Approaches to Risk-Informed and Performance-Based Requirements for Nuclear Power Reactors, 26267-26275 [E6-6745]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 86 (Thursday, May 4, 2006)]
[Proposed Rules]
[Pages 26267-26275]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-6745]


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Proposed Rules
                                                Federal Register
________________________________________________________________________

This section of the FEDERAL REGISTER contains notices to the public of 
the proposed issuance of rules and regulations. The purpose of these 
notices is to give interested persons an opportunity to participate in 
the rule making prior to the adoption of the final rules.

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Federal Register / Vol. 71, No. 86 / Thursday, May 4, 2006 / Proposed 
Rules

[[Page 26267]]



NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50 and 53

RIN 3150-AH81


Approaches to Risk-Informed and Performance-Based Requirements 
for Nuclear Power Reactors

AGENCY: Nuclear Regulatory Commission.

ACTION: Advance notice of proposed rulemaking (ANPR).

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

SUMMARY: The Nuclear Regulatory Commission (NRC) is considering 
modifying its approach to develop risk-informed and performance-based 
requirements applicable to nuclear power reactors. The NRC is 
considering an approach that, in addition to the ongoing effort to 
revise some specific regulations to make them risk-informed and 
performance-based, would establish a comprehensive set of risk-informed 
and performance-based requirements applicable for all nuclear power 
reactor technologies as an alternative to current requirements. This 
new rule would take advantage of operating experience, lessons learned 
from the current rulemaking activities, advances in the use of risk-
informed technology, and would focus NRC and industry resources on the 
most risk-significant aspects of plant operations to better ensure 
public health and safety. The set of new alternative requirements would 
be intended primarily for new power reactors although they would be 
available to existing reactor licensees.
    At the conclusion of this ANPR phase and taking into consideration 
public comment, the NRC will determine how to proceed regarding making 
the requirements for nuclear power plants risk-informed and 
performance-based.

DATES: The comment period expires December 29, 2006. This time period 
allows public comment on the proposals in this ANPR.
    Comments on the general proposals in this ANPR would be most 
beneficial to the NRC if submitted within 90 days of issuance of the 
ANPR. Comments on any periodic updates will be most beneficial if 
submitted within 90 days of their respective issuance. Periodic updates 
that are issued will be placed on the NRC's interactive rulemaking Web 
site, Ruleforum, (https://ruleforum.llnl.gov), for information or 
comment. Supplements to this ANPR are anticipated to be issued and will 
request additional public comments.
    Comments received after the above date will be considered if it is 
practical to do so, but the Commission is able to assure consideration 
only for comments received on or before the above date.

ADDRESSES: You may submit comments by any one of the following methods. 
Please include the following number RIN 3150-AH81 in the subject line 
of your comments. Comments on this ANPR submitted in writing or in 
electronic form will be made available for public inspection. Because 
your comments will not be edited to remove any identifying or contact 
information, the NRC cautions you against including information such as 
social security numbers and birth dates in your submission.
    Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
    E-mail comments to: SECY@nrc.gov. If you do not receive a reply e-
mail confirming that we have received your comments, contact us 
directly at (301) 415-1966. You may also submit comments via the NRC's 
rulemaking Web site at https://ruleforum.llnl.gov. Address questions 
about our rulemaking Web site to Carol Gallagher (301) 415-5905; e-mail 
cag@nrc.gov. Comments can also be submitted via the Federal eRulemaking 
Portal https://www.regulations.gov.
    Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 
20852, between 7:30 a.m. and 4:15 p.m. Federal workdays. (Telephone 
(301) 415-1966).
    Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 
(301) 415-1101.
    Publicly available documents related to this ANPR may be viewed 
electronically on the public computers located at the NRC's Public 
Document Room (PDR), O1 F21, One White Flint North, 11555 Rockville 
Pike, Rockville, Maryland. The PDR reproduction contractor will copy 
documents for a fee. Selected documents, including comments, may be 
viewed and downloaded electronically via the NRC rulemaking Web site at 
https://ruleforum.llnl.gov.
    Publicly available documents created or received at the NRC after 
November 1, 1999, are available electronically at the NRC's Electronic 
Reading Room at https://www.nrc.gov/reading-rm/adams.html. From this 
site, the public can gain entry into the NRC's Agencywide Document 
Access and Management System (ADAMS), which provides text and image 
files of NRC's public documents. If you do not have access to ADAMS or 
if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

FOR FURTHER INFORMATION CONTACT: Joseph Birmingham, Office of Nuclear 
Reactor Regulation (NRR), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001; telephone (301) 415-2829, e-mail: 
jlb4@nrc.gov; or Mary Drouin, Office of Nuclear Regulatory Research 
(RES), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; 
telephone: (301) 415-6675, e-mail: mxd@nrc.gov.

SUPPLEMENTARY INFORMATION:

Background

    The NRC is considering developing a comprehensive set of risk-
informed, performance-based, and technology neutral requirements for 
licensing nuclear power reactors. These requirements would be included 
in NRC regulations as a new 10 CFR Part 53 and could be used as an 
alternative to the existing requirements in 10 CFR Part 50.
    The Commission directed the NRC staff to develop an ANPR to 
facilitate early stakeholder participation in this effort. The 
Commission also directed the NRC staff to: (1) Incorporate in the ANPR 
a formal program plan for risk-informing 10 CFR Part 50, as well as 
other related risk-informed efforts, (2) integrate safety, security, 
and preparedness throughout the effort and (3) include the effort to 
develop risk-informed and performance-based alternatives to the single 
failure criterion (ADAMS Accession Numbers

[[Page 26268]]

ML051290351, ML052570437, and ML052640492).
    The NRC has conducted public meetings and workshops to engage 
interested stakeholders in dialogue on the merits of various approaches 
to risk-inform and performance-base the requirements for nuclear power 
reactors. In particular, the NRC conducted (1) a workshop on March 14-
16, 2005, to discuss the staff's work in development of a technology-
neutral framework in support of a regulatory structure for new plant 
licensing, and (2) a public meeting on August 25, 2005, to discuss 
plans for a risk-informed and performance-based revision to 10 CFR Part 
50. Meeting minutes were taken and are available to the public (ADAMS 
Accession Numbers ML050900045 and ML052500385, respectively). At the 
above workshop and meeting, the NRC discussed the desirability of 
various approaches for risk-informing the requirements for nuclear 
power reactors and particularly for new reactors of diverse types. The 
NRC discussed approaches such as (1) developing an integrated set of 
risk-informed requirements using a technology-neutral framework as a 
basis for regulation, and (2) continuing to risk-inform 10 CFR Part 50 
on an issue-by-issue basis.
    The NRC also plans to continue the ongoing efforts to revise 
specific regulations in 10 CFR Part 50 as described in SECY-98-300, 
``Options for Risk-Informed Revisions to 10 CFR Part 50--Domestic 
Licensing of Productions and Utilization Facilities'' (ML992870048). 
The Commission proposes to focus resources in the near-term on 
completion and subsequent implementation of the ongoing risk-informed 
rulemaking efforts for current operating reactors and not to initiate 
new efforts to risk-inform and performance-base other regulations at 
this time, unless specific regulations or guidance documents are 
identified that could enhance the efficiency and effectiveness of NRC 
reviews of near-term applications.
    Although the NRC conducted the meetings discussed above to get a 
sense of stakeholder interest and to ascertain the desired path 
forward, the NRC is issuing this ANPR to obtain additional comment on 
the proposed approaches, to ensure that the Commission's intent is 
known to all stakeholders, and to allow the NRC to proceed to risk-
inform the requirements for power reactors in an open, integrated, and 
transparent manner.

Proposed Plan

    The NRC has developed a proposed plan to develop an integrated 
risk-informed and performance-based alternative to 10 CFR Part 50 that 
would cover power reactor applications including non-LWR reactor 
designs. Safety, security, and preparedness will be integrated into 
this effort to provide one cohesive structure. This structure will 
ensure that the reactor regulations, and staff processes and programs 
are built on a unified safety concept and are properly integrated so 
that they complement one another. Based on the above, the overall 
objectives of a risk-informed and performance-based alternative to 10 
CFR Part 50 are to: (1) Enhance safety and security by focusing NRC and 
licensee resources in areas commensurate with their importance to 
public health and safety, (2) provide NRC with a framework that uses 
risk information in an integrated manner, (3) use risk information to 
provide flexibility in plant design and operation while maintaining or 
enhancing safety and security, (4) ensure that risk-informed activities 
are coherently and properly integrated such that they complement one 
another and continue to meet the 1995 Commission's PRA Policy 
Statement, and (5) allow for different reactor technologies in a manner 
that will promote stability and predictability in the long term.
    The approach addresses risk-informed power reactor activities and 
the associated guidance documents. Risk-informed activities addressing 
non-power reactors, nuclear materials and waste are not addressed.
    The NRC's proposed approach is to create an entire new Part in 10 
CFR (referred to as ``10 CFR Part 53'') that can be applied to any 
reactor technology and that is an alternative to 10 CFR Part 50. Two 
major tasks are proposed: (1) Develop the technical basis for 
rulemaking for 10 CFR Part 53, and (2) develop the regulations and 
associated guidance for 10 CFR Part 53.

Task 1: Development of Technical Basis

    The objective of this task is to develop the technical basis for a 
risk-informed and performance-based 10 CFR Part 53. The technical basis 
provides the criteria and guidelines for development and implementation 
of the regulations to be included in Part 53. Current activities 
associated with developing the technical basis are described in SECY-
05-0006 (ADAMS accession number ML043560093).
    As the technical basis is being developed, it is anticipated that 
additional issues will be identified for which stakeholder input is 
desired. Therefore, it is envisioned that supplemental issues will be 
added to this ANPR over time.
    At the end of the ANPR phase, the Commission will decide whether to 
proceed to formal rulemaking.

Task 2: Rule Development

    The objective of this task is to develop and issue the regulations 
for 10 CFR Part 53. If upon completion of the technical basis the 
Commission directs the NRC staff to proceed to rulemaking, the NRC 
staff will follow its normal rule development process. The NRC staff 
will develop proposed rule text, interact with stakeholders in an 
appropriate forum (e.g., posting on web, public workshops), and provide 
a proposed rule package to the Commission for consideration.
    In development of the rulemaking, the necessary guidance documents 
to meet the regulations in 10 CFR Part 53 will also be developed.

Specific Considerations

    Before determining whether to develop a proposed rule, the NRC is 
seeking comments on this matter from all interested persons. Specific 
areas on which the Commission is requesting comments are discussed in 
the following sections. Comments, accompanied by supporting reasons, 
are particularly requested on the questions contained in each section.

A. Plan

    The NRC is seeking comments on the proposed described above:
    1. Is the proposed plan to make a risk-informed and performance-
based alternative to 10 CFR Part 50 reasonable? Is there a better 
approach than to create an entire new 10 CFR Part 53 to achieve a risk-
informed and performance-based regulatory framework for nuclear power 
reactors? If yes, please describe the better approach?
    2. Are the objectives, as articulated above in the proposed plan 
section, understandable and achievable? If not, why not? Should there 
be additional objectives? If so, please describe the additional 
objectives and explain the reasons for including them.
    3. Would the approach described above in the proposed plan section 
accomplish the objectives? If not, why not and what changes to the 
approach would allow for accomplishing the objectives?
    4. Would existing licensees be interested in using risk-informed 
and performance-based alternative regulations to 10 CFR Part 50 as 
their licensing basis? If not, why not? If so, please discuss the main 
reasons for doing so.
    5. Should the alternative regulations be technology-neutral (i.e., 
applicable to

[[Page 26269]]

all reactor technologies, e.g., light water reactor or gas cooled 
reactor), or be technology-specific? Please discuss the reasons for 
your answer. If technology-specific, which technologies should receive 
priority for development of alternative regulations?
    6. When would alternative regulations and supporting documents need 
to be in place to be of most benefit? Is it premature to initiate 
rulemaking for non-LWR technologies? If so, when should such an effort 
be undertaken? Could supporting guidance be developed later than the 
alternative regulations, e.g. phased in during plant licensing and 
construction?
    7. The NRC encourages active stakeholder participation through 
development of proposed supporting documents, standards, and guidance. 
In such a process, the proposed documents, standards, and guidance 
would be submitted to and reviewed by NRC staff, and the NRC staff 
could endorse them, if appropriate. Is there any interest by 
stakeholders to develop proposed supporting documents, standards, or 
guidance? If so, please identify your organization and the specific 
documents, standards, or guidance you are interested in taking the lead 
to develop?

B. Integration of Safety, Security, and Emergency Preparedness

    The Commission believes that safety, security, and emergency 
preparedness should be integrated in developing a risk-informed and 
performance-based set of requirements for nuclear power reactors (i.e., 
in this context, 10 CFR Part 53). The NRC has proposed to establish 
security performance standards for new reactors (see SECY-05-0120, 
ADAMS Accession Number ML051100233). Under the proposed approach, 
nuclear plant designers would analyze and establish, at an earlier 
stage of design, security design aspects such that there would be a 
more robust and effective (intrinsic) security posture and less 
reliance on operational (extrinsic) security programs (guns, guards and 
gates). This approach takes advantage of making plants more secure by 
design rather than security components being added on after design.
    As part of this approach, the NRC is seeking comment on the 
following issues:
    8. In developing the requirements for this alternative regulatory 
framework, how should safety, security, and emergency preparedness be 
integrated? Does the overall approach described in the technology-
neutral framework clearly express the appropriate integration of 
safety, security, and preparedness? If not, how could it better do so?
    9. What specific principles, concepts, features or performance 
standards for security would best achieve an integrated safety and 
security approach? How should they be expressed? How should they be 
measured?
    10. The NRC is considering rulemaking to require that safety and 
security be integrated so as to allow an easier and more thorough 
understanding of the effects that changes in one area would have on the 
other and to ensure that changes with unacceptable impacts are not 
implemented. How can the safety-security interface be better integrated 
in design and operational requirements?
    11. Should security requirements be risk-informed? Why or why not? 
If so, what specific security requirements or analysis types would most 
benefit from the use of Probabilistic Risk Assessment (PRA) and how?
    12. Should emergency preparedness requirements be risk-informed? 
Why or why not? How should emergency preparedness requirements be 
modified to be better integrated with safety and security?

C. Level of Safety

    The staff, in SECY-05-0130 (ADAMS Accession Number ML051670388), 
proposed options for establishing a regulatory standard that would be 
applied during licensing to enhance safety for new plants consistent 
with the Commission's policy statement for Regulation of Advanced 
Nuclear Power Plants. Four options were evaluated which included: (1) 
Perform a case-by-case review, (2) use the Quantitative Health 
Objectives (QHOs) in the Commission's policy statement on ``Safety 
Goals for the Operation of Nuclear Power Plants'' (ADAMS Accession 
Number ML051580401), (3) develop other risk objectives for the 
acceptable level of safety, and (4) develop new QHOs. The NRC is 
soliciting stakeholder views on these options.
    Subsidiary risk objectives could also be developed to implement the 
Commission's expectation regarding enhanced safety for new plants. Such 
subsidiary risk objectives could be a useful way to:
     Focus more on plant design,
     Provide quantitative criteria for accident prevention and 
mitigation, and
     Provide high level goals to assist in establishing plant 
system and equipment reliability and availability targets.
    Currently, subsidiary risk objectives of 10-5/plant year 
and 10-6/plant year that could be applicable to all reactor 
designs are being considered for accident prevention and accident 
mitigation, respectively, where:
     Accident prevention refers to preventing major fuel 
damage, and
     Accident mitigation refers to preventing releases of 
radioactive material offsite such that no early fatalities occur (i.e., 
from acute radiation doses).
    Feedback is sought specifically on the following:
    13. Which of the options in SECY-05-0130 with respect to level of 
safety should be pursued and why? Are there alternative options? If so, 
please discuss the alternative options and their benefits.
    14. Should the staff pursue developing subsidiary risk objectives? 
Why or why not? Are there other uses of subsidiary risk objectives that 
are not specified above? If so, what are they?
    15. Are the subsidiary risk objectives specified above reasonable 
surrogates for the QHOs for all reactor designs?
    16. Should the latent fatality QHO be met by preventive measures 
alone without credit for mitigative measures, or is this too 
restrictive?
    17. Are there other subsidiary risk objectives applicable to all 
reactor designs that should be considered? What are they and what would 
be their basis?
    18. Should a mitigation goal be associated with the early fatality 
QHO or should it be set without credit for preventive measures (i.e., 
assuming major fuel damage has occurred)?
    19. Should other factors be considered in accident mitigation 
besides early fatalities, such as latent fatalities, late containment 
failure, land contamination, and property damage? If so, what should be 
the acceptance criteria and why?
    20. Would a level 3 PRA analysis (i.e., one that includes 
calculation of offsite health and economic effects) still be needed if 
subsidiary risk objectives can be developed? For a specific technology, 
can practical subsidiary risk objectives be developed without the 
insights provided by level 3 PRAs?

D. Integrated Risk

    For new plant licensing, potential applicants have indicated 
interest in locating new plants at new and existing sites. In addition, 
potential applicants have indicated interest in locating multiple (or 
modular) reactor units at new and existing sites. The NRC is evaluating 
the issue of integrated risk. The staff, in SECY-05-0130, evaluated

[[Page 26270]]

three options which included: (1) No consideration of integrated risk, 
(2) quantification of integrated risk at the site only from new 
reactors (i.e., the integrated risk would not consider existing 
reactors), and (3) quantification of integrated site risk for all 
reactors (new and existing) at that site. Another aspect of this issue 
is the level of safety associated with the integrated risk. The NRC is 
presently considering whether the integrated risk should be restricted 
to the same level that would be applied to a single reactor. If this 
approach were adopted, for an entity who proposed to add multiple 
reactors to an existing site, the integrated risk would not be allowed 
to exceed the level of safety expressed by the QHOs in the Commission's 
Safety Goal Policy Statement.
    The NRC is soliciting stakeholder views on these or other options.
    Feedback is sought specifically on the following:
    21. Which of the options in SECY-05-0130 with respect to integrated 
risk should be pursued and why? Are there alternative options? If so, 
what are they?
    22. Should the integrated risk from multiple reactors be 
considered? Why or why not?
    23. If integrated risk should be considered, should the risk meet a 
minimum threshold specified in the regulations? Why or why not?

E. ACRS Views on Level of Safety and Integrated Risk

    In a letter dated September 21, 2005, the Advisory Committee on 
Reactor Safeguards (ACRS) raised a number of questions related to new 
plant licensing. The ACRS discussed issues related to requiring 
enhanced safety and how the risk from multiple reactors at a single 
site should be accounted for. The details of the ACRS discussion are in 
the September 21, 2005 letter which is attached to this ANPR. The 
Commission, in a September 14, 2005 SRM, directed the staff to consider 
ACRS comments in developing a subsequent notation vote paper addressing 
these policy issues.
    Feedback is sought specifically on the following:
    24. Should the views raised in the ACRS letter and by various 
members of the Committee be factored into the resolution of the issues 
of level of safety and integrated risk? Why or why not?

F. Containment Functional Performance Standards

    The Commission has directed the staff to develop options for 
containment functional performance requirements and criteria which take 
into account such features as core, fuel, and cooling system design. In 
developing these options, the NRC is seeking stakeholder views on the 
following aspects:
    25. How should containment be defined and what are its safety 
functions? Are the safety functions different for different designs? If 
so, how?
    26. Should the containment functional performance standards be 
design and technology specific? Why or why not?
    27. What approach should be taken to develop technology-neutral 
containment performance standards that would be applicable to all 
reactor designs and technologies? Should containment performance be 
defined in terms of the integrated performance capability of all 
mechanistic barriers to radiological release or in terms of the 
performance capability of a means of limiting or controlling 
radiological releases separate from the fuel and reactor pressure 
boundary barriers?
    28. What plant physical security functions should be associated 
with containment and what should be the related functional performance 
standards?
    29. How should PRA information and insights be combined with 
traditional deterministic approaches and defense-in-depth in 
establishing the proposed containment functional performance 
requirements and criteria for controlling radiological releases?
    30. How should the rare events in the range 10-4 to 
10-7 per year be considered in developing the containment 
functional performance requirements and criteria? Should events less 
than 10-7 per year in frequency be considered in developing 
the containment functional performance requirements and criteria?

G. Technology-Neutral Framework

    In support of determining the requirements for these alternative 
regulations, the NRC is developing a technology-neutral framework. This 
framework provides one approach in the form of criteria and guidelines 
that could serve as the technical basis for 10 CFR Part 53 that is 
technology-neutral, risk-informed, and performance-based. A working 
draft of this framework was issued for public review and comment in 
SECY-05-0006, dated January 7, 2005 (ML043560093). The latest working 
draft of the framework document is on the Ruleforum website. An updated 
version with additional information will be placed on the Ruleforum 
website in July 2006. The framework provides the criteria and 
guidelines for the following:
     Safety, security, and emergency preparedness expectations.
     Defense-in-depth and treatment of uncertainties.
     Licensing basis events (LBEs) identification and 
selection.
     Safety classification of structures, systems, and 
components.
     PRA technical acceptability.
    The NRC is seeking stakeholder views of the following aspects:
    31. Is the overall top-down organization of the framework, as 
illustrated in Figure 2-6 a suitable approach to organize the approach 
for licensing new reactors? Does it meet the objectives and principles 
of Chapter 1? Can you describe a better way to organize a new licensing 
process?
    32. Do you agree that the framework should now be applied to a 
specific reactor design? If not, why not? Which reactor design concept 
would you recommend?
    33. The unified safety concept used in the framework is meant to 
derive regulations from the Safety Goals and other safety principles 
(e.g., defense-in-depth). Does this approach result in the proper 
integration of reactor regulations and staff processes and programs 
such that regulatory coherence is achieved? If not, why not?
    34. The framework is proposing an approach for the technical basis 
for an alternative risk-informed and performance-based 10 CFR Part 50. 
The scope of 10 CFR Part 50 includes sources of radioactive material 
from reactor and spent fuel pool operations. Similarly, the framework 
is intended to apply to this same scope. Is it clear that the framework 
is intended to apply to all of these sources? If not, how should the 
framework be revised to make this intention clear?
    The Commission believes that safety, security, and emergency 
preparedness should be integrated. The approach in the framework to 
achieve this integration is to define the safety, security, and 
preparedness expectations that are needed and to define protective 
strategies and defense-in-depth principles for each area in an 
integrated manner.
    35. What role should the following factors play in integrating 
emergency preparedness requirements (as contained in 10 CFR 50.47) in 
the overall framework for future plants:
     The range of accidents that should be considered?
     The extent of defense-in-depth?
     Operating experience?
     Federal, state, and local authority input and acceptance?
     Public acceptance?
     Security-related events?

[[Page 26271]]

    36. What should the emergency preparedness requirements for future 
plants be? Should they be technology-specific or generic regardless of 
the reactor type?
    The core of the NRC's safety philosophy has always been the concept 
of defense-in-depth, and defense-in-depth remains basic to the safety, 
security, and preparedness expectations of the technology-neutral 
framework. Defense-in-depth is the mechanism used to compensate for 
uncertainty. This includes uncertainty in the type and magnitude of 
challenges to safety, as well as in the measures taken to assure 
safety.
    37. Is the approach used in the framework for how defense-in-depth 
treats uncertainties well described and reasonable? If not, how should 
it be improved?
    38. Are the defense-in-depth principles discussed in the framework 
clearly stated? If not, how could they be better stated? Are additional 
principles needed? If so, what would they be? Are one or more of the 
stated principles unnecessary? If so, which principles are unnecessary 
and why are they unnecessary?
    39. The framework emphasizes that sufficient margins are an 
essential part of defense-in-depth measures. The framework also 
provides some quantitative margin guidance with respect to LBEs in 
Chapter 6. Should the framework provide more quantitative guidance on 
margins in general in a technology-neutral way? What would be the 
nature of this guidance?
    40. The framework stresses that all of the Protective Strategies 
must be included in the design of a new reactor but it does not discuss 
the relative emphasis placed on each strategy compared to the others. 
Are there any conditions under which any of these protective strategies 
would not be necessary? Should the framework contain guidelines as to 
the relative importance of each strategy to the whole defense-in-depth 
application?
    41. Are the protective strategies well enough defined in terms of 
the challenges they defend against? If not, why not? Are there 
challenges not protected by these five protective strategies? If so, 
what would they be?
    In the framework, risk information is used in two basic parts of 
the licensing process: (1) Identification and selection of those events 
that are used in the design to establish the licensing basis, and (2) 
the safety classification of selected systems, structures, and 
components.
    42. Is the approach to and the basis for the selection LBEs 
reasonable? If not, why not? Is the cut-off for the rare event 
frequency at 1E-7 per year acceptable? If not, why not? Should the cut-
off be extended to a lower frequency?
    43. Is the approach used to select and to safety classify 
structures, systems, and components reasonable? If not, what would be a 
better approach?
    44. Is the approach and basis to the construction of the proposed 
frequency-consequence (F-C) curve reasonable? If not, why not?
    45. Are the deterministic criteria proposed for the LBEs in the 
various frequency categories reasonable from the standpoint of assuring 
an adequate safety margin? In particular, are the deterministic dose 
criteria for the LBEs in the infrequent and rare categories reasonable? 
If not, why not?
    46. Is it reasonable to use a 95% confidence value for the 
mechanistic source term for both the PRA sequences and the sequences 
designated as LBEs to provide margin for uncertainty? If not, why not? 
Is it reasonable to use a conservative approach for dispersion to 
calculate doses? If not, why not?
    The approach proposed in the framework requires a full-scope 
``living'' PRA that would incorporate operating experience and 
performance-based requirements in the periodic re-examination of events 
designated as LBEs that were originally selected based on the design, 
and structures, systems, and components that were characterized as 
safety-significant.
    47. The approach proposed in the framework does not predefine a set 
of LBEs to be addressed in the design. The LBEs are plant specific and 
identified and selected from the risk-significant events based on the 
plant-specific PRA. Because the plant design and operation may change 
over time, the risk-significant events may change over time. The 
licensee would be required to periodically reassess the risk of the 
plant and, as a result, the LBEs may change. This reassessment could be 
performed under a process similar to the process under 10 CFR 50.59. Is 
this approach reasonable? If not, why not?
    48. The framework provides guidance for a technically acceptable 
full-scope PRA. Is the scope and level of detail reasonable? If not, 
why not? Should it be expanded and if so, in what way?
    49. Because a PRA (including the supporting analyses) will be used 
in the licensing process, should it be subject to a 10 CFR Part 50 
Appendix B approach to quality assurance? If not, why not?
    Chapter 8 describes and applies a process to identify the topics 
which the requirements must address to ensure the success of the 
protective strategies and administrative controls. This process is 
based upon:
     Developing and applying a logic diagram for each 
protective strategy to identify the pathways that can lead to failure 
of the strategy and then, through a series of questions, identify what 
needs to be done to prevent the failure;
     Applying the defense-in-depth principles from Chapter 4 to 
each protective strategy;
     Developing and applying a logic diagram to identify the 
needed administrative controls; and
     Providing guidance on how to write the requirements.
    50. Is this process clear, understandable, and adequate? If not, 
why not? What should be done differently?
    51. Is the use of logic diagrams to identify the topics that need 
to be addressed in the requirements reasonable? If not, what should be 
used?
    52. Is the list of topics identified for the requirements adequate? 
Is the list complete? If not, what should be changed (added, deleted, 
modified) and why?
    53. A completeness check was made on the topics for which 
requirements need to be developed for the new 10 CFR Part 53 
(identified in Chapter 8) by comparing them to 10 CFR Part 50, NEI 02-
02, and the International Atomic Energy Agency (IAEA) safety standards 
for design and operation. Are there other completeness checks that 
should be made? If so, what should they be?
    54. The results of the completeness check comparison are provided 
in Appendix G. The comparison identified a number of areas that are not 
addressed by the topics but that are covered in the IAEA standards. 
Should these areas be included in the framework? If so, why should they 
be included? If not, why not?

H. Defense-in-Depth

    In SECY-03-0047 (ML030160002), the staff recommended that the 
Commission approve the development of a policy statement or description 
(e.g., white paper) on defense-in-depth for nuclear power plants to 
describe: The objectives of defense-in-depth (philosophy); the scope of 
defense-in-depth (design, operation, etc.); and the elements of 
defense-in-depth (high level principles and guidelines). The policy 
statement or description would be technology-neutral and risk-informed 
and would be useful in providing consistency in other regulatory 
programs (e.g., Regulatory Analysis Guidelines). In the SRM on SECY-03-
0047, the Commission directed the staff to consider whether it can 
accomplish

[[Page 26272]]

the same goals in a more efficient and effective manner by updating the 
Commission Policy Statement on Use of Probabilistic Risk Assessment 
Methods in Nuclear Regulatory Activities to include a more explicit 
discussion of defense-in-depth, risk-informed regulation, and 
performance-based regulation. The NRC is interested in stakeholder 
comment on a policy statement on defense-in-depth.
    55. Would development of a better description of defense-in-depth 
be of any benefit to current operating plants, near-term designs, or 
future designs? Why or why not? If so, please discuss any specific 
benefits.
    56. If the NRC undertakes developing a better description of 
defense-in-depth, would it be more effective and efficient to 
incorporate it into the Commission's Policy Statement on PRA or should 
it be provided in a separate policy statement? Why?
    57. RG 1.174 assumes that adequate defense-in-depth exists and 
provides guidance for ensuring it is not significantly degraded by a 
change to the licensing basis. Should RG 1.174 be revised to include a 
better description of defense-in-depth? Why or why not? If so, would a 
change to RG 1.174 be sufficient instead of a policy statement? Why or 
why not?
    58. How should defense-in-depth be addressed for new plants?
    59. Should development of a better description of defense-in-depth 
(whether as a new policy statement, a revision to the PRA policy 
statement, or as an update to RG 1.174) be completed on the same 
schedule as 10 CFR Part 53? Why or why not?

I. Single Failure Criterion

    In SECY-05-0138 (ML051950619), the staff forwarded to the 
Commission a draft report entitled ``Technical Report to Support 
Evaluation of a Broader Change to the Single Failure Criterion'' and 
recommended to the Commission that any followup activities to risk-
inform the Single Failure Criterion (SFC) should be included in the 
activities to risk-inform the requirements of 10 CFR Part 50. The 
Commission directed the staff to seek additional stakeholder 
involvement. The report provides the following options: (1) Maintain 
the SFC as is, (2) risk-inform the SFC for design bases analyses, (3) 
risk-inform SFC based on safety significance, and (4) replace SFC with 
risk and safety function reliability guidelines. The NRC is soliciting 
stakeholder feedback with regard to the proposed alternatives.
    60. Are the proposed options reasonable? If not, why not?
    61. Are there other options for risk-informing the SFC? If so, 
please discuss these options.
    62. Which option, if any, should be considered?
    63. Should changes to the SFC in 10 CFR Part 50 be pursued separate 
from or as a part of the effort to create a new 10 CFR Part 53? Why or 
why not?

J. Continue Individual Rulemakings to Risk-Inform 10 CFR Part 50

    The NRC has for some time been revising certain provisions of 10 
CFR Part 50 to make them more risk-informed and performance-based. 
Examples are: (1) A revision to 10 CFR 50.65, ``Requirements for 
Monitoring the Effectiveness of Maintenance at Nuclear Power Plants;'' 
(2) a revision of 10 CFR 50.48 to allow licensees to voluntarily adopt 
National Fire Protection Association (NFPA) Standard 805, 
``Performance-Based Standard for Fire Protection for Light Water 
Reactor Electric Generating Plants, 2001 Edition,'' (NFPA 805); and (3) 
issuance of 10 CFR 50.69, ``Risk-Informed Categorization and Treatment 
of Structures, Systems, and Components for Nuclear Power Reactors,'' as 
a voluntary alternative set of requirements. These actions have been 
effective but required extensive NRC and industry efforts to develop 
and implement.
    The NRC plans to continue the current risk-informed rulemaking 
actions, e.g., 10 CFR 50.61 on pressurized thermal shock and 10 CFR 
50.46 on redefinition of the emergency core cooling system break size, 
that are ongoing, and would undertake new risk-informed rulemaking only 
on an as-needed basis.
    The NRC is seeking comment on the following issues:
    64. Should the NRC continue with the ongoing current rulemaking 
efforts and not undertake any effort to risk-inform other regulations 
in 10 CFR Part 50, or should the NRC undertake new risk-informed 
rulemaking on a case-by-case priority basis? Why?
    65. If the NRC were to undertake new risk-informed rulemakings, 
which regulations would be the most beneficial to revise? What would be 
the anticipated safety benefits?
    66. In addition to revising specific regulations, are there any 
particular regulations that do not need to be revised, but whose 
associated regulatory guidance documents, could be revised to be more 
risk-informed and performance-based? What are the safety benefits 
associated with revising these guides? Which ones in particular are 
stakeholders interested in having revised and why?
    67. If additional regulations and/or associated regulatory guidance 
documents were to be revised, when should the NRC initiate these 
efforts, e.g., immediately or after having started implementation of 
current risk-informed 10 CFR Part 50 regulations?
    At the end of the ANPR phase, the NRC will assess whether to adjust 
its approach to risk-inform the requirements for nuclear power reactors 
including existing and new plants.

List of Subjects in 10 CFR Part 50

    Classified information, Criminal penalties, Fire protection, 
Intergovernmental relations, Nuclear power plants and reactors, 
Radiation protection, Reactor siting criteria, Reporting and 
recordkeeping requirements.

    The authority citation for this document is 42 U.S.C. 2201.

    Dated at Rockville, Maryland, this 28th day of April, 2006.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.

Attachment--Letter From G. B. Wallis, Chairman ACRS, dated September 
21, 2005, ``Report on Two Policy Issues Related to New Plant 
Licensing,'' ADAMS Accession Number ML052640580

[ACRSR-2149]
September 21, 2005.
The Honorable Nils J. Diaz, Chairman, U.S. Nuclear Regulatory 
Commission, Washington, DC.

Subject: Report on Two Policy Issues Related to New Plant Licensing

Dear Chairman Diaz: During the 523rd meeting of the Advisory 
Committee on Reactor Safeguards, June 1-3, 2005, we met with the NRC 
staff and discussed two policy issues related to new plant 
licensing. We also discussed this matter during our 524th, July 6-8, 
2005, and 525th, September 8-10, 2005 meetings. We had the benefit 
of the documents referenced.
    These policy issues were:
     What shall be the minimum level of safety that new 
plants need to meet to achieve enhanced safety?
     How shall the risk from multiple reactors at a single 
site be accounted for?
    In SECY-05-0130, the staff recommends that the expectation for 
enhanced safety be met by requiring that new plants meet the 
Quantitative Health Objectives (QHOs), i.e., by applying the QHOs to 
individual plants. The staff maintains that this would represent an 
enhancement in safety over current plants, which are now required to 
meet adequate protection, but may not meet the QHOs. The staff 
argues that this position is consistent with the Commission's Policy 
Statement on Regulation of Advanced Nuclear Power Plants.

[[Page 26273]]

    The staff proposes to address the risk of multiple reactors at a 
single site by requiring that the integrated risk associated with 
only new reactors (i.e., modular or multiple reactors) at a site not 
exceed the risk expressed by the QHOs. The risk from existing 
plants, which may already exceed the QHOs, is not considered.
    We discussed these issues and concluded that use of the existing 
QHOs is not sufficient to resolve either of these issues. In 
considering the overall scope of the issues raised by the staff, we 
found it more apt and effective to reframe the two issues into the 
following questions:
    1. What are the appropriate measures of safety to use in the 
consideration of the certification of a new reactor design?
    2. Should quantitative criteria for these measures be imposed to 
define the minimum level of safety?
    3. How should these measures be applied to modular designs?
    4. How should risk from multiple reactors at a site be combined 
for evaluation by suitable criteria?
    5. How should the combination of new and old reactors at a site 
be evaluated by these criteria?
    6. What should these criteria be?
    7. How should compliance with these criteria be demonstrated?

Discussion

Question 1. What are the appropriate measures of safety to use in 
the consideration of the certification of a new reactor design?

    The QHOs are criteria for the risk at a site and thus involve 
not only the design and operation of the reactor(s), but also the 
site characteristics, the number and power level of plants on the 
site, meteorological conditions, population distribution, and 
emergency planning measures. By themselves, the QHOs do not express 
the defense-in-depth philosophy that the Commission seeks to limit 
not only the risk from accidents, but also the frequency of 
accidents.
    Although core damage frequency (CDF) and large, early release 
frequency (LERF) have been viewed by the NRC as light water reactor 
(LWR)-specific surrogates for the QHOs, they have come to be 
accepted as metrics to gauge the acceptable level of safety of 
certified designs and the acceptability of proposed changes in the 
licensing basis. They are measures of reactor design safety that 
incorporate a defense-in-depth balance between prevention and 
mitigation. Currently used values of these metrics have been derived 
from the QHOs. If they were no longer to be viewed as surrogates, 
acceptance values for these metrics could be independently specified 
and need not be derived from the QHOs. Thus, they would be 
fundamental characteristics of reactor design independent of siting 
and emergency planning requirements.
    If these measures are no longer viewed as surrogates for the 
QHOs, the appropriate measure of a large release need not be 
restricted to ``early'' but could be a ``large release frequency'' 
(LRF) which would apply to the summation of all large release 
frequencies regardless of the time of occurrence. The LRF would thus 
have broader applicability to designs in which the release is likely 
to occur over an extended period.
    A majority of the Committee members favors the use of CDF and 
LRF as fundamental measures of the enhanced safety of new reactor 
designs and not simply as surrogates for the QHOs.
    In SECY-05-0130, the staff argues that it will be difficult to 
derive such measures for different technologies, although the staff 
proposes to include them as subsidiary goals in their technology-
neutral framework document. Although the processes and mechanisms 
for failure and release will differ greatly for different reactor 
technologies, technology-neutral definitions in terms of a release 
from the fuel (the accident prevention/CDF goal) and from the 
containment/confinement (the large release goal) seem feasible to 
us. For example, the CDF of a Pebble Bed Modular Reactor (PBMR), 
would be an indicator of the success criteria for the design 
measures intended to prevent release from the fuel of that module. 
It could be defined in terms of the frequency of exceeding a fuel 
temperature of 1600 [deg]C.

Question 2. Should quantitative criteria for these measures be 
imposed to define the minimum level of safety?

    In the current Policy Statement on the Regulation of Advanced 
Nuclear Power Plants, the Commission decided not to set numerical 
criteria for enhanced safety but rather focused on aspects which 
might make designs more robust. In addition, the Safety Goal Policy 
Statement was intended to provide a definition of ``how safe is safe 
enough.'' If a plant would meet the QHOs at a proposed site, then 
the additional risk it imposes is already very low compared to other 
risk in society. It now seems possible to build economically 
competitive reactors with risks at most sites that would be much 
lower than implied by the QHOs. The Electric Power Research 
Institute (EPRI) and European Utility Requirements Documents specify 
CDF and LERF values that would provide large margins to the QHOs for 
virtually all sites. An explicit commitment to lower values of CDF 
and LRF would be responsive to the Commission's desire for enhanced 
safety and may have significant impact on public perceptions and 
confidence.
    We considered the following alternatives, identifying arguments 
in favor of each. Since such a decision has broad practical 
implementation and policy implications, we recommend that the staff 
further explore the consequences of these (and possibly other) 
choices as a basis for an eventual Commission decision.
    a. Set maximum values for CDF and LRF at 10-5/yr and 
10-6/yr for new reactor designs. This would make more 
explicit the Commission's stated expectation that future reactors 
provide enhanced safety. This could also provide a basis for 
establishing multinational design approval (as these would now be 
independent of U.S. QHOs). The suggested values are consistent with 
those in the EPRI and the European Utility Requirements Documents, 
the EPR Safety Document, and those used in the certification of 
advanced reactors (the ABWR, AP600 and CE-System 80+). These values 
are also consistent with the generic values for an accident 
prevention frequency and a LRF in the staff's draft technology-
neutral framework document.
    b. Leave the values unspecified. CDF and LRF would be considered 
along with other aspects of the design, such as defense-in-depth and 
passive safety features, in reaching a decision about design 
certification. This would give the staff more flexibility to respond 
to technology-specific features.
    On a preliminary basis, the majority of the Committee members 
favor Alternative (a), but is not ready to make a recommendation 
until more is understood about the likely consequences and policy 
implications of the decision.

Question 3. How should these measures be applied to modular 
designs?

    The staff's considerations of integrated risk do not distinguish 
between criteria for modular reactor designs and criteria for the 
risk due to multiple plants on a site. Thus, the staff treats CDF 
and LRF (or LERF) for modular designs and/or multiple plants on a 
site as still being QHO risk surrogates. In our view, the CDF and 
LRF metrics are design criteria that are to be ``imposed'' at the 
plant design certification stage independent of any site 
considerations.
    New reactors could include PBMR, AP600, AP1000, Economic and 
Simplified Boiling Water Reactor (ESBWR), and EPR, and the number of 
new reactors at a site could vary by an order of magnitude.
    Some Committee members believe that to get consistency in 
expectations of enhanced safety in all cases, the integrated risk 
from all new reactors on a site is the appropriate measure. This is 
true both for the risk metric LRF and the defense-in-depth accident 
prevention metric CDF. Thus, for the PBMR, which is proposed in 
terms of an eight-module package, the CDF and LRF goals (e.g., 
10-5/ry and 10-6/ry) would be applied to the 
package. In effect each module would have to have a somewhat lower 
CDF and LRF. Because of the potential for interactions, analysis of 
individual modules may not be meaningful and the analysis should 
focus on the ``eight pack.''
    Other Committee members prefer CDF and LRF design specifications 
that are independent of the number of modules. These members believe 
the specified acceptable CDF for enhanced safety (e.g. 
10-5/yr) should be applied to each module at the design 
stage and would be an indicator of the success criteria for the 
design measures provided for each module intended to prevent release 
from the fuel of that module. Similarly, LRF would be on a modular 
basis. As it may be possible to restrict the total power of a given 
module to a level that the quantity of fission products releasable 
cannot exceed the acceptance LRF value (e.g. 10-6/yr), a 
modular design implicitly represents a kind of defense-in-depth 
(given appropriate consideration of common-mode failures and module 
interactions).

[[Page 26274]]

Question 4. How should risk from multiple reactors at a site be 
combined for evaluation by suitable criteria?

    The QHOs address the risk to individuals that live in the 
vicinity of a site. Logically, the risk to these individuals should 
be determined by integrating the risk from all the units at the 
site. The manner by which the risks of different units at a site are 
to be integrated must address the treatment of modular designs, 
units with differing power levels, and accidents involving multiple 
units.

Question 5. How should the combination of new and old reactors at a 
site be evaluated by these criteria?

    Any new plant that meets the independent safety criteria 
discussed in Questions 1 through 3 would be expected to add 
substantially less risk to an existing site than that already 
provided by existing plants on the site. If a proposed site already 
exceeds the QHOs, it should not be approved for new plants. For 
existing sites not being proposed for the addition of new plants, 
there would be no need to assess their risk status because they 
provide adequate protection. These sites would, thus, be 
grandfathered in the new framework.

Question 6. What should these criteria be?

    Use of the QHOs for evaluating the site suitability for new 
reactors is attractive because the QHOs represent a fundamental 
statement about risk independent of any particular technology. The 
current QHOs (prompt and latent fatalities), however, only address 
individual risk and do not directly address societal risks such as 
total deaths, injuries, non-fatal cancers, and land contamination. 
These societal impacts are addressed somewhat in the current 
regulations by the siting criteria on population.
    Some ACRS members believe that measures of societal risk need to 
be an explicit part of any new technology-neutral framework. The 
staff argues in the technology-neutral framework document that the 
limits proposed there for CDF and LRF limit societal risks such as 
land contamination and dose to the total population. However, these 
members recognize that CDF and LRF are not equivalent to risk and 
disagree with the staff's position.
    Other ACRS members believe that the current siting criteria have 
served to limit societal risks. In addition, societal risks are 
considered in the environmental impact assessments of license 
renewal. The estimates presented in NUREG-1437 Vol. 1 indicate that 
the risk of early and latent fatalities from current nuclear power 
plants is small. The predicted early and latent fatalities from all 
plants (that is, the risk to the population of the United States 
from all nuclear power plants) is approximately one additional early 
fatality per year and approximately 90 additional latent fatalities 
per year, which is a small fraction of the approximately 100,000 
accidental and 500,000 cancer fatalities per year from other 
sources. The evaluation of Severe Accident Mitigation Alternatives 
(SAMAs) as part of the license renewal process also considers 
societal risk measures and monetizes them to perform cost benefit 
studies. Based on current NRC regulatory analysis guidance, very few 
of these SAMAs appear cost beneficial.
    Environmental impact statements (EISs) also assess the societal 
costs of probabilistic accidents at the current sites. The results, 
although very approximate, indicate that the societal costs at many 
current reactor sites would likely exceed a reasonable societal cost 
risk acceptance criterion. For example, these would exceed the cost 
associated with 0.1% of the above noted 100,000 early fatalities due 
to all accidents.
    Thus, the inclusion of a quantitative societal risk acceptance 
measure appears important and could add to greater public confidence 
and understanding of the risks of nuclear power. It may be 
worthwhile for the staff to consider supplementing the current QHOs 
with additional risk acceptance measures that relate directly to 
societal risks.

Question 7. How should compliance with these criteria be 
demonstrated?

    The establishment of goals or criteria of various kinds cannot 
be divorced from the ability to demonstrate compliance. Considerable 
improvement in PRA practice will be needed to provide confidence 
that the goals on CDF and LRF for future plants will be met in a 
meaningful way. Operating experience has been crucial for the 
analysts to appreciate the significance of potential errors/faults. 
For example, before TMI, it was assumed that operators would not 
have problems diagnosing what is going on under certain conditions.
    Some of the challenges that new plants will create for PRA 
analysts are:
    i. Operating experience on component failure rate distributions 
and frequencies developed for light-water reactors has limited 
applicability to other reactor types.
    ii. Some designs are considering components, e.g., microturbines 
and fuel cells, for which reliability data are nearly non-existent.
    iii. Digital Instrumentation and Control systems are expected to 
be an integral part of future reactor designs. The risk consequences 
of such practice are difficult to quantify at this time.
    Thus, in addition to the imposition of design goals for low CDF 
and LRF, it will be important to maintain sufficient defense-in-
depth in the technology-neutral framework.
    We look forward to additional discussion with the staff on these 
issues.

    Sincerely,

    Graham B. Wallis, Chairman.

Additional Comments From ACRS Members Dana A. Powers and John D. Sieber

    We disagree with our colleagues on the matter of this letter. 
The Commission has indicated a laudable expectation that future 
reactors will be safer than current reactors. The question that our 
colleagues should have addressed first is whether a quantitative 
metric is needed to substantiate this expectation. It is by no means 
obvious that such a metric is essential. We can well imagine future 
plants designed in conjunction with far more comprehensive 
probabilistic safety analyses that realistically address all known 
accident hazards during all modes of operation to a depth far 
greater than is attempted now for elements of the fleet of operating 
reactors. Our experience has been that whenever improvements are 
made in quantitative risk analysis methods, unforeseen, hazardous, 
plant configurations, systems interactions and operations become 
apparent. Hidden, these configurations, interactions and operations 
may arise unexpectedly with undesirable consequences. Revealed, they 
can be avoided often with modest efforts. This is exploitation of 
the full potential of quantitative risk analysis to achieve greater 
safety in nuclear power plants. It contrasts with the more effete 
pursuit of the ``bottomline'' results of PRA to compare with 
arbitrarily proliferated safety metrics.
    Our objective should be to foster the voluntary development of 
quantitative risk analysis methods both in scope and depth in order 
to improve the safety of nuclear power plants. Fostering voluntary 
development of methods by nuclear community is especially important 
now when methods developments have stagnated at NRC relative to the 
situation a decade ago.
    Our colleagues seem to presume it essential that future reactors 
meet the Quantitative Health Objectives (QHOs). These QHOs define a 
very stringent safety level that has always been viewed as an 
``aiming point'' or a benchmark and not as some minimum standard 
that cannot be exceeded. Indeed, the definition of the QHOs was 
undertaken to define ``how safe is safe enough'' so that no 
additional regulatory requirements for greater safety would be 
needed. Requiring such a stringent standard as the QHOs as a minimum 
level of safety for advanced reactors appears to go well beyond the 
authority granted by the Atomic Energy Act that requires adequate 
protection of the public health and safety. We are unaware that the 
Commission has made such a demand for advanced reactors. Were the 
Commission to make such a demand, we would question the wisdom of 
doing so. By demanding such a stringent level of safety, our 
colleagues appear to be willing to forego great strides in safety 
that can be achieved with advanced plants if these plants fail to 
live up to what can only be viewed as an extreme safety standard.
    The demands our colleagues appear to make on the safety of 
advanced reactors lack a critical dimension of practicality since we 
do not believe the technology now exists to do the calculations 
needed to compare a plant's safety profile to the QHOs. By the very 
definitions of the QHOs, such calculations would entail analyses of 
modes of operation only very crudely addressed today by most (fire 
risk, shutdown risk and natural phenomena risk) and the conduct of 
uncertainty analyses dealing with both parameters and models that to 
our knowledge have been done by no one.
    Because of the limitations of risk assessment technology 
available today for the evaluation of the current fleet of nuclear 
power plants, surrogate metrics such as core damage frequency (CDF) 
and large early

[[Page 26275]]

release frequency (LERF) have been introduced and widely used. Our 
colleagues seem to believe that there are known critical values of 
these surrogate metrics that mark the point at which a plant meets 
the QHOs. We know of no defensible analysis that establishes such 
critical values of these surrogate metrics. We are, of course, quite 
aware of very limited analyses considering only risk during normal 
operations that purport to show existing reactors meet the QHOs. 
Such limited analyses are simply not pertinent. They do not meet the 
exacting standards required by the definitions of the QHOs. Should 
defensible analyses ever be done, we are sure that they will show 
the critical values of the surrogate metrics are technology 
dependent. Indeed, more defensible analyses will show in all 
likelihood that better surrogate measures can be defined for 
advanced reactor technologies.
    Our colleagues are sufficiently enamored with the existing 
surrogate metrics that they recommend these surrogates be enshrined 
on a level equivalent to QHOs. More remarkable, our colleagues want 
to establish critical values of the metrics that are a factor of ten 
less than the values they assert mark a plant meeting the rather 
stringent level of safety defined by the QHOs. They do this, 
apparently, for no other reason than the fact that clever engineers 
can design plants meeting these smaller values at least for a 
limited number of operational states. While we are willing to 
congratulate the engineers on their designs, we can see no reason 
why such stringent safety requirements should be made regulatory 
requirements to be imposed on the designers' efforts. Again, we 
worry that doing so may create unnecessary burdens that cause our 
society to sacrifice for practical reasons great improvements in 
power reactor safety simply because these improvements fall short of 
our colleagues unreasonably high safety expectations.
    Though surrogate metrics have been useful, it is important to 
remember that they are only expedients. The full promise of risk-
informed safety assessment will not be realized until it is possible 
to do routinely risk assessments of sufficient scope and depth so it 
is possible to dispense with surrogate metrics. Enshrining these 
surrogates along with the QHOs will only delay efforts to reach this 
preferred status.
    The potential of our colleagues recommendations have to stifle 
new technology and forego improved safety reaches a crisis when they 
speak to the location of modern, safer plants on sites with older 
but still adequately safe plants. Our colleagues have no tolerance 
for a single older plant if a newer, safer plant is to be collocated 
on the site. They are willing to tolerate any number of similarly 
old plants on a site if a new, safer plant is not added to this 
site. We find this remarkable. Our colleagues' recommendations give 
no credit for experience with a site. They fail to recognize the 
finite life of older plants even when licenses have been renewed. We 
fear that our colleagues have failed to assess the integral safety