[Federal Register Volume 71, Number 86 (Thursday, May 4, 2006)]
[Proposed Rules]
[Pages 26267-26275]
From the Federal Register Online via the Government Publishing 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).
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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, (http://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: [email protected]. 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 http://ruleforum.llnl.gov. Address questions
about our rulemaking Web site to Carol Gallagher (301) 415-5905; e-mail
[email protected]. Comments can also be submitted via the Federal eRulemaking
Portal http://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
http://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 http://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 [email protected].
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:
[email protected]; or Mary Drouin, Office of Nuclear Regulatory Research
(RES), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001;
telephone: (301) 415-6675, e-mail: [email protected].
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
consequences of their stringent demands on this matter. A very great
concern is that our colleagues pursuit of ideals in risk avoidance
may well arrest the current, healthy quest for improved safety among
those exploring advanced reactor designs.
References
1. U.S. Nuclear Regulatory Commission, SECY-05-130,'' Policy
Issues Related to New Plant Licensing and Status of the Technology
Neutral Framework for New Plant Licensing,'' dated July 21, 2005.
2. U.S. Nuclear Regulatory Commission, ``Safety Goals for the
Operations of Nuclear Power Plants, Policy Statement,'' Federal
Register, Vol. 51, (51 FR 30028), August 4, 1986.
3. U.S. Nuclear Regulatory Commission, ``Commission's Policy
Statement on the Regulation of Advanced Nuclear Power Plants,'' 59
FR 35461, July 12, 1994.
4. U.S. Nuclear Regulatory Commission, NUREG-1437, Volume 1,
``Generic Environmental Impact Statement for License Renewal of
Nuclear Plants,'' May 1996.
[FR Doc. E6-6745 Filed 5-3-06; 8:45 am]
BILLING CODE 7590-01-P