5228 Federal Register / Vol. 70, No. 20 / Tuesday, February 1, 2005 / Notices 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer of 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(i)–(viii). 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) e-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HEARINGDOCKET@NRC.GOV; or (4) 

facsimile transmission addressed to the 

Office of the Secretary, U.S. Nuclear 

Regulatory Commission, Washington, 

DC, Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

transmission to (301) 415–3725 or by email to OGCMailCenter@nrc.gov. A copy 

of the request for hearing and petition 

for leave to intervene should also be 

sent to the General Counsel, Tennessee 

Valley Authority, ET 11A, 400 West 

Summit Hill Drive, Knoxville, TN 

37902, attorney for the licensee. 

For further details with respect to this 

action, see the application for 

amendment dated September 15, 2004, 

which is available for public inspection 

at the Commission’s PDR, located at 

One White Flint North, File Public Area 

O1 F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. Publicly 

available records will be accessible from 

the Agencywide Documents Access and 

Management System’s (ADAMS) Public 

Electronic Reading Room on the Internet 

at the NRC Web site, http://

www.nrc.gov/reading-rm/adams.html. 

Persons who do not have access to 

ADAMS or who encounter problems in 

accessing the documents located in 

ADAMS, should contact the NRC PDR 

Reference staff by telephone at 1–800–

397–4209, (301) 415–4737, or by e-mail 

to pdr@nrc.gov.

Dated in Rockville, Maryland, this 25th 

day of January 2005.

For the Nuclear Regulatory Commission. 

Douglas V. Pickett, 

Senior Project Manager, Section II, Project 

Directorate II, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation.

[FR Doc. 05–1771 Filed 1–31–05; 8:45 am] 

BILLING CODE 7590–01–P

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 technologyneutral 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 

NRCREP@nrc.gov. 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 CAG@nrc.gov.

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 

AXS3@nrc.gov.

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 AXS3@nrc.gov, if 

you plan to attend the workshop so that 

you can be pre-registered. Preregistration 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 

Federal Register / Vol. 70, No. 20 / Tuesday, February 1, 2005 / Notices 5229

have adequate time to be processed 

through security. 

Please notify Mr. A. Singh at (301) 

415–0250 or e-mail AXS3@nrc.gov 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 technologyneutral 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—

• A safety philosophy. 

• Protective strategies. 

• Risk, design, construction, and 

operational objectives. 

• Treatment of uncertainties. 

• A process for defining the scope of 

requirements. 

• 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 

• 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) 

• 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 

• 8:30 a.m. to 11 a.m.—Open discussion 

with stakeholders on implementation 

and other issues (includes example of 

applying the framework) 

• 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* 

• 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 technologyneutral 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. 

• Is it appropriate to use the 

Commission’s Safety Goal Policy 

Quantitative Health Objectives (QHO ) 

as the level of safety the technologyneutral 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 defensein-depth approach that will ensure 

5230 Federal Register / Vol. 70, No. 20 / Tuesday, February 1, 2005 / Notices 

uncertainties in performance do not 

compromise achieving overall plant 

safety objectives. 

• 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? 

• What is meant by each protective 

strategy? For example, for Barrier 

Integrity protective strategy, what 

constitutes or defines a barrier? 

• Is the use of protective strategies a 

reasonable approach for defining highlevel safety functions? If not, what other 

approach(es) should be considered? 

• Is the use of a deductive analysis of 

each protective strategy, to identify 

technology-neutral requirements and 

performance-based measures, a 

reasonable approach? 

• 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? 

• 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. 

• 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? 

• 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? 

• 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?) 

• Should the F–C curve shown in 

Figure 4–1 be expressed in terms of dose 

or curies released? 

• 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? 

• Can specific regions under the F–C 

curve be related to safety margins so as 

to facilitate implementation of safety 

decision-making? 

• 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? 

• Are the proposed technologyneutral risk guidelines appropriate? If 

not, what should be used? 

• 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? 

• Is the proposed approach for 

protection of the environment 

appropriate and adequate? If not, what 

is appropriate? 

• 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? 

• 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-designbasis 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). 

• 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? 

• 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?

• 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? 

• What risk criteria and process are 

appropriate for non-LWR concepts (e.g., 

high temperature gas reactors) to 

address accident prevention and safety 

classification? 

• 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 bestestimate 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. 

• Should the consideration of 

integrated risk be applied to all reactors 

on a site, not just modular reactors? 

• 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.

Federal Register / Vol. 70, No. 20 / Tuesday, February 1, 2005 / Notices 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. 

• Are the types of uncertainty 

adequately described? If not, what 

should be changed or added? 

• 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?

• Are there additional defense-indepth principles that should be adhered 

to? If so, what are they? 

• Is the proposed defense-in-depth 

criteria for containment appropriate? If 

not, what should be used? 

• 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? 

• Is the implementation of the 

defense-in-depth model described in the 

report appropriate? If not, how it should 

be changed? 

• Are incompleteness uncertainties 

reasonably accounted for? If not, how 

should they be dealt with?

• 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. 

• Does the proposed functional 

performance requirement and criterion 

for containment take into account such 

features as the fuel, core, and cooling 

system design? 

• Are the proposed performance 

requirement and criterion performancebased? 

• Are the proposed performance 

requirement and criterion riskinformed? 

• Does the proposed performance 

requirement and criterion adequately 

account for uncertainties, including 

completeness uncertainties? 

• Would the proposed performance 

requirement and criterion result in 

excessive regulatory burden, including 

containment design, construction and 

operating costs? 

• Does the proposed performance 

requirement and criterion provide for 

public confidence? 

• 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. 

• Should the technology-neutral 

requirements be developed as an 

independent alternative to licensing 

under 10 CFR part 50? 

• Is there a near-term (i.e., 3–5 years) 

need for the framework? 

• 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? 

• The approach for obtaining the 

needed administrative requirements is 

5232 Federal Register / Vol. 70, No. 20 / Tuesday, February 1, 2005 / Notices 

being developed. Is the process 

described so far reasonable? Are the 

discussions on analysis methods and 

qualification, and on research and 

development appropriate? 

• 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)? 

• 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? 

• Are the described checks for 

completeness of the framework 

adequate? What other checks could be 

performed? 

• Is it reasonable and practical to 

maintain a living PRA, which would be 

used to periodically reclassify reactor 

accidents as operating experience 

accrues? 

• 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? 

• 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? 

• At what stage should the 

technology-specific regulatory guides be 

developed and to what level of detail? 

Currently, it is envisioned, prior to preapplication 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. 

• 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 technologyneutral regulations? 

• 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. 

• Will the identified set of 

appendices be helpful? Should any be 

dropped or redirected? 

• 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 performancebased requirements. Identifies the steps 

in this process including the need for 

safety margin.

—Are there additional performancebased 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. 

• Have the appropriate terms been 

identified? If not, what terms should be 

deleted or added? 

• Are the definitions reasonable? If 

not, why? 

• 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 axs3@NRC.GOV.

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