[Federal Register Volume 64, Number 146 (Friday, July 30, 1999)]
[Proposed Rules]
[Pages 41338-41357]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-19363]
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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.
ADDRESSES: Submit comments to: Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC, 20555-0001, Attention:
Rulemakings and Adjudications Staff.
Deliver comments to: 11555 Rockville Pike, Rockville, Maryland,
between 7:30 a.m. and 4:15 p.m. on Federal workdays.
You may also provide comments via NRC's interactive rulemaking
website through the NRC home page (http://www.nrc.gov). From the home
page, select ``Rulemaking'' from the tool bar at the bottom of the
page. The interactive rulemaking website can then be accessed by
selecting ``Rulemaking Forum.'' This site provides the ability to
upload comments as files (any format), if your web browser supports
that function. For information about the interactive rulemaking
website, contact Ms. Carol Gallagher by telephone at (301) 415-5905 or
e-mail cag@nrc.gov.
FOR FURTHER INFORMATION CONTACT: Theodore S. Sherr, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC, 20555-0001, telephone (301) 415-7218; e-mail
tss@nrc.gov.
SUPPLEMENTARY INFORMATION:
I. Background
II. Description of Proposed Action
I. Background
A near-criticality incident at a low enriched fuel fabrication
facility in May 1991 prompted NRC to review its safety regulations for
licensees that possess and process large quantities of SNM. [See NUREG-
1324, ``Proposed Method
[[Page 41339]]
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
(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
(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:
(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
(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
[[Page 41340]]
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. 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.
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
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. 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
[[Page 41341]]
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.
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. 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. 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. 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). 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
[[Page 41342]]
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)]
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.
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). 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. 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
(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
[[Page 41343]]
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.
ERPG criteria are widely used by those involved in assessing or
responding to the release of hazardous chemicals including ``* * *
community emergency planners and response specialists, air dispersion
modelers, industrial process safety engineers, implementers of
environmental regulations such as the Superfund Amendment and
Reauthorization Act, industrial hygienists, and toxicologists,
transportation safety engineers, fire protection specialists, and
government agencies. * * *'' (DOE Risk Management Quarterly, 1997).
Despite their general acceptance, there are currently only
approximately 80 ERPG criteria available, and some chemicals of
importance (e.g., nitric acid) are not covered.
The qualitative language in the performance requirement allows the
applicant/licensee to propose and adopt an appropriate standard, which
may be an AEGL or ERPG standard, or where there is no AEGL or ERPG
value available, the applicant may develop or adopt a criterion that is
comparable in severity to those that have been established for other
chemicals. For example, for the worker performance requirement,
existing criteria that can be used by licensees to define appropriate
concentration levels to satisfy the performance requirement are the
AEGL-3 and ERPG-3. AEGL-3 is defined as ``The airborne concentration
(expressed in ppm or mg/m3) of a substance at or above which
it is predicted that the general population, including susceptible, but
excluding hypersusceptible, individuals, 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. 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. 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). 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
[[Page 41344]]
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. 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. 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.
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. 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. 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.
[[Page 41345]]
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. 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.
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. 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. 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
[[Page 41346]]
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.
(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. 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.
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. 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. 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.
[[Page 41347]]
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.
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. 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.
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
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. 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
[[Page 41348]]
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. 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.
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. 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. 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
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
[[Page 41349]]
would be reviewed at one time. However, with the proposed ``living
license,'' changes to the facility will be reviewed by NRC either
before changes are made, or relatively shortly thereafter. As a result,
review of the license renewal application is expected to be performed
with minimal additional review of the licensee's safety program. This
approval would be contingent on the licensee satisfying any
requirements associated with the National Environmental Policy Act of
1969 as implemented in 10 CFR Part 51.
Section 70.74 Additional Reporting Requirements
The new requirements that would be incorporated in the proposed
amendments to Part 70 would revise the reporting of events to NRC. This
new approach, based on consideration of the risk and consequences
established in 10 CFR 70.61(b) is intended to replace and expand on the
approach licensees have currently been using for reporting criticality
events under Bulletin 91-01. The new approach would cover all types of
events, not just criticality events, and establish a timeframe for
reporting that is 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. 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. 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
Graig, D.K., et al., ``Alternative Guideline Limits for Chemicals
Without Environmental Response Planning Guidelines,'' American
Industrial Hygiene Association Journal, 1995.
Fisher, D.R., Hui, T.E., Yurconic, M., and Johnson, J.R., ``Uranium
Hexafluoride Public Risk,'' Pacific Northwest National Laboratory,
PNL-10065, Richland, WA, August 1994.
National Council on Radiation Protection and Measurements (NCRP),
``Basic Radiation Protection Criteria,'' NCRP Report No. 39,
Washington, DC, 1971.
National Council on Radiation Protection and Measurements (NCRP),
``Recommendations on Limits for Exposure to Ionizing Radiation,''
NCRP Report No. 91, Washington, DC, 1987.
U.S. Nuclear Regulatory Commission, ``Proposed Methods for
Regulating Major Materials Licensees,'' NUREG-1324, Washington, DC,
February 1992.
U.S. Nuclear Regulatory Commission/ Occupational Safety and Health
Administration (OSHA), ``Memorandum of Understanding Between NRC and
OSHA; Worker Protection at NRC-Licensed Facilities'' (53 FR 43950;
October 31, 1988).
U.S. Nuclear Regulatory Commission, ``Certification of Gaseous
Diffusion Plants'' (59 FR 48944; September 23, 1994).
U.S. Nuclear Regulatory Commission, ``Abnormal Occurrence Reports:
Implementation of Section 208 of Energy Reorganization Act of 1974''
(61 FR 67072; December 19, 1996).
U.S. Nuclear Regulatory Commission, ``Site Decommissioning
Management Plan,'' NUREG-1444, Washington, DC, October 1993.
U.S. Nuclear Regulatory Commission, ``Strategic Plan, Fiscal Year
1997--Fiscal Year 2002,'' NUREG-1614, Washington, DC, September
1997.
U.S. Environmental Protection Agency, ``Manual of Protective Action
Guides and Protective Actions for Nuclear Incidents,'' EPA-400-R-92-
001, May 1992.
U.S. Nuclear Regulatory Commission, ``Instruction Concerning Risks
from Occupational Radiation Exposure,'' Regulatory Guide 8.29, Rev.
1, February 1996.
Theide, L., ``Emergency Information Where It's Needed,'' DOE Risk
Management Quarterly, Vol 5, No 2, Richland, WA, May 1997.
These documents are available for inspection and copying for a fee
at the NRC Public Document Room, 2120 L Street, 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
[[Page 41350]]
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. 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. 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.
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. 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: btm1@nrc.gov.
Paperwork Reduction Act Statement
This proposed rule amends information collection requirements that
are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501, et
seq.). This rule has been submitted to the Office of Management and
Budget (OMB) for review and approval of the paperwork requirements.
The public reporting burden for this information collection is
estimated to average 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. How can the burden of the information collection be minimized,
including the use of automated collection techniques?
Send comments on any aspect of this proposed information
collection, including suggestions for reducing the burden, to the
Records Management Branch (T-6-F33), U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, or by Internet electronic mail
at bjs1@nrc.gov; and to the Desk Officer, Office of Information and
Regulatory Affairs, NEOB-10202 (3150-0009), Office of Management and
Budget, Washington, DC 20503.
Comments to OMB on the information collections or on the above
issues should be submitted by 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.
[[Page 41351]]
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: btm1@nrc.gov.
The Commission requests public comment on the draft Regulatory
Analysis. Comments on the draft analysis may be submitted to NRC as
indicated under the ADDRESSES heading.
Regulatory Flexibility Certification
As required by the Regulatory Flexibility Act, as amended, 5 U.S.C.
605(b), the Commission certifies that this proposed rule, if adopted,
would not have a significant economic impact on a substantial number of
small entities. This proposed rule would affect facilities that are
authorized to possess a critical mass of SNM and who are engaged in one
of the following activities: (a) enriched uranium processing; (b)
fabrication of uranium fuel or fuel assemblies; (c) uranium enrichment;
(d) enriched uranium hexafluoride conversion; (e) plutonium processing;
(f) fabrication of mixed-oxide fuel or fuel assemblies; (g) scrap
recovery of special nuclear material; or (h) any other activity
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
Criminal penalties, Hazardous materials transportation, Material
control and accounting, Nuclear materials, Packaging and containers,
Radiation protection, Reporting and recordkeeping requirements,
Scientific equipment, Security measures, Special nuclear material.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 553, NRC is proposing to adopt
the following amendments to Part 70.
PART 70--DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL
1. The authority citation for part 70 continues to read as follows:
Authority: Secs. 51, 53, 161, 182, 183, 68 Stat. 929, 930, 948,
953, 954, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.
2071, 2073, 2201, 2232, 2233, 2282, 2297f); secs. 201, as amended,
202, 204, 206, 88 Stat. 1242, as amended, 1244, 1245, 1246 (42
U.S.C. 5841, 5842, 5845, 5846). Sec. 193, 104 Stat. 2835, as amended
by Pub. L. 104-134, 110 Stat. 1321, 1321-349 (42 U.S.C. 2243).
Sections 70.1(c) and 70.20a(b) also issued under secs. 135, 141,
Pub. L. 97-425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
Section 70.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 70.21(g) also issued under sec. 122,
68 Stat. 939 (42 U.S.C. 2152). Section 70.31 also issued under sec.
57d, Pub. L. 93-377, 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36
and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42
U.S.C. 2234). Section 70.61 also issued under secs. 186, 187, 68
Stat. 955 (42 U.S.C. 2236, 2237). Section 70.62 also issued under
sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).
2. The undesignated center heading ``GENERAL PROVISIONS'' is
redesignated as ``Subpart A--General Provisions.''
3. In 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.
* * * * *
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
[[Page 41352]]
unlikely, independent, and concurrent changes in process conditions
before a criticality accident is possible.
* * * * *
Hazardous chemicals produced from licensed materials means
substances having licensed material as precursor compound(s) or
substances that physically or chemically interact with licensed
materials; that are toxic, explosive, flammable, corrosive, or reactive
to the extent that they can endanger life or health if not adequately
controlled. These include substances commingled with licensed material,
and include substances such as hydrogen fluoride that is produced by
the reaction of uranium hexafluoride and water, but do not include
substances prior to process addition to licensed material or after
process separation from licensed material.
Integrated safety analysis (ISA) means a systematic analysis to
identify plant and external hazards and their potential for initiating
accident sequences, the potential accident sequences, their likelihood
and consequences, and the items relied on for safety. As used here,
integrated means joint consideration of, and protection from, all
relevant hazards, including radiological, nuclear criticality, fire,
and chemical. However, with respect to compliance with the regulations
of this part, the NRC requirement is limited to consideration of the
effects of all relevant hazards on radiological safety, prevention of
nuclear criticality accidents, or chemical hazards directly associated
with NRC licensed radioactive material.
Integrated safety analysis summary means the document submitted
with the license application, license amendment application, or license
renewal application that provides a synopsis of the results of the
integrated safety analysis and contains the information specified in
Sec. 70.65(b).
Items relied on for safety means structures, systems, equipment,
components, and activities of personnel that are relied on to prevent
potential accidents at a facility that could exceed the performance
requirements in Sec. 70.61 or to mitigate their potential consequences.
This does not limit the licensee from identifying additional
structures, systems, equipment, components, or activities of personnel
(i.e., beyond those in the minimum set necessary for compliance with
the performance requirements) as items relied on for safety.
* * * * *
Management measures mean the functions performed by the licensee,
generally on a continuing basis, that are applied to items relied upon
for safety, to ensure the items are available and reliable to perform
their functions when needed. Management measures include configuration
management, maintenance, training and qualifications, procedures,
audits and assessments, incident investigations, records management,
and other quality assurance elements.
* * * * *
Unacceptable performance deficiencies mean deficiencies in the
items relied on for safety or the management measures that need to be
corrected to ensure an adequate level of protection as defined in 10
CFR 70.61(b), (c), or (d).
* * * * *
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.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Secs. 70.9, 70.17, 70.19, 70.20a, 70.20b, 70.21,
70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, 70.39, 70.42, 70.50,
70.51, 70.52, 70.53, 70.57, 70.58, 70.59, 70.61, 70.62, 70.64, 70.65,
70.72, 70.73, 70.74 and Appendix A.
* * * * *
5. The undesignated center heading ``EXEMPTIONS'' is redesignated
as ``Subpart B--Exemptions.''
Secs. 70.13a and 70.14 [Redesignated]
6. Sections 70.13a and 70.14 are redesignated as Secs. 70.14 and
70.17, respectively.
7. The undesignated center heading ``GENERAL LICENSES'' is
redesignated as ``Subpart C--General Licenses.''
8. The undesignated center heading ``LICENSE APPLICATIONS'' is
redesignated as ``Subpart D--License Applications.''
9. The undesignated center heading ``LICENSES'' is redesignated as
``Subpart E--Licenses.''
10. The undesignated center heading ``ACQUISITION, USE AND TRANSFER
OF SPECIAL NUCLEAR MATERIAL, CREDITORS' RIGHTS,'' is redesignated as
``Subpart F--Acquisition, Use, and Transfer of Special Nuclear
Material, Creditors' Rights.''
11. The undesignated center heading ``SPECIAL NUCLEAR MATERIAL
CONTROL RECORDS, REPORTS AND INSPECTIONS'' is redesignated as ``Subpart
G--Special Nuclear Material Control Records, Reports, and
Inspections.''
12. In Sec. 70.50 paragraph (c) is revised and paragraph (d) is
added to read as follows.
Sec. 70.50 Reporting requirements.
* * * *
(c) Preparation and submission of reports. Reports made by
licensees in response to the requirements of this section must be made
as follows:
(1) Licensees shall make reports required by paragraphs (a) and (b)
of this section, and by Sec. 70.74 and appendix A of this part if
applicable, by telephone to the NRC Operations Center.3 To
the extent that the information is available at the time of
notification, the information provided in these reports must include:
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\3\ The commercial telephone number for the NRC Operations
Center is (301) 816-5100.
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(i) Caller's name, position title and call back telephone number;
(ii) Date, time, and exact location 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,
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;
[[Page 41353]]
(viii) Notifications related to the event that were made or are
planned to any local, State, or other Federal agencies;
(ix) Status of any press releases related to the event that were
made or are planned.
(2) Written report. Each licensee who makes a report required by
paragraph (a) or (b) of this section, or by Sec. 70.74 and appendix A
of this part if applicable, shall submit a written follow-up report
within 30 days of the initial report. Written reports prepared pursuant
to other regulations may be submitted to fulfill this requirement if
the report contains all of the necessary information and the
appropriate distribution is made. These written reports must be sent to
the U.S. Nuclear Regulatory Commission, Document Control Desk,
Washington, DC 20555, with a copy to the appropriate NRC regional
office listed in appendix D of 10 CFR part 20. The reports must include
the following:
(i) Complete applicable information required by Sec. 70.50(c)(1);
(ii) The probable cause of the event, including all factors that
contributed to the event and the manufacturer and model number (if
applicable) of any equipment that failed or malfunctioned;
(iii) Corrective actions taken or planned to prevent occurrence of
similar or identical events in the future and the results of any
evaluations or assessments; and
(iv) For licensees subject to subpart H of this part, whether the
event was identified and evaluated in the Integrated Safety Analysis.
(d) The provisions of Sec. 70.50 do not apply to licensees subject
to Sec. 50.72. They do apply to those part 50 licensees possessing
material licensed under part 70 who are not subject to the notification
requirements in Sec. 50.72.
13. The undesignated center heading ``MODIFICATION AND REVOCATION
OF LICENSES'' is redesignated as ``Subpart I--Modification and
Revocation of Licenses.''
Secs. 70.61 and 70.62 [Redesignated]
14. Sections 70.61 and 70.62 are redesignated as Secs. 70.81 and
70.82, respectively.
15. The undesignated center heading ``ENFORCEMENT'' is redesignated
as ``Subpart J--Enforcement.''
Secs. 70.71 and 70.72 [Redesignated]
16. Sections 70.71 and 70.72 are redesignated as Secs. 70.91 and
70.92, respectively.
17. In part 70, a new subpart H (Secs. 70.60-70.74) is added to
read as follows:
Subpart H--Additional Requirements for Certain Licensees Authorized
to Possess a Critical Mass of Special Nuclear Material
Sec.
70.60 Applicability.
70.61 Performance requirements.
70.62 Safety program and integrated safety analysis.
70.64 Requirements for new facilities or new processes at existing
facilities.
70.65 Additional content of applications.
70.66 Additional requirements for approval of license application.
70.72 Facility changes and change process.
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.
(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:
(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:
(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
[[Page 41354]]
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.
(f) Each licensee must establish a controlled area, as defined in
Sec. 20.1003, in which the licensee retains the authority to determine
all activities, including exclusion or removal of personnel and
property from the area. For the purpose of complying with the
performance requirements of this section, individuals who are not
workers, as defined in Sec. 70.4, may be permitted to perform ongoing
activities (e.g., at a facility not related to the licensed activities)
in the controlled area, if the licensee:
(1) Demonstrates and documents, in the integrated safety analysis,
that the risk for those individuals at the location of their activities
does not exceed the performance requirements of paragraphs (b)(2),
(b)(3), (b)(4)(ii), (c)(2), and (c)(4)(ii) of this section; or
(2) Provides: training in accordance with 10 CFR 19.12(a)(1)-(5) to
these individuals to ensure that they are aware of the risks associated
with accidents involving the licensed activities as determined by the
integrated safety analysis, and conspicuously posts and maintains
notices stating where the information in 10 CFR 19.11(a) may be
examined by these individuals. Under these conditions, the performance
requirements for workers specified in paragraphs (b) and (c) of this
section may be applied to these individuals.
Sec. 70.62 Safety program and integrated safety analysis.
(a) Safety program. (1) Each licensee shall establish and maintain
a safety program that demonstrates compliance with the performance
requirements of Sec. 70.61. The safety program may be graded such that
management measures applied are commensurate with the reduction of the
risk attributable to that item. 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. The log must be initiated at the
time of discovery and updated promptly upon the conclusion of each
investigation of a failure of an item relied on for safety or
management measure.
(b) Process safety information. Each licensee or applicant shall
maintain process safety information to enable the performance of an
integrated safety analysis. This process safety information must
include information pertaining to the hazards of the materials used or
produced in the process, information pertaining to the technology of
the process, and information pertaining to the equipment in the
process.
(c) Integrated safety analysis. (1) Each licensee or applicant
shall conduct an integrated safety analysis, that is of appropriate
detail for the complexity of the process, that identifies:
(i) Radiological hazards related to possessing or processing
licensed material at its facility;
(ii) Chemical hazards of licensed material and hazardous chemicals
produced from licensed material;
(iii) Facility hazards which could affect the safety of licensed
materials and thus present an increased radiological risk;
(iv) Potential accident sequences caused by process deviations or
other events internal to the plant and credible external events,
including natural phenomena;
(v) The consequence and the likelihood of occurrence of each
potential accident sequence identified pursuant to paragraph (c)(1)(iv)
of this section, and the methods used to determine the consequences and
likelihoods; and
(vi) Each item relied on for safety identified pursuant to
Sec. 70.61(e) of this part, the characteristics of its preventive,
mitigative, or other safety function, and the assumptions and
conditions under which the item is relied upon to support compliance
with the performance requirements of Sec. 70.61.
(2) Integrated safety analysis team qualifications. In order to
assure the adequacy of the integrated safety analysis, the analysis
must be performed by a team with expertise in engineering and process
operations. The team shall include at least one person who has
experience and knowledge specific to each process being evaluated, and
persons who have experience in nuclear criticality safety, radiation
safety, fire safety, and chemical process safety. One member of the
team must be knowledgeable in the specific integrated safety analysis
methodology being used.
(3) Requirements for existing licensees. Notwithstanding other
provisions regarding the effective date for part 70, subpart H,
requirements, licensees shall comply with the provisions in paragraphs
(c)(3)(i), (ii), and (iii) of this section beginning on [the date of
publication of the final rule]. Individuals holding an NRC license on
[the date of publication of the final rule] shall, with regard to
existing licensed activities:
(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
[[Page 41355]]
plan submitted under paragraph (c)(3)(i) of this section.
(iii) Pending the correction of unacceptable performance
deficiencies identified during the conduct of the integrated safety
analysis, the licensee shall implement appropriate compensatory
measures to ensure adequate protection.
(d) Management measures. Each applicant or licensee shall establish
management measures to provide continuing assurance of compliance with
the performance requirements of Sec. 70.61. The measures applied to a
particular engineered or administrative control or control system may
be commensurate with the reduction of the risk attributable to that
control or control system. The management measures shall ensure that
engineered and administrative controls and control systems that are
identified as items relied on for safety pursuant to Sec. 70.61(e) of
this part are designed, implemented, and maintained, as necessary, to
ensure they are available and reliable to perform their function when
needed, in the context of compliance with the performance requirements
of Sec. 70.61 of this part.
Sec. 70.64 Requirements for new facilities or new processes at
existing facilities.
(a) Baseline design criteria. Each prospective applicant or
licensee shall address the following baseline design criteria in the
design of new facilities. Each existing licensee shall address the
following baseline design criteria in the design of new processes at
existing facilities that require a license amendment under Sec. 70.72.
The baseline design criteria must be applied to the design of new
facilities and new processes, but do not require retrofits to existing
facilities or existing processes (e.g., those housing or adjacent to
the new process); however, all facilities and processes must comply
with the performance requirements in Sec. 70.61. Licensees shall
maintain the application of these criteria unless the evaluation
performed pursuant to paragraph (c) of this section demonstrates that a
given item is not relied on for safety or does not require adherence to
the specified criteria.
(1) Quality standards and records. The design must be developed and
implemented in accordance with management measures, to provide adequate
assurance that items relied on for safety will be available and
reliable to perform their function when needed. Appropriate records of
these items must be maintained by or under the control of the licensee
throughout the life of the facility.
(2) Natural phenomena hazards. The design must provide for adequate
protection against natural phenomena with consideration of the most
severe documented historical events for the site.
(3) Fire protection. The design must provide for adequate
protection against fires and explosions.
(4) Environmental and dynamic effects. The design must provide for
adequate protection from environmental conditions and dynamic effects
associated with normal operations, maintenance, testing, and postulated
accidents that could lead to loss of safety functions.
(5) Chemical protection. The design must provide for adequate
protection against chemical risks produced from licensed material,
plant conditions which affect the safety of licensed material, and
hazardous chemicals produced from licensed material.
(6) Emergency capability. The design must provide for emergency
capability to maintain control of:
(i) Licensed material;
(ii) Evacuation of personnel; and
(iii) Onsite emergency facilities and services that facilitate the
use of available offsite services.
(7) Utility services. The design must provide for continued
operation of essential utility services.
(8) Inspection, testing, and maintenance. The design of items
relied on for safety must provide for adequate inspection, testing, and
maintenance, to ensure their availability and reliability to perform
their function when needed.
(9) Criticality control. The design must provide for criticality
control including adherence to the double contingency principle.
(10) Instrumentation and controls. The design must provide for
inclusion of instrumentation and control systems to monitor and control
the behavior of items relied on for safety.
(b) Facility and system design and plant layout must be based on
defense-in-depth practices.4 The design process must
incorporate, to the extent practicable:
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\4\ As used in Sec. 70.64, defense-in-depth practices means a
design philosophy, applied from the outset and through completion of
the design, that is based on providing successive levels of
protection such that health and safety will not be wholly dependent
upon any single element of the design, construction, maintenance, or
operation of the facility. The net effect of incorporating defense-
in-depth practices is a conservatively designed facility and system
that will exhibit greater tolerance to failures and external
challenges. The risk insights obtained through performance of the
integrated safety analysis can be then used to supplement the final
design by focusing attention on the prevention and mitigation of the
higher-risk potential accidents.
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(1) Preference for the selection of engineered controls over
administrative controls to increase overall system reliability; and
(2) Features that enhance safety by reducing challenges to items
relied on for safety.
Sec. 70.65 Additional content of applications.
(a) In addition to the contents required by Sec. 70.22, each
application must include a description of the applicant's safety
program established under Sec. 70.62, including the integrated safety
analysis summary and a description of the management measures.
(b) The integrated safety analysis summary must be submitted with
the license or renewal application (and amendment application as
necessary), but shall not be incorporated in the license. However,
changes to the integrated safety analysis summary shall meet the
conditions of Sec. 70.72. The integrated safety analysis summary must
contain:
(1) A general description of the site with emphasis on those
factors that could affect safety (i.e., meteorology, seismology);
(2) A general description of the facility with emphasis on those
areas that could affect safety, including an identification of the
controlled area boundaries;
(3) A description of each process (defined as a single reasonably
simple integrated unit operation within an overall production line)
analyzed in the integrated safety analysis in sufficient detail to
understand the theory of operation; and, for each process, the hazards
that were identified in the integrated safety analysis pursuant to
Sec. 70.62(c)(1)(i)-(iii) and a general description of the types of
accident sequences;
(4) Information that demonstrates the licensee's compliance with
the performance requirements of Sec. 70.61; the requirements for
criticality monitoring and alarms in Sec. 70.24; and, if applicable,
the requirements of Sec. 70.64;
(5) A description of the team, qualifications, and the methods used
to perform the integrated safety analysis;
(6) A list briefly describing all items relied on for safety which
are identified pursuant to Sec. 70.61(e) in sufficient detail to
understand their functions in relation to the performance requirements
of Sec. 70.61;
(7) A description of the proposed quantitative standards used to
assess the consequences from acute chemical
[[Page 41356]]
exposure to licensed material or chemicals produced from licensed
materials which are on-site, or expected to be on-site as described in
Sec. 70.61(b)(4) and (c)(4);
(8) A descriptive list that identifies all items relied on for
safety that are the sole item preventing or mitigating an accident
sequence that exceeds the performance requirements of Sec. 70.61; and
(9) A description of the definitions of likely, unlikely, highly
unlikely, and credible as used in the evaluations in the integrated
safety analysis.
Sec. 70.66 Additional requirements for approval of license
application.
An application for a license from an applicant subject to subpart H
will be approved if the Commission determines that the applicant has
complied with the requirements of Sec. 70.21, Sec. 70.22, Sec. 70.23
and Sec. 70.60 through Sec. 70.65.
Sec. 70.72 Facility changes and change process.
(a) The licensee shall establish a configuration management system
to evaluate, implement, and track each change to the site, structures,
processes, systems, equipment, components, computer programs, and
activities of personnel. This system must be documented in written
procedures and must assure that the following are addressed prior to
implementing any change:
(1) The technical basis for the change;
(2) Impact of the change on safety and health or control of
licensed material;
(3) Modifications to existing operating procedures including any
necessary training or retraining before operation;
(4) Authorization requirements for the change;
(5) For temporary changes, the approved duration (e.g., expiration
date) of the change; and
(6) The impacts or modifications to the integrated safety analysis,
integrated safety analysis summary, or other safety program
information, developed in accordance with Sec. 70.62.
(b) Any change to site, structures, processes, systems, equipment,
components, computer programs, and activities of personnel must be
evaluated by the licensee as specified in paragraph (a) of this
section, before the change is implemented. The evaluation of the change
must determine, before the change is implemented, if an amendment to
the license is required to be submitted in accordance with Sec. 70.34.
(c) The licensee may make changes to the site, structures,
processes, systems, equipment, components, computer programs, and
activities of personnel, without prior Commission approval, if the
change:
(1) Does not:
(i) Create new types \5\ of accident sequences that, unless
mitigated or prevented, would exceed the performance requirements of
Sec. 70.61 and that have not previously been described in the
integrated safety analysis summary; or
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\5\ Any change in the defining characteristics of the elements
of an accident sequence may change the ``type'' of the accident
sequence for a given process. For example, a new type of accident
could involve a different initiator, significant changes in the
consequence, or a change in the safety function of a control (e.g.,
temperature limiting device versus a flow limiting device).
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(ii) Use new processes, technologies, or control systems for which
the licensee has no prior experience;
(2) Does not remove, without at least an equivalent replacement of
the safety function, an item relied on for safety that is listed in the
integrated safety analysis summary;
(3) Does not alter any item relied on for safety, listed in the
integrated safety analysis summary, that is the sole item preventing or
mitigating an accident sequence that exceeds the performance
requirements of Sec. 70.61; and
(4) Is not otherwise prohibited by this section, license condition,
or order.
(d)(1) For any changes that affect the integrated safety analysis
summary, as submitted in accordance with Sec. 70.65, but do not require
NRC pre-approval, the licensee shall submit revised pages to the
integrated safety analysis summary, to NRC, within 90 days of the
change.
(2) For changes that require pre-approval under Sec. 70.72, the
licensee shall submit an amendment request to the NRC in accordance
with Sec. 70.34 and Sec. 70.65.
(3) A brief summary of all changes to the records required by
Sec. 70.62(a)(2) of this part, that are made without prior Commission
approval, must be submitted to NRC every 12 months.
(e) If a change covered by Sec. 70.72 is made, the affected on-site
documentation must be updated promptly.
(f) The licensee shall maintain records of changes to its facility
carried out under this section. These records must include a written
evaluation that provides the bases for the determination that the
changes do not require prior Commission approval under paragraph (c) or
(d) of this section. These records must be maintained until termination
of the license.
Sec. 70.73 Renewal of licenses.
Applications for renewal of a license must be filed in accordance
with Secs. 2.109, 70.21, 70.22, 70.33, 70.38, and 70.65. Information
contained in previous applications, statements, or reports filed with
the Commission under the license may be incorporated by reference,
provided that these references are clear and specific.
Sec. 70.74 Additional reporting requirements.
(a) Reports to NRC Operations Center. (1) Each licensee shall
report to the NRC Operations Center the events described in appendix A
to part 70.
(2) Reports must be made by a knowledgeable licensee representative
and by any method that will ensure compliance with the required time
period for reporting.
(3) The information provided must include a description of the
event and other related information as described in Sec. 70.50(c)(1).
(4) Follow-up information to the reports must be provided until all
information required to be reported in Sec. 70.50(c)(1) of this part is
complete.
(5) Each licensee shall provide reasonable assurance that reliable
communication with the NRC Operations Center is available during each
event.
(b) Written reports. Each licensee who makes a report required by
paragraph (a)(1) of this section shall submit a written follow-up
report within 30 days of the initial report. The written report must
contain the information as described in Sec. 70.50(c)(2).
18. Appendix A to part 70 is added to read as follows:
Appendix A to Part 70--Reportable Safety Events
As required by 10 CFR 70.74, licensees subject to the
requirements in subpart H of part 70, shall report:
(a) One hour reports. Events to be reported to the NRC
Operations Center within 1 hour of discovery, supplemented with the
information in 10 CFR 70.50(c)(1) as it becomes available, followed
by a written report within 30 days:
(1) An inadvertent nuclear criticality.
(2) An acute intake by an individual of 30 mg or greater of
uranium in a soluble form.
(3) An acute chemical exposure to an individual from licensed
material or hazardous chemicals produced from licensed material that
exceeds the quantitative standards established to satisfy the
requirements in Sec. 70.61(b)(4).
(4) An event or condition such that no items relied on for
safety, as documented in the Integrated Safety Analysis summary,
remain available and reliable, in an accident sequence evaluated in
the Integrated Safety Analysis, to perform their function:
(i) In the context of the performance requirements in
Sec. 70.61(b) and Sec. 70.61(c), or
[[Page 41357]]
(ii) Prevent a nuclear criticality accident (i.e., loss of all
controls in a particular sequence).
(5) Loss of controls such that only one item relied on for
safety, as documented in the Integrated Safety Analysis summary,
remains available and reliable to prevent a nuclear criticality
accident, and has been in this state for greater than eight hours.
(b) Twenty-four hour reports. Events to be reported to the NRC
Operations Center within 24 hours of discovery, supplemented with
the information in 10 CFR 70.50(c)(1) as it becomes available,
followed by a written report within 30 days:
(1) Any event or condition that results in the facility being in
a state that was not analyzed, was improperly analyzed, or is
different from that analyzed in the Integrated Safety Analysis, and
which results in failure to meet the performance requirements of
Sec. 70.61.
(2) Loss or degradation of items relied on for safety that
results in failure to meet the performance requirement of
Sec. 70.61.
(3) An acute chemical exposure to an individual from licensed
material or hazardous chemicals produced from licensed materials
that exceeds the quantitative standards that satisfy the
requirements of Sec. 70.61(c)(4).
(4) Any natural phenomenon or other external event, including
fires internal and external to the facility, that has affected or
may have affected the intended safety function or availability or
reliability of one or more items relied on for safety.
(5) An occurrence of an event or process deviation that was
considered in the Integrated Safety Analysis and:
(i) Was dismissed due to its likelihood; or
(ii) Was categorized as unlikely and whose associated
unmitigated consequences would have exceeded those in Sec. 70.61(b)
had the item(s) relied on for safety not performed their safety
function(s).
(c) Concurrent Reports. Any event or situation, related to the
health and safety of the public or onsite personnel, or protection
of the environment, for which a news release is planned or
notification to other government agencies has been or will be made,
shall be reported to the NRC Operations Center concurrent to the
news release or other notification.
For the Nuclear Regulatory Commission.
Dated at Rockville, Maryland, this 23rd day of July, 1999.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 99-19363 Filed 7-29-99; 8:45 am]
BILLING CODE 7590-01-P
