considers together consequences and likelihood. For regulatory

purposes, each performance requirement is considered an equivalent

level of risk. For example, the acceptable likelihood of intermediate-

consequence events is allowed to be greater than the acceptable

likelihood for high-consequence events.

A risk-informed approach must consider not only the consequences of

potential accidents, but also their likelihood of occurrence. As

mentioned above, the performance requirements rely on the terms

``unlikely'' and ``highly unlikely'' to focus on the risk of accidents.

However, the Commission has decided not to include quantitative

definitions ``unlikely'' and ``highly unlikely'' in the proposed rule,

because a single definition for each term, that would apply to all the

facilities regulated by Part 70, may not be appropriate. Depending on

the type of facility and its complexity, the number of potential

accidents and their consequences could differ markedly

However, the Commission has decided not to include quantitative

definitions ``unlikely'' and ``highly unlikely'' in the proposed rule,

because a single definition for each term, that would apply to all the

facilities regulated by Part 70, may not be appropriate. Depending on

the type of facility and its complexity, the number of potential

accidents and their consequences could differ markedly. Therefore, to

ensure that the overall facility risk from accidents is acceptable for

different types of facilities, the rule requires applicants to develop,

for NRC approval (see Sec. 70.65), the meaning of ``unlikely'' and

``highly unlikely'' specific to their processes and facility. To

accommodate this development, the Commission believes that the SRP is

the appropriate document to include guidelines for licensees to use. A

draft ``Standard Review Plan for the Review of a License Application

for a Fuel Cycle Facility'' has been developed. The draft SRP provides

one acceptable approach for the meaning of ``unlikely'' and ``highly

unlikely'' that can be applied to existing fuel cycle facilities.

The general approach for complying with the performance

requirements is that, at the time of licensing, each hazard (e.g.,

fire, chemical, electrical, industrial) that can potentially affect

radiological safety is identified and evaluated, in an ISA, by the

licensee. The impact of accidents, both internal and external,

associated with these hazards is compared with the three performance

requirements

ral approach for complying with the performance

requirements is that, at the time of licensing, each hazard (e.g.,

fire, chemical, electrical, industrial) that can potentially affect

radiological safety is identified and evaluated, in an ISA, by the

licensee. The impact of accidents, both internal and external,

associated with these hazards is compared with the three performance

requirements. Any (and all) structures, systems, components, or human

actions, for which credit is taken in the ISA for mitigating (reducing

the consequence of) or preventing (reducing the likelihood of) the

accident such that all three performance requirements are satisfied,

must be identified as an ``item relied on for safety.'' ``Items relied

on for safety'' is a term that is defined in 10 CFR 70.4, and in this

approach, the applicant has a great deal of flexibility in selecting

and identifying the actual ``items.'' For example, they can be defined

at the systems-level, component-level, or sub-component-level.

``Management measures'' [see discussion in 10 CFR 70.62(d)] are applied

to each item in a graded fashion to ensure that it will perform its

safety function when needed. The combination of the set of ``items

relied on for safety'' and the ``management measures'' applied to each

item will determine the extent of the licensee's programmatic and

design requirements, consistent with the facility risk, and will ensure

that at any given time, the facility risk is maintained safe and

protected from accidents (viz., satisfies the performance

requirements).

The proposed performance requirements also address certain chemical

hazards that result from the processing of licensed nuclear material.

The question of the extent of NRC's authority to regulate chemical

hazards at its fuel cycle facilities was raised after an accident in

1986 at a Part 40 licensed facility, in which a cylinder of uranium

hexafluoride ruptured and resulted in a worker fatality

e proposed performance requirements also address certain chemical

hazards that result from the processing of licensed nuclear material.

The question of the extent of NRC's authority to regulate chemical

hazards at its fuel cycle facilities was raised after an accident in

1986 at a Part 40 licensed facility, in which a cylinder of uranium

hexafluoride ruptured and resulted in a worker fatality. The cause of

the worker's death was the inhalation of hydrogen fluoride gas, which

was produced from the chemical reaction of uranium hexafluoride and

water (humidity in air). Partly as a result of the coordinated Federal

response and resulting Congressional investigation into that accident,

NRC and the OSHA entered into an MOU, in 1988, that clarified the

agencies' interpretations of their respective responsibilities for the

regulation of chemical hazards at nuclear facilities. The MOU

identified the following four areas of responsibility. Generally, NRC

covers the first three areas, whereas OSHA covers the fourth area:

(1) Radiation risk produced by radioactive materials;

(2) Chemical risk produced by radioactive materials;

(3) Plant conditions that affect the safety of radioactive

materials; and

(4) Plant conditions that result in an occupational risk, but do

not affect the safety of licensed radioactive materials.

One goal of the performance requirements in Sec. 70.61 is to be

consistent with the NRC-OSHA MOU. Therefore, the performance

requirements in Sec. 70.61 include explicit standards for the MOU's

first two areas of responsibility. In addition, the third MOU area of

responsibility is specifically evaluated by licensees under the ISA

requirements of Sec. 70.62(c)(1)(iii)

ed radioactive materials.

One goal of the performance requirements in Sec. 70.61 is to be

consistent with the NRC-OSHA MOU. Therefore, the performance

requirements in Sec. 70.61 include explicit standards for the MOU's

first two areas of responsibility. In addition, the third MOU area of

responsibility is specifically evaluated by licensees under the ISA

requirements of Sec. 70.62(c)(1)(iii). As an example of the third MOU

area, if the failure of a chemical system adjacent to a nuclear system

could affect the safety of the nuclear system such that the radiation

dose (and associated likelihood of that accident) exceeded a

performance requirement, the chemical system failure would be within

the scope of the ISA and the means to prevent the chemical system

failure from impacting

the nuclear system would be within NRC's regulatory purview.

OSHA provided comments, by a letter dated February 1, 1999, on a

draft of the rule that had been revised to be consistent with the MOU.

In that letter, OSHA expressed concerns that the rule language would

preempt OSHA from enforcing any of its standards, rules or other

requirements with respect to chemical hazards at the facilities covered

by the NRC draft rule. This concern is based on case law under the OSH

Act. The pertinent provision in the OSH Act states:

``(b)(1) Nothing in this chapter shall apply to working

conditions of employees with respect to which other Federal

agencies, and State agencies acting under section 2021 of title 42,

exercise statutory authority to prescribe or enforce standards or

regulations affecting occupational safety or health.'' [29 U.S.C.

653(b)(1)]

under the OSH

Act. The pertinent provision in the OSH Act states:

``(b)(1) Nothing in this chapter shall apply to working

conditions of employees with respect to which other Federal

agencies, and State agencies acting under section 2021 of title 42,

exercise statutory authority to prescribe or enforce standards or

regulations affecting occupational safety or health.'' [29 U.S.C.

653(b)(1)]

NRC staff subsequently met with OSHA officials on February 25,

1999, and some clarifications and further information were provided at

that meeting. As a result of the meeting discussions, some changes were

made to the rule language to more clearly specify the scope of NRC

involvement. However, these changes do not fully resolve the basic

preemption issue. The problems identified with the rule are not unique,

i.e., the preemption issue is generic and may already exist for any

NRC-licensed facilities where there are requirements to analyze

hazards. At the February 25 meeting, OSHA confirmed that the rule

language is consistent with the October 21, 1988 MOU; indicated that

they have no suggested changes to the MOU; and indicated that they are

not opposed to the proposed rule. The Commission's view is that the

proposed rule is consistent with NRC responsibilities and authority

under the Atomic Energy Act, and consistent with the OSHA MOU. The only

resolution of the preemption issue appears to be a legislative

modification of the OSH Act. Public comments would be appreciated on

any options that may have been overlooked.

Within each performance requirement, NRC recognizes that the

proposed radiological standards are more restrictive, in terms of acute

health effects to workers or the public, than the chemical standards

for a given consequence (high or intermediate) and that this is

consistent with current regulatory practice. The choice of each

criterion is discussed below in a paragraph-by-paragraph discussion of

Sec. 70.61

rmance requirement, NRC recognizes that the

proposed radiological standards are more restrictive, in terms of acute

health effects to workers or the public, than the chemical standards

for a given consequence (high or intermediate) and that this is

consistent with current regulatory practice. The choice of each

criterion is discussed below in a paragraph-by-paragraph discussion of

Sec. 70.61.

The use of any of the performance requirements is not intended to

imply that the specified worker or public radiation dose or chemical

exposure constitutes an acceptable criterion for an emergency dose to a

worker or the public. Rather, these values have been proposed in this

section as a reference value, to be used by licensees in the ISA (a

forward-looking analysis) to establish controls (i.e., items relied on

for safety and associated management measures) necessary to protect

workers from potential accidents with low or exceedingly low

probabilities of occurrence that are not expected to occur during the

operating life of the facility.

Section 70.61(b). This section addresses performance requirements

for high-consequence events.

The consequences identified in Sec. 70.61(b) of the proposed rule

are referred to as ``high-consequence events'' and include accidental

exposure of a worker or an individual located outside of the controlled

area to high levels of radiation or hazardous chemicals. These

accidents, if they occurred, would represent radiation doses to a

worker or an individual located outside of the controlled area at

levels with clinically observable biological damage or concentrations

of hazardous chemicals produced from licensed material at which death

or life-threatening injury could occur. The goal is to ensure an

acceptable level of risk by limiting the combination of the likelihood

of occurrence and the identified consequences. Thus, high-consequence

events must be sufficiently mitigated to a lower consequence or

prevented such that the event is highly unlikely (or lower)

azardous chemicals produced from licensed material at which death

or life-threatening injury could occur. The goal is to ensure an

acceptable level of risk by limiting the combination of the likelihood

of occurrence and the identified consequences. Thus, high-consequence

events must be sufficiently mitigated to a lower consequence or

prevented such that the event is highly unlikely (or lower). The

application of ``items relied on for safety'' provides this prevention

or mitigation function.

Section 70.61(b)(1). An acute exposure of a worker to a radiation

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

(TEDE) is considered to be a high-consequence event. According to the

National Council on Radiation Protection and Measurements (NCRP, 1971),

life-saving actions--including the ``* * * search for and removal of

injured persons, or entry to prevent conditions that would probably

injure numbers of people''--should be undertaken only when the ``* * *

planned dose to the whole body shall not exceed 100 rems.'' This is

consistent with a later NCRP position (NCRP, 1987) on emergency

occupational exposures, that states ``* * * when the exposure may

approach or exceed 1 Gy (100 rad) of low-LET [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.''

Section 70.61(b)(2). The exposure of an individual located outside

of the controlled area to a radiation dose of 0.25 Sv (25 rem) or

greater TEDE is considered a high-consequence event

ody, 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.''

Section 70.61(b)(2). The exposure of an individual located outside

of the controlled area to a radiation dose of 0.25 Sv (25 rem) or

greater TEDE is considered a high-consequence event. This is generally

consistent with the criterion established in 10 CFR 100.11,

``Determination of exclusion area, low population zone, and population

center distance,'' and 10 CFR 50.34, ``Contents of applications;

technical information,'' where a whole-body dose of 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 70.61(b)(3). The intake of 30 mg of soluble uranium by an

individual located outside of the controlled area is considered a high-

consequence event. This choice, which is based on a review of the

available literature [Pacific Northwest Laboratories (PNL), 1994], is

consistent with the selection of 30 mg of uranium as a criterion that

was discussed during the Part 76 rulemaking, ``Certification of Gaseous

Diffusion Plants.'' In particular, the final rule that established Part

76 (59 FR 48944; September 23, 1994) stated that ``The NRC will

consider whether the potential consequences of a reasonable spectrum of

postulated accident scenarios exceed * * * 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 70.61(b)(4). An acute chemical exposure to hazardous

chemicals produced from licensed material at concentrations that either

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 70.61(b)(4). An acute chemical exposure to hazardous

chemicals produced from licensed material at concentrations that either

(1) could cause death or life-threatening injuries to a worker; or (2)

could cause irreversible health effects to an individual located

outside of the controlled area, is considered a high-consequence event.

Chemical consequence criteria corresponding to anticipated adverse

health effects to humans from acute exposures (i.e., a single exposure

or multiple exposures occurring within a short time--24 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 EPA

and the National Academy of Sciences, has identified a priority list of

approximately 85 chemicals. Consequence criteria for 12 of these have

currently been developed and criteria for approximately 30 additional

chemicals per year are expected. Another set of chemical consequence

criteria, the Emergency Response Planning Guidelines (ERPGs), has been

developed by the American Industrial Hygiene Association to provide

estimates of concentration ranges where defined adverse health effects

might be observed because of short exposures to hazardous chemicals

ped and criteria for approximately 30 additional

chemicals per year are expected. Another set of chemical consequence

criteria, the Emergency Response Planning Guidelines (ERPGs), has been

developed by the American Industrial Hygiene Association to provide

estimates of concentration ranges where defined adverse health effects

might be observed because of short exposures to hazardous chemicals.

ERPG criteria are widely used by those involved in assessing or

responding to the release of hazardous chemicals including ``* * *

community emergency planners and response specialists, air dispersion

modelers, industrial process safety engineers, implementers of

environmental regulations such as the Superfund Amendment and

Reauthorization Act, industrial hygienists, and toxicologists,

transportation safety engineers, fire protection specialists, and

government agencies. * * *'' (DOE Risk Management Quarterly, 1997).

Despite their general acceptance, there are currently only

approximately 80 ERPG criteria available, and some chemicals of

importance (e.g., nitric acid) are not covered.

The qualitative language in the performance requirement allows the

applicant/licensee to propose and adopt an appropriate standard, which

may be an AEGL or ERPG standard, or where there is no AEGL or ERPG

value available, the applicant may develop or adopt a criterion that is

comparable in severity to those that have been established for other

chemicals. For example, for the worker performance requirement,

existing criteria that can be used by licensees to define appropriate

concentration levels to satisfy the performance requirement are the

AEGL-3 and ERPG-3

or where there is no AEGL or ERPG

value available, the applicant may develop or adopt a criterion that is

comparable in severity to those that have been established for other

chemicals. For example, for the worker performance requirement,

existing criteria that can be used by licensees to define appropriate

concentration levels to satisfy the performance requirement are the

AEGL-3 and ERPG-3. AEGL-3 is defined as ``The airborne concentration

(expressed in ppm or mg/m3) of a substance at or above which

it is predicted that the general population, including susceptible, but

excluding hypersusceptible, individuals, 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.'' Similarly, for the

public, 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,''

and ERPG-2 is defined as ``The maximum airborne concentration below

which it is believed that nearly all individuals could be exposed for

up to 1 hour without experiencing or developing irreversible or other

health effects or symptoms that could impair an individual's ability to

take protective action.''

Section 70.61(c). This section addresses performance requirements

for intermediate-consequence events.

The consequences identified in Sec

ncentration 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.''

Section 70.61(c). This section addresses performance requirements

for intermediate-consequence events.

The consequences identified in Sec. 70.61(c) of the proposed rule

are referred to as ``intermediate-consequence events'' and include

accidental exposure of a worker or an individual outside of the

controlled area to levels of radiation or hazardous chemicals that

generally correspond to permanent injury to a worker, transient injury

to a non-worker, or significant releases of radioactive material to the

environment. The goal is to ensure an acceptable level of risk by

limiting the combination of the likelihood of occurrence and the

identified consequences. Thus, ``intermediate-consequence events'' must

be sufficiently mitigated to a lower consequence or prevented such that

the event is unlikely (or lower). The application of ``items relied on

for safety'' provides this prevention or mitigation function.

Section 70.61(c)(1). A worker radiation dose between 0.25 Sv (25

rem) and 1 Sv (100 rem) TEDE is considered an intermediate-consequence

event [over 1 Sv (100 rem) is a high-consequence event]. This value was

chosen because of the use of 0.25 Sv (25 rem) as a criterion in

existing NRC regulations. For 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

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,

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.

Section 70.61(c)(2). A dose to any individual located outside of

the controlled area between 0.05 Sv (5 rem) and 0.25 Sv (25 rem) is

considered an intermediate-consequence event. NRC has used a 0.05-Sv

(5-rem) exposure criterion in a number of its existing regulations. For

example, 10 CFR 72.106, ``Controlled area of an ISFSI or MRS,'' states

that ``Any 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 the above-ground portion of the geologic

repository, 10 CFR 60.136, states that ``. . . for [accidents], no

individual located on or beyond any point on the boundary of the

preclosure controlled area will receive . . . a total effective dose

equivalent of 5 rem. . . .'' A TEDE of 0.05 Sv (5 rem) is also the

upper limit of EPA's Protective Action Guides of between 0.01 to 0.05

Sv (1 to 5 rem) for emergency evacuation of members of the public in

the event of an accidental release that could result in inhalation,

ingestion, or absorption of radioactive materials.

Section 70.61(c)(3)

ed area will receive . . . a total effective dose

equivalent of 5 rem. . . .'' A TEDE of 0.05 Sv (5 rem) is also the

upper limit of EPA's Protective Action Guides of between 0.01 to 0.05

Sv (1 to 5 rem) for emergency evacuation of members of the public in

the event of an accidental release that could result in inhalation,

ingestion, or absorption of radioactive materials.

Section 70.61(c)(3). The release of radioactive material to the

environment 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

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 annually). In particular, AO reporting criterion 1.B.1

requires the reporting of an event that involves ``. . . the release of

radioactive material to an unrestricted area in 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

days. In addition, a release of radioactive material, from a facility,

resulting in these concentrations, would be expected to cause some

environmental contamination in the area affected by the release. This

contamination would pose a longer-term hazard to the environment and

members of the public until it was properly remediated

out 6.8 mSv (680 mrem), or 5000 x 0.5 mSv/yr x 1 day x 1 yr/365

days. In addition, a release of radioactive material, from a facility,

resulting in these concentrations, would be expected to cause some

environmental contamination in the area affected by the release. This

contamination would pose a longer-term hazard to the environment and

members of the public until it was properly remediated. Depending on

the extent of environmental 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.

Section 70.61(c)(4). An acute chemical exposure to hazardous

chemicals produced from licensed material at concentrations that

either; (a) to a worker, could cause irreversible health effects (but

at concentrations below those which could cause death or life-

threatening effects); or (b) to an individual located outside of the

controlled area, could cause notable discomfort (but at concentrations

below those which could cause irreversible effects), is considered an

intermediate-consequence event. Chemical consequence criteria

corresponding to anticipated adverse health effects to humans from

acute exposures (i.e., a single exposure or multiple exposures

occurring within a short time--24 hours or less) have been developed,

or are under development, by a number of organizations. Of particular

interest, two existing standards, AEGL-2 and ERPG-2, can be used to

define the concentration level for irreversible health effects, and two

existing standards, AEGL-1 and ERPG-1, can be used to define the

concentration level for notable discomfort

osures

occurring within a short time--24 hours or less) have been developed,

or are under development, by a number of organizations. Of particular

interest, two existing standards, AEGL-2 and ERPG-2, can be used to

define the concentration level for irreversible health effects, and two

existing standards, AEGL-1 and ERPG-1, can be used to define the

concentration level for notable discomfort. The qualitative language in

the performance requirement allows the applicant/licensee to adopt and

propose an appropriate standard, which may be an AEGL or ERPG standard,

or where there is no AEGL or ERPG value available, the applicant may

develop or adopt a criterion that is comparable in severity to those

that have been established for other chemicals.

Section 70.61(d). This section addresses performance requirements

for an accidental nuclear criticality.

The third performance requirement states that the risk of nuclear

criticality accidents must be limited by assuring that under normal and

credible abnormal conditions, all nuclear processes are subcritical,

including use of an approved margin of subcriticality for safety. It

also requires that preventive controls and measures shall be the

primary means of protection against nuclear criticality accidents.

Although detecting and mitigating the consequences of a nuclear

criticality are important objectives (e.g., for establishing alarm

systems), the prevention of a criticality is a primary NRC objective.

The basis for this provision is the NRC strategic plan (NUREG-1614,

Vol. 1), which, for nuclear materials safety, states NRC's performance

goal of ``. . . no accidental criticality involving licensed

material.'' The language chosen for this performance requirement

closely follows the language of the applicable industry standard, ANSI/

ANS Standard 8.1-1983, ``Nuclear Criticality Safety in Operations with

Fissionable Materials Outside Reactors.''

Section 70.61(e). This section addresses items relied on for safety

and management measures

``. . . no accidental criticality involving licensed

material.'' The language chosen for this performance requirement

closely follows the language of the applicable industry standard, ANSI/

ANS Standard 8.1-1983, ``Nuclear Criticality Safety in Operations with

Fissionable Materials Outside Reactors.''

Section 70.61(e). This section addresses items relied on for safety

and management measures.

Paragraph 70.61(e) would require that each engineered or

administrative control or control system that is needed to meet the

performance requirements be designated as an item relied on for safety.

This means that any control or control system that is necessary to

maintain the acceptable combination of consequence and likelihood for

an accident is designated an item relied on for safety. The importance

of this section is that, once a control is designated as an item relied

on for safety, it falls into the envelope of the safety program

required by section 70.62. For example, records will be kept regarding

the item, and management measures such as the configuration control

program are applied to the item and to changes that affect the item, to

ensure that the item will be available and reliable to perform its

function when needed.

The failure of an item relied on for safety does not necessarily

mean that an accident will occur which will cause one of the

consequences listed in the performance requirements to be exceeded.

Some control systems may have parallel (redundant or diverse) control

systems that would continue to prevent the accident. The need for such

defense-in-depth and single-failure resistance would ideally be based

on the severity and likelihood of the potential accident

ean that an accident will occur which will cause one of the

consequences listed in the performance requirements to be exceeded.

Some control systems may have parallel (redundant or diverse) control

systems that would continue to prevent the accident. The need for such

defense-in-depth and single-failure resistance would ideally be based

on the severity and likelihood of the potential accident. In other

cases, the failure of an item may mean that the particular accident

sequence is no longer ``highly unlikely'', or ``unlikely.'' In these

cases, the performance requirement is not met, and the expectation

would be that a management measure would exist (possibly in the form of

an operating procedure) that ensured that the facility would not

operate in a condition that exceeds the performance requirement. For

example, a facility that relies on emergency power could not operate

for an extended time in the absence of an emergency power source even

if grid power is available. In this manner, the items relied on for

safety and the management measures complement each other to ensure

adequate protection from accidents at any given time.

Section 70.61(f). This section addresses the term ``controlled

area'' used in the performance requirements.

Section 70.61(f) requires licensees to identify a controlled area

consistent with the use of that term in Part 20, and provides

clarification regarding the activities that may occur inside the

controlled area. The function of this term is to delimit an area over

which the licensee exercises control of activities. Control includes

the power to exclude individuals, if necessary. The size of the

controlled area is not specified in the regulation because it will be

dependent upon the particular activities that are conducted at the site

and their relationship to the licensed activities. [Within the

controlled area will be a restricted area (as defined in Sec

ch the licensee exercises control of activities. Control includes

the power to exclude individuals, if necessary. The size of the

controlled area is not specified in the regulation because it will be

dependent upon the particular activities that are conducted at the site

and their relationship to the licensed activities. [Within the

controlled area will be a restricted area (as defined in Sec. 20.1003),

access to which is controlled by the licensee for purposes of radiation

safety.]

Individuals who do not receive an occupational dose (as that term

is used in Part 20) in the controlled area will be subject to the dose

limits for members of the public in 10 CFR 20.1301.

However, the Commission recognizes that certain licensees may have

ongoing activities at their site (i.e., within the controlled area)

that are not related to the licensed activities. For example, a non-

nuclear facility may be adjacent to the nuclear facility but both are

within the controlled area (which may be defined similar to the site

boundary). This raises a question regarding the appropriate accident

standard for these individuals. Protection of the individuals at the

non-nuclear facility must consider that the nature of many potential

accidents at a fuel cycle facility is such that there may not be

sufficient time during which to take action to exclude individuals from

the controlled area. Therefore, for purposes of the ISA accident

evaluation, the rule explicitly contains two options for these

individuals (as well as an implicit third option). In the first option,

the licensee evaluates, in the ISA, the risk at its location (as

opposed to that at any point at or beyond the controlled area boundary)

and determines that it meets the performance requirements for members

of the public. In the second option, performance requirements for

workers may be applied to individuals in the controlled area if the

provisions of Sec. 70.61(f)(2) are satisfied

option,

the licensee evaluates, in the ISA, the risk at its location (as

opposed to that at any point at or beyond the controlled area boundary)

and determines that it meets the performance requirements for members

of the public. In the second option, performance requirements for

workers may be applied to individuals in the controlled area if the

provisions of Sec. 70.61(f)(2) are satisfied. These conditions ensure

that the individuals are aware of the risks to them from the potential

accidents at the nuclear facility and have received appropriate

training and access to information. This parallels and is consistent

with the use of the term, ``Exclusion area'', by 10 CFR Parts 50 and

100, which states, ``Activities unrelated to operation of the reactor

may be permitted in an exclusion area under appropriate limitations,

provided that no significant hazards to the public health and safety

will result.'' The implied third option is to define (or redefine) a

controlled area such that within it only activities associated with the

licensed nuclear facility are permitted.

The Commission's intent is that the ISA does not evaluate

compliance with the accident standards for individuals who make

infrequent visits to the controlled area and restricted area (e.g.,

visitors). Use of the ISA to determine the risks to these individuals

would need to consider second-order effects such as the probability of

the individual being present at the time that the unlikely (or highly

unlikely) accident occurred. This level of detail is unnecessary to

accomplish the purpose of this rule (viz., to document and maintain the

safety basis of the facility design and operations). Application of the

Part 20 regulations provides adequate protection for these individuals.

In addition, the provisions (i.e., performance requirements) to protect

workers and non-workers during accidents should, implicitly, provide a

degree of protection to the infrequently present individuals.

this rule (viz., to document and maintain the

safety basis of the facility design and operations). Application of the

Part 20 regulations provides adequate protection for these individuals.

In addition, the provisions (i.e., performance requirements) to protect

workers and non-workers during accidents should, implicitly, provide a

degree of protection to the infrequently present individuals.

Section 70.62 Safety Program and Integrated Safety Analysis

This paragraph addresses the safety program, that includes process

safety information, ISA, and management measures. The performance of an

ISA, and the establishment of measures to ensure the availability and

reliability of items relied on for safety when needed, are the means by

which licensees demonstrate an adequate level of protection at their

facilities. The ISA is a systematic analysis to identify plant and

external hazards and their potential for initiating accident sequences;

the potential accident sequences and their consequences; and the site,

structures, systems, equipment, components, and activities of personnel

relied on for safety. As used here, ``integrated'' means joint

consideration of, and protection from, all relevant hazards, including

radiological, criticality, fire, and chemical. The structure of the

safety program recognizes the critical role that the ISA plays in

identifying potential accidents and the items relied on for safety.

However, it also recognizes that the performance of the ISA, by itself,

will not ensure adequate protection. Instead, an effective management

system is needed to ensure that the items relied on for safety are

available and reliable to perform their function when needed. Detailed

requirements for each part of the safety program are included in this

section.

Section 70.62(a). Each licensee would be required to establish and

maintain a safety program that demonstrates compliance with the

performance requirements of Sec. 70.61

gement

system is needed to ensure that the items relied on for safety are

available and reliable to perform their function when needed. Detailed

requirements for each part of the safety program are included in this

section.

Section 70.62(a). Each licensee would be required to establish and

maintain a safety program that demonstrates compliance with the

performance requirements of Sec. 70.61. Although the ISA would be the

primary tool in identifying the potential accidents requiring

consequence mitigation and accident prevention, process safety

information would be used to develop the ISA, and management measures

would be used to ensure the availability and reliability of items

relied on for safety identified through the ISA. The management

measures may be graded according to the risk importance associated with

an item relied on for safety.

The licensee is also required to establish and maintain records

demonstrating that it has, and continues to meet, the requirement of

this section. These records serve two major purposes. First, they can

supplement information that has been submitted as part of the license

application. Second, records are often needed to demonstrate licensee

compliance with applicable regulations and license commitments. It is

important, therefore, that an appropriate system of recordkeeping be

implemented to allow easy retrieval of required information.

Finally, each licensee would also be required to establish and

maintain a log documenting each discovery that an item relied on for

safety has failed to perform its function either in the context of the

performance requirements of Sec. 70.61 or on demand. The phrase ``* * *

in the context of the performance requirements of Sec. 70.61'' means

that items relied on for safety that fail would require logging even if

their failures did not result in process upsets or accidents but could

have resulted in the accident conditions they are protecting against,

had all conditions been optimum for the accident

ance requirements of Sec. 70.61 or on demand. The phrase ``* * *

in the context of the performance requirements of Sec. 70.61'' means

that items relied on for safety that fail would require logging even if

their failures did not result in process upsets or accidents but could

have resulted in the accident conditions they are protecting against,

had all conditions been optimum for the accident. This would not

include failures during times, such as routine maintenance on an item,

when the item or measure was clearly documented to not be available.

The log must contain: (a) The identity of the item that failed and the

safety function affected; (b) date of discovery of the failure; (c)

duration of time that the item was unable to perform its function; (d)

any other affected items relied on for safety and their safety

function; (e) affected processes; (f) the cause of the failure; (g)

whether the failure was in the context of performance requirements, or

on demand, or both; and (h) any corrective or compensatory actions

taken. The log should be initiated at the time of discovery and updated

promptly at the completion of each investigation of a failure of an

item relied on for safety. The purpose of the log is to assist NRC in

determining whether items relied on for safety are, in fact, available

and reliable and in detecting system problems that may impact ISA

evaluations.

Section 70.62(b). This paragraph would require the licensee to

maintain process-safety information pertaining to the hazards of the

materials used or produced in the process, the technology of the

process, and the equipment in the process. NRC confidence in the margin

of safety at its licensed facilities depends, in part, on the ability

of licensees to maintain a set of current, accurate, and complete

records available for NRC inspection. The process-safety information

should be used in support of development of an ISA.

Section 70.62(c). This paragraph proposes requirements for

conducting

ess, and the equipment in the process. NRC confidence in the margin

of safety at its licensed facilities depends, in part, on the ability

of licensees to maintain a set of current, accurate, and complete

records available for NRC inspection. The process-safety information

should be used in support of development of an ISA.

Section 70.62(c). This paragraph proposes requirements for

conducting

an ISA. There are four major steps in performing an ISA:

(1) Identify all hazards at the facility, including both

radiological and non-radiological hazards. Hazardous materials, their

location, and quantities, should be identified, as well as all

hazardous conditions, such as high temperature and high pressure. In

addition, any interactions that could result in the generation of

hazardous materials or conditions should be identified.

(2) Analyze the hazards to identify how they might result in

potential accidents. These accidents could be caused by process

deviations or other events internal to the plant, or by credible

external events, including natural phenomena such as floods,

earthquakes, etc. To accomplish the task of identifying potential

accidents, the licensee needs to ensure that detailed and accurate

information about plant processes is maintained and made available to

the personnel performing the ISA.

(3) Determine the consequences of each accident that has been

identified. For an accident with consequences at a ``high'' or

``intermediate level,'' as defined in 10 CFR 70.61, the likelihood of

such an accident must be shown to be commensurate with the

consequences, as required in 10 CFR 70.61.

ation 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 70.61, the likelihood of

such an accident must be shown to be commensurate with the

consequences, as required in 10 CFR 70.61.

(4) Identify the items relied on for safety (i.e., those items that

are relied on to prevent accidents or to mitigate their consequences,

identified in the ISA). These items are needed to reduce the

consequences or likelihood of the accidents to acceptable levels. The

identification of items relied on for safety is required only for

accidents with consequences at a high or intermediate level, as defined

in 10 CFR 70.61.

It is expected that the licensee or applicant would perform the ISA

using a ``team'' of individuals with expertise in engineering and

process operations related to the system being evaluated; the team

should include persons with experience in nuclear criticality safety,

radiation safety, fire safety, and chemical process safety, as

warranted by the materials and potential hazards associated with the

process being evaluated. At least one member of the ISA team should be

an individual who has experience and knowledge that is specific to the

process being evaluated. Finally, at least one individual in the team

must be knowledgeable in the specific ISA methodology being used.

Current Part 70 licensees, for whom the rule applies, would be

required to develop plans and submit them to NRC within 6 months of the

effective date of the rule. Each 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

specific ISA methodology being used.

Current Part 70 licensees, for whom the rule applies, would be

required to develop plans and submit them to NRC within 6 months of the

effective date of the rule. Each 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 ISA within 4

years of the effective date of the rule; correct any unacceptable

vulnerabilities identified; and submit the results to NRC for approval

in the form of an ISA summary that contains the information required by

10 CFR 70.65(b). Pending the correction of any unacceptable

vulnerabilities, licensees would be expected to implement appropriate

compensatory measures to ensure adequate protection until the

vulnerability can be more appropriately corrected.

Applicants for licenses to operate new facilities or new processes

at existing facilities would be expected to design their facilities or

processes to protect against the occurrence of the adverse consequences

identified in 10 CFR 70.61, using the baseline design criteria 10 CFR

70.64(a). Before operation, applicants would be expected to update

their ISAs, based on as-built conditions and submit the results to NRC

as ISA summaries, along with the applications, following the

requirements in 10 CFR 70.65(b).

The Commission believes that sufficient flexibility is permitted in

the ISA methodology chosen to be able to accommodate a wide range of

technologies. However, to assure that sufficient flexibility exists,

the Commission is requesting comments on this matter.

Section 70.62(d). Although the ISA would play a critical role in

identifying potential accidents and the items relied on for safety, the

performance of an ISA would not, by itself, ensure adequate protection

ISA methodology chosen to be able to accommodate a wide range of

technologies. However, to assure that sufficient flexibility exists,

the Commission is requesting comments on this matter.

Section 70.62(d). Although the ISA would play a critical role in

identifying potential accidents and the items relied on for safety, the

performance of an ISA would not, by itself, ensure adequate protection.

In addition, as would be provided for in 10 CFR 70.62(d), an effective

management system would be needed to ensure that the items relied on

for safety are available and reliable to perform their function when

needed. As stated before, management measures may be graded to better

implement the results of the ISA.

Management measures are functions performed by the licensee, in

general on a continuing basis, that are applied to items relied on for

safety. Management measures include: (a) Configuration management; (b)

maintenance; (c) training and qualifications; (d) procedures; (e)

audits and assessments; (f) incident investigations; (g) records

management; and (h) other quality assurance elements. Changes in the

configuration of the facility need to be carefully controlled to ensure

consistency among the facility design and operational requirements, the

physical configuration, and the facility documentation. Maintenance

measures must be in place to ensure the availability and reliability of

all hardware, identified as items relied on for safety, to perform

their function when needed. Training measures must be established to

ensure that all personnel relied on for safety are appropriately

trained to perform their safety functions. Periodic audits and

assessments of licensee safety programs must be performed to ensure

that facility operations are conducted in compliance with NRC

regulations and protect the worker and the public health and safety and

the environment

needed. Training measures must be established to

ensure that all personnel relied on for safety are appropriately

trained to perform their safety functions. Periodic audits and

assessments of licensee safety programs must be performed to ensure

that facility operations are conducted in compliance with NRC

regulations and protect the worker and the public health and safety and

the environment. When abnormal events occur, investigations of those

events must be carried out to determine the root cause and identify

corrective actions to prevent their recurrence and to ensure that they

do not lead to more serious consequences. Finally, to demonstrate

compliance with NRC regulations, records that document safety program

activities must be maintained for the life of the facility.

This section also would require that the safety program ensure that

each item relied on for safety would perform its intended function when

needed and in the context of the performance requirements of this

section. The utility of the two modifying requirements, ``when

needed,'' and ``in the context of the performance requirements of this

section,'' is clarified as follows:

The phrase ``when needed'' is used to acknowledge that a particular

safety control need not be continuously functioning. For example, it

may not be operational during maintenance or calibration testing, or

may not be required when the process is not operational or when special

nuclear material is not present. However, the phrase, when needed, does

not relieve a licensee from compliance with the performance

requirements. For example, if a particular component is out for

maintenance, the licensee must consider credible event sequences in

developing the ISA and identifying items relied on for safety--a high-

consequence event sequence still has to be highly unlikely

l

nuclear material is not present. However, the phrase, when needed, does

not relieve a licensee from compliance with the performance

requirements. For example, if a particular component is out for

maintenance, the licensee must consider credible event sequences in

developing the ISA and identifying items relied on for safety--a high-

consequence event sequence still has to be highly unlikely. Compliance

with the performance requirements in these cases can be established by

various means including identification of additional items relied on

for safety (and application of safety program management measures to

them), or by limiting operations or placing the plant in a different

operating mode during the maintenance of the item relied on for safety.

To illustrate, a loss of offsite power during a one-week

maintenance outage of the emergency diesel generator that is relied on

for safety would still be a credible event sequence. If the loss of

power, combined with the generator's inoperable status, could result in

a combination of dose and likelihood that exceeds a performance

requirement, then the licensee would not be in compliance with the

performance requirements of Sec. 70.61. A licensee cannot claim, after

the maintenance, that since the power was not lost, the generator was

available when needed. The concept is that the ISA is used as a risk-

informed, forward-look at the credible facility hazards and their

effects on plant systems and modes of operation. The rule would require

that each item necessary to comply with the performance requirements be

identified as important to safety and placed under the safety program

management controls. In identifying each item, the ISA must consider

various modes of operation and the likelihood that a given safety

control will be inoperable (e.g., because of being off-line for

maintenance) during credible event sequences

require

that each item necessary to comply with the performance requirements be

identified as important to safety and placed under the safety program

management controls. In identifying each item, the ISA must consider

various modes of operation and the likelihood that a given safety

control will be inoperable (e.g., because of being off-line for

maintenance) during credible event sequences.

The section would also require that the safety control perform its

function ``* * *in the context of the performance requirements of this

section.'' This phrase indicates that the function of interest is the

one credited in the ISA to meet certain consequence criteria with a

certain frequency. Second, this phrase would require that additional

safety controls be defined in cases where one control does not result

in compliance with the performance requirement or has periods when it

is inoperable. Using the loss of offsite power example again, a

licensee would still be required to meet the risk-informed performance

requirements of the rule when an emergency diesel generator used as an

item relied on for safety is not operable or out of service for

maintenance.

Section 70.64 Requirements for New Facilities or New Processes at

Existing Facilities

This section deals with baseline design criteria for new facilities

or new processes at existing facilities.

A major feature of the proposed amendments to Part 70 is the

requirement that licensees and applicants for a license perform an ISA

and use the ISA process to develop risk-informed decisions regarding

facility safety. The ISA process is applied to existing designs to

identify risk insights on those areas that warrant additional

preventive or mitigative measures. For new facilities, the proposed

rule would require the performance of the ISA before construction [see

the existing Sec. 70.21(f) and Sec. 70.23(a)(7)], and the updating of

the ISA before beginning operations

decisions regarding

facility safety. The ISA process is applied to existing designs to

identify risk insights on those areas that warrant additional

preventive or mitigative measures. For new facilities, the proposed

rule would require the performance of the ISA before construction [see

the existing Sec. 70.21(f) and Sec. 70.23(a)(7)], and the updating of

the ISA before beginning operations. 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. In addition, a

fundamental element of NRC's safety philosophy is that designs and

operations should provide for defense-in-depth protection against

accidents. Therefore, the Commission has specified baseline design

criteria in Sec. 70.64 that are similar in use to the general design

criteria in Part 50 Appendix A; Part 72, Subpart F; and 10 CFR 60.131.

The baseline design criteria identify 10 initial safety design

considerations, including: (a) Quality standards and records; (b)

natural phenomena hazards; (c) fire protection; (d) environmental and

dynamic effects 2; (e) chemical protection; (f) emergency

capability; (g) utility services; (h) inspection, testing, and

maintenance; (i) criticality control; and (j) instrumentation and

controls. The baseline design criteria do not provide relief from

compliance with the safety performance requirements of Sec. 70.61. The

baseline design criteria are generally an acceptable set of initial

design safety considerations, which may not be sufficient to ensure

adequate safety for all new processes and facilities. The ISA process

is intended to identify additional safety features that may be needed.

On the other hand, the Commission recognizes that there may be

processes or facilities for which some of the baseline design criteria

may not be necessary or appropriate, based on the results of the ISA

nsiderations, which may not be sufficient to ensure

adequate safety for all new processes and facilities. The ISA process

is intended to identify additional safety features that may be needed.

On the other hand, the Commission recognizes that there may be

processes or facilities for which some of the baseline design criteria

may not be necessary or appropriate, based on the results of the ISA.

For these processes and facilities, any design features that are

inconsistent with the baseline design criteria should be identified and

justified.

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\2\ Environmental and dynamic effects are effects that could be

caused by ambient conditions. For example, an item relied on for

safety will need to function within its expected environment (i.e.,

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).

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Using the baseline design criteria and considering defense-in-depth

practices in the design should result in a new facility design that is

based on providing successive levels of protection such that health and

safety will not be wholly dependent on any single element of the

design, construction, maintenance, or operation of the facility. The

net effect of incorporating defense-in-depth practices is a

conservatively designed facility and system that will exhibit greater

tolerance for failures and external challenges. The risk insights

obtained through performance of the ISA can be then used to supplement

the final design by focusing attention on the prevention and mitigation

of the potential accidents having higher-risk.

Section 70.65 Additional Content of Applications

epth practices is a

conservatively designed facility and system that will exhibit greater

tolerance for failures and external challenges. The risk insights

obtained through performance of the ISA can be then used to supplement

the final design by focusing attention on the prevention and mitigation

of the potential accidents having higher-risk.

Section 70.65 Additional Content of Applications

In addition to the information that currently must be submitted to

NRC, under Sec. 70.22, for a license application, this section requires

additional information to be submitted to demonstrate compliance with

the proposed new subpart. In particular, this additional information

would need to include a description of the applicant's safety program

established under Sec. 70.62, a description of the management measures,

and an ISA summary.

The ISA summary would contain: (a) A description of the site and

the facility; (b) a description of the team qualifications and ISA

methodology; (c) the processes analyzed in the ISA and the maximum

consequences of each; (d) a demonstration of how the licensee meets the

requirements for criticality monitoring and alarms in Sec. 70.24; (e) a

demonstration of how the licensee meets the performance requirements of

Sec. 70.61 and, if applicable, Sec. 70.64; (f) a list of items relied

on for safety and a description of their safety function; (g) a

description of the proposed standards used to assess the consequences

from acute chemical exposures; and (h) the definitions of ``likely'',

``unlikely'', ``highly unlikely'', and ``credible'' as used in the ISA.

The plant and process descriptions, ISA team qualifications and

methods, and definitions of terms used in the ISA, are all needed to

fully understand the facility and the ISA and how it was developed.

Although some of the facility information is also requested in

Sec

ical exposures; and (h) the definitions of ``likely'',

``unlikely'', ``highly unlikely'', and ``credible'' as used in the ISA.

The plant and process descriptions, ISA team qualifications and

methods, and definitions of terms used in the ISA, are all needed to

fully understand the facility and the ISA and how it was developed.

Although some of the facility information is also requested in

Sec. 70.22, there may be information about the facility which would be

too detailed for inclusion in the general site description, but would

be needed to be included here to understand the ISA and ISA results.

The demonstration of how the licensee meets Secs. 70.24, 70.61, and

70.64 is a critical element in determining whether the applicant

understands and complies with the regulations and can operate the

facility safely. Another critical element is the applicant's

identification of the items relied on for safety. Through the ISA

process, the applicant should have identified potential accidents that

can occur in individual processes and in the facility as a whole. As

discussed earlier, these accidents are prevented or their consequences

mitigated using controls that are identified in the ISA summary as

items relied on for safety. It is important for NRC staff to review the

items relied on for safety, that were identified as such by the

applicant or licensee, to determine whether potential accidents are

adequately prevented or mitigated. Since items relied on for safety

play a key role in assuring that the performance requirements are met,

and because the applicant has great flexibility in selecting and

identifying what the actual ``items'' are (as discussed in relation to

Sec. 70.61), the items relied on for safety would be clearly and

unambiguously identified on a list. This list of items is then managed

and controlled by the applicant through the management measures in

Sec. 70.61 to ensure that they continue to perform the safety function

required

t has great flexibility in selecting and

identifying what the actual ``items'' are (as discussed in relation to

Sec. 70.61), the items relied on for safety would be clearly and

unambiguously identified on a list. This list of items is then managed

and controlled by the applicant through the management measures in

Sec. 70.61 to ensure that they continue to perform the safety function

required. By evaluating the ISA methodology, and the ISA summary,

supplemented by reviewing the ISA and other information, as needed, at

the licensee's facility, the staff can better understand the potential

hazards at the facility, how the applicant plans to address these

hazards, and thereby have confidence in the safety basis on which the

license will be issued.

The ISA summary would be required to be submitted on the docket in

conjunction with the license application but would not be considered

part of the license. The ISA, on which the ISA summary is based, would

be maintained current at the licensee's facility and available for NRC

review, but it would not be submitted and docketed. The information and

commitments contained in the license application that are incorporated

into the license conditions cannot be changed without prior review and

approval of NRC staff, at which time a license amendment is issued.

Although the ISA summary will be on the docket, since it is not part of

the license it can be changed without a license amendment, unless it

reflects a change that cannot be made without prior approval per

Sec. 70.72(c). However, the information used to perform the ISA, and

the ISA summary, both form integral parts of the safety basis for

issuance of the license and therefore must be maintained to adequately

represent the current status of the facility. So that NRC knows the

current status of the facility, changes to these documents, on which

NRC based its safety conclusion, are to be submitted to NRC, as

discussed in Sec. 70.72.

on used to perform the ISA, and

the ISA summary, both form integral parts of the safety basis for

issuance of the license and therefore must be maintained to adequately

represent the current status of the facility. So that NRC knows the

current status of the facility, changes to these documents, on which

NRC based its safety conclusion, are to be submitted to NRC, as

discussed in Sec. 70.72.

Section 70.66 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 new subpart, 10 CFR 70.60 through 70.66, will be satisfied.

Section 70.72 Facility Changes and Change Process

This section deals with changes to site, structures, systems,

equipment, components, and activities of personnel after a license

application has been approved.

Past incidents at fuel cycle facilities have often resulted from

changes not fully analyzed, not authorized by licensee 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 safety at that facility. This section would require the

licensee to establish and use a system to evaluate changes and the

potential impacts of those changes before implementing them. By using

this system to evaluate, implement and track changes to the facility,

the licensee can make certain changes without NRC pre-approval. If the

change affects information contained in the ISA summary, the licensee

would be required to notify NRC within 90 days of the change by

submitting updated ISA summary pages in that time. For changes that

affect the on-site documentation, such as the ISA, management measures

or process-safety information, the licensee would be required to notify

NRC within 12 months of the change

proval. If the

change affects information contained in the ISA summary, the licensee

would be required to notify NRC within 90 days of the change by

submitting updated ISA summary pages in that time. For changes that

affect the on-site documentation, such as the ISA, management measures

or process-safety information, the licensee would be required to notify

NRC within 12 months of the change. This update frequency would allow

NRC staff to review the changes being made to the facility in enough

time to ensure that the licensee's evaluations of potential impacts to

health and safety were accurate. It also allows NRC staff to maintain

relatively current facility and safety information on the docket at all

times. In addition, maintaining the license and ISA summary so that

they reflect the current configuration of the facility would facilitate

a relatively simple, cost-effective license renewal process. The

Commission is particularly interested in comments concerning the 90 day

time period for submitting updated ISA summary pages that reflect

changes to a facility's site, structures, systems, equipment,

components, and activities of personnel.

Some changes, however, would require NRC pre-approval before they

can be implemented. These are changes that are considered major and

could have a significant impact on health and safety. The staff

considered two options for the types of changes that would require NRC

pre-approval. Option 1 is consistent with the types of changes that

have required pre-approval at Part 70 licensees in the past, and which

the staff believes would require NRC pre-approval for only a relatively

few significant changes. Option 2 is consistent with the change control

process required for Part 50 licensees (power reactors) and which the

staff believes would require more requests for NRC pre-approval.

The advantages of Option 1 are that it focuses on the most

significant changes to the facility and is equivalent to looking at the

highest risk changes

pre-approval for only a relatively

few significant changes. Option 2 is consistent with the change control

process required for Part 50 licensees (power reactors) and which the

staff believes would require more requests for NRC pre-approval.

The advantages of Option 1 are that it focuses on the most

significant changes to the facility and is equivalent to looking at the

highest risk changes. It contains very little subjective criteria and

is therefore easier to implement and inspect. It also would likely only

result in a few license amendments a year which is generally consistent

with the past practice at these facilities. Since Option 1 would permit

more changes without NRC pre-approval, a relatively short timeframe (90

days) for submitting updated ISA summary pages is required in order for

NRC to have information that reflects the current status of the

facility and to be confident that adequate protection is still provided

with the changes, as reflected in the ISA summary. The advantages of

Option 2 are that NRC would have more control over the changes at the

facilities, i.e., staff expects that more changes would be reviewed by

the staff before being implemented; thus, it would be less likely that

NRC would have a concern with a change after the fact; and it is

consistent with the change control process at power reactors, where

changes are reported only after 12 months.

The proposed rule language reflects Option 1.

Section 70.73 Renewal of Licenses

ilities, i.e., staff expects that more changes would be reviewed by

the staff before being implemented; thus, it would be less likely that

NRC would have a concern with a change after the fact; and it is

consistent with the change control process at power reactors, where

changes are reported only after 12 months.

The proposed rule language reflects Option 1.

Section 70.73 Renewal of Licenses

Under the proposed amendments to Part 70, changes to site,

structures, systems, equipment, components, and activities of personnel

made by the licensee pursuant to Sec. 70.72 would be documented on a

continuing basis on-site. A description of those changes would also be

sent to NRC periodically. This process is intended to keep the

documents, which support the license, current and thereby establish a

``living'' license. In the past, the license renewal process was

burdensome to NRC and the licensee because all changes made to the

facility since the last license renewal

would be reviewed at one time. However, with the proposed ``living

license,'' changes to the facility will be reviewed by NRC either

before changes are made, or relatively shortly thereafter. As a result,

review of the license renewal application is expected to be performed

with minimal additional review of the licensee's safety program. This

approval would be contingent on the licensee satisfying any

requirements associated with the National Environmental Policy Act of

1969 as implemented in 10 CFR Part 51.

Section 70.74 Additional Reporting Requirements

vely shortly thereafter. As a result,

review of the license renewal application is expected to be performed

with minimal additional review of the licensee's safety program. This

approval would be contingent on the licensee satisfying any

requirements associated with the National Environmental Policy Act of

1969 as implemented in 10 CFR Part 51.

Section 70.74 Additional Reporting Requirements

The new requirements that would be incorporated in the proposed

amendments to Part 70 would revise the reporting of events to NRC. This

new approach, based on consideration of the risk and consequences

established in 10 CFR 70.61(b) is intended to replace and expand on the

approach licensees have currently been using for reporting criticality

events under Bulletin 91-01. The new approach would cover all types of

events, not just criticality events, and establish a timeframe for

reporting that is scaled according to risk. The new reporting

requirements are intended to supplement the requirements in the

existing Parts 20 and 70 and elsewhere in the regulations. A more

detailed discussion of the new requirements is found in the following

discussion of Appendix A to Part 70.

Appendix A Reportable Events

The reporting of events supports NRC's need to be aware of

conditions that could result in an imminent danger to the worker or to

public health and safety or to the environment. In particular, NRC

needs to be aware of licensee efforts to address potential emergencies.

Further, once safe conditions have been restored after an event, NRC

has an interest in disseminating information on the event to the

nuclear industry and other interested parties, to reduce the likelihood

that the event will occur in the future. Also, in the event of an

accident, NRC must be able to respond accurately to requests for

information by the public and the media

tial emergencies.

Further, once safe conditions have been restored after an event, NRC

has an interest in disseminating information on the event to the

nuclear industry and other interested parties, to reduce the likelihood

that the event will occur in the future. Also, in the event of an

accident, NRC must be able to respond accurately to requests for

information by the public and the media. Finally, NRC must evaluate the

performance of individual licensees and the industry as a whole to

fulfill its statutory mandate to protect the health and safety of the

worker and the public and the environment.

Licensee reporting of events would consist of two reporting classes

based on the hazard--reports that must be made in 1 hour and those to

be reported within 24 hours. According to this approach, licensees

would report events based on two criteria: (1) Whether actual

consequences have occurred or whether a potential for such consequences

exists; and (2) the seriousness of the consequences. The events that

must be reported within the shortest timeframe (1 hour) are high-

consequence events. These events encompass unintended criticalities and

loss of criticality controls, and loss of chemical controls or the

occurrence of chemical exposures that exceed the performance

requirements in Sec. 70.61(b).

Less serious events or failure to meet the performance requirements

for reasons not otherwise specifically stated, that have occurred shall

be reported within 24 hours. These include chemical exposure to

licensed material or hazardous chemicals that exceed the lower

threshold limits in Sec. 70.61(c)(4), and events that were dismissed in

the ISA based on likelihood.

Events that could potentially lead to exceeding the performance

requirements in Sec. 70.61 should also be reported. External events,

such as a hurricane, tornado, earthquake, flood, or fire, either

internal or external to the plant, that affected or could have affected

a facility, must be reported within 24 hours

ec. 70.61(c)(4), and events that were dismissed in

the ISA based on likelihood.

Events that could potentially lead to exceeding the performance

requirements in Sec. 70.61 should also be reported. External events,

such as a hurricane, tornado, earthquake, flood, or fire, either

internal or external to the plant, that affected or could have affected

a facility, must be reported within 24 hours. This reporting

requirement would capture, for example, a tornado that strikes a

facility, an earthquake motion experienced by a facility, or any type

of fire. Since these events could have affected a facility, NRC would

want to know about such events to assess a licensee's conclusion of

whether any detrimental effects did in fact occur, or could have

occurred in the absence of controls that were present but not part of

the safety basis. Another category of potential events that would be

reported is one that involves the existence of an unsafe condition that

is not identified in the ISA. This condition could be caused by a

deviation from established safe operating conditions, by an

unanticipated and unanalyzed set of circumstances, or by an improper

analysis. This type of event would be reported within 24 hours.

The proposed rule also would require concurrent reporting of events

when a news release is made or if other Government agencies are

notified, as is done under 10 CFR Part 50.72, to support NRC's ability

to be responsive to questions concerning the safety of NRC-licensed

facilities.

References

of circumstances, or by an improper

analysis. This type of event would be reported within 24 hours.

The proposed rule also would require concurrent reporting of events

when a news release is made or if other Government agencies are

notified, as is done under 10 CFR Part 50.72, to support NRC's ability

to be responsive to questions concerning the safety of NRC-licensed

facilities.

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

, ``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, NW (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 may be

requested by writing to U.S

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 may be

requested by writing to U.S. Nuclear Regulatory Commission,

Reproduction and Distribution Services, Washington, DC 20555-0001:

``Standard Review Plan for the Review of a License Application for a

Fuel Cycle Facility'' (Draft NUREG-1520); and ``Integrated Safety

Analysis Guidance Document'' (Draft NUREG-1513).

Plain Language

The Presidential Memorandum dated June 1, 1998, entitled ``Plain

Language in Government Writing,'' directed that the Federal

government's writing be in plain language. The NRC requests comments on

this proposed rule specifically with respect to the clarity and

effectiveness of the language used. Comments should be sent to the

address listed above.

Finding of No Significant Environmental Impact: Availability

The Commission has determined, under the National Environmental

Policy Act of 1969, as amended, and the Commission's regulations in

Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a

major Federal action significantly affecting the quality of the human

environment, and therefore an environmental impact statement is not

required.

The proposed amendments to Part 70 are intended to provide

increased confidence in the margin of safety at certain facilities that

possess a critical mass of SNM

ulations in

Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a

major Federal action significantly affecting the quality of the human

environment, and therefore an environmental impact statement is not

required.

The proposed amendments to Part 70 are intended to provide

increased confidence in the margin of safety at certain facilities that

possess a critical mass of SNM. To accomplish this objective, the

amendments: (1) Identify appropriate consequence criteria and the level

of protection needed to prevent or mitigate accidents that exceed such

criteria; (2) require affected licensees to perform an integrated

safety analysis (ISA) to identify potential accidents at the facility

and the items relied on for safety; (3) require the implementation of

measures to ensure that the items relied on for safety are available

and reliable to perform their function when needed; and (4) require the

inclusion of the safety bases, as reflected in the ISA summary, in the

license application. The language, in the proposed rule, that defines

an environmental consequence of concern, is relevant to the question of

environmental 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

n 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 on 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

ks 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, DC and the Part 70 website.

Single copies of the environmental assessment are available from Barry

Mendelsohn, Office of Nuclear Material Safety and Safeguards, U.S.

Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone

(301) 415-7262; e-mail: [email protected].

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 99 hours per response, and the recordkeeping

burden is estimated to average 560 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. NRC is seeking public comment on the

potential impact of the information collections contained in the

proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper

performance of NRC's function? Will the information have practical

utility?

2. Is the burden estimate accurate?

3. Is there a way to enhance the quality, utility, and clarity of

the information to be collected?

4

the

potential impact of the information collections contained in the

proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper

performance of NRC's function? Will the information have practical

utility?

2. Is the burden estimate accurate?

3. Is there a way to enhance the quality, utility, and clarity of

the information to be collected?

4. How can the burden of the information collection be minimized,

including the use of automated collection techniques?

Send comments on any aspect of this proposed 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 [email protected]; 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 August 30, 1999. Comments received after

this date will be considered if it is practical to do so, but assurance

of consideration cannot be given to comments received after this date.

Public Protection Notification

If a means used to impose an information collection does not

display a currently valid OMB control number, the NRC may not conduct

nor sponsor, and a person is not required to respond to, the

information collection.

Regulatory Analysis

considered if it is practical to do so, but assurance

of consideration cannot be given to comments received after this date.

Public Protection Notification

If a means used to impose an information collection does not

display a currently valid OMB control number, the NRC may not conduct

nor sponsor, and a person is not required to respond to, the

information collection.

Regulatory Analysis

The Commission has prepared a draft Regulatory Analysis on this

proposed regulation. The analysis examines the benefits and costs of

the alternatives considered by the Commission. The draft Regulatory

Analysis is available for inspection in the NRC Public Document Room,

2120 L Street NW (Lower Level), Washington, DC and the Part 70 website.

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: [email protected].

The Commission requests public comment on the draft Regulatory

Analysis. Comments on the draft analysis may be submitted to NRC as

indicated under the ADDRESSES heading.

Regulatory Flexibility Certification

As required by the Regulatory Flexibility Act, as amended, 5 U.S.C.

605(b), the Commission certifies that this proposed rule, if adopted,

would not have a significant economic impact on a substantial number of

small entities. This proposed rule would affect facilities that are

authorized to possess a critical mass of SNM and who are engaged in one

of the following activities: (a) enriched uranium processing; (b)

fabrication of uranium fuel or fuel assemblies; (c) uranium enrichment;

(d) enriched uranium hexafluoride conversion; (e) plutonium processing;

t economic impact on a substantial number of

small entities. This proposed rule would affect facilities that are

authorized to possess a critical mass of SNM and who are engaged in one

of the following activities: (a) enriched uranium processing; (b)

fabrication of uranium fuel or fuel assemblies; (c) uranium enrichment;

(d) enriched uranium hexafluoride conversion; (e) plutonium processing;

(f) fabrication of mixed-oxide fuel or fuel assemblies; (g) scrap

recovery of special nuclear material; or (h) any other activity

involving a critical mass of SNM that the Commission determines could

significantly affect public health and safety or the environment. These

licensees do not fall within the scope of the definition of ``small

entities'' set forth in the Regulatory Flexibility Act, nor the size

standards published by NRC (10 CFR 2.810).

Voluntary Consensus Standards

The National Technology Transfer Act of 1995, Pub. L. 104-113,

requires that Federal Agencies use technical standards that are

developed or adopted by voluntary consensus standards bodies unless the

use of such a standard is inconsistent with applicable law or otherwise

impractical. In this proposed rule, the NRC proposes to use the

following voluntary consensus standard, ANSI/ANS Standard 8.1-1983,

``Nuclear Criticality Safety in Operations with Fissionable Material

Outside Reactors,'' developed by the American Nuclear Society. Portions

of the standard were used in the definition of double contingency and

in Sec. 70.61(d). The NRC invites comment on the applicability and use

of other standards.

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

1(d). The NRC invites comment on the applicability and use

of other standards.

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)

95-601, sec. 10, 92 Stat.

2951 (42 U.S.C. 5851). Section 70.21(g) also issued under sec. 122,

68 Stat. 939 (42 U.S.C. 2152). Section 70.31 also issued under sec.

57d, Pub. L. 93-377, 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36

and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42

U.S.C. 2234). Section 70.61 also issued under secs. 186, 187, 68

Stat. 955 (42 U.S.C. 2236, 2237). Section 70.62 also issued under

sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).

2. The undesignated center heading ``GENERAL PROVISIONS'' is

redesignated as ``Subpart A--General Provisions.''

3. In Sec. 70.4, the definitions of Acute, Available and reliable

to perform their function when needed, Configuration management,

Critical mass of special nuclear material, Double contingency,

Hazardous chemicals produced from licensed material, Integrated safety

analysis (ISA), Integrated safety analysis summary, Items relied on for

safety, Management measures, Unacceptable performance deficiencies, and

Worker are added, in alphabetical order, as follows:

Sec. 70.4 Definitions.

* * * * *

Acute as used in this part means a single radiation dose or

chemical exposure event or multiple radiation dose or chemical exposure

events occurring within a short time (24 hours or less).

* * * * *

Available and reliable to perform their function when needed as

used in subpart H of this part means that, based upon the analyzed,

credible conditions in the integrated safety analysis, items relied on

for safety will perform their intended safety function when needed and

management measures will be implemented that ensure continuous

compliance with the performance requirements of Sec. 70.61 of this

part, considering factors such as necessary maintenance, operating

limits, common cause failures, and the likelihood and consequences of

failure or degradation of the items and measures

lied on

for safety will perform their intended safety function when needed and

management measures will be implemented that ensure continuous

compliance with the performance requirements of Sec. 70.61 of this

part, considering factors such as necessary maintenance, operating

limits, common cause failures, and the likelihood and consequences of

failure or degradation of the items and measures.

* * * * *

Configuration management (CM) means ensuring, as part of the safety

program, oversight and control of design information, safety

information, and modifications (both temporary and permanent) that

might impact the ability of items relied on for safety to perform their

function when needed.

* * * * *

Critical mass of special nuclear material (SNM) means special

nuclear material in a quantity exceeding 700 grams of contained

uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1500

grams of contained uranium-235, if no uranium enriched to more than 4

percent by weight of uranium-235 is present; 450 grams of any

combination thereof; or one-half such quantities if massive moderators

or reflectors made of graphite, heavy water, or beryllium may be

present.

* * * * *

Double contingency means a process design that incorporates

sufficient factors of safety to require at least two

unlikely, independent, and concurrent changes in process conditions

before a criticality accident is possible.

* * * * *

Hazardous chemicals produced from licensed materials means

substances having licensed material as precursor compound(s) or

substances that physically or chemically interact with licensed

materials; that are toxic, explosive, flammable, corrosive, or reactive

to the extent that they can endanger life or health if not adequately

controlled

before a criticality accident is possible.

* * * * *

Hazardous chemicals produced from licensed materials means

substances having licensed material as precursor compound(s) or

substances that physically or chemically interact with licensed

materials; that are toxic, explosive, flammable, corrosive, or reactive

to the extent that they can endanger life or health if not adequately

controlled. These include substances commingled with licensed material,

and include substances such as hydrogen fluoride that is produced by

the reaction of uranium hexafluoride and water, but do not include

substances prior to process addition to licensed material or after

process separation from licensed material.

Integrated safety analysis (ISA) means a systematic analysis to

identify plant and external hazards and their potential for initiating

accident sequences, the potential accident sequences, their likelihood

and consequences, and the items relied on for safety. As used here,

integrated means joint consideration of, and protection from, all

relevant hazards, including radiological, nuclear criticality, fire,

and chemical. However, with respect to compliance with the regulations

of this part, the NRC requirement is limited to consideration of the

effects of all relevant hazards on radiological safety, prevention of

nuclear criticality accidents, or chemical hazards directly associated

with NRC licensed radioactive material.

Integrated safety analysis summary means the document submitted

with the license application, license amendment application, or license

renewal application that provides a synopsis of the results of the

integrated safety analysis and contains the information specified in

Sec. 70.65(b).

Items relied on for safety means structures, systems, equipment,

components, and activities of personnel that are relied on to prevent

potential accidents at a facility that could exceed the performance

requirements in Sec. 70.61 or to mitigate their potential consequences

des a synopsis of the results of the

integrated safety analysis and contains the information specified in

Sec. 70.65(b).

Items relied on for safety means structures, systems, equipment,

components, and activities of personnel that are relied on to prevent

potential accidents at a facility that could exceed the performance

requirements in Sec. 70.61 or to mitigate their potential consequences.

This does not limit the licensee from identifying additional

structures, systems, equipment, components, or activities of personnel

(i.e., beyond those in the minimum set necessary for compliance with

the performance requirements) as items relied on for safety.

* * * * *

Management measures mean the functions performed by the licensee,

generally on a continuing basis, that are applied to items relied upon

for safety, to ensure the items are available and reliable to perform

their functions when needed. Management measures include configuration

management, maintenance, training and qualifications, procedures,

audits and assessments, incident investigations, records management,

and other quality assurance elements.

* * * * *

Unacceptable performance deficiencies mean deficiencies in the

items relied on for safety or the management measures that need to be

corrected to ensure an adequate level of protection as defined in 10

CFR 70.61(b), (c), or (d).

* * * * *

Worker means an individual whose assigned duties in the course of

employment involve exposure to radiation and/or radioactive material

from licensed and unlicensed sources of radiation (i.e., an individual

who is subject to an occupational dose as in 20 CFR 20.1003).

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

Sec. 70.8 Information collection requirements: OMB approval.

* * * * *

r means an individual whose assigned duties in the course of

employment involve exposure to radiation and/or radioactive material

from licensed and unlicensed sources of radiation (i.e., an individual

who is subject to an occupational dose as in 20 CFR 20.1003).

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

Sec. 70.8 Information collection requirements: OMB approval.

* * * * *

(b) The approved information collection requirements contained in

this part appear in Secs. 70.9, 70.17, 70.19, 70.20a, 70.20b, 70.21,

70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, 70.39, 70.42, 70.50,

70.51, 70.52, 70.53, 70.57, 70.58, 70.59, 70.61, 70.62, 70.64, 70.65,

70.72, 70.73, 70.74 and Appendix A.

* * * * *

5. The undesignated center heading ``EXEMPTIONS'' is redesignated

as ``Subpart B--Exemptions.''

Secs. 70.13a and 70.14 [Redesignated]

6. Sections 70.13a and 70.14 are redesignated as Secs. 70.14 and

70.17, respectively.

7. The undesignated center heading ``GENERAL LICENSES'' is

redesignated as ``Subpart C--General Licenses.''

8. The undesignated center heading ``LICENSE APPLICATIONS'' is

redesignated as ``Subpart D--License Applications.''

9. The undesignated center heading ``LICENSES'' is redesignated as

``Subpart E--Licenses.''

10. The undesignated center heading ``ACQUISITION, USE AND TRANSFER

OF SPECIAL NUCLEAR MATERIAL, CREDITORS' RIGHTS,'' is redesignated as

``Subpart F--Acquisition, Use, and Transfer of Special Nuclear

Material, Creditors' Rights.''

11. The undesignated center heading ``SPECIAL NUCLEAR MATERIAL

CONTROL RECORDS, REPORTS AND INSPECTIONS'' is redesignated as ``Subpart

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

Inspections.''

12. In Sec. 70.50 paragraph (c) is revised and paragraph (d) is

added to read as follows.

Sec. 70.50 Reporting requirements.

* * * *

of Special Nuclear

Material, Creditors' Rights.''

11. The undesignated center heading ``SPECIAL NUCLEAR MATERIAL

CONTROL RECORDS, REPORTS AND INSPECTIONS'' is redesignated as ``Subpart

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

Inspections.''

12. In Sec. 70.50 paragraph (c) is revised and paragraph (d) is

added to read as follows.

Sec. 70.50 Reporting requirements.

* * * *

(c) Preparation and submission of reports. Reports made by

licensees in response to the requirements of this section must be made

as follows:

(1) Licensees shall make reports required by paragraphs (a) and (b)

of this section, and by Sec. 70.74 and appendix A of this part if

applicable, by telephone to the NRC Operations Center.3 To

the extent that the information is available at the time of

notification, the information provided in these reports must include:

---------------------------------------------------------------------------

\3\ The commercial telephone number for the NRC Operations

Center is (301) 816-5100.

---------------------------------------------------------------------------

(i) Caller's name, position title and call back telephone number;

(ii) Date, time, and exact location of the event;

(iii) Description of the event, including;

(A) Radiological or chemical hazards involved including isotopes,

quantities, and chemical and physical form of any material released;

(B) Actual or potential health and safety consequences to the

workers, the public, and the environment, including relevant chemical

and radiation data for actual personnel exposures to radiation or

radioactive materials or chemicals (e.g., level of radiation exposure,

concentration of chemicals, and duration of exposure);

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

degradation or failure of structures, systems, equipment, components,

and activities of personnel relied on to prevent potential accidents or

mitigate their consequences; and

(D) Whether the remaining structures, systems, equipment,

co

ls (e.g., level of radiation exposure,

concentration of chemicals, and duration of exposure);

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

degradation or failure of structures, systems, equipment, components,

and activities of personnel relied on to prevent potential accidents or

mitigate their consequences; and

(D) Whether the remaining structures, systems, equipment,

components, and activities of personnel relied on to prevent potential

accidents or mitigate their consequences are available and reliable to

perform their function.

(iv) External conditions affecting the event;

(v) Additional actions taken by the licensee in response to the

event;

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

was terminated);

(vii) Current and planned site status, including any declared

emergency class;

(viii) Notifications related to the event that were made or are

planned to any local, State, or other Federal agencies;

(ix) Status of any press releases related to the event that were

made or are planned.

(2) Written report. Each licensee who makes a report required by

paragraph (a) or (b) of this section, or by Sec. 70.74 and appendix A

of this part if applicable, shall submit a written follow-up report

within 30 days of the initial report. Written reports prepared pursuant

to other regulations may be submitted to fulfill this requirement if

the report contains all of the necessary information and the

appropriate distribution is made. These written reports must be sent to

the U.S. Nuclear Regulatory Commission, Document Control Desk,

Washington, DC 20555, with a copy to the appropriate NRC regional

office listed in appendix D of 10 CFR part 20. The reports must include

the following:

y be submitted to fulfill this requirement if

the report contains all of the necessary information and the

appropriate distribution is made. These written reports must be sent to

the U.S. Nuclear Regulatory Commission, Document Control Desk,

Washington, DC 20555, with a copy to the appropriate NRC regional

office listed in appendix D of 10 CFR part 20. The reports must include

the following:

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

(ii) The probable cause of the event, including all factors that

contributed to the event and the manufacturer and model number (if

applicable) of any equipment that failed or malfunctioned;

(iii) Corrective actions taken or planned to prevent occurrence of

similar or identical events in the future and the results of any

evaluations or assessments; and

(iv) For licensees subject to subpart H of this part, whether the

event was identified and evaluated in the Integrated Safety Analysis.

(d) The provisions of Sec. 70.50 do not apply to licensees subject

to Sec. 50.72. They do apply to those part 50 licensees possessing

material licensed under part 70 who are not subject to the notification

requirements in Sec. 50.72.

13. The undesignated center heading ``MODIFICATION AND REVOCATION

OF LICENSES'' is redesignated as ``Subpart I--Modification and

Revocation of Licenses.''

Secs. 70.61 and 70.62 [Redesignated]

14. Sections 70.61 and 70.62 are redesignated as Secs. 70.81 and

70.82, respectively.

15. The undesignated center heading ``ENFORCEMENT'' is redesignated

as ``Subpart J--Enforcement.''

Secs. 70.71 and 70.72 [Redesignated]

16. Sections 70.71 and 70.72 are redesignated as Secs. 70.91 and

70.92, respectively.

17. In part 70, a new subpart H (Secs. 70.60-70.74) is added to

read as follows:

Subpart H--Additional Requirements for Certain Licensees Authorized

to Possess a Critical Mass of Special Nuclear Material

g ``ENFORCEMENT'' is redesignated

as ``Subpart J--Enforcement.''

Secs. 70.71 and 70.72 [Redesignated]

16. Sections 70.71 and 70.72 are redesignated as Secs. 70.91 and

70.92, respectively.

17. In part 70, a new subpart H (Secs. 70.60-70.74) is added to

read as follows:

Subpart H--Additional Requirements for Certain Licensees Authorized

to Possess a Critical Mass of Special Nuclear Material

Sec.

70.60 Applicability.

70.61 Performance requirements.

70.62 Safety program and integrated safety analysis.

70.64 Requirements for new facilities or new processes at existing

facilities.

70.65 Additional content of applications.

70.66 Additional requirements for approval of license application.

70.72 Facility changes and change process.

70.73 Renewal of licenses.

70.74 Additional reporting requirements.

Sec. 70.60 Applicability.

The regulations in Sec. 70.61 through Sec. 70.74 apply, in addition

to other applicable Commission regulations, to each applicant or

licensee that is or plans to be: authorized to possess greater than a

critical mass of special nuclear material, and engaged in enriched

uranium processing, fabrication of uranium fuel or fuel assemblies,

uranium enrichment, enriched uranium hexafluoride conversion, plutonium

processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap

recovery of special nuclear material, or any other activity that the

Commission determines could significantly affect public health and

safety. The regulations in Sec. 70.61 through Sec. 70.74 do not apply

to decommissioning activities performed pursuant to other applicable

Commission regulations including Sec. 70.25 and Sec. 70.38 of this

Part. Also, the regulations in Sec. 70.61 through Sec. 70.74 do not

apply to activities that are certified by the Commission pursuant to

Part 76 of this chapter or licensed by the Commission pursuant to other

parts of this chapter.

Sec. 70.61 Performance requirements.

ssioning activities performed pursuant to other applicable

Commission regulations including Sec. 70.25 and Sec. 70.38 of this

Part. Also, the regulations in Sec. 70.61 through Sec. 70.74 do not

apply to activities that are certified by the Commission pursuant to

Part 76 of this chapter or licensed by the Commission pursuant to other

parts of this chapter.

Sec. 70.61 Performance requirements.

(a) Each applicant or licensee shall evaluate, in the integrated

safety analysis performed in accordance with Sec. 70.62, its compliance

with the performance requirements in paragraphs (b), (c), and (d) of

this section.

(b) The risk of each credible high-consequence event must be

limited, unless the event is highly unlikely, through the application

of engineered controls, administrative controls, or both, that reduce

the likelihood of occurrence of the event or its consequence.

Application of additional controls is not required for those high-

consequence events demonstrated to be highly unlikely. High-consequence

events are those internally or externally initiated events that result

in:

(1) An acute worker dose of 1 Sv (100 rem) or greater total

effective dose equivalent;

(2) An acute dose of 0.25 Sv (25 rem) or greater total effective

dose equivalent to any individual located outside the controlled area

identified pursuant to paragraph (f) of this section;

(3) An intake of 30 mg or greater of uranium in soluble form by any

individual located outside the controlled area identified pursuant to

paragraph (f) of this section; or

(4) An acute chemical exposure to an individual from licensed

material or hazardous chemicals produced from licensed material that:

located outside the controlled area

identified pursuant to paragraph (f) of this section;

(3) An intake of 30 mg or greater of uranium in soluble form by any

individual located outside the controlled area identified pursuant to

paragraph (f) of this section; or

(4) An acute chemical exposure to an individual from licensed

material or hazardous chemicals produced from licensed material that:

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

(ii) Could lead to irreversible or other serious, long-lasting

health effects to any individual located outside the controlled area

identified pursuant to paragraph (f) of this section. If an applicant

possesses or plans to possess quantities of material capable of such

chemical exposures, then the applicant shall propose appropriate

quantitative standards for these health effects, as part of the

information submitted pursuant to Sec. 70.65 of this part.

(c) The risk of each credible intermediate-consequence event must

be limited, unless the event is unlikely, through the application of

engineered controls, administrative controls, or both, that reduce the

likelihood of occurrence of the event or its consequence. Application

of additional controls is not required for those intermediate-

consequence events demonstrated to be unlikely. Intermediate-

consequence events are those internally or externally initiated events,

that are not high-consequence events, that result in:

(1) An acute worker dose of 0.25 Sv (25 rem) or greater total

effective dose equivalent;

(2) An acute dose of 0.05 Sv (5 rem) or greater total effective

dose equivalent to any individual located outside the controlled area

identified pursuant to paragraph (f) of this section;

(3) A 24-hour averaged release of radioactive material outside the

restricted area in concentrations exceeding 5000 times the values in

table 2 of appendix B to 10 CFR part 20; or

(4) An acute chemical exposure to an individual from licensed

material or hazardous chemicals produced from licensed material that:

outside the controlled area

identified pursuant to paragraph (f) of this section;

(3) A 24-hour averaged release of radioactive material outside the

restricted area in concentrations exceeding 5000 times the values in

table 2 of appendix B to 10 CFR part 20; or

(4) An acute chemical exposure to an individual from licensed

material or hazardous chemicals produced from licensed material that:

(i) Could lead to irreversible or other serious, long-lasting

health effects to a worker, or

(ii) Could cause mild transient health effects to any individual

located outside the controlled area as specified in

paragraph (f) of this section. If an applicant possesses or plans to

possess quantities of material capable of such chemical exposures, then

the applicant shall propose appropriate quantitative standards for

these health effects, as part of the information submitted pursuant to

Sec. 70.65 of this part.

(d) In addition to complying with paragraphs (b) and (c) of this

section, the risk of nuclear criticality accidents must be limited by

assuring that under normal and credible abnormal conditions, all

nuclear processes are subcritical, including use of an approved margin

of subcriticality for safety. Preventive controls and measures must be

the primary means of protection against nuclear criticality accidents.

(e) Each engineered or administrative control or control system

necessary to comply with paragraphs (b), (c), or (d) of this section

shall be designated as an item relied on for safety. The safety

program, established and maintained pursuant to Sec. 70.62 of this

part, shall ensure that each item relied on for safety will be

available and reliable to perform its intended function when needed and

in the context of the performance requirements of this section.

essary to comply with paragraphs (b), (c), or (d) of this section

shall be designated as an item relied on for safety. The safety

program, established and maintained pursuant to Sec. 70.62 of this

part, shall ensure that each item relied on for safety will be

available and reliable to perform its intended function when needed and

in the context of the performance requirements of this section.

(f) Each licensee must establish a controlled area, as defined in

Sec. 20.1003, in which the licensee retains the authority to determine

all activities, including exclusion or removal of personnel and

property from the area. For the purpose of complying with the

performance requirements of this section, individuals who are not

workers, as defined in Sec. 70.4, may be permitted to perform ongoing

activities (e.g., at a facility not related to the licensed activities)

in the controlled area, if the licensee:

(1) Demonstrates and documents, in the integrated safety analysis,

that the risk for those individuals at the location of their activities

does not exceed the performance requirements of paragraphs (b)(2),

(b)(3), (b)(4)(ii), (c)(2), and (c)(4)(ii) of this section; or

(2) Provides: training in accordance with 10 CFR 19.12(a)(1)-(5) to

these individuals to ensure that they are aware of the risks associated

with accidents involving the licensed activities as determined by the

integrated safety analysis, and conspicuously posts and maintains

notices stating where the information in 10 CFR 19.11(a) may be

examined by these individuals. Under these conditions, the performance

requirements for workers specified in paragraphs (b) and (c) of this

section may be applied to these individuals.

Sec. 70.62 Safety program and integrated safety analysis.

as determined by the

integrated safety analysis, and conspicuously posts and maintains

notices stating where the information in 10 CFR 19.11(a) may be

examined by these individuals. Under these conditions, the performance

requirements for workers specified in paragraphs (b) and (c) of this

section may be applied to these individuals.

Sec. 70.62 Safety program and integrated safety analysis.

(a) Safety program. (1) Each licensee shall establish and maintain

a safety program that demonstrates compliance with the performance

requirements of Sec. 70.61. The safety program may be graded such that

management measures applied are commensurate with the reduction of the

risk attributable to that item. The three elements of the safety

program; namely, process safety information, integrated safety

analysis, and management measures, are described in paragraphs (b)

through (d) of this section.

(2) Each licensee shall establish and maintain records that

demonstrate compliance with the requirements of paragraphs (b) through

(d) of this section.

(3) Each licensee shall establish and maintain a log, available for

NRC inspection, documenting each discovery that an item relied on for

safety or management measure has failed to perform its function either

in the context of the performance requirements of Sec. 70.61 or upon

demand. This log must identify the item relied on for safety or

management measure that has failed and the safety function affected,

the date of discovery, date (or estimated date) of the failure,

duration (or estimated duration) of the time that the item was unable

to perform its function, any other affected items relied on for safety

or management measures and their safety function, affected processes,

cause of the failure, whether the failure was in the context of the

performance requirements or upon demand or both, and any corrective or

compensatory action that was taken

failure,

duration (or estimated duration) of the time that the item was unable

to perform its function, any other affected items relied on for safety

or management measures and their safety function, affected processes,

cause of the failure, whether the failure was in the context of the

performance requirements or upon demand or both, and any corrective or

compensatory action that was taken. The log must be initiated at the

time of discovery and updated promptly upon the conclusion of each

investigation of a failure of an item relied on for safety or

management measure.

(b) Process safety information. Each licensee or applicant shall

maintain process safety information to enable the performance of an

integrated safety analysis. This process safety information must

include information pertaining to the hazards of the materials used or

produced in the process, information pertaining to the technology of

the process, and information pertaining to the equipment in the

process.

(c) Integrated safety analysis. (1) Each licensee or applicant

shall conduct an integrated safety analysis, that is of appropriate

detail for the complexity of the process, that identifies:

(i) Radiological hazards related to possessing or processing

licensed material at its facility;

(ii) Chemical hazards of licensed material and hazardous chemicals

produced from licensed material;

(iii) Facility hazards which could affect the safety of licensed

materials and thus present an increased radiological risk;

(iv) Potential accident sequences caused by process deviations or

other events internal to the plant and credible external events,

including natural phenomena;

facility;

(ii) Chemical hazards of licensed material and hazardous chemicals

produced from licensed material;

(iii) Facility hazards which could affect the safety of licensed

materials and thus present an increased radiological risk;

(iv) Potential accident sequences caused by process deviations or

other events internal to the plant and credible external events,

including natural phenomena;

(v) The consequence and the likelihood of occurrence of each

potential accident sequence identified pursuant to paragraph (c)(1)(iv)

of this section, and the methods used to determine the consequences and

likelihoods; and

(vi) Each item relied on for safety identified pursuant to

Sec. 70.61(e) of this part, the characteristics of its preventive,

mitigative, or other safety function, and the assumptions and

conditions under which the item is relied upon to support compliance

with the performance requirements of Sec. 70.61.

(2) Integrated safety analysis team qualifications. In order to

assure the adequacy of the integrated safety analysis, the analysis

must be performed by a team with expertise in engineering and process

operations. The team shall include at least one person who has

experience and knowledge specific to each process being evaluated, and

persons who have experience in nuclear criticality safety, radiation

safety, fire safety, and chemical process safety. One member of the

team must be knowledgeable in the specific integrated safety analysis

methodology being used.

(3) Requirements for existing licensees. Notwithstanding other

provisions regarding the effective date for part 70, subpart H,

requirements, licensees shall comply with the provisions in paragraphs

(c)(3)(i), (ii), and (iii) of this section beginning on [the date of

publication of the final rule]. Individuals holding an NRC license on

[the date of publication of the final rule] shall, with regard to

existing licensed activities:

s. Notwithstanding other

provisions regarding the effective date for part 70, subpart H,

requirements, licensees shall comply with the provisions in paragraphs

(c)(3)(i), (ii), and (iii) of this section beginning on [the date of

publication of the final rule]. Individuals holding an NRC license on

[the date of publication of the final rule] shall, with regard to

existing licensed activities:

(i) Within 6 months of the effective date of the rule, submit for

NRC approval, a plan that describes the integrated safety analysis

approach that will be used, the processes that will be analyzed, and

the schedule for completing the analysis of each process.

(ii) Within 4 years of the effective date of the rule, complete an

integrated safety analysis, correct all unacceptable performance

deficiencies, and submit an integrated safety analysis summary in

accordance with Sec. 70.65 or the approved

plan submitted under paragraph (c)(3)(i) of this section.

(iii) Pending the correction of unacceptable performance

deficiencies identified during the conduct of the integrated safety

analysis, the licensee shall implement appropriate compensatory

measures to ensure adequate protection.

(d) Management measures. Each applicant or licensee shall establish

management measures to provide continuing assurance of compliance with

the performance requirements of Sec. 70.61. The measures applied to a

particular engineered or administrative control or control system may

be commensurate with the reduction of the risk attributable to that

control or control system. The management measures shall ensure that

engineered and administrative controls and control systems that are

identified as items relied on for safety pursuant to Sec. 70.61(e) of

this part are designed, implemented, and maintained, as necessary, to

ensure they are available and reliable to perform their function when

needed, in the context of compliance with the performance requirements

of Sec. 70.61 of this part.

shall ensure that

engineered and administrative controls and control systems that are

identified as items relied on for safety pursuant to Sec. 70.61(e) of

this part are designed, implemented, and maintained, as necessary, to

ensure they are available and reliable to perform their function when

needed, in the context of compliance with the performance requirements

of Sec. 70.61 of this part.

Sec. 70.64 Requirements for new facilities or new processes at

existing facilities.

(a) Baseline design criteria. Each prospective applicant or

licensee shall address the following baseline design criteria in the

design of new facilities. Each existing licensee shall address the

following baseline design criteria in the design of new processes at

existing facilities that require a license amendment under Sec. 70.72.

The baseline design criteria must be applied to the design of new

facilities and new processes, but do not require retrofits to existing

facilities or existing processes (e.g., those housing or adjacent to

the new process); however, all facilities and processes must comply

with the performance requirements in Sec. 70.61. Licensees shall

maintain the application of these criteria unless the evaluation

performed pursuant to paragraph (c) of this section demonstrates that a

given item is not relied on for safety or does not require adherence to

the specified criteria.

(1) Quality standards and records. The design must be developed and

implemented in accordance with management measures, to provide adequate

assurance that items relied on for safety will be available and

reliable to perform their function when needed. Appropriate records of

these items must be maintained by or under the control of the licensee

throughout the life of the facility.

(2) Natural phenomena hazards. The design must provide for adequate

protection against natural phenomena with consideration of the most

severe documented historical events for the site.

safety will be available and

reliable to perform their function when needed. Appropriate records of

these items must be maintained by or under the control of the licensee

throughout the life of the facility.

(2) Natural phenomena hazards. The design must provide for adequate

protection against natural phenomena with consideration of the most

severe documented historical events for the site.

(3) Fire protection. The design must provide for adequate

protection ag

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