[Federal Register Volume 65, Number 181 (Monday, September 18, 2000)]
[Rules and Regulations]
[Pages 56211-56231]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-23354]



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  Federal Register / Vol. 65, No. 181 / Monday, September 18, 2000 / 
Rules and Regulations  

[[Page 56211]]



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: Final rule.

-----------------------------------------------------------------------

SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations governing the domestic licensing of special nuclear 
material (SNM) for licensees authorized to possess a critical mass of 
SNM, that are engaged in one of the following activities: enriched 
uranium processing; fabrication of uranium fuel or fuel assemblies; 
uranium enrichment (other than certified existing gaseous diffusion 
plants); enriched uranium hexafluoride conversion; plutonium 
processing; fabrication of mixed-oxide fuel or fuel assemblies; scrap 
recovery of SNM; or any other activity involving a critical mass of SNM 
that the Commission determines could significantly affect public health 
and safety or the environment. The amendments establish performance 
requirements, require affected licensees to perform an integrated 
safety analysis (ISA) to identify potential accidents at the facility 
and the items relied on for safety necessary to prevent these potential 
accidents and/or mitigate their consequences; require the 
implementation of measures to ensure that the items relied on for 
safety are available and reliable to perform their function when 
needed; require the safety bases to be maintained, and changes reported 
to NRC; allow for licensees to make certain changes to their safety 
program and facilities without prior NRC approval; require reporting of 
certain events; and require the NRC to perform a backfit analysis under 
specified circumstances.

EFFECTIVE DATE: The final rule, with the exception of Sec. 70.76, is 
effective October 18, 2000. Section 70.76 will become effective after 
the issuance of staff guidance for the implementation of that 
provision. Once such guidance has been developed, the NRC will publish 
a Federal Register notice specifying the effective date of Sec. 70.76.

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

SUPPLEMENTARY INFORMATION:
I. Background
II. Public Comments on Proposed Rule
III. Changes from the Proposed Rule
IV. Section-by-Section Analysis of Part 70 Amendments
V. Finding of No Significant Environmental Impact: Availability
VI. Paperwork Reduction Act Statement
VII. Public Protection Notification
VIII. Regulatory Analysis
IX. Regulatory Flexibility Certification
X. Voluntary Consensus Standards
XI. Backfit Statement
XII. Small Business Regulatory Enforcement Fairness Act
XIII. List of Subjects

I. Background

    On July 30, 1999 (64 FR 41338), the Commission published a proposed 
rule for public comment that would amend its regulations governing the 
domestic licensing of SNM for certain licensees authorized to possess a 
critical mass of SNM. The Commission's action was in response to a 
Petition for Rulemaking, (PRM)-70-7, submitted by the Nuclear Energy 
Institute (NEI), which was published on November 26, 1996 (61 FR 
60057). The proposed rule was intended to grant the NEI PRM in part and 
would modify the petitioner's proposal. The majority of the proposed 
modifications to Part 70 were included in a proposed new Subpart H, 
``Additional Requirements for Certain Licensees Authorized to Possess a 
Critical Mass of Special Nuclear Material.'' These modifications were 
proposed in order to increase confidence in the margin of safety at the 
facilities affected by the rule.
    In developing the proposed rule, the Commission sought to achieve 
its objectives through a risk-informed and performance-based regulatory 
approach that included: (1) The identification of performance 
requirements for prevention of accidents or mitigation of their 
consequences; (2) the performance of an ISA to identify potential 
accidents at the facility and the items relied on for safety; (3) the 
implementation of measures to ensure that the items relied on for 
safety are available and reliable to perform their function when 
needed; (4) the maintenance of the safety bases, including the 
reporting of changes to the NRC; and (5) the allowance for licensees to 
make certain changes to their safety program and facilities without 
prior NRC approval.
    The 75-day public comment period on the proposed rule ended on 
October 13, 1999. During and after the public comment period, the NRC 
staff posted on the NRC web site revised versions of the draft Standard 
Review Plan (SRP) that would implement the proposed requirements (i.e., 
on August 4, 1999, a complete draft SRP was posted, and revised 
chapters, taking into account comments received, were posted during the 
period March 16-April 3, 2000). In addition, three stakeholder meetings 
were held to discuss the SRP (September 14-15, 1999, February 9, 2000, 
and April 18-19, 2000).

II. Public Comments on Proposed Rule

    In preparing the final rule, the NRC staff carefully reviewed and 
considered more than 90 comments on the rule, included in 9 individual 
letters filed during the public comment period. In addition, 13 
submittals containing more than 200 specific comments were received on 
the SRP. To simplify the analysis, the NRC staff grouped all written 
comments on the rule into the following major topic areas: Performance 
Requirements and Design Criteria; Content of Applications and ISA 
Summary; Safety Program; Change Process, License Renewal and Backfit; 
Definitions; and Miscellaneous. A more detailed analysis is also 
documented as

[[Page 56212]]

an attachment to SECY-00-0111. A review of the comments and NRC staff's 
responses follow:

A. Performance Requirements and Design Criteria

    Comment A.1: One commenter stated that the proposed rule should 
specify dose limits for anticipated occurrences similar to those in 
Secs. 72.104 and 72.106. This part of the rule would then cover the 
range of likelihood (anticipated, likely, unlikely, and highly 
unlikely) of potential accidents that could occur at nuclear fuel cycle 
facilities. This could result in an increase in the number of 
structures, systems, and components (SSCs) relied on for safety and 
would impact the design, operation, and licensing of the mixed-oxide 
(MOX) facility.
    Response: No change in the rule language has been made. The dose 
limits for normal operation are contained in 10 CFR Part 20 [viz., 0.05 
Sv (5 rem) Total Effective Dose Equivalent (TEDE)/yr for a trained 
worker]. The NRC staff views ``anticipated occurrences'' to be 
conditions of normal operations, and believes that the measures 
currently used by Part 70 licensees to comply with Part 20 have been 
and will continue to be successful in protecting workers and the public 
during normal operations. Thus, there is insufficient justification to 
require identification of ``items'' to demonstrate compliance with Part 
20 during normal operations.
    Comment A.2: One commenter proposed that the NRC maintain 
consistency with past precedent (i.e., the Commission's rationale in 
Part 60) and eliminate the specific worker dose limits in Part 70.
    Response: No change in rule language has been made. The regulatory 
experience and industry events that initiated the effort to add a 
systematic accident analysis to Part 70 primarily involved health 
impacts to workers as opposed to the public. The NRC staff believes 
that the rule's focus on both the potential impacts on workers and the 
public is appropriate. Based on the discussions and correspondence with 
the industry and public during development of the proposed rule, and 
all other comments on the proposed rule, there appears to be general 
consensus on this approach.
    Comment A.3: One commenter stated, in response to the Federal 
Register notice request for comments on the clarity and effectiveness 
of the language used (per June 1, 1998, Presidential Memorandum), that 
the language in Sec. 70.61(b) and (c) could be substantially clearer; 
the commenter provided an alternative plain language version of this 
section.
    Response: The language in the proposed rule was written in response 
to public comments to focus on risk (i.e., likelihood times 
consequence) of accidents. The language has been changed, in response 
to the comment, to provide additional clarity. The proposed revisions 
provided by the commenter, however, are not merely editorial but 
represent substantive changes. They appear to have eliminated the 
concept of limiting risk, and instead, focused on the likelihood of 
accidents. The revised language in the rule attempts to retain the 
emphasis on controlling accident risks within appropriate performance 
requirements.
    Comment A.4: Three commenters expressed concerns about how the 
worker dose limits in Sec. 70.61(f) would be applied to ``co-located 
workers.'' One commenter suggested that the performance requirements in 
Sec. 70.61 consider the individuals working in the nearby facilities as 
public when performing an accident analysis to determine the 
consequences of the accidents that may occur at the facility. The 
commenter concluded that this would result in a more stringent 
application of safety requirements for the protection of workers (e.g., 
additional items relied on for safety) at the MOX Fuel Fabrication 
Facility, Pit Disassembly, Conversion Facility, Immobilization 
Facility, and any other nearby DOE facilities, and would also have a 
substantial impact on the cost of the MOX facility. A second commenter 
agreed with this assessment, noting that a worker (as defined in 
Sec. 70.4) who leaves the controlled area to perform a work-related 
function would have to be treated as a member of the public when 
performing an ISA and would be subject to the more stringent public 
radiation exposure limits. Outside of the controlled area the TEDE 
limit of 1 mSv (0.1 rem) for members of the public would apply [cf. 10 
CFR 20.1301(a)(1)] rather than the annual TEDE occupational dose limit 
of 50 mSv (5 rems) (10 CFR 20.1201). According to this commenter, this 
problem has already arisen at the Hanford Tank Waste Remediation System 
where co-located workers are governed by the appreciably lower public 
dose limits. A third commenter agreed with the above positions and also 
stated that the NRC intends to consider individuals outside of the 
controlled boundary as workers if they are subject to Part 20 
requirements. The commenter noted, as did the first commenter, that DOE 
requirements in 10 CFR Part 835 provide an equivalent level of 
protection, such that co-located workers--who are subject to the 
requirements of either Part 20 or 10 CFR Part 835--should be considered 
``workers,'' provided the licensee can demonstrate the ability to 
provide management measures (e.g., notification, evacuation, etc.) in 
the event of an emergency.
    Response: NRC regulations do not specifically address personnel 
designated as ``co-located'' workers. In response to the comments, the 
first sentence in Sec. 70.61(f) was changed to read as follows: ``Each 
licensee must establish a controlled area, as defined in Sec. 20.1003. 
In addition, the licensee must retain the authority to exclude or 
remove personnel and property from the area.'' The licensee can set the 
controlled area at any location around its facility as long as it 
maintains control of that area as specified in Part 20 and retains the 
authority to exclude or remove personnel and property from the area. If 
the controlled area included the nearby Department of Energy (DOE) 
facilities, then NRC would consider the personnel working at those 
facilities to be ``workers'' for the purposes of the performance 
requirements of Sec. 70.61, provided the conditions of Sec. 70.61(f)(2) 
are met. The DOE and its contractors could satisfy these conditions by 
documenting their compliance with the requirements of 10 CFR 
19.12(a)(1)-(5). To emphasize that the Sec. 70.61(f)(2) requirements, 
regarding 10 CFR Part 19 training, can be satisfied in combination with 
existing training, rather than separate training solely devoted to 10 
CFR Part 19, 10 CFR 70(f)(2) has been changed to read: ``Provides 
training that satisfies 10 CFR 19.12(a)(1)-(5)''. To emphasize that the 
training provided to satisfy Sec. 70.61(f)(2) requirements includes 
making individuals aware of the risks associated with accidents 
involving the licensed activities as determined by the ISA, the word 
``to'' was changed to ``and,'' so that it now reads ``to these 
individuals and ensures that they are aware of the risks associated 
with accidents''.
    Regarding the concern about the worker who leaves the controlled 
area, the risk levels of Sec. 70.61 for the public pertain to any 
individual, including workers, outside the controlled area. On the 
other hand, with respect to the applicability of the Part 20 
occupational dose limit of 0.05 Sv (5 rem)/yr TEDE, a worker can 
receive an occupational dose and be subject to the Part 20 occupational 
limit, regardless of his location--including activities outside the 
controlled area. The ``assigned duties performed in the course of 
employment'' is the distinguishing factor for radiation workers 
consistent

[[Page 56213]]

with the definition of ``occupational exposure'' in 10 CFR 20. The 
changes to Part 70, including the ``worker'' definition, do not affect 
this. In this comment, the relationship between Part 20 annual limits 
for radiation exposure and the Sec. 70.61 standards for a forward-
looking severe accident assessment have been misinterpreted. Part 70 
revisions do not limit doses outside a controlled area to 1 mSv (0.1) 
rem/yr.
    Comment A.5: One commenter recommended that baseline criterion (8) 
be rewritten as follows: ``the design of items relied on for safety 
must provide for adequate inspection, testing, and maintenance, or 
adequate training, testing and qualification for personnel whose 
activities are relied on for safety, to ensure their availability and 
reliability to perform their function when needed.''
    Response: No change in rule language has been made. The baseline 
design criteria are applied from the outset of new design work and are 
primarily focused on physical design and facility features. The intent 
is to achieve a conservatively designed facility tolerant of both 
upsets and human errors. Adequate training, testing, and qualification, 
as noted in the comment, will be required as management measures under 
Sec. 70.62, but the NRC does not see a need for the facility physical 
design to incorporate such training, testing, and qualification of 
personnel.
    Comment A.6: One commenter stated that the baseline criterion on 
environmental and dynamic effects [Sec. 70.64a(4)] is unclear. For 
example, the commenter questioned the need for a formal Equipment 
Environmental Qualification Program, similar to that required under 10 
CFR 50.49 and Regulatory Guide 1.89. According to the commenter, the 
NRC should clarify this requirement and should not impose requirements 
that may not be appropriate or necessary because of the nature of the 
processes at non-reactor nuclear facilities.
    Response: No change in rule language has been made. The baseline 
design criterion on environmental and dynamic effects does not require 
a formal Equipment Environmental Qualification Program, similar to that 
required under 10 CFR 50.49 and Regulatory Guide 1.89. This criterion 
applies only to new facilities and new processes and is intended to 
ensure that potential ambient conditions are considered during the 
design of the facility.
    Comment A.7: Two commenters had concerns regarding the defense-in-
depth definition in Sec. 70.64. One commenter stated that the 
definition does not reflect the defense-in-depth design philosophy as 
defined in WASH-1250, ``The Safety of Power Reactor and Related 
facilities,'' which outlined three levels of safety concepts in the 
design of a nuclear facility. According to the commenter, the 
definition presented in Secs. 70.64(b)(1) and (2) oversimplifies and 
does not adequately represent the implementation of the defense-in-
depth philosophy in the design. In particular, the commenter noted that 
the preference for engineered controls over administrative controls and 
features that reduce challenges to items relied on for safety are only 
partially implemented in the concept [of defense-in-depth]. Another 
commenter agreed, stating that Sec. 70.64(b)(1) appeared unnecessarily 
prescriptive by discouraging a licensee from using anything but an 
engineered safety control. According to this commenter, as long as the 
licensee can satisfactorily demonstrate that an administrative safety 
control or a system of administrative and engineered controls will 
enable the performance criteria to be satisfied, the choice of items 
relied on for safety and the nature of ``defense-in-depth'' practices 
that is applied should be flexible. The commenter's view is that this 
flexibility, in the grading of defense-in-depth safety concepts, would 
be consistent with the ability granted a licensee to grade all aspects 
of its safety program [cf. Sec. 70.62(a)].
    Response: With respect to the footnote to Sec. 70.64(b) that 
describes defense-in-depth, which applies to new facilities and new 
processes at existing facilities, the NRC staff believes that it does 
reflect the defense-in-depth design philosophy as defined in WASH-1250. 
Further, it reflects the Commission's current guidance on the 
relationship between defense-in-depth and risk-informed regulation that 
is discussed in the Commission policy white paper, ``Risk-Informed and 
Performance-Based Regulation.'' With respect to Secs. 70.64(b)(1) and 
(2), the NRC staff did not mean to imply that these provisions 
encompassed the defense-in-depth philosophy.
    Comment A.8: One commenter recommended that the emergency 
capability baseline design criterion in Sec. 70.64(a)(6)(ii) address 
on-site personnel (rather than all personnel). The commenter suggested 
that the rule language be rewritten as ``Evacuation of on-site 
personnel; and * * *.''
    Response: The NRC agrees with the comment. The proposed change has 
been made and is consistent with the intent of the original rule 
language.
    Comment A.9: One commenter stated that the criticality performance 
objective in Sec. 70.61(d) is not related to Secs. 70.61(b) or 
70.61(c); yet, the three conditions are all linked together. The 
commenter suggested that subpart (d) should be segregated from (b) and 
(c) if (d) is preserved as an independent entry (as would seem 
preferable). Otherwise (d) should be subsumed under (b) and/or (c), and 
the regulatory basis for criticality prevention should be predicated on 
the risks and/or consequences of the accidents, rather than the 
presence of initiator precursor, per se. 
    Response: No change in the rule language has been made. The NRC 
believes that a separate performance requirement for nuclear 
criticality prevention is appropriate. The NRC staff recognizes that 
many (but not all) nuclear criticality accidents would reasonably be 
expected to result in worker doses that exceed the high- and 
intermediate-consequence standards in Sec. 70.61(b) or (c). However, 
regardless of the dose directly resulting from the accident, an 
inadvertent nuclear criticality should be avoided. This is consistent 
with the Commission's goal to prevent inadvertent criticalities, as 
reflected in the NRC Strategic Plan (NUREG-1614).

B. Content of Applications and ISA Summary

    Comment B.1: One commenter stated that the rule should not 
prescribe an acceptable level of detail required in the application, 
but should defer this issue to the SRP. The commenter noted that, 
although progress has been made in certain areas (e.g., use of language 
such as ``* * * types of accident sequences * * *.''), in 
Sec. 70.65(b)(6), which requires the applicant to list all items relied 
on for safety for high- and intermediate-consequence accidents, the 
required level of descriptive detail for items relied on for safety 
(''sufficient detail'') remains vague. The commenter recommends that 
information at the ``systems level'' should be required, rather than at 
the ``component'' or ``sub-component'' level.
    Response: The NRC disagrees with the comment. The current language 
permits the description of information at a systems level provided that 
there is enough detail to understand the function of the system in 
relation to the performance requirements. The degree of detail provided 
in the ISA Summary, with the other information available to NRC staff, 
must be sufficient for the NRC staff to make the determination 
specified in Sec. 70.66 (i.e., that the performance requirements of the 
regulation are satisfied).

[[Page 56214]]

    Comment B.2: One commenter stated that the list of items relied on 
for safety should not include procedures that the personnel must 
follow. According to the commenter, since procedures are constantly 
being adjusted, revised, and improved, their inclusion in the list of 
items relied on for safety would necessitate frequent revisions, to the 
ISA Summary, that may have little if any safety significance.
    Response: Sec. 70.65(b)(6) requires a list, in the ISA Summary, 
briefly describing each item relied on for safety. It does not require 
procedures to be listed in the ISA Summary. Therefore, the rule 
language permits the approach described in the comment. Typically, the 
actual personnel action would be regarded as an item relied on for 
safety and this would be expected to be addressed in the ISA Summary.
    Comment B.3: Two commenters had concerns about the relationship 
among the ISA, the ISA Summary, and the safety program. One commenter 
recommended that the NRC clarify the relationship of the ISA Summary to 
the license and the safety basis to ensure consistency throughout the 
rule with the intent expressed in Sec. 70.65(b). The commenter was 
concerned that the wording of Sec. 70.65(a) is inconsistent with the 
idea, presented in Sec. 70.65(b), that the ISA Summary will not be 
incorporated in the license. The commenter suggested removing the 
language in Sec. 70.65(a) that references the inclusion of the ISA 
Summary in the license application, since that requirement is 
adequately covered in Sec. 70.65(b). The commenter also recommended 
that a discussion of management measures be included as part of the ISA 
Summary. The second commenter stated that the rule implies that the ISA 
Summary, as part of the safety program, is part of the license. 
Further, the same commenter stated that the ``Statement of 
Considerations'' erroneously states that the results of the ISA must be 
submitted for NRC approval.
    Response: The NRC generally agrees with the comment. The rule 
language in Sec. 70.65(a) has been changed to remove the reference to 
the ISA Summary and management measures. This removes the implication 
that the ISA Summary is part of the license. With respect to the 
relationship of the ISA Summary to the management measures, although 
under the proposed rule, the elements of the ISA Summary did not 
explicitly include management measures, one of the elements 
[70.65(b)(4)] of the ISA Summary required information that demonstrates 
compliance with the performance requirements. Such a demonstration 
requires information about management measures. As suggested in the 
comment, the language in Sec. 70.65(b)(4) has been clarified to 
explicitly include a description of the management measures. With 
regard to the comment that the ``Statement of Considerations'' 
erroneously ``states that the results of the ISA must be submitted for 
approval'', the assertion that the ``Statement of Considerations'' is 
erroneous is incorrect--the ``Statement of Considerations'' is 
accurate. In response to this comment, and to clarify the role of the 
ISA Summary in licensing determinations, changes have been made to 
Sec. 70.62(c)(3)(ii) and Sec. 70.66. In particular, 
Sec. 70.62(c)(3)(ii) has been modified to specifically state that the 
ISA Summary is submitted for approval consistent with the ``Statement 
of Considerations'' for the proposed rule. Section 70.66 states that 
this submission will be approved if the Commission determines that 
``the applicant has complied with the requirements of Sec. 70.21, 
Sec. 70.22, Sec. 70.23, and Sec. 70.60 through Sec. 70.65.'' The degree 
of detail provided in the ISA Summary [contents of the ISA Summary are 
described in Sec. 70.65(b)] and the other information available, must 
be sufficient for the NRC staff to make the determination specified in 
Sec. 70.66. To supplement staff understanding of information submitted, 
NRC may visit the facility during the licensing review to ensure a 
sufficient safety basis for operation.
    Comment B.4: Two commenters were concerned with the broad nature of 
the requirement in Sec. 70.65(b)(3) that seeks information on each 
process analyzed in the ISA, regardless of the risk associated with the 
process. According to one commenter, the ISA Summary should only 
address those processes for which accident sequences have been 
identified that would produce consequences that exceed the performance 
criteria of Sec. 70.61.
    Response: The NRC staff needs some information on each process 
analyzed in the ISA to assess completeness and quality of the 
licensee's ISA process and to understand and assess the completeness 
and functions of the items relied on for safety. The degree of detail 
provided in the ISA Summary, together with the other information 
available, must be sufficient for the NRC staff to make the 
determination specified in Sec. 70.66. In addition the information is 
useful in confirming the adequacy of emergency planning.
    Comment B.5: According to one commenter, Sec. 70.65(b) implies that 
the ISA Summary is a single document. In practice, the commenter noted 
that it will be a sequence of documents that cover the facility, and if 
multiple documents are submitted, they should all be in the same 
format.
    Response: The NRC agrees that multiple ISA Summaries are allowed 
for each facility resulting in the possibility that the ISA Summary 
will be a sequence of documents rather than consisting of a single 
document. The definition of ISA Summary in Sec. 70.4 has been changed 
to reflect this process. NRC staff approval of individual documents 
(e.g., on a process basis) will be conditioned to allow for subsequent 
identification of system interaction effects that may be identified 
through the review of other ISA summary documents submitted.
    Comment B.6: One commenter stated that the requirement, in 
Sec. 70.65(b)(7), to provide information on the locations of onsite 
chemicals, is unnecessary.
    Response: No change in the rule language has been made. Section 
70.65(b)(7) does not require information on the locations of onsite 
chemicals to be submitted to the NRC. The regulation requires a 
description of the proposed quantitative standards used to assess the 
consequences to an individual from acute chemical exposure to licensed 
material or chemicals produced from licensed material. This information 
is necessary to ensure safety and is consistent with NRC's Memorandum 
of Understanding (MOU) with the Occupational Safety and Health 
Administration (OSHA).
    Comment B.7: One commenter objected to the requirement to provide 
process descriptions, noting that American Institute of Chemical 
Engineers (AIChE) guidelines may result in the process being broken 
into ``nodes'' or ``segments.'' The commenter suggested that the rule 
should specify the descriptions of segments or nodes that could only be 
combined into a process if the boundaries established for the hazard 
analysis match.
    Response: The NRC agrees with the comment, but does not believe a 
change in the rule language is needed. The intent of the 
Sec. 70.65(b)(3) requirement is to provide process information so that 
the NRC staff can understand: What activities are performed at the site 
that involve hazardous materials associated with or produced from 
licensed radiological material, including any use, storage, 
manufacturing, or handling of those materials; what was analyzed in the 
ISA; and the hazards identified in the ISA. The AIChE guidelines use 
the term ``process nodes'' with respect to Hazard and Operability 
Analysis (HAZOP) and define it as ``sections of equipment with definite 
boundaries

[[Page 56215]]

* * * within which process parameters are investigated for deviations * 
* *.'' In HAZOP analyses, the term ``node'' designates a pipeline or 
vessel that has a common design intent. In meeting the Sec. 70.65(b)(3) 
requirement, several nodes may be combined.

C. Safety Program

    Comment C.1: Three commenters questioned the narrow definition of 
the safety program that is presented in Sec. 70.62(a) and recommended 
deleting it from the rule language. According to the commenters, the 
safety program is broader than the three elements identified in 
Sec. 70.62(a)(1) as process safety information, ISA, and management 
measures. The commenters noted that fuel cycle facility safety programs 
encompass the three elements identified plus all the other topics 
addressed in the license application. This includes, for example, 
radiation safety, criticality safety, chemical safety, and fire 
protection, in addition to the three elements directly associated with 
the ISA.
    Response: The NRC staff agrees in principle with the comment. The 
term ``safety program,'' as used in Sec. 70.62 (a), is related to the 
elements needed to demonstrate compliance with the performance 
requirements in Sec. 70.61. This safety program consists of process 
safety information, ISA, and management measures. There is no intent to 
indicate that these elements represent the total safety program at the 
facility. Therefore, the rule language was clarified by changing ``The 
three elements of the safety program; namely process safety 
information, integrated safety analysis, and management measures, are 
described in paragraph (b) through (d) of this section * * *'' to 
``Three elements of this safety program; namely process safety 
information, integrated safety analysis, and management measures, are 
described in paragraph (b) through (d) of this section.''
    Comment C.2: One commenter stated that the current proposed rule 
offers sufficient flexibility in selecting ISA methodology so that a 
broad spectrum of facilities can be addressed and such that licensees 
have flexibility to interface with their site processes, procedures and 
resources.
    Response: The Commission agrees with the comment; therefore, no 
change was made to the rule language with respect to ISA methodologies. 
The final rule offers sufficient flexibility in selecting an ISA 
methodology that can be used to analyze a facility's site, processes 
and procedures.
    Comment C.3: Two commenters were concerned about the implementation 
of the final rule, and, in particular, the time frame for compliance 
with those aspects of the rule not related to the completion of the ISA 
and the submittal of the ISA Summary. One commenter, citing the 
experience when Part 20 was revised, recommended an effective date 
sufficiently far into the future so programmatic changes could be 
implemented at the operating facilities and any necessary conforming 
license amendments could be completed. Regarding the latter issue, both 
commenters cited 10 CFR 20.1008 as an example of how potential 
contradictions between license applications and regulations could be 
addressed. One commenter recommended including an additional provision 
of this type, especially in light of license conditions that have been 
added to licenses recently renewed by the NRC.
    Response: The NRC agrees with the comment. In Sec. 70.76(a), it 
states that ``this provision shall apply for subpart H requirements as 
soon as the NRC approves that licensee's ISA Summary pursuant to 
Sec. 70.66. For requirements other than Subpart H, this provision 
applies regardless of the status of the approval of a licensee's ISA 
Summary.'' In addition, Appendix A was revised to include the 
following: ``Licensees must comply with reporting requirements in this 
appendix, except for (a)(1), (a)(2), and (b)(4), after they have 
submitted an ISA Summary in accordance with Sec. 70.62(c)(3)(ii). 
Licensees must comply with (a)(1), (a)(2), and (b)(4) after October 18, 
2000.'' In addition, Sec. 70.62(c)(3)(ii) was revised to further 
clarify implementation schedules for existing licensees.
    Comment C.4: Two commenters stated that a graded approach should be 
used in determining the management measures that need to be applied to 
items relied on for safety. One commenter recommended that the language 
in Sec. 70.62(d) should be changed as follows: ``The measures applied 
to a particular engineered or administrative control or control system 
may be graded commensurate with the reduction of the risk attributable 
to that control or control system.'' The other commenter recommended 
that other factors besides risk including consequences, life cycle, and 
magnitude of hazard involved, should be used to determine appropriate 
management measures.
    Response: The NRC agrees with the comment and has made the 
suggested change to the rule language in Sec. 70.62(d). Regarding the 
question of considering other factors besides ``risk,'' the NRC notes 
that the grading of measures to consequences, life cycle, and magnitude 
of hazard, is part of grading the measures to risk. The phrase used in 
the rule--``commensurate with the reduction of risk attributable to 
that item''--does not imply requiring a quantitative determination of 
the risk significance of any particular item relied on for safety. The 
rule is non-prescriptive regarding the grading approach and criteria to 
be used, allowing applicants to propose such details.
    Comment C.5: One commenter stated that the 4-year period for 
conducting the ISA and for modifying the facility to address any 
identified unacceptable performance deficiencies may be too short and 
recommended a 5-year period instead. According to the commenter, a 5-
year time-frame would be consistent with the time allowed for existing 
licensees that have committed, by license condition, to perform ISAs. 
The commenter also recommended that the period should start on the date 
when the NRC approves the plan required in Sec. 70.62(3)(i), noting 
that if the clock starts on the effective date of the rule and the NRC 
takes one year to approve the ISA plan, the licensee will be unduly 
hampered. In addition, the commenter stated that there should be some 
incentive for the NRC to complete its approval process in a timely 
manner and recommended imposition of a 90-day limit for NRC to issue a 
decision on the acceptability of a licensee's ISA approach. The 
commenter also recommended that appropriate and sufficient time be 
allowed for the licensee to present a plan to the NRC and to implement 
the plan to correct any identified unacceptable performance 
deficiencies.
    Response: Regarding the proposal for a 5-year period for conducting 
the ISA and correcting all unacceptable deficiencies, the NRC believes 
that the 4-year period proposed in the proposed rule is reasonable. 
However, NRC recognizes that there may be some instances where 
modifications resulting from the ISA cannot be completed within the 4 
years specified and has modified Sec. 70.62(c)(3)(ii) to accommodate 
these instances by clarifying that NRC may approve extensions for 
reasons that are beyond the control of the licensee. Regarding the 
licensee being unduly hampered because of the time required for the NRC 
staff to approve the plan required by Sec. 70.62(c)(3)(i), the NRC 
staff expects to complete the licensing review within 90 days, assuming 
that the information submitted is complete. However, the time it takes 
the NRC to approve the

[[Page 56216]]

plan will depend on the quality of the plan submitted by the licensee. 
In addition, current industry development of an ISA Summary guidance 
document should facilitate the licensing review process.
    Comment C.6: One commenter stated that the plan required in 
Sec. 70.62(c) which should be submitted within 6 months of the 
effective date of the rule, should pertain only if a licensee has not 
already completed the actions outlined in Sec. 70.62(c)(3)(ii).
    Response: The implementation plan and the ISA must satisfy the 
requirements in the final rule. If the actions outlined in 
Sec. 70.62(c)(3)(ii) have been completed, then all that would be 
required to satisfy Sec. 70.62(c)(3)(i) is submission of a description 
of any additional work that must be performed to meet the requirements 
in Subpart H of the rule, or a confirmation that the work submitted 
meets the requirements in Subpart H of the rule.
    Comment C.7: Four commenters disagreed with the requirement in 
Sec. 70.62(a)(3) to establish and maintain a log of failures of items 
relied on for safety. One commenter stated that the requirement should 
be rewritten to be performance-based rather than prescriptive. The 
commenter noted that most licensees have an incident reporting and 
corrective action system, which is used for all activities at the 
facility. As long as these systems meet the performance objective, it 
seems unnecessary for the rule language to be prescriptive in how it is 
met. Another commenter agreed, stating that it is inappropriate to 
impose this extra record-keeping burden on the licensee, because the 
licensee already has to generate records of this nature to manage its 
business and another different log is unnecessary work. Another 
commenter noted that because of the reporting requirements of 
Sec. 70.62(a)(2) and Sec. 70.74(a)(1), the NRC will already possess all 
of the information sought in the ``log'' of Sec. 70.62(a)(3).
    Response: The NRC generally agrees with the comment that 
maintenance of the failure log would be unnecessarily prescriptive. 
Regarding the concern about prescriptiveness, the rule has been revised 
to eliminate the requirement for licensees to establish and maintain a 
specific log of information developed and maintained elsewhere. 
However, the final rule requires that data be readily retrievable and 
available. This information is necessary to evaluate the reliability 
and availability of items relied on for safety, the likelihood of 
failure of the items, and the effectiveness of management measures 
implemented by the licensee. The NRC also anticipates this information 
will be reviewed during periodic inspections by NRC as part of the 
revised oversight process that is being developed. Regarding the 
redundancy of reporting, the rule currently requires the licensee to 
report only any loss or degradation of items relied on for safety that 
results in failure to meet the performance requirements of Sec. 70.61. 
The requirements of Sec. 70.62(a)(3) include a much broader set of 
items, including all items relied on for safety or management measures 
that have failed to perform their function.

D. Change Process, License Renewal, and Backfit

    Comment D.1: Five commenters were concerned about the requirement 
in Sec. 70.72(d)(1) to submit changes to the ISA Summary every 90 days. 
Two commenters stated that an annual update [similar to the annual 
Final Safety Analysis Report updates for reactors per 10 CFR 50.71(e)] 
should suffice, considering that the potential consequences of reactor 
accidents are significantly greater than those at fuel cycle 
facilities. One commenter stated that an annual update to the ISA 
Summary would be consistent with the reporting requirements (for 
changes to records) of Sec. 70.72(d)(3). Another commenter stated that 
the 90-day reporting of changes is entirely too frequent, which would 
mean that the facility and the NRC would always have change reporting 
in progress. According to the commenter, there is no need for NRC to 
have this ``real-time'' knowledge; rather, it is only important that 
the licensee have ``real-time'' knowledge. The commenter noted that the 
NRC only needs reasonably current knowledge, because the current ISA is 
available and accessible at the site. The commenter believes that a 12-
month to 24-month update for reporting, as used in other places, is 
satisfactory and more efficient, noting that this seems clearly 
justified based on the fact that all the information is available at 
the site and accessible to the NRC at any time.
    Response: The NRC agrees with the comment that submitting updates 
to the ISA Summary to the NRC can be less frequent than required in the 
proposed rule. The final rule requires only annual reporting within 30 
days after the end of the calender year during which the changes 
occurred.
    Comment D.2: One commenter noted that, under Sec. 70.72, the NRC 
should define ``periodically'' in the context of reporting of changes 
made to SSCs etc.
    Response: The NRC determined that no change to the reporting 
requirements is necessary in response to the comment. The comment 
referenced language in the ``Statement of Considerations'' not the 
rule. The specific reporting requirements were defined in the proposed 
rule and are included, as revised, in the final rule.
    Comment D.3: Two commenters were concerned about the footnote in 
Sec. 70.72(c) that attempts to explain new types of accident sequences. 
Both commenters stated that the language in the footnote would require 
nearly all process changes to be approved by NRC through a license 
amendment, which would be in conflict with the overall objectives for 
the proposed rule. Both commenters recommended that the footnote be 
deleted.
    Response: The NRC agrees that the footnote did not successfully 
clarify the definition of ``new types of accident sequences.'' Thus, 
the footnote has been deleted from the final rule. The NRC staff will 
develop a guidance document, with input from stakeholders, to describe 
an acceptable change process that meets the requirements of the final 
rule in more detail. The degree of detail provided in the ISA Summary, 
together with the other information available, must be sufficient for 
the NRC staff to make the determination specified in Sec. 70.66. In 
addition, the NRC staff had added a discussion to Chapter 3 of the SRP 
to describe an acceptable level of detail in the identification of the 
types of accident sequences.
    Comment D.4: Three commenters were concerned about the 
requirements, in Sec. 70.72, regarding configuration management and the 
overly broad process for making changes at licensed facilities. One 
commenter stated that the requirements, as written, apply to all site, 
structures, processes, systems, equipment, components, computer 
programs, and activities of personnel, regardless of safety 
significance. The commenter noted that compliance with these 
requirements would appear to require configuration management and 
change control applied to everything on the site of the licensed 
facility; that could include the wastewater treatment facility, a laser 
facility, the administration building, maintenance of the shrubbery, 
etc. Every change would require an evaluation and a summary submitted 
to the NRC, even though inclusion in the change control process would 
make no contribution to the safety of licensed operations and would 
impose an undue burden on the licensee. To remedy this, the commenter 
recommended that the configuration and change process be limited to any 
``changes to the site, processes or items relied on for safety as 
described in the

[[Page 56217]]

ISA Summary.'' Another commenter agreed, stating that the requirement 
is too broad and all-encompassing and would require configuration 
management evaluation of changes having no or absolutely minimal effect 
on health and safety (e.g., office remodeling, planting of shrubbery, 
changing paint colors). The commenter suggested that rather than 
control every change by means of configuration management, the licensee 
should first rely on internal procedures to screen any proposed changes 
initially for their potential safety significance.
    Response: No change in the rule language has been made. The 
emphasis of this requirement is clearly on licensed operations and the 
associated safety controls. If a licensee has established a 
configuration management system in accordance with Sec. 70.72(a), it is 
important the licensee use the system to evaluate every change made at 
a facility that could affect safety (i.e., generally not shrubbery, 
paint color) to ensure that any impacts from those changes on the 
safety of operations is identified, considered, documented, before 
implementing the change. In some cases, the analysis would be trivial 
because no known hazards would be involved in the change (e.g., certain 
changes in the administration building, or changes to shrubbery). Often 
it is clear that there are no safety implications associated with the 
proposed change. However, there may be special cases in which 
apparently minor changes could adversely affect safety, such as 
installation of a drinking water fountain in a radiological control 
area. In addition, every change which is assessed in the configuration 
management system does not need to be submitted to the NRC. Section 
70.72(d)(2) states that only those changes to records required by 
Sec. 70.62(a)(2) need to be submitted. These would include changes to 
the process safety information, ISA, and management measures. In 
addition, with respect to the use of an ``initial screening'' 
mechanism, the NRC staff considers an initial screening to assess the 
safety impact of a change to be part of an evaluation, as called for in 
Sec. 70.72(a). In some cases, this screening will be sufficient.
    Comment D.5: One commenter stated that Secs. 70.72(c)(1)(i), 
(c)(2), and (c)(3) are wrong to use the ISA Summary as the decision-
making document. The commenter noted that the ISA, the detailed 
licensee-generated information and evaluations that the licensee uses 
to manage its program, comprises the information base for decisions. 
Summaries only provide a general level of information about the more 
important elements of the safety system for operations as determined 
under the licensed program.
    Response: No change in the rule language has been made. The ISA 
Summary is prepared based on the ISA, and contains key information that 
is directly related to facility safety, such as a list of items relied 
on for safety, a description of hazards identified in the ISA, and a 
general description of the types of accident sequences. The contents of 
the ISA Summary are described in Sec. 70.65(b). The NRC staff could 
review the adequacy of changes using the ISA instead of the ISA 
Summary, but this approach would require submission of a greater amount 
of information to NRC and would pose an unnecessary burden on the 
licensee. (Also, see response to Comment B.3.)
    Comment D.6: Two commenters are concerned about the annual 
requirement in Sec. 70.72(d)(3) to submit a brief summary of all 
changes to the records required by Sec. 70.62(a)(2). According to one 
commenter, the submittal would cover process safety information 
[Sec. 70.62(b)] including procedures, drawings, and detailed equipment 
lists. The commenter does not believe the NRC requires a summary of 
changes to this type information. A second commenter agreed, stating 
that the wording of this section will inadvertently and significantly 
expand the information that would have to be reported. In particular, 
the view was expressed that Sec. 70.72(d) would require the licensees 
to submit voluminous information that could include the update to 
process safety information, including drawings, flow process diagrams, 
and piping and instrumentation diagrams. The commenter suggested that 
this section should be reworded to read: ``a brief summary of all 
changes to the integrated safety analysis and ISA summary, that are 
made without prior Commission approval, must be submitted to the NRC 
every 12 months * * *.''
    Response: No change in the rule language has been made. The 
regulation currently requires submission of ``* * * a brief summary of 
all the changes to the records required by Sec. 70.62(a)(2) * * *'' 
This does not require the submittal of actual charts and drawings but a 
written summary of the changes made. For the reasons cited in the 
response to comment D.1, it is important that the NRC be knowledgeable 
of changes made to this information.
    Comment D.7: One commenter noted that, unlike Sec. 50.59, the 
requirements of Sec. 70.72 do not call for the submittal of a brief 
description and summary safety evaluation for each change. The 
commenter believes that the NRC would benefit from a description of 
changes made to the ISA Summary. Accordingly, Sec. 70.72 should require 
brief descriptions and summary safety evaluations of each change made 
pursuant to Sec. 70.72 and require that an updated ISA Summary be 
provided on a biennial basis.
    Response: No change in the rule language has been made. The brief 
summaries of changes submitted under the requirements of 
Sec. 70.72(d)(2) would be expected to include an explanation of each 
change, the reasons why the change was made, and why it did not require 
pre-approval. This information will be included in a guidance document 
to be developed. The NRC staff views this as sufficient and does not 
anticipate the need for licensees to submit a summary safety evaluation 
for each change, as long as each change has been made in accordance 
with the final rule and the approved process.
    Comment D.8: Two commenters questioned the current timeframe (10 
years) or the need for renewal of licenses, suggesting that the new 
rule, in effect, resulted in a ``living license.'' One commenter stated 
that if a ``living license'' is truly the outcome as described in the 
Supplementary Information, renewal periods as long as 20 years would be 
appropriate. The other commenter noted that with updates required every 
12 months there is no real need for the NRC to renew the license--it 
only becomes a maintenance chore to confirm periodically that the 
licensing basis remains intact. The commenter believes that the living 
license concept provides advantages for the NRC and the licensee.
    Response: Although the NRC generally agrees with those comments, no 
change in the rule language has been made. A specific time period for 
renewals is not specified in Part 70 and to establish one in the rule 
would require consideration of many factors, such as compliance with 
the National Environmental Policy Act and the impact of the loss of 
commitments linked to license renewal, that were not addressed in the 
current rulemaking (e.g., financial assurance for decommissioning). 
Establishment of a new term for licenses (e.g., 20 years) in 10 CFR 70 
would require an analysis of these factors and an opportunity for 
public comment. The NRC staff will evaluate whether a longer term for 
fuel cycle licenses is appropriate in light of the new requirements in 
Subpart H. In any case, even if NRC ultimately declines to extend the 
term of the fuel cycle licenses (nominally 10 years), the burden of 
license renewal should be

[[Page 56218]]

significantly reduced because the licensee will be required to maintain 
current the ISA Summary, items relied on for safety, and management 
measures.
    Comment D.9: Five commenters recommended that a backfit provision 
similar to that in 10 CFR 50.109 or 10 CFR 76.76 should be included in 
the final rule. One commenter stated that the backfit provision should 
apply to current proposed changes at existing facilities. Another 
commenter stated that the backfit provision should be immediately 
effective for those processes or parts of an existing facility for 
which an ISA has been completed. A third commenter favored an 
immediately effective backfit provision. However, as an alternative, 
the commenter would make the provision effective for facilities or 
systems for which the ISA has been completed and the ISA Summary 
submitted to the NRC. A fourth commenter stated that deferring 
consideration of a backfit provision would be evading an extremely 
important issue, expressing the view that it is vital that a formal, 
systematic, and disciplined review of new, changed, or differing 
positions that could backfit existing facilities be applied to increase 
regulatory certainty. According to the commenter, no change to the 
backfit language in 10 CFR 50.109, which has been used successfully to 
control backfits at power reactors in the past, is needed to allow for 
qualitative analysis. 10 CFR 50.109, which the commenter endorses, is 
viewed as neither a quantitative nor a qualitative backfit provision. 
In contrast to the statement made in the Statement of Considerations of 
the proposed rule, the commenter does not believe that a comprehensive 
risk baseline is necessary before reasoned judgments can be made on the 
benefits and risks of a proposed backfit.
    Response: The Commission agrees that regulatory stability and 
certainty can be improved by establishing a backfit provision for fuel 
cycle facilities covered by Subpart H of the final rule. Consequently, 
NRC has included a backfit provision in the rule in Sec. 70.76. The 
wording of Sec. 70.76 is similar to the current language in Sec. 76.76. 
For requirements other than Subpart H, this provision will apply 
immediately after NRC publication of backfit guidance. For Subpart H 
requirements, this provision shall apply for a licensee as soon as the 
NRC approves that licensee's ISA Summary pursuant to Sec. 70.66. The 
NRC will publish guidance that will address the qualitative versus 
quantitative analysis issue and consideration of chemical risks. The 
NRC staff anticipates completing this guidance within six months of the 
publication of the final rule. Under the Sec. 70.76 backfit provision, 
a backfit analysis is not required for modifications necessary to bring 
the facility into compliance with the rule, including the performance 
requirements in Subpart H. The subject of backfit is discussed in more 
detail in an attachment to SECY-00-0111.

E. Definitions

    Comment E.1: One commenter recommended a change (from 4 percent to 
5 percent enrichment) in the definition of a critical mass of SNM to 
reflect the higher enrichments that are currently in use.
    Response: The definition of critical mass of SNM in Part 70 is used 
solely to determine when Subpart H applies. To emphasize this point, 
the definition was changed to include the phrase, ``for purposes of 
subpart H.'' The definition, including the 4 percent figure, is 
identical to that used in Sec. 70.24, which requires criticality 
accident alarms and other related measures.
    Comment E.2: Regarding the issue of ``reasonable assurance,'' two 
commenters stated that, in the definition of available and reliable to 
perform their function when needed, the use of the term ``ensure'' 
implies a level of certainty that is unrealistically high. Both 
commenters recommended replacing the term ``ensure'' with the term 
``provide reasonable assurance.'' One commenter also recommended 
removing the word ``continuous'' from the definition, which would now 
read ``* * * means that * * * items relied on for safety will perform 
their intended safety function when needed and management measures will 
be implemented to provide reasonable assurance of compliance with the 
performance requirements of Sec. 70.61.''
    Response: The definition was revised to remove the word 
``continuous,'' but no change was made regarding ``ensure.'' With 
respect to ``ensure,'' the proposed rule language does not indicate a 
level of certainty that is unrealistic. The term ``ensure'' is used 
extensively throughout NRC's regulations in the context of a licensee's 
obligations to connote ``make sure'' or ``make certain.'' Specifically, 
elsewhere in Part 70 alone, the term is used in this context eight 
times: Secs. 70.24(a)(3), 70.32(j), 70.38(g)(4)(iii), 70.51(a)(10), 
70.52(c), 70.57(b)(3), 70.57(b)(4), and 70.57(b)(6). Whereas, the term 
``reasonable assurance'' is used just once in Part 70, in 
Sec. 70.23(b), to describe the level of assurance that the Commission 
must find in order to approve construction. The use of ``ensure'' in 
the definition of ``available and reliable to perform their function 
when needed''' in Sec. 70.4 is appropriate. In short, licensees 
``ensure'' and the Commission determines ``with reasonable assurance.'' 
Regarding the issue of ``continuous compliance,'' the definition of 
``available and reliable'' in Sec. 70.4 has been modified to delete the 
word ``continuous.'' This change recognizes the concept that a failure 
of an item relied on for safety does not automatically infer a failure 
to meet the performance requirements of Sec. 70.61. In addition, the 
NRC recognizes that items relied on for safety may temporarily not be 
available (i.e., not continuous) when taken out of service for 
maintenance or functional testing; however, the performance 
requirements must still be met. A discussion has been added to Chapter 
3 in the SRP to address the relationship of failures of items relied on 
for safety to meeting the performance requirements.
    Comment E.3: One commenter stated that there is a ``disconnect'' 
regarding the definition of the term items relied on for safety and 
recommends that the term be replaced by the term Measures relied on for 
safety.
    Response: The reason for the comment is not clear, but perhaps the 
commenter objects to the use of the term ``item'' to refer to a 
personnel action. Part 70 does, in fact, allow human actions to be 
items relied on for safety and permits flexibility in determining how 
the items and measures are defined. Consequently, the Commission has 
retained the original text in the final rule. (See related Comments 
B.2. and E.4.)
    Comment E.4: One commenter was concerned that the term items relied 
on for safety includes ``activities of personnel,'' and proposed 
changing the definition in 70.4 to limit items relied on for safety to 
``structures, systems, equipment, and components.'' According to the 
commenter, it is reasonably straightforward to classify physical items 
as being relied upon for safety, and to apply graded quality assurance 
controls, including management measures, to design, construction, 
operation, and maintenance, etc., of those physical items, based on 
their respective safety functions. The commenter stated that it can be 
confusing to try and classify and grade items when they include 
``personnel activities,'' since an activity has little importance 
absent the context of its influence on a physical item's safety 
function. Removing ``personnel activities'' from the definition of 
items relied on for safety would not limit their

[[Page 56219]]

importance but rather, would put activities in context with the 
structures, systems, equipment, or components to which they are 
related, without necessitating a change in the balance of the proposed 
rule. The commenter stated that removing personnel activities from the 
definition of items relied on for safety will also help address the 
concern raised (in comment B.2) regarding the treatment of procedures 
as items relied on for safety.
    Response: No change was made to the rule language. Human actions 
that are relied on to prevent an accident (i.e., administrative 
controls) are as important as the ``physical items'' needed to prevent 
an accident. Just as there are measures (e.g., maintenance, 
configuration management) needed to ensure the availability and 
reliability of physical controls, there are analogous measures (e.g., 
training, procedures) needed to ensure the availability and reliability 
of human actions. Graded approaches that can be applied to the 
maintenance of a physical control depending on the risk significance of 
the control could also be applied to the training of workers who 
perform safety functions, depending on the risk significance of the 
human's actions. Although the reliability of engineered controls may be 
higher than administrative controls, the final rule allows licensees 
the flexibility to employ both engineered as well as administrative 
controls.
    Comment E.5: One commenter stated that the NRC should define the 
terms likely, unlikely, highly unlikely, and credible in the rule so 
that there will be one set of definitions applied to all nuclear fuel 
facilities. The commenter stated that this will minimize the 
interpretation and application of these terms in the ISA.
    Response: No change in the rule language has been made. Part 70 
applies to different types of fuel cycle facilities, some of which are 
more complex and have more accident sequences than others. Accordingly, 
since the application of the terms in the rule will be necessarily 
specific to the individual context in which they are applied, the 
development of a definition for these terms in the rule language is 
impracticable. The Commission, however, will provide general guidance 
on the application of the terms unlikely and highly unlikely in the SRP 
to aid licensees in implementing the provisions of the rule.
    Comment E.6: One commenter recommended a change in the definition 
of worker. In particular, the language ``* * * exposure to radiation 
and /or radioactive material from licensed and unlicensed sources of 
radiation'' would be replaced with ``* * * exposure to radiation and /
or radioactive material from licensed sources of radiation, and 
radiation from man-made non-regulated sources (e.g., an individual).'' 
As originally defined, persons who are subject to occupational doses 
from natural sources of radiation, (e.g., airline pilots and astronauts 
subject to high cosmic background might be included, whereas workers 
involved with the possession or use of unlicensed radioactive materials 
might not be). The commenter stated that the proposed change removes 
this source of confusion.
    Response: The NRC staff agrees in principle with the comment. 
However, the commenter's proposed change does not eliminate the 
confusion (e.g., some man-made unlicensed sources of radiation are part 
of background or otherwise not included in occupational doses as 
defined in NRC's radiation protection standards in 10 CFR 20). Instead, 
in response to the comment, the definition in Sec. 70.4 was changed to: 
Worker, as used in Subpart H, means an individual who receives an 
occupational dose as defined in 10 CFR 20.1003.

F. Miscellaneous

    Comment F.1: One commenter recommended that the criticality 
requirements of Sec. 70.24 be revised to permit alternate criticality 
control provisions to be accepted for DOE facilities without requiring 
an exemption.
    Response: Comments on Sec. 70.24 are outside the scope of the 
rulemaking.
    Comment F.2: Two commenters recommended changes in the 
decommissioning requirements of Secs. 70.22(a)(9) and 70.38. In 
particular, one commenter recommended that the timeliness and schedule 
provisions in the decommissioning requirements of Sec. 70.38 be revised 
to include separate requirements for DOE facilities.
    Response: Comments on Secs. 70.22(a)(9) and 70.38 are outside the 
scope of the rulemaking.
    Comment F.3: One commenter expressed concern with the language in 
Sec. 70.23(b), which states that the Commission will approve 
construction of a plutonium processing and fuel fabrication facility 
only after determining that the design bases of SSCs, and the attendant 
quality assurance program are adequate to protect against natural 
phenomena and the consequences of potential accidents. In particular, 
the commenter stated that this provision, as written, seems contrary to 
other changes being proposed under the draft rule, because it addresses 
consequences of potential accidents, as opposed to the risk associated 
with credible accidents.
    Response: Section 70.23(b) has not been modified in this 
rulemaking. The reference to ``consequences'' in the rule language does 
not preclude a risk-informed approach in satisfying this requirement. 
The NRC will need to consider the risk, and thus the likelihood of 
consequences of potential accidents occurring, in order to determine 
whether there is reasonable assurance of protection against such 
consequences. This consideration of risk will be important in 
determining the need for (and the ability of) the applicant to reduce 
the likelihood of accidents and to mitigate their consequences.
    Comment F.4: One commenter recommended that Sec. 70.11 be revised 
to reflect the applicability of NRC authority over a MOX fuel 
fabrication facility owned by the DOE, pursuant to changes in law last 
year.
    Response: The NRC agrees with the comment, but believes a separate 
rulemaking is required. Since October 17, 1998, when the amendment to 
Section 202 of the Energy Reorganization Act of 1974 was enacted, 
Sec. 70.11, as well as several other subsections of the regulations, 
need to be updated to reflect this legislative change. However, to 
address this subsection and all the other instances, and to avoid the 
necessity for potential future revisions of this type, the NRC intends 
to institute an administrative-type rule amendment to conform all of 
the references to Section 202 in the regulations, including Sec. 70.11, 
to merely cite Section 202, rather than repeat the text of that 
section. Because this rule change affects various parts of the 
regulations, it will be conducted independently of the current Part 70 
amendments.
    Comment F.5: One commenter stated that as additional DOE facilities 
are licensed by the NRC under Part 70, the NRC should ensure that the 
requirements address the full range of fissionable and fissile 
materials at these facilities.
    Response: This issue is beyond the scope of the rulemaking. It will 
be addressed, if necessary, in the future.
    Comment F.6: One commenter agreed that the proposed rule is 
entirely consistent with the U.S. Environmental Protection Agency's 
Risk Management Program regulations and the general duty clause of the 
Clean Air Act, and contains appropriate complementary safety measures 
for facilities possessing a critical mass of SNM.

[[Page 56220]]

    Response: No response necessary.
    Comment F.7: One commenter strongly recommended that the NRC adopt, 
by reference, the 1998 edition of National Fire Protection Association 
(NFPA) 801, ``Facilities Handling Radioactive Materials.'' NFPA 801 
would apply to Sec. 70.62, ``Safety program and integrated safety 
analysis,'' that addresses protection from all relevant hazards, 
including radiological, criticality, fire, and chemical. The NFPA 
standard would also apply to Sec. 70.64, ``Requirements for new 
facilities or new processes at existing facilities,'' that addresses 
fire protection. The reference to NFPA 801 is in keeping with the 
requirements of Public Law 104-113, ``National Technology Transfer and 
Advancement Act,'' that requires Federal agencies to use private 
sector-developed national consensus technical standards in carrying out 
public policy, wherever appropriate.
    Response: The suggested change would be an unnecessarily 
prescriptive rule requirement. Instead, the NRC identifies the 
standards in NFPA 801 and 600 as an acceptable approach for 
demonstrating compliance with 10 CFR Part 70 in the SRP.
    Comment F.8: One commenter noted that the proposed rule 
incorporates the current terms of the MOU between the NRC and OSHA. 
This should avoid misunderstanding and result in more effective 
implementation for all concerned parties.
    Response: Although the rule is consistent with the NRC-OSHA MOU, 
the rule itself does not incorporate the terms of the MOU. 
Nevertheless, the NRC agrees with the spirit of the comment.
    Comment F.9: Two commenters expressed concern over those portions 
of Secs. 70.22 and 70.23 of the existing rule that address the 
regulation of plutonium processing and fuel fabrication facilities. One 
commenter asked if Sec. 70.22 (f) should be coordinated with 
Sec. 70.65. The commenter noted that it is not clear if the 
requirements are collateral, complementary, or redundant. The same 
commenter stated that Sec. 70.23(b) should be examined to clarify the 
need for this requirement in light of similar information being 
submitted pursuant to Sec. 70.65. The second commenter agreed, stating 
that Sec. 70.22(f) requires plutonium-related applicants to provide 
information on the facility site and design basis of principal SSCs, 
etc., as part of the license application. The commenter believed that 
this information is also required in other sections of the revised 
rule, and thus is redundant.
    Response: No change in the rule language has been made. The 
requirements are not viewed as redundant, considering: the timeframe 
for submittal of information required by the two sections could be 
different; and Sec. 70.23(b) contains a requirement for NRC 
construction approval before the start of construction.
    Comment F.10: Two commenters were concerned about the construction 
authorization provisions in Secs. 70.23(b) and 70.23(a)(7). According 
to one commenter, irrespective of Sec. 70.65, the construction 
authorization provision in Sec. 70.23(b) appears to be an unnecessary 
step and should be considered for deletion by the NRC. If the NRC 
chooses to retain Sec. 70.23(b), the NRC should clarify how the 
authorization process would be conducted, given that the procedural 
step has never been exercised. Furthermore, the NRC should identify how 
the ``design basis'' authorization is defined, why it is necessary, and 
how it relates to the ISA. The second commenter noted that 
Sec. 70.23(a)(7), which applies to other Part 70 licensees, allows 
construction to commence based on a conclusion by the Director, NRC 
Office of Nuclear Material Safety and Safeguards, that environmental 
impacts have been appropriately addressed. The commenter stated that 
this discretion afforded the NRC under Sec. 70.23(a)(7)--i.e., NRC's 
authority over construction associated with ``any * * * activity which 
the Commission determines will significantly affect the quality of the 
environment'' is adequate to ensure the sufficiency of information 
provided to the NRC to authorize or disallow construction. The 
commenter proposes that Sec. 70.23(a)(7) be clarified for applicability 
to plutonium facilities, and that Secs. 70.22(f), 70.23(a)(7),and 
70.23(b) be eliminated. Doing so would avoid the preconception that, 
irrespective of design features and material composition, plutonium is 
``more special'' than other SNM.
    Response: No change in the rule language has been made. The Atomic 
Energy Commission specifically established these requirements (see 36 
FR 9786; May 28, 1971 and 36 FR 17573; September 2, 1971) for plutonium 
facilities in recognition of the potential exposures and ground-
contamination levels that may result if only a small fraction of the 
dispersible plutonium in process were released (see SECY-R 188, March 
17, 1971). The current revisions to Part 70 do not impact this section 
and therefore, the suggested change is outside the scope of the 
rulemaking. Regarding the authorization process, the NRC staff has 
clarified this process in a letter to Duke, Cogema, Stone & Webster, 
dated September 10, 1999. The design basis was also identified in this 
letter. NRC provided additional guidance on this process in the draft 
Standard Review Plan for a Mixed Oxide Fuel Fabrication Facility. In 
addition, the NRC staff is currently assessing the opportunities for 
hearings associated with the review of a license application for a 
plutonium processing facility and may offer additional guidance on this 
topic later in 2000.
    Comment F.11: One commenter noted that, under Sec. 70.23(a)(8), the 
NRC will approve a plutonium facility's license application only after 
construction of principal SSCs has been completed in accordance with 
the application. Certainly this is not a requirement unique to 
plutonium facilities. The NRC already has the authority to grant 
licenses conditional on successful completion of certain actions (such 
as successful start-up testing, training, etc.). Completion of 
construction in accordance with the license application seems such an 
obvious condition that this specific provision seems redundant and 
therefore unnecessary.
    Response: The NRC established Sec. 70.23(a)(8) specifically for 
plutonium processing facilities. Because the current revisions to Part 
70 do not impact this section, comments regarding this section are 
outside the scope of the rulemaking. See response to Comment F.10.
    Comment F.12: One commenter noted that the terminology in Appendix 
A (b)(1) clearly ties the failure to the performance requirements. The 
phrase, ``and which results in failure to meet the performance 
requirements of Sec. 70.61'' is very clear. This phrase should be 
consistently included in Appendix A (b)(2)-(5) using the exact same 
wording.
    Response: No change in the rule language has been made. The linkage 
to the failure to meet the performance requirements is already included 
in Appendix A (b)(2) and (b)(3). For the events described in Appendix A 
(b)(4) and (5), the NRC staff desires to be informed when such events 
occur, regardless of the licensee's determination with respect to the 
performance requirements. In these cases, the NRC staff will 
independently confirm the licensee's assessment of whether the 
performance requirements were met, on the basis of the information 
reported.
    Comment F.13: One commenter stated that the reporting requirements 
of Sec. 70.50 continue to misrepresent the principles of the 1988 NRC-
OSHA MOU. Section 70.50(c)(1)(iii)(A)

[[Page 56221]]

requires the reporting of chemical hazards and Sec. 70.50(c)(1)(iii)(B) 
requires the reporting of personnel exposures to chemicals. According 
to the commenter, although the MOU principles have been correctly 
incorporated into other proposed revisions to Part 70 (e.g., 
Secs. 70.4, 70.61(b), 70.62(c), 70.64(a), 70.74, and Appendix A), they 
were incorrectly referenced in Sec. 70.50. MOU principle (2) limits NRC 
jurisdiction to regulation of chemical hazards of licensed material and 
hazardous chemicals produced from licensed material. The two 
aforementioned sections of Sec. 70.50 should be corrected to properly 
incorporate the MOU principles.
    Response: The rule was revised in response to the comment to 
reflect more precisely the language in the NRC-OSHA MOU.
    Comment F.14: One commenter noted that applicants for licenses to 
operate new facilities or new processes at existing facilities would be 
expected (``Statements of Consideration,'' 64 FR 41346) to update their 
ISAs, based on as-built conditions, and submit the results to the NRC 
before operation. The process for uranium enrichment facilities that 
must comply with Sec. 70.23a would differ from this description. 
Uranium enrichment facilities would submit a complete license 
application, including an ISA Summary, for construction and operation. 
This application would be the basis for NRC review, and culminate in 
issuance of a license for construction and operation. After issuance of 
the license, the licensee would institute change control under 
Sec. 70.72. The licensee would then be required to submit summaries of 
changes and ISA Summary updates as required by Sec. 70.72. An 
inspection would verify that the facility has been constructed in 
accordance with the license, before operation, as required by 
Sec. 70.32(k). No pre-operational submittal and review of an updated 
ISA Summary are anticipated for uranium enrichment facilities because 
their configuration would be controlled after issuance of the 
construction and operation license.
    Response: As recognized by the commenter, no changes to the rule 
are necessary. The differences in the licensing process for enrichment 
facilities other than the gaseous diffusion plants (regulated under 10 
CFR Part 76) reflect the process mandated in Section 193 of the Atomic 
Energy Act of 1954, et. seq.

III. Changes from the Proposed Rule

Subpart A--General Provisions

Authority
    This section has been changed to reflect the redesignations of 
Secs. 70.61 and 70.62 as Secs. 70.81 and 70.82, respectively.

Section 70.4  Definitions

    The definition of ``available and reliable to perform their 
function when needed'' has been modified to eliminate the need to 
maintain ``continuous'' compliance with the performance requirements of 
Sec. 70.61. The definition of ``configuration management'' has been 
modified to clarify its role as a ``management measure.'' The 
definition of ``critical mass of special nuclear material'' has been 
modified to emphasize that the definition is only for the purposes of 
Subpart H. The definition of ``double contingency'' has been changed to 
provide minor clarification. The definition for ``worker'' has been 
clarified and has been revised to emphasize that the definition is only 
for purposes of Subpart H. The definitions of ``ISA'' and ``ISA 
Summary'' have been changed to indicate that the ISA can be performed 
on a process by process basis and the ISA Summary can be submitted in 
multiple documents that cover all the portions of the facility.

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

Section 70.50  Reporting Requirements

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

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

Section 70.60  Applicability

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

Section 70.61  Performance Requirements

    The performance requirements in Secs. 70.61(b) and (c) have been 
revised to provide clarification. The requirement to establish a 
controlled area in Sec. 70.61(f) has been revised to clarify the 
conditions for establishing the controlled area, and to clarify the 
applicability of the performance requirements to individuals within the 
controlled area. The requirement in Sec. 70.61(f)(2) to provide 
training ``in accordance with'' 10 CFR 19.12(a)(1)-(5) has been revised 
to clarify that equivalent training is acceptable. The new language 
specifies that this training must ``satisfy'' 10 CFR 19.12(a)(1)-(5).

Section 70.62  Safety program and integrated safety analysis

    The requirement to establish and maintain a safety program in 
Sec. 70.62(a)(1) has been revised to clarify that the safety program 
referred to in this section is focused on the safety program for 
satisfying the new Subpart H requirements, and that the application of 
management measures may be graded according to risk. Section 
70.62(a)(3) has been modified to make it performance-based by 
eliminating the prescriptive requirement to maintain a log of failures 
for items relied on for safety. Section 70.62(c)(3) has been revised to 
clarify the schedule for planning and performing an ISA, correcting all 
performance deficiencies, and submitting the ISA Summary to the NRC for 
approval. The ISA Summary can be submitted as a single document, or as 
a sequence of documents (e.g., on a process basis). The approval 
process for the ISA Summary will require the issuance of a license 
amendment; however, a license condition will be established that allows 
the licensee to make changes in accordance with Sec. 70.72, including 
certain changes that do not require prior NRC approval. In addition, a 
provision has been added to the schedule for complying with the 
requirements in subpart H for factors beyond control of the licensee. 
This would allow additional time for correcting a performance 
deficiency if the NRC approves. Also Sec. 70.62(d) has been modified to 
reflect that management measures may be graded commensurate with the 
reduction in risk.

Section 70.64  Requirements for New Facilities or New Processes at 
Existing Facilities

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

Section 70.65  Additional Content of Applications

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

[[Page 56222]]

management measures, is submitted with the license application.

Section 70.66  Additional Requirements for Approval of License 
Application

    Section 70.66(b) has been added to clarify, for existing licensees, 
the basis for Commission approval of the ISA plan, submitted under 
Sec. 70.62(c)(3)(i), and the ISA Summary, submitted under 
Sec. 70.62(c)(3)(ii).

Section 70.72  Facility Changes and Change Process

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

Section 70.76  Backfitting

    Section 70.76 was added to include requirements for performing a 
backfit analysis. The wording in Sec. 70.76 is similar to the current 
language in Sec. 76.76 for gaseous diffusion plants with one exception, 
70.76(a)(4)(i). The exception in Sec. 70.76(a)(4)(i) relates to the 
backfit requirements being inapplicable to changes associated with 
bringing the facility in compliance with the requirements of the new 
subpart H. The backfit section includes a provision stating that it 
shall apply for subpart H requirements as soon as NRC approves that 
licensee's ISA Summary (contents of ISA Summary described in 
Sec. 70.65(b)) pursuant to Sec. 70.66 and, for requirements other than 
Subpart H, it shall apply immediately after NRC publication of backfit 
guidance.

Section 70.92  Criminal Penalties

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

IV. Section-by-Section Analysis of Part 70 Amendments

Authority
    This section has been changed to reflect the redesignations of 
Secs. 70.61 and 70.62 as Secs. 70.81 and 70.82, respectively.

Subpart A--General Provisions

Section 70.4  Definitions

    Definitions of the following 12 terms have been added to this 
section to clarify the meaning of certain terms and phrases used in the 
new Subpart H: ``Acute,'' ``Available and reliable to perform their 
function when needed,'' ``Configuration management,'' ``Critical mass 
of SNM,'' ``Double contingency principle,'' ``Hazardous chemicals 
produced from licensed materials,'' ``Integrated safety analysis,'' 
``Integrated safety analysis summary,'' ``Items relied on for safety,'' 
``Management measures,'' ``Unacceptable performance deficiencies,'' and 
``Worker.''

Section 70.14  Foreign Military Aircraft

    This section reflects an administrative change to redesignate this 
section, formerly Sec. 70.13a.

Section 70.17  Specific Exemptions

    This section reflects an administrative change to redesignate this 
section, formerly Sec. 70.14.

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

Section 70.50  Reporting Requirements

    Paragraph (c) has been reworded to include information to be 
transmitted when making verbal or written reports to the NRC. The new 
information derives from the specifics of the new Subpart H, such as 
sequence of events and whether the event was evaluated in the ISA. To 
the extent the new information is also applicable to licensees not 
subject to Subpart H, the information was added with no differentiation 
noted. The new information that would only apply to Subpart H licensees 
is noted.

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

Section 70.60  Applicability

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

Section 70.61  Performance Requirements

    This section identifies the performance requirements that licensees 
subject to Part 70, Subpart H must satisfy. These performance 
requirements explicitly address the risks to workers or members of the 
public and the environmental releases caused by accidents. Because 
accidents are unanticipated events that usually occur over a relatively 
short period of time, the Part 70 changes seek to ensure adequate 
protection of workers, members of the public, and the environment by 
limiting the risk (product of likelihood and consequence) of such 
accidents. If, without the implementation of controls, a high 
consequence event under Sec. 70.61(b) is highly unlikely, then it is 
not necessary for the licensee to apply the engineered or 
administrative controls mentioned in the rule. Similarly, if, without 
the implementation of controls an intermediate consequence event under 
Sec. 70.61(c) is unlikely, then it is not necessary for the licensee to 
apply the engineered or administrative controls mentioned in the rule.

Section 70.62  Safety Program and Integrated Safety Analysis

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

Section 70.64  Requirements for New Facilities or New Processes at 
Existing Facilities

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

Section 70.65  Additional Content of Applications

    In addition to the information that currently must be submitted to 
the NRC under Sec. 70.22, for a license application, this section 
requires additional information to be submitted to demonstrate 
compliance with the new subpart H requirements. In particular, this 
additional information includes a description of the applicant's safety 
program established under Sec. 70.62. This information will be 
incorporated in the license, as appropriate. The ISA Summary must be 
submitted with the license application or in accordance with 
70.62(c)(3)(ii), but will not be incorporated in the license.

[[Page 56223]]

Section 70.66  Additional Requirements for Approval of License 
Application

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

Section 70.72  Facility Changes and Change Process

    This section contains requirements that govern changes to site, 
structures, systems, equipment, components, and activities of personnel 
after a license application has been approved. It requires the licensee 
to establish and use a configuration management system to evaluate 
changes and the potential impacts of those changes before implementing 
them. The regulation permits the licensee to make certain changes 
without NRC pre-approval, but requires the licensee to submit a brief 
summary of the changes plus updated ISA Summary pages annually within 
30 days after the end of the calender year during which the changes 
occurred.

Section 70.73  Renewal of Licenses

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

Section 70.74  Additional Reporting Requirements

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

Section 70.76  Backfitting

    This section contains requirements for performing a backfit 
analysis that are based on those in 10 CFR 76.76. It will become 
effective after NRC publication of backfit guidance. NRC staff will 
work with stakeholders to develop backfit guidance which will include 
making clear that an adequate demonstration can be based on 
quantitative or qualitative evaluations of the nature of the increase 
in the overall health and safety protection of the public. The NRC will 
publish a separate Federal Register notice upon publication of the 
guidance to indicate the effectiveness of Sec. 70.76. After that notice 
is published, this provision will not be applied to Subpart H 
requirements until the NRC approves the licensee's ISA Summary pursuant 
to Sec. 70.66. If the approved ISA Summary is one of a sequence of 
approvals (e.g., on a process basis), the backfit provision will apply 
to the portion of the facility covered by that ISA Summary document. 
However, the backfit provision does not apply to changes to those 
portions of the facility that are required by the NRC staff to address 
system interaction effects identified through the review of other ISA 
Summary submissions for that facility.

Subpart J--Enforcement

Section 70.92  Criminal Penalties

    This section has been changed to reflect the redesignations of 
Secs. 70.13a, 70.14, 70.61, 70.62, 70.71, and 70.72 as Secs. 70.14, 
70.17, 70.81, 70.82, 70.91, and 70.92, respectively, and the addition 
of Secs. 70.66, 70.73, and 70.76.
Appendix A--Reportable Safety Events
    This appendix contains a list of events that licensees must report 
to the NRC. These events are categorized according to their 
consequences (or potential consequences) and fall into two classes: a 
1-hour or 24-hour reporting timeframe. The emphasis on consequences, 
rather than risk, is appropriate in this case because the event has 
already occurred. Appendix A also requires concurrent reporting of 
events when a news release is made or if other Government agencies are 
notified, as is done under 10 CFR 50.72, to enhance coordination and 
support NRC's ability to respond to questions concerning the safety of 
NRC-licensed facilities.

V. 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 is not a major Federal 
action significantly affecting the quality of the human environment and 
therefore, an environmental impact statement is not required.
    The amendments to 10 CFR Part 70 are intended to provide increased 
confidence in the margin of safety at certain facilities that possess a 
critical mass of SNM. To accomplish this objective, the amendments: (1) 
Identify appropriate performance requirements and the level of 
protection needed to prevent or mitigate accidents that exceed such 
requirements; (2) require affected licensees to perform an ISA to 
identify potential accidents at the facility and the items relied on 
for safety; (3) require the implementation of measures to ensure that 
the items relied on for safety are available and reliable to perform 
their functions when needed; (4) require the safety bases to be 
maintained, and changes reported to the NRC; (5) allow for licensees to 
make certain changes to their safety program and facilities without 
prior NRC approval; (6) require reporting of certain events; and (7) 
require a backfit analysis under specified conditions.
    The rule language that defines the performance requirements is 
relevant to the question of environmental impact. Licensees are 
required to protect against the occurrence of or to mitigate the 
consequences of accidents that could adversely affect workers, the 
public, or the environment. For example, licensees are required to 
provide an adequate level of protection against a ``release of 
radioactive material to the environment outside the restricted area in 
concentrations that, if averaged over 24 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', NRC's, other 
governmental agencies', and the public's understanding of the risks at 
these facilities and licensees' ability to ensure that those risks are 
appropriately controlled. For existing licensees, any deficiencies 
identified in the ISA (which must be completed within 4 years) will 
need to be promptly addressed. For new licensees, operations will not 
begin unless licensees demonstrate an adequate level of protection 
against potential accidents identified in the ISA. As a result, the 
safety and environmental impact of the new amendments is positive. 
There will be less potential adverse impact on the environment from 
licensed operations carried out under the final rule than if those 
operations were carried out under the existing Part 70 regulation. 
Thus, the Commission has determined, based on

[[Page 56224]]

the Environmental Assessment that supports the rule, that there will be 
no significant impact on the human environment from this action.
    The NRC requested public comments on any environmental justice 
considerations that may be related to this rule. No comments were 
received in response to this request.
    The NRC also requested the States' views on the environmental 
assessment for this rule. No comments were received in response to this 
request.
    The Environmental Assessment is available for inspection at the NRC 
Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC. 
Single copies of the Environmental Assessment are available from Barry 
Mendelsohn, Office of Nuclear Material Safety and Safeguards, U.S. 
Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone 
(301) 415-7262; e-mail: [email protected].

VI. Paperwork Reduction Act Statement

    This final rule amends information collection requirements that are 
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). These requirements were approved by the Office of Management and 
Budget, approval number 3150-0009.
    The public reporting burden for this information collection is 
estimated to average 92 hours per response and the recordkeeping burden 
is estimated to average 548 hours, including the time for reviewing 
instructions, searching existing data sources, gathering and 
maintaining the data needed, and completing and reviewing the 
information collection. Send comments on any aspect of this information 
collection, including suggestions for reducing the burden, to the 
Records Management Branch (T-6 E6), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, or by Internet electronic mail at 
[email protected]; and to the Desk Officer, Office of Information and 
Regulatory Affairs, NEOB-10202, (3150-0009), Office of Management and 
Budget, Washington, DC 20503.

VII. 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 
or sponsor, and a person is not required to respond to, the information 
collection.

VIII. Regulatory Analysis

    The Commission has prepared a regulatory analysis on this final 
regulation. The analysis examines the costs and benefits of the 
alternatives considered by the Commission. The analysis is available 
for inspection in the NRC Public Document Room, 2120 L Street NW. 
(Lower Level), Washington, DC. Single copies of the environmental 
assessment are available from Barry Mendelsohn, Office of Nuclear 
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 
Washington, DC, 20555-0001, telephone (301) 415-7262; e-mail: 
[email protected].

IX. Regulatory Flexibility Certification

    As required by the Regulatory Flexibility Act, as amended, 5 U.S.C. 
605(b), the Commission certifies that this rule does not have a 
significant economic impact on a substantial number of small entities. 
The regulation affects facilities that are authorized to possess a 
critical mass of SNM and that are engaged in one of the following 
activities: enriched uranium processing; fabrication of uranium fuel or 
fuel assemblies; uranium enrichment; enriched uranium hexafluoride 
conversion; plutonium processing; fabrication of mixed-oxide fuel or 
fuel assemblies; scrap recovery of SNM or any other activity involving 
a critical mass of SNM that the Commission determines could 
significantly affect public health and safety or the environment. These 
licensees do not fall within the scope of the definition of ``small 
entities'' set forth in the Regulatory Flexibility Act, nor the size 
standards published by the NRC (10 CFR 2.810).

X. Voluntary Consensus Standards

    The National Technology Transfer Act of 1995, Pub. L. 104-113, 
requires that Federal agencies use technical standards developed or 
adopted by voluntary consensus standards bodies unless the use of such 
a standard is inconsistent with applicable law or otherwise 
impractical. In this regulation, the NRC will use the following 
voluntary consensus standard--ANSI/ANS Standard 8.1-1983, ``Nuclear 
Criticality Safety in Operations with Fissionable Material Outside 
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). A consensus standard with the complete scope of the 
requirements established in this rulemaking does not exist. The 
Commission will reference ANS 8.1 and other consensus standards, as 
appropriate, as acceptable approaches to demonstrate compliance with 
specific portions of the final rule. This will be addressed in the 
Standard Review Plan that is being established with the rule.

XI. Backfit Statement

    The NRC has determined that the backfit rule does not apply to this 
final rule; therefore, a backfit analysis is not required for this 
final rule because these amendments do not involve any provisions that 
would impose backfits as defined in 10 CFR Chapter I. However, future 
changes to the requirements in subpart H or NRC requirements that apply 
to facilities covered by subpart H will be subject to the backfit 
requirements in Sec. 70.76 established in this rule.

XII. Small Business Regulatory Enforcement Fairness Act

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

List of Subjects in 10 CFR Part 70

    Hazardous materials transportation, Nuclear materials, Packaging 
and containers, Radiation protection, Reporting and recordkeeping 
requirements, Scientific equipment, Security measures.

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 552 and 553; the NRC is adopting 
the following amendments to Part 70.

PART 70--DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL

    1. The authority citation for part 70 is revised 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 70.81 also issued under secs. 186, 187, 68 
Stat. 955 (42 U.S.C. 2236, 2237). Section 70.82 also issued under 
sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).


[[Page 56225]]



    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 
principle, Hazardous chemicals produced from licensed materials, 
Integrated safety analysis, Integrated safety analysis summary, Items 
relied on for safety, Management measures, Unacceptable performance 
deficiencies, and Worker 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 on 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 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 a management measure that 
provides oversight and control of design information, safety 
information, and records of modifications (both temporary and 
permanent) that might impact the ability of items relied on for safety 
to perform their functions when needed.
* * * * *
    Critical mass of special nuclear material (SNM), as used in Subpart 
H, means special nuclear material in a quantity exceeding 700 grams of 
contained uranium-235; 520 grams of uranium-233; 450 grams of 
plutonium; 1500 grams of contained uranium-235, if no uranium enriched 
to more than 4 percent by weight of uranium-235 is present; 450 grams 
of any combination thereof; or one-half such quantities if massive 
moderators or reflectors made of graphite, heavy water, or beryllium 
may be present.
* * * * *
    Double contingency principle means that process designs should 
incorporate sufficient factors of safety to require at least two 
unlikely, independent, and concurrent changes in process conditions 
before a criticality accident is possible.
* * * * *
    Hazardous chemicals produced from licensed materials means 
substances having licensed material as precursor compound(s) or 
substances that physically or chemically interact with licensed 
materials; and that are toxic, explosive, flammable, 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 facility and external hazards and their potential for 
initiating accident sequences, the potential accident sequences, their 
likelihood and consequences, and the items relied on for safety. As 
used here, integrated means joint consideration of, and protection 
from, all relevant hazards, including radiological, nuclear 
criticality, fire, and chemical. However, with respect to compliance 
with the regulations of this part, the NRC requirement is limited to 
consideration of the effects of all relevant hazards on radiological 
safety, prevention of nuclear criticality accidents, or chemical 
hazards directly associated with NRC licensed radioactive material. An 
ISA can be performed process by process, but all processes must be 
integrated, and process interactions considered.
    Integrated safety analysis summary means a document or documents 
submitted with the license application, license amendment application, 
license renewal application, or pursuant to Sec. 70.62(c)(3)(ii) that 
provides a synopsis of the results of the integrated safety analysis 
and contains the information specified in Sec. 70.65(b). The ISA 
Summary can be submitted as one document for the entire facility, or as 
multiple documents that cover all portions and processes of the 
facility.
    Items relied on for safety mean 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 on 
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, when used in Subpart H of this Part, means an individual 
who receives an occupational dose as defined in 10 CFR 20.1003.

    4. In 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

[[Page 56226]]

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.\1\ To the extent 
that the information is available at the time of notification, the 
information provided in these reports must include:
---------------------------------------------------------------------------

    \1\ The commercial telephone number for the NRC Operations 
Center is (301) 816-5100.
---------------------------------------------------------------------------

    (i) Caller's name, position title, and call-back telephone number;
    (ii) Date, time, and exact location of the event;
    (iii) Description of the event, including:
    (A) Radiological or chemical hazards involved, including isotopes, 
quantities, and chemical and physical form of any material released;
    (B) Actual or potential health and safety consequences to the 
workers, the public, and the environment, including relevant chemical 
and radiation data for actual personnel exposures to radiation or 
radioactive materials or hazardous chemicals produced from licensed 
materials (e.g., level of radiation exposure, concentration of 
chemicals, and duration of exposure);
    (C) The sequence of occurrences leading to the event, including 
degradation or failure of structures, systems, equipment, components, 
and activities of personnel relied on to prevent potential accidents or 
mitigate their consequences; and
    (D) Whether the remaining structures, systems, equipment, 
components, and activities of personnel relied on to prevent potential 
accidents or mitigate their consequences are available and reliable to 
perform their function;
    (iv) External conditions affecting the event;
    (v) Additional actions taken by the licensee in response to the 
event;
    (vi) Status of the event (e.g., whether the event is on-going or 
was terminated);
    (vii) Current and planned site status, including any declared 
emergency class;
    (viii) Notifications, related to the event, that were made or are 
planned to any local, State, or other Federal agencies;
    (ix) Status of any press releases, related to the event, that were 
made or are planned.
    (2) Written report. Each licensee that 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 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 that 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 and Amended]

    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.''


Sec. 70.71  [Redesignated]

    16. Section 70.71 is redesignated as Sec. 70.91.

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


Sec. 70.92  Criminal Penalties

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

    18. In part 70, a new subpart H (Secs. 70.60-70.76) 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.
70.76   Backfitting


Sec. 70.60  Applicability.

    The regulations in Sec. 70.61 through Sec. 70.76 apply, in addition 
to other applicable Commission regulations, to each applicant or 
licensee that is or plans to be authorized to possess greater than a 
critical mass of special nuclear material, and engaged in enriched 
uranium processing, fabrication of uranium fuel or fuel assemblies, 
uranium enrichment, enriched uranium hexafluoride conversion, plutonium 
processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap 
recovery of special nuclear material, or any other activity that the 
Commission determines could significantly affect public health and 
safety. The regulations in Sec. 70.61 through Sec. 70.76 do not apply 
to decommissioning activities performed

[[Page 56227]]

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. 70.76 do not apply to activities that are 
certified by the Commission pursuant to part 76 of this chapter or 
licensed by the Commission pursuant to other parts of this chapter. 
Unless specifically addressed in Sec. 70.61 through Sec. 70.76, 
implementation by current licensees of the Subpart H requirements shall 
be completed no later than the time of the ISA Summary submittal 
required in Sec. 70.62(c)(3)(ii).


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. Engineered controls, administrative controls, or both, shall 
be applied to the extent needed to reduce the likelihood of occurrence 
of the event so that, upon implementation of such controls, the event 
is highly unlikely or its consequences are less severe than those in 
paragrahs (b)(1)-(4) of this section. 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 subpart.
    (c) The risk of each credible intermediate-consequence event must 
be limited. Engineered controls, administrative controls, or both shall 
be applied to the extent needed so that, upon implementation of such 
controls, the event is unlikely or its consequences are less than those 
in paragraphs (c)(1)-(4) of this section. 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 Part 20; or
    (4) An acute chemical exposure to an individual from licensed 
material or hazardous chemicals produced from licensed material that:
    (i) Could lead to irreversible or other serious, long-lasting 
health effects to a worker, or
    (ii) Could cause mild transient health effects to any individual 
located outside the controlled area as specified in paragraph (f) of 
this section. If an applicant possesses or plans to possess quantities 
of material capable of such chemical exposures, then the applicant 
shall propose appropriate quantitative standards for these health 
effects, as part of the information submitted pursuant to Sec. 70.65 of 
this subpart.
    (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 
subpart, 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 addition, the licensee must retain the authority to 
exclude or remove personnel and property from the area. For the purpose 
of complying with the performance requirements of this section, 
individuals who are not workers, as defined in 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 that satisfies 10 CFR 19.12(a)(1)-(5) to 
these individuals and ensures that they are aware of the risks 
associated with accidents involving the licensed activities as 
determined by the integrated safety analysis, and conspicuously posts 
and maintains notices stating where the information in 10 CFR 19.11(a) 
may be examined by these 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 or applicant 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 graded commensurate 
with the reduction of the risk attributable to that item. Three 
elements of this safety program; namely, process safety information, 
integrated safety analysis, and management measures, are described in 
paragraphs (b) through (d) of this section.
    (2) Each licensee or applicant shall establish and maintain records 
that demonstrate compliance with the requirements of paragraphs (b) 
through (d) of this section.
    (3) Each licensee or applicant shall maintain records of failures 
readily retrievable and available for NRC inspection, documenting each 
discovery that an item relied on for safety or management measure has 
failed to perform its function upon demand or has degraded such that 
the performance requirements of Sec. 70.61 are not satisfied. These 
records must identify the item relied on for safety or management 
measure that has failed and the safety function affected, the date of 
discovery, date (or estimated date) of the failure, duration (or 
estimated duration) of the

[[Page 56228]]

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. A failure must be recorded at the time of discovery and the 
record of that failure updated promptly upon the conclusion of each 
failure investigation of an item relied on for safety or management 
measure.
    (b) Process safety information. Each licensee or applicant shall 
maintain process safety information to enable the performance and 
maintenance of an integrated safety analysis. This process safety 
information must include information pertaining to the hazards of the 
materials used or produced in the process, information pertaining to 
the technology of the process, and information pertaining to the 
equipment in the process.
    (c) Integrated safety analysis. (1) Each licensee or applicant 
shall conduct and maintain an integrated safety analysis, that is of 
appropriate detail for the complexity of the process, that identifies:
    (i) Radiological hazards related to possessing or processing 
licensed material at its facility;
    (ii) Chemical hazards of licensed material and hazardous chemicals 
produced from licensed material;
    (iii) Facility hazards that could affect the safety of licensed 
materials and thus present an increased radiological risk;
    (iv) Potential accident sequences caused by process deviations or 
other events internal to the facility and credible external events, 
including natural phenomena;
    (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 subpart, 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. 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. Individuals holding an NRC 
license on September 18, 2000 shall, with regard to existing licensed 
activities:
    (i) By April 18, 2001, 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) By October 18, 2004, or in accordance with the approved plan 
submitted under Sec. 70.62(c)(3)(i), complete an integrated safety 
analysis, correct all unacceptable performance deficiencies, and 
submit, for NRC approval, an integrated safety analysis summary, 
including a description of the management measures, in accordance with 
Sec. 70.65. The Commission may approve a request for an alternative 
schedule for completing the correction of unacceptable performance 
deficiencies if the Commission determines that the alternative is 
warranted by consideration of the following:
    (A) Adequate compensatory measures have been established;
    (B) Whether it is technically feasible to complete the correction 
of the unacceptable performance deficiency within the allotted 4-year 
period;
    (C) Other site-specific factors which the Commission may consider 
appropriate on a case-by-case basis and that are beyond the control of 
the licensee.
    (iii) Pending the correction of unacceptable performance 
deficiencies identified during the conduct of the integrated safety 
analysis, the licensee shall implement appropriate compensatory 
measures to ensure adequate protection.
    (d) Management measures. Each applicant or licensee shall establish 
management measures to ensure compliance with the performance 
requirements of Sec. 70.61. The measures applied to a particular 
engineered or administrative control or control system may be graded 
commensurate with the reduction of the risk attributable to that 
control or control system. The management measures shall ensure that 
engineered and administrative controls and control systems that are 
identified as items relied on for safety pursuant to Sec. 70.61(e) of 
this subpart are designed, implemented, and maintained, as necessary, 
to ensure they are available and reliable to perform their function 
when needed, to comply with the performance requirements of Sec. 70.61 
of this subpart.


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 analysis 
performed pursuant to Sec. 70.62(c) demonstrates that a given item is 
not relied on for safety or does not require adherence to the specified 
criteria.
    (1) Quality standards and records. The design must be developed and 
implemented in accordance with management measures, to provide adequate 
assurance that items relied on for safety will be available and 
reliable to perform their function when needed. Appropriate records of 
these items must be maintained by or under the control of the licensee 
throughout the life of the facility.
    (2) Natural phenomena hazards. The design must provide for adequate 
protection against natural phenomena with consideration of the most 
severe documented 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, 
facility conditions which affect the safety of licensed material, and 
hazardous chemicals produced from licensed material.

[[Page 56229]]

    (6) Emergency capability. The design must provide for emergency 
capability to maintain control of:
    (i) Licensed material and hazardous chemicals produced from 
licensed material;
    (ii) Evacuation of on-site 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 facility layout must be based on 
defense-in-depth practices. \1\ The design must incorporate, to the 
extent practicable:
---------------------------------------------------------------------------

    \1\ As used in Sec. 70.64, Requirements for new facilities or 
new processes at existing facilities, defense-in-depth practices 
means a design philosophy, applied from the outset and through 
completion of the design, that is based on providing successive 
levels of protection such that health and safety will not be wholly 
dependent upon any single element of the design, construction, 
maintenance, or operation of the facility. The net effect of 
incorporating defense-in-depth practices is a conservatively 
designed facility and system that will exhibit greater tolerance to 
failures and external challenges. The risk insights obtained through 
performance of the integrated safety analysis can be then used to 
supplement the final design by focusing attention on the prevention 
and mitigation of the higher-risk potential accidents.
---------------------------------------------------------------------------

    (1) Preference for the selection of engineered controls over 
administrative controls to increase overall system reliability; and
    (2) Features that enhance safety by reducing challenges to items 
relied on for safety.


Sec. 70.65  Additional content of applications.

    (a) In addition to the contents required by Sec. 70.22, each 
application must include a description of the applicant's safety 
program established under Sec. 70.62.
    (b) The integrated safety analysis summary must be submitted with 
the license or renewal application (and amendment application as 
necessary), but shall not be incorporated in the 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, including a description of 
the management measures; 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 each item relied on for safety which 
is 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 to an individual from acute chemical exposure 
to licensed material or chemicals produced from licensed materials 
which are on-site, or expected to be on-site as described in 
Sec. 70.61(b)(4) and (c)(4);
    (8) A descriptive list that identifies all items relied on for 
safety that are the sole item preventing or mitigating an accident 
sequence that exceeds the performance requirements of Sec. 70.61; and
    (9) A description of the definitions of 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.

    (a) An application for a license from an applicant subject to 
subpart H will be approved if the Commission determines that the 
applicant has complied with the requirements of Secs. 70.21, 70.22, 
70.23, and 70.60 through 70.65.
    (b) Submittals by existing licensees in accordance with 
Sec. 70.62(c)(3)(i) will be approved if the Commission determines that:
    (1) The integrated safety analysis approach is in accordance with 
the requirements of Secs. 70.61, 70.62(c)(1), and 70.62(c)(2); and
    (2) The schedule is in compliance with Sec. 70.62(c)(3)(ii).
    (c) Submittals by existing licensees in accordance with 
Sec. 70.62(c)(3)(ii) will be approved if the Commission determines 
that:
    (1) The requirements of Sec. 70.65(b) are satisfied; and
    (2) The performance requirements in Sec. 70.61 (b), (c) and (d) are 
satisfied, based on the information in the ISA Summary, together with 
other information submitted to NRC or available to NRC at the 
licensee's site.


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:

[[Page 56230]]

    (1) Does not:
    (i) Create new types 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
    (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 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 of this chapter.
    (2) For changes that do not require pre-approval under Sec. 70.72, 
the licensee shall submit to NRC annually, within 30 days after the end 
of the calendar year during which the changes occurred, a brief summary 
of all changes to the records required by Sec. 70.62(a)(2) of this 
subpart.
    (3) For all changes that affect the integrated safety analysis 
summary, the licensee shall submit to NRC annually, within 30 days 
after the end of the calendar year during which the changes occurred, 
revised integrated safety analysis summary pages.
    (e) If a change covered by 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 of this 
chapter. 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 subpart 
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 that makes a report required by 
paragraph (a)(1) of this section shall submit a written follow-up 
report within 30 days of the initial report. The written report must 
contain the information as described in Sec. 70.50(c)(2).


Sec. 70.76  Backfitting.

    (a) For each licensee, this provision shall apply to Subpart H 
requirements as soon as the NRC approves that licensee's ISA Summary 
pursuant to Sec. 70.66. For requirements other than Subpart H, this 
provision applies regardless of the status of the approval of a 
licensee's ISA Summary.
    (1) Backfitting is defined as the modification of, or addition to, 
systems, structures, or components of a facility; or to the procedures 
or organization required to operate a facility; any of which may result 
from a new or amended provision in the Commission rules or the 
imposition of a regulatory staff position interpreting the Commission 
rules that is either new or different from a previous NRC staff 
position.
    (2) Except as provided in paragraph (a)(4) of this section, the 
Commission shall require a systematic and documented analysis pursuant 
to paragraph (b) of this section for backfits which it seeks to impose.
    (3) Except as provided in paragraph (a)(4) of this section, the 
Commission shall require the backfitting of a facility only when it 
determines, based on the analysis described in paragraph (b) of this 
section, that there is a substantial increase in the overall protection 
of the public health and safety or the common defense and security to 
be derived from the backfit and that the direct and indirect costs of 
implementation for that facility are justified in view of this 
increased protection.
    (4) The provisions of paragraphs (a)(2) and (a)(3) of this section 
are inapplicable and, therefore, backfit analysis is not required and 
the standards in paragraph (a)(3) of this section do not apply where 
the Commission finds and declares, with appropriately documented 
evaluation for its finding, any of the following:
    (i) That a modification is necessary to bring a facility into 
compliance with Subpart H of this part;
    (ii) That a modification is necessary to bring a facility into 
compliance with a license or the rules or orders of the Commission, or 
into conformance with written commitments by the licensee;
    (iii) That regulatory action is necessary to ensure that the 
facility provides adequate protection to the health and safety of the 
public and is in accord with the common defense and security; or
    (iv) That the regulatory action involves defining or redefining 
what level of protection to the public health and safety or common 
defense and security should be regarded as adequate.
    (5) The Commission shall always require the backfitting of a 
facility if it determines that the regulatory action is necessary to 
ensure that the facility provides adequate protection to the health and 
safety of the public and is in accord with the common defense and 
security.
    (6) The documented evaluation required by paragraph (a)(4) of this 
section must include a statement of the objectives of and reasons for 
the modification and the basis for invoking the exception. If immediate 
effective regulatory action is required, then the documented evaluation 
may follow, rather than precede, the regulatory action.
    (7) If there are two or more ways to achieve compliance with a 
license or the rules or orders of the Commission, or with written 
license commitments, or there are two or more ways to reach an adequate 
level of protection, then ordinarily the licensee is free to choose the 
way that best suits its purposes. However, should it be necessary or 
appropriate for the Commission to prescribe a specific way to comply 
with its requirements or to achieve adequate protection, then cost may 
be a factor in selecting the way, provided that the objective of 
compliance or adequate protection is met.
    (b) In reaching the determination required by paragraph (a)(3) of 
this section, the Commission will consider how the backfit should be 
scheduled in light of other ongoing regulatory

[[Page 56231]]

activities at the facility and, in addition, will consider information 
available concerning any of the following factors as may be appropriate 
and any other information relevant and material to the proposed 
backfit:
    (1) Statement of the specific objectives that the proposed backfit 
is designed to achieve;
    (2) General description of the activity that would be required by 
the licensee in order to complete the backfit;
    (3) Potential change in the risk to the public from the accidental 
release of radioactive material and hazardous chemicals produced from 
licensed material;
    (4) Potential impact on radiological exposure or exposure to 
hazardous chemicals produced from licensed material of facility 
employees;
    (5) Installation and continuing costs associated with the backfit, 
including the cost of facility downtime;
    (6) The potential safety impact of changes in facility or 
operational complexity, including the relationship to proposed and 
existing regulatory requirements;
    (7) The estimated resource burden on the NRC associated with the 
proposed backfit and the availability of such resources;
    (8) The potential impact of differences in facility type, design, 
or age on the relevancy and practicality of the proposed backfit; and
    (9) Whether the proposed backfit is interim or final and, if 
interim, the justification for imposing the proposed backfit on an 
interim basis.
    (c) No license will be withheld during the pendency of backfit 
analyses required by the Commission's rules.
    (d) The Executive Director for Operations shall be responsible for 
implementation of this section, and all analyses required by this 
section shall be approved by the Executive Director for Operations or 
his or her designee.

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

Appendix A to Part 70--Reportable Safety Events

    Licensees must comply with reporting requirements in this 
appendix, except for (a)(1), (a)(2), and (b)(4), after they have 
submitted an ISA Summary in accordance with Sec. 70.62(c)(3)(ii). 
Licensees must comply with (a)(1), (a)(2), and (b)(4) after October 
18, 2000. 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
    (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.

    Dated at Rockville, Maryland, this 6th day of September, 2000.

    For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 00-23354 Filed 9-15-00; 8:45 am]
BILLING CODE 7590-01-P