[Federal Register Volume 70, Number 20 (Tuesday, February 1, 2005)]
[Notices]
[Pages 5228-5232]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-1770]


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

NUCLEAR REGULATORY COMMISSION


Workshop on Regulatory Structure for New Plant Licensing, Part 1: 
Technology-Neutral Framework

    The U.S. Nuclear Regulatory Commission (NRC) has issued a working 
draft of a NUREG report ``Regulatory Structure for New Plant Licensing, 
Part 1: Technology-Neutral Framework'' (draft NUREG-3-2005) for public 
review and comment. The purpose of this working draft NUREG is to 
provide an approach, scope, and acceptance criteria that could be used 
by the NRC staff to develop a technology-neutral set of requirements 
for future plant licensing. At the present time, the material contained 
in the working draft NUREG is preliminary and does not represent a 
final staff position, but rather is an interim product issued for the 
purpose of engaging stakeholders early in the development of the 
document and to support a workshop to be held in March 2005. As such, 
certain sections of this document are incomplete and are planned to be 
completed following receipt of initial stakeholder feedback. It is the 
staff's intent to complete this document in late 2005 and issue it as a 
final draft for stakeholder review and comment.
    The work represented in this document is, however, considered 
sufficiently developed to illustrate one possible way to establish a 
technology-neutral approach to future plant licensing and to identify 
the key technical and policy issues which must be addressed; 
accordingly, it can serve as a useful vehicle for engaging stakeholders 
and facilitating discussion.
    The NRC staff has issued a working draft NUREG on ``Regulatory 
Structure for New Plant Licensing, Part 1: Technology-Neutral 
Framework.'' The NRC staff requests comments within 90 days from the 
issuing date of this Federal Register Notice. Comments may be 
accompanied by relevant information or supporting data. Please mention 
draft NUREG-3-2005 in the subject line of your comments. You may submit 
comments by any one of the following methods.
    Mail comments to Rules and Directives Branch, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001.
    E-mail comments to <a href="/cdn-cgi/l/email-protection#99d7cbdacbdcc9d9f7ebfab7fef6ef"><span class="__cf_email__" data-cfemail="38766a7b6a7d6878564a5b165f574e">[email protected]</span></a>. You may also submit comments via 
the NRC's rulemaking Web site at http://ruleforum.llnl.gov. Address 
questions about our rulemaking Web site to Carol Gallagher (301) 415-
5905; e-mail <a href="/cdn-cgi/l/email-protection#f7b4b6b0b7998594d9909881"><span class="__cf_email__" data-cfemail="41020006012f33226f262e37">[email protected]</span></a>.
    Hand deliver comments to: Rules and Directives Branch, Office of 
Administration, U.S. Nuclear Regulatory Commission at (301) 415-5144.
    Requests for information about the draft NUREG may be directed to 
Mr. A. Singh at (301) 415-0250 or e-mail <a href="/cdn-cgi/l/email-protection#07465f54344769756429606871"><span class="__cf_email__" data-cfemail="c889909bfb88a6baabe6afa7be">[email protected]</span></a>.
    Comments will be most helpful if received by April 22, 2005. 
Comments received after this date will be considered if it is practical 
to do so, but the NRC is able to ensure consideration only for comments 
received on or before this date.
    The NRC intends to conduct a workshop on March 14-16, 2005, to help 
facilitate the review and comment process. This workshop will be held 
in the auditorium at NRC headquarters, 11545 Rockville Pike, Rockville, 
Maryland.
    Please notify Mr. A. Singh at (301) 415-0250 or e-mail 
<a href="/cdn-cgi/l/email-protection#46071e15750628342568212930"><span class="__cf_email__" data-cfemail="4f0e171c7c0f213d2c61282039">[email protected]</span></a>, if you plan to attend the workshop so that you can be 
pre-registered. Pre-registration will help facilitate your entry into 
the NRC facility for the workshop. In addition, please arrive at NRC 
headquarters 45 minutes prior to the start of the workshop so that you

[[Page 5229]]

have adequate time to be processed through security.
    Please notify Mr. A. Singh at (301) 415-0250 or e-mail <a href="/cdn-cgi/l/email-protection#44051c1777042a36276a232b32"><span class="__cf_email__" data-cfemail="6120393252210f13024f060e17">[email protected]</span></a> 
if you would like to make a formal presentation at the workshop. Once 
all the presenters have been identified, you will be notified with the 
time allocated for your presentation.

Background

    The Commission, in its Policy Statement on Regulation of Advanced 
Nuclear Power Plants, stated its intention to ``improve the licensing 
environments for advanced nuclear power reactors to minimize complexity 
and uncertainty in the regulatory process.'' The staff noted in its 
Advanced Reactor Research Plan to the Commission, (SECY-03-0059, 
ML023310534) that a risk-informed regulatory structure applied to 
license and regulate new reactors, regardless of their technology, 
could enhance consistency and efficiency of NRC's regulatory process 
across reactors with radically different concepts. As such, this new 
process, if implemented, could be available for use later in the 
decade.
    The NRC's past light-water reactor (LWR) experience, especially the 
recent efforts to risk-inform the regulations, has provided insight 
into the potential value of following a top-down approach for the 
development of a regulatory structure for a new generation of reactors. 
Such an approach could also facilitate the implementation of 
performance-based regulation and make the regulations for new reactors 
more coherent.
    The development of a technology-neutral regulatory structure will 
help ensure that a systematic approach is used to develop the 
regulations that will govern the design, construction, and operation of 
new reactors. This structure will ensure uniformity, consistency, and 
defensibility in the development of the regulations, particularly when 
addressing the unique design and operational aspects of new reactors.

Discussion

    A working draft of NUREG-3-2005, ``Regulatory Structure for New 
Plant Licensing, Part 1: Technology-Neutral Framework,'' has been 
issued for stakeholder review and comment. The objective of the 
regulatory structure for new plant licensing is to provide a 
technology-neutral approach to enhancing the effectiveness and 
efficiency of new plant licensing in the longer term (beyond the 
advanced designs currently in the pre-application stage). This 
regulatory structure has four major parts:
    (1) A technology-neutral framework.
    (2) A set of technology-neutral requirements.
    (3) A technology-specific framework.
    (4) Technology-specific regulatory guides.
    Currently, only work related to Part 1 of the regulatory structure 
for new plant licensing, the technology-neutral framework, has 
proceeded. Work has not been initiated on the other three parts. The 
staff has done enough work to demonstrate the feasibility of developing 
a technology-neutral framework. The framework is a hierarchal structure 
that combines deterministic and probabilistic criteria for developing 
technology-neutral requirements to ensure the protection of the public 
health and safety. The framework contains criteria for developing--
    <bullet> A safety philosophy.
    <bullet> Protective strategies.
    <bullet> Risk, design, construction, and operational objectives.
    <bullet> Treatment of uncertainties.
    <bullet> A process for defining the scope of requirements.
    <bullet> Performance-based concepts.
    For each of these items, the staff has developed preliminary 
``working'' criteria that demonstrate the feasibility of a technology-
neutral framework in sufficient detail to start soliciting stakeholder 
input. However, difficult technical and policy issues associated with 
these items are being addressed by the staff that must be resolved 
before the framework can be completed and implemented. These issues 
will be discussed in detail at the workshop (see below).

Workshop Agenda

    A final agenda will be provided at the workshop. The preliminary 
agenda is as follows:

Monday, March 14, 2005

<bullet> 8:30 a.m. to 10 a.m.--Introduction and NRC presentation 
(Overview of Regulatory Structure for New Plant Licensing, and Policy 
and Technical Issues)
<bullet> 10 a.m. to 5:30 p.m.--Open discussion with stakeholders on 
policy and technical issues (Safety Philosophy, Protective Strategies, 
Risk Objectives, Design, Construction, Operational Objectives, 
Treatment of Uncertainties and Defense-in-Depth, Performance-Based 
Concepts)

Tuesday, March 15, 2005

<bullet> 8:30 a.m. to 11 a.m.--Open discussion with stakeholders on 
implementation and other issues (includes example of applying the 
framework)
<bullet> 12:15 p.m. to 5:30 p.m.--Breakout Sessions (Small, parallel 
group discussions on various policy and technical issues, to be 
identified)

Wednesday, March 16, 2005*

<bullet> 8:30 a.m. to 12:30 p.m.--Specific comments on the working 
draft NUREG and formal stakeholder presentations

    *The workshop may be extended into the afternoon if additional time 
is needed to accommodate stakeholder presentations.

Policy and Technical Issues

    The staff is soliciting comments on the issues associated with 
development and implementation of the framework document. These issues 
include, but are not limited to, the following topics:

1. Safety Philosophy (Level of Safety)

    An issue for Commission consideration with respect to developing a 
new regulatory structure is defining the goal in the technology-neutral 
requirements for achieving enhanced safety. The Advanced Reactor Policy 
states that the Commission ``expects that advanced reactor designs will 
comply with the Commission's Safety Goal Policy'' and that ``advanced 
reactors will provide enhanced margins of safety.'' The framework 
proposes a safety philosophy that will define a level of safety that 
will meet the expectation of enhanced safety. In the framework, the 
staff proposes a safety philosophy directly tied to the Commission's 
1986 Safety Goal Policy (51 FR 28044); that is, the staff proposes that 
the technology-neutral requirements be written to achieve the level of 
safety defined by the Safety Goal Policy Quantitative Health 
Objectives.
    <bullet> Is it appropriate to use the Commission's Safety Goal 
Policy Quantitative Health Objectives (QHO ) as the level of safety the 
technology-neutral regulations should be written to achieve? If not, 
what should be used?

2. Protective Strategies

    Protective strategies are identified that define the safety 
fundamentals for safe nuclear power plant design, construction, and 
operation. They are the fundamental building blocks for developing 
technology-neutral requirements and regulations. Acceptable performance 
in these protective strategies provides reasonable assurance that the 
overall mission of adequate protection of public health and safety is 
met. Moreover, the protective strategies implicitly require a defense-
in-depth approach that will ensure

[[Page 5230]]

uncertainties in performance do not compromise achieving overall plant 
safety objectives.
    <bullet> Is the process described for the development of a 
technology-neutral regulatory structure reasonable? Is it complete? Is 
the relationship between the different pieces of the framework 
understandable? If not, where is it not understandable?
    <bullet> What is meant by each protective strategy? For example, 
for Barrier Integrity protective strategy, what constitutes or defines 
a barrier?
    <bullet> Is the use of protective strategies a reasonable approach 
for defining high-level safety functions? If not, what other 
approach(es) should be considered?
    <bullet> Is the use of a deductive analysis of each protective 
strategy, to identify technology-neutral requirements and performance-
based measures, a reasonable approach?
    <bullet> Are the protective strategies described in Chapter 3, 
``Safety Fundamentals: Protective Strategies'' reasonable? Are they 
complete? If not, what strategies are missing or not reasonable?
    <bullet> Are the basic principles of a performance-based approach 
presented in Chapter 3 sufficiently clear and reasonable? If not, where 
are they not clear or not reasonable?

3. Quantitative Risk Objectives and Criteria, Design, Construction, and 
Operational Objectives and Criteria

    The risk objectives and the design, construction, and operational 
objectives complement the protective strategies. The risk and design 
objectives provide a safety approach for meeting safety and risk goals 
for all facilities, that is parallel to protective strategies. This 
approach ensure that worker risk and environment is maintained within 
acceptable levels, and sets specific design expectations that provide 
defense-in-depth requirements at the design level.
    <bullet> Is meeting a frequency consequence (F-C) curve an 
appropriate way to achieve enhanced safety for new reactors? If so, how 
should the F-C curve be interpreted? How could this interpretation be 
done on a practical basis? Should another approach be used? If so, what 
should it be?
    <bullet> The Top Level Regulatory Criteria (TLRC) is another curve, 
which represents exposure at the site boundary under various 
conditions. What are the advantages and disadvantages of these two 
curves?
    <bullet> With respect to implementing the F-C curve, where and how 
should the consequences be evaluated? (For example: evaluated at a 
particular site and its boundary? Averaged over all weather or for a 
conservatively defined weather?)
    <bullet> Should the F-C curve shown in Figure 4-1 be expressed in 
terms of dose or curies released?
    <bullet> Should the F-C curve be used as the acceptance criteria 
for all event sequences analyzed? If so, how should the cumulative 
effects of all event sequences be considered? Or, should the F-C curve 
frequency represent a cumulative frequency of all event sequences 
leading to a defined consequence?
    <bullet> Can specific regions under the F-C curve be related to 
safety margins so as to facilitate implementation of safety decision-
making?
    <bullet> Are the International Commission on Radiation Protection 
(ICRP) guidelines the appropriate criteria to use for specifying 
radiological limits for new reactors? Should other guidelines be used? 
If so, what are they?
    <bullet> Are the proposed technology-neutral risk guidelines 
appropriate? If not, what should be used?
    <bullet> Is the proposed use of 10 CFR part 20 and GDC 19 of 
appendix A to 10 CFR part 50 appendix A appropriate for worker 
protection? If not, what is appropriate?
    <bullet> Is the proposed approach for protection of the environment 
appropriate and adequate? If not, what is appropriate?
    <bullet> Are the objectives and issues identified in the discussion 
of construction objectives appropriate? Are they sufficiently complete? 
What additional considerations will be important for new reactor 
designs?
    <bullet> Are the operational objectives appropriate? What issues 
are not discussed that likely to be important for new reactors? Are any 
of the identified issues unnecessary for new reactors?
    Commission approved the use of probabilistic criteria for 
identifying events that must be considered for the design, in the 
safety classification of Structures, Systems and Components (SSCs) and 
to replace the single failure criterion. The approach proposed in the 
framework involves identifying event sequence categories by frequency 
to define abnormal operational occurrences (AOOs), design basis 
accidents (DBAs), and beyond-design-basis events, classifying SSCs as 
either risk-significant or non-risk-significant based on the SSCs' 
quantified risk importance and criteria consistent with the work done 
in support of the 10 CFR 50.69 rulemaking; and replace the single-
failure criterion with event sequences from the design-specific 
probabilistic risk assessment (PRA).
    <bullet> Is the proposed approach for the selection of AOOs and 
DBAs reasonable? Should another approach be used? If so, what should it 
be? Are the acceptance criteria reasonable?
    <bullet> Can a technology-neutral definition of accident prevention 
be developed? If so, what should it be? If not, what technology-
specific definitions should be used?
    <bullet> Should a risk-informed safety classification process build 
upon the risk criteria and process contained in 10 CFR 50.69? If not, 
what risk criteria and process should be used?
    <bullet> What risk criteria and process are appropriate for non-LWR 
concepts (e.g., high temperature gas reactors) to address accident 
prevention and safety classification?
    <bullet> What acceptance criteria should be used to reflect 
uncertainties? Should they be set at a defined level of confidence; or 
should evaluation of uncertainty in both the challenge and the 
capability be required?
    The Commission approved the use of scenario-specific source terms, 
provided that the staff understands the fission product behavior, and 
plant conditions and performance. In the framework, the staff used a 
flexible, performance-based approach to establish scenario-specific 
licensing source terms. The key features of this approach are: (1) 
Scenarios are to be selected from a design-specific PRA; (2) source 
term calculations are based on verified analytical tools; (3) source 
terms for compliance should be 95% confidence level values, based on 
best-estimate calculations; and (4) source terms for licensing 
decisions should reflect scenario-specific timing, form, and magnitude 
of the release.
    The approach used for selecting DBAs may result in smaller source 
terms than used for LWR safety analyses. Is this approach reasonable 
for siting? Or should siting be based on a large source term?
    The Commission asked the staff to provide further details on the 
options for, and associated impacts of, requiring that modular reactor 
designs account for the integrated risk posed by multiple reactors.
    <bullet> Should the consideration of integrated risk be applied to 
all reactors on a site, not just modular reactors?
    <bullet> If integrated risk is to be considered on a per site 
basis, how should it be accounted for?

--limit the number of reactors on a site?
--site specific criteria?
--nationwide criteria?
--other criteria?

    Note: See ACRS letter of April 22, 2004 for additional 
considerations.


[[Page 5231]]


    The Commission approved the staff proposal that no change to 
emergency preparedness requirements is needed in the near term. The 
Commission also approved, for the longer term, the staff developing 
guidelines for assessing possible modifications to emergency 
preparedness requirements as part of the work to develop a description 
of defense-in-depth.
    What should the role of emergency preparedness in defense-in-depth 
be, as it relates to possible simplification of the emergency planning 
requirements; e.g., reduction in the size of the emergency planning 
zones (EPZs) for reactors that are designed with greater safety margins 
than the current light water reactors?
    In considering possible changes to the existing emergency 
preparedness regulations or guidance, should factors other than reactor 
size and location, level of safety (i.e., likelihood of release), 
magnitude and chemical form of release, and timing of release be 
addressed? Is consideration of these factors adequate and reasonable? 
If not, why? In addition, should the changes address considerations 
beyond the following; and if so, what are they?

1. Consideration of the full range of accidents.
2. Use of the defense-in-depth philosophy.
3. Prototype operating experience.
4. Acceptance by Federal, State, and local agencies.
5. Acceptance by the public.

4. Treatment of Uncertainties and Defense-in-Depth

    The Commission approved the staff recommendation for developing a 
definition of defense-in-depth that would be incorporated into a policy 
statement. In licensing future reactors, the treatment of uncertainties 
will play a key role in ensuring safety limits are met and the design 
is robust with respect to unanticipated factors. In general, 
uncertainties associated with new plants will tend to be larger than 
uncertainties associated with existing plants due to new technologies 
being used, the lack of operating experience or, in the case of some 
proposed LWRs, new design features (e.g., increased use of passive 
systems). Any licensing approach for new plants must account for the 
treatment of these uncertainties. The aim is to develop an approach for 
future reactors which can be reconciled with past practices used for 
operating reactors, but which improves on past practices by being more 
consistent and by making use of quantitative information where 
possible. The approach recommended for dealing with uncertainties when 
ensuring the safety of new plants is the concept of multiple successive 
layers of barriers and lines of defense against undesirable 
consequences. This approach is usually referred to as defense-in-depth. 
The concept of defense-in-depth is fundamental to the treatment of 
uncertainties.
    <bullet> Are the types of uncertainty adequately described? If not, 
what should be changed or added?
    <bullet> A major reason for including a deterministic 
(structuralist) component in the defense-in-depth model (i.e., the 
protective strategies) is to address the unknown contributors 
(initiating events, failure mechanisms, physical performance, etc.) to 
accidents. The deterministic component of the model requires that each 
protective strategy is implemented, however, the extent or degree to 
which each strategy is implemented is tempered by the associated risk 
(which is the probabilistic or rationalist component of the model).

    --What approaches to determining the degree of defense-in-depth 
provided by each protective strategy would be appropriate?
    --How relevant is the rationalist approach, given the uncertainty 
associated with the unknown contributors?
    --Are expert judgment approaches appropriate? What caveats and 
controls would be needed?
    --Are there ways to structure the uncertainty associated with 
``unknown'' aspects of the risk that can be helpful? Could these be 
used to provide a qualitative description of the uncertainty that would 
provide a basis for assessment?
    --What other possibilities are there?

    <bullet> Are there additional defense-in-depth principles that 
should be adhered to? If so, what are they?
    <bullet> Is the proposed defense-in-depth criteria for containment 
appropriate? If not, what should be used?
    <bullet> Is the defense-in-depth model advocated in the report 
appropriate? Does it achieve the proper balance between structuralist 
and rationalist aspects? If not, how should it be changed?
    <bullet> Is the implementation of the defense-in-depth model 
described in the report appropriate? If not, how it should be changed?
    <bullet> Are incompleteness uncertainties reasonably accounted for? 
If not, how should they be dealt with?
    <bullet> Are the proposed factors for considering changes to 
existing emergency preparedness regulations or guidance appropriate? If 
not, what should be used?
    The Commission asked the staff to develop containment functional 
performance requirements and criteria, working closely with industry 
experts (e.g., designers, Electric Power Research Institute, etc.) and 
other stakeholders regarding options in this area, and to take into 
account such features as core, fuel, and cooling systems design. The 
Commission also stated that the staff should pursue the development of 
functional performance standards, and then submit options and 
recommendations to the Commission on this important policy decision.
    <bullet> Does the proposed functional performance requirement and 
criterion for containment take into account such features as the fuel, 
core, and cooling system design?
    <bullet> Are the proposed performance requirement and criterion 
performance-based?
    <bullet> Are the proposed performance requirement and criterion 
risk-informed?
    <bullet> Does the proposed performance requirement and criterion 
adequately account for uncertainties, including completeness 
uncertainties?
    <bullet> Would the proposed performance requirement and criterion 
result in excessive regulatory burden, including containment design, 
construction and operating costs?
    <bullet> Does the proposed performance requirement and criterion 
provide for public confidence?
    <bullet> How should the options, including the proposed option, be 
revised in consideration of the above questions?

5. Process for Defining Scope of Requirements (and General 
Implementation Issues)

    A deductive process will be developed to identify and define the 
scope and content of detailed technical and administrative requirements 
that are necessary to ensure the safety objectives and criteria are 
met.
    <bullet> Should the technology-neutral requirements be developed as 
an independent alternative to licensing under 10 CFR part 50?
    <bullet> Is there a near-term (i.e., 3-5 years) need for the 
framework?
    <bullet> The derivation of detailed technical requirements is being 
developed. Is the process described (and illustrated with the barrier 
integrity example) for the identification of the scope and content of 
the detailed technical requirements from the protective strategies 
reasonable? How could it be improved?
    <bullet> The approach for obtaining the needed administrative 
requirements is

[[Page 5232]]

being developed. Is the process described so far reasonable? Are the 
discussions on analysis methods and qualification, and on research and 
development appropriate?
    <bullet> Should the technology-neutral requirements build upon and 
utilize 10 CFR part 50 requirements as much as possible (i.e., whenever 
10 CFR 50 requirements are technology neutral they should be 
incorporated)?
    <bullet> Are the desired characteristics of a technology-neutral 
regulatory structure listed in Sections 1.4 and 6.3 of the framework 
reasonable? Is the list complete? If not, what characteristic(s) is 
missing?
    <bullet> Are the described checks for completeness of the framework 
adequate? What other checks could be performed?
    <bullet> Is it reasonable and practical to maintain a living PRA, 
which would be used to periodically reclassify reactor accidents as 
operating experience accrues?
    <bullet> From a regulatory perspective, in terms of enforceability, 
is it practical to include the technology-specific details in a 
regulatory guide, although included as part of the license, or directly 
in a regulation?
    <bullet> Would performance-based requirements developed according 
to appendix A to CFR 10 part 50, sufficiently address enforceability, 
given that prescriptive requirements are easier to enforce?
    <bullet> At what stage should the technology-specific regulatory 
guides be developed and to what level of detail? Currently, it is 
envisioned, prior to pre-application or pre-certification, to develop 
the technology-specific regulatory guides for each technology type, not 
for each applicant. The technology-specific regulatory guide would 
specify how to interpret such statements in the technology-neutral 
regulation as fuel damage, accident prevention.
    <bullet> It is envisioned that these new technology-neutral 
regulations would be a voluntary alternative to 10 CFR part 50. Should 
these regulations be voluntary or mandatory? What would be the 
motivation for an applicant to use this alternative? Should a licensee 
be allowed to seek an exemption to 10 CFR part 50 to propose an 
alternative approach based on the technology-neutral regulations?
    <bullet> Is a technology-neutral framework desirable for licensing 
future reactors? What are the advantages of using a technology-neutral 
framework? What are the difficulties of using such a framework?

6. Appendices

    The following appendices have been identified to provide further 
detailed information in understanding the criteria and guidelines in 
the framework document.
    <bullet> Will the identified set of appendices be helpful? Should 
any be dropped or redirected?
    <bullet> Would additional appendices be helpful? If yes, what 
should be the topic and to what level should it be written?
    A. Guidance for the Formulation of Performance-Based Requirements: 
Provides an explanation of how the topics that must be addressed to 
provide defense-in-depth protection via the protective strategies can 
be implemented through performance-based requirements. Identifies the 
steps in this process including the need for safety margin.

--Are there additional performance-based considerations that should be 
included in appendix A?

    B. Current Quantitative Guidelines for LWRs: The Framework 
discusses the possibility of using surrogates to demonstrate that the 
risk objectives of the frequency-consequence curve have been met. 
Appendix B illustrates how core damage frequency and large early 
release frequency are used for current LWRs as surrogates for the risk 
objectives expressed by the latent cancer QHO and early fatality QHO, 
respectively.

--Are there additional examples of the use of surrogates to achieve 
higher level risk objectives that would be useful here?

    C. Safety Characteristics of New Reactors: Brief summary 
descriptions of a number of possible new reactor concepts. Includes a 
discussion of safety features (and vulnerabilities, if identified) 
structured to make clear the linkage to the Framework.

--Are there additional characteristics/features/attributes of the 
various innovative designs that should receive special attention in 
appendix C?

    D. Probabilistic Risk Assessment Quality Needs for New Reactors: 
There are now standards for PRA of LWRs. This appendix will define PRA 
in a technology-neutral manner (e.g., core damage frequency as a 
definition for Level 1 is technology-specific), identify extensions and 
changes that may be needed for some new reactors, and will describe how 
PRA is related to the development of regulatory requirements for new 
reactors (e.g., development of a living PRA and what a living PRA 
entails).

--What should be the scope and depth of this appendix? At a higher 
level and look to professional organization to develop standard?

    E. Assessment of 10 CFR Part 50 for New Reactors: A review of 10 
CFR Part 50 requirements against a specific new reactor design. 
Identifies where current requirements are directly applicable, which 
requirements are not applicable, which requirements need to be adapted 
to the new design concept, and what design features and uncertainties 
call for new requirements.
    F. Completeness Check: A review of other work being performed in 
this area to identify any significant holes. Review and compare against 
the NEI-02-02 framework and the technical document being prepared by 
IAEA relating to technology-neutral regulations.

--Are there other sources that should be reviewed?

7. Glossary

    A glossary is being developed with a standard set of definitions of 
terms, in order to provide a common understanding, and to help 
facilitate discussions and communication regarding the regulatory 
structure for new plant licensing.
    <bullet> Have the appropriate terms been identified? If not, what 
terms should be deleted or added?
    <bullet> Are the definitions reasonable? If not, why?
    <bullet> Should the definitions be standardized? Can the 
definitions be used elsewhere? If not, which definitions can not be 
standardized, and why?
    Information about the working draft NUREG and the workshop may be 
directed to Mr. A. Singh at (301) 415-0250 or e-mail <a href="/cdn-cgi/l/email-protection" class="__cf_email__" data-cfemail="f6978e85c5b6b8a4b5d8b1b9a0">[email protected]</a>.
    Although a time limit is given for comments on this draft document, 
comments and suggestions in connection with items for inclusion in 
guides currently being developed, or improvements in all published 
guides, are encouraged at any time.

(5 U.S.C. 552(a))

    Dated at Rockville, Maryland, this 25th day of January 2005.

    For the Nuclear Regulatory Commission.
Charles E. Ader,
Director, Division of Risk Analysis and Applications, Office of Nuclear 
Regulatory Research.
[FR Doc. 05-1770 Filed 1-31-05; 8:45 am]
BILLING CODE 7590-01-P


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