[Federal Register Volume 69, Number 224 (Monday, November 22, 2004)]
[Rules and Regulations]
[Pages 68008-68048]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-25665]
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Part II
Nuclear Regulatory Commission
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10 CFR Part 50
Risk-Informed Categorization and Treatment of Structures, Systems and
Components for Nuclear Power Reactors; Final Rule
Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 /
Rules and Regulations
[[Page 68008]]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AG42
Risk-Informed Categorization and Treatment of Structures, Systems
and Components for Nuclear Power Reactors
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to provide an alternative approach for establishing the
requirements for treatment of structures, systems and components (SSCs)
for nuclear power reactors using a risk-informed method of categorizing
SSCs according to their safety significance. The amendment revises
requirements with respect to ``special treatment,'' that is, those
requirements that provide increased assurance (beyond normal industrial
practices) that SSCs perform their design basis functions. This
amendment permits licensees (and applicants for licenses) to remove
SSCs of low safety significance from the scope of certain identified
special treatment requirements and revise requirements for SSCs of
greater safety significance. In addition to the rulemaking and its
associated analyses, the Commission is also issuing a regulatory guide
(RG) to implement the rule.
EFFECTIVE DATE: December 22, 2004.
ADDRESSES: The final rule and related documents are available on NRC's
rulemaking Web site at http://ruleforum.llnl.gov. For information about
the interactive rulemaking Web site contact Ms. Carol Gallagher, (301)
415-5905 (e-mail: [email protected]).
FOR FURTHER INFORMATION CONTACT: Mr. Timothy Reed, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone (301) 415-1462; e-mail: [email protected].
SUPPLEMENTARY INFORMATION:
Table of Contents
I. Background
II. Comments on Proposed Rule
III. Final Rule
IV. Pilot Activities
V. Section by Section Analysis
VI. Guidance
VII. Criminal Penalties
VIII. Compatibility of Agreement State Regulations
IX. Availability of Documents
X. Voluntary Consensus Standards (Public Law 104-113)
XI. Finding of No Significant Environmental Impact
XII. Paperwork Reduction Act Statement
XIII. Regulatory Analysis
XIV. Regulatory Flexibility Act Certification
XV. Backfit Analysis
XVI. Small Business Regulatory Enforcement Fairness Act
I. Background
I.1 History and General Background
The NRC has established a set of regulatory requirements for
commercial nuclear reactors to ensure that a reactor facility does not
impose an undue risk to the health and safety of the public, thereby
providing reasonable assurance of adequate protection to public health
and safety. The current body of NRC regulations and their
implementation are largely based on a ``deterministic'' approach.
This deterministic approach establishes requirements for
engineering margin and quality assurance in design, manufacture, and
construction. In addition, it assumes that adverse conditions can exist
(e.g., equipment failures and human errors) and establishes a specific
set of design basis events (DBEs). The deterministic approach contains
implied elements of probability (qualitative risk considerations), from
the selection of accidents to be analyzed (e.g., reactor vessel rupture
is considered too improbable to be included) to the system level
requirements for emergency core cooling (e.g., safety train redundancy
and protection against single failure). The deterministic approach then
requires that the licensed facility include safety systems capable of
preventing and/or mitigating the consequences of those DBEs to protect
public health and safety. Those SSCs necessary to defend against the
DBEs are defined as ``safety-related,'' and these SSCs are the subject
of many regulatory requirements designed to ensure that they are of
high quality and high reliability, and have the capability to perform
during postulated design basis conditions. Typically, the regulations
establish the scope of SSCs that receive special treatment using one of
three different terms: ``safety-related,'' ``important to safety,'' or
``basic component.'' The terms ``safety-related `` and ``basic
component'' are defined in the regulations, while ``important to
safety,'' used principally in the general design criteria (GDC) of
Appendix A to 10 CFR part 50, is not explicitly defined.
These prescriptive requirements as to how licensees are to treat
SSCs, especially those that are defined as ``safety-related,'' are
referred to in the rulemaking as ``special treatment requirements.''
These requirements were developed to provide greater assurance that
these SSCs would perform their functions under particular conditions
(e.g., seismic events or harsh environments), with high quality and
reliability, for as long as they are part of the plant. These include
particular examination techniques, testing strategies, documentation
requirements, personnel qualification requirements, independent
oversight, etc. In many instances, these ``special treatment''
requirements were developed as a means to gain assurance when more
direct measures (e.g., testing under design basis conditions or routine
operation) could not show that SSCs were functionally capable.
Special treatment requirements are imposed on nuclear reactor
applicants and licensees through numerous regulations that have been
issued since the 1960's. These requirements specify different scopes of
equipment for different special treatment requirements depending on the
specific regulatory concern, but are derived from consideration of the
deterministic DBEs.
Treatment for an SSC, as a general term and as it will be used in
this rulemaking, refers to activities, processes, and/or controls that
are performed or used in the design, installation, maintenance, and
operation of SSCs as a means of:
(1) Specifying and procuring SSCs that satisfy performance
requirements;
(2) Verifying over time that performance is maintained;
(3) Controlling activities that could impact performance; and
(4) Providing assessment and feedback of results to adjust
activities as needed to meet desired outcomes.
Treatment includes, but is not limited to, quality assurance,
testing, inspection, condition monitoring, assessment, evaluation, and
resolution of deviations. The distinction between ``treatment'' and
``special treatment'' is the degree of NRC specification as to what
must be implemented for particular SSCs or for particular conditions.
Defense-in-depth is an element of the NRC's safety philosophy that
employs successive measures to prevent accidents or mitigate damage if
a malfunction, accident, or naturally caused event occurs at a nuclear
facility. Defense-in-depth is a philosophy used by the NRC to provide
redundancy as well as the philosophy of a multiple-barrier approach
against fission product releases. The defense-in-depth philosophy
ensures that safety will not be wholly dependent on any single element
of the design, construction,
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maintenance, or operation of a nuclear facility. The net effect of
incorporating defense-in-depth into design, construction, maintenance,
and operation is that the facility or system in question tends to be
more tolerant of failures and external challenges.
A probabilistic approach to regulation enhances and extends the
traditional deterministic approach by allowing consideration of a
broader set of potential challenges to safety, providing a logical
means for prioritizing these challenges based on safety significance,
and allowing consideration of a broader set of resources to defend
against these challenges. Until the accident at Three Mile Island
(TMI), the NRC only used probabilistic criteria in specialized areas,
such as for certain man-made hazards and for natural hazards (with
respect to initiating event frequency). The major investigations of the
TMI accident recommended that probabilistic risk assessment (PRA)
techniques be used more widely to augment traditional non-probabilistic
methods of analyzing plant safety.
In contrast to the deterministic approach, PRAs address credible
initiating events by assessing the event frequency. Mitigating system
reliability is then assessed, including the potential for common cause
failures. The probabilistic treatment goes beyond the single failure
requirements used in the deterministic approach. The probabilistic
approach to regulation is therefore considered an extension and
enhancement of traditional regulation by considering risk in a more
coherent and complete manner.
The primary need for improving the implementation of defense-in-
depth in a risk-informed regulatory system is guidance to determine how
many measures are appropriate and how good these should be. Instead of
merely relying on bottom-line risk estimates, defense-in-depth is
invoked as a strategy to ensure public safety given there exists both
unquantified and unquantifiable uncertainty in engineering analyses
(both deterministic and risk assessments).
Risk insights can make the elements of defense-in-depth clearer by
quantifying them to the extent practicable. Although the uncertainties
associated with the importance of some elements of defense may be
substantial, the fact that these elements and uncertainties have been
quantified can aid in determining how much defense is appropriate from
a regulatory perspective. Decisions on the adequacy of, or the
necessity for, elements of defense should reflect risk insights gained
through identification of the individual performance of each defense
system in relation to overall performance.
The Commission published a Policy Statement on the ``Use of
Probabilistic Risk Assessment'' on August 16, 1995 (60 FR 42622). In
the policy statement, the Commission stated that the use of PRA
technology should be increased in all regulatory matters to the extent
supported by the state of the art in PRA methods and data, and in a
manner that supports the NRC's traditional defense-in-depth philosophy.
The policy statement also stated that, in making regulatory judgments,
the Commission's safety goals for nuclear power reactors and subsidiary
numerical objectives (on core damage frequency and containment
performance) should be used with appropriate consideration of
uncertainties.
To implement this Commission policy, the NRC staff developed
guidance on the use of risk information for reactor license amendments
and issued Regulatory Guide (RG) 1.174, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis.'' This RG provided guidance on
an acceptable approach to risk-informed decision-making consistent with
the Commission's policy, including a set of key principles. These
principles include:
(1) Be consistent with the defense-in-depth philosophy;
(2) Maintain sufficient safety margins;
(3) Any changes allowed must result in only a small increase in
core damage frequency or risk, consistent with the intent of the
Commission's Safety Goal Policy Statement; and,
(4) Incorporate monitoring and performance measurement strategies.
RG 1.174 states that consistency with the defense-in-depth
philosophy will be preserved by ensuring that:
(1) A reasonable balance is preserved among prevention of
accidents, prevention of barrier failure, and mitigation of
consequences;
(2) An over-reliance on programmatic activities to compensate for
weaknesses in equipment or device design is avoided;
(3) System redundancy, independence, and diversity are preserved
commensurate with the expected frequency, consequences of challenges to
the system, and uncertainties (e.g., no risk outliers);
(4) Defenses against potential common cause failures are preserved,
and the potential for the introduction of new common cause failure
mechanisms is assessed;
(5) The independence of barriers is not degraded; and,
(6) Defenses against human errors are preserved.
I.2 Rule Initiation
In addition to RG 1.174, the NRC also issued other regulatory
guides on risk-informed approaches for specific types of applications.
These included RG 1.175, Risk-informed Inservice Testing, RG 1.176,
Graded Quality Assurance, RG 1.177, Risk-informed Technical
Specifications, and RG 1.178, Risk-informed Inservice Inspection. In
this respect, the Commission has been successful in developing and
implementing a regulatory means for considering risk insights into the
current regulatory framework. One such risk-informed application, the
South Texas Project (STP) submittal on graded quality assurance, is
particularly noteworthy.
In March 1996, STP Nuclear Operating Company (STPNOC) requested
that the NRC approve a revised Operations Quality Assurance Program
(OQAP) that incorporated the methodology for grading quality assurance
(QA) based on PRA insights. The STP graded QA proposal was an extension
of the existing regulatory framework. Specifically, the STP approach
continued to use the traditional safety-related categorization, but
allowed for gradation of safety significance within the ``safety-
related'' categorization (consistent with 10 CFR part 50 appendix B)
through use of a risk-informed process. Following extensive discussions
with the licensee and substantial review, the NRC staff approved the
proposed revision to the OQAP on November 6, 1997. Subsequent to NRC's
approval, STPNOC identified implementation difficulties associated with
the graded QA program. Despite the reduced QA requirement applied for a
large number of SSCs in which the licensee judged to be of low safety
significance, other regulatory requirements such as environmental
qualification, the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (BPV), or seismic requirements,
continued to impose substantial burdens. As a result, the replacement
of a low safety significant component needed to satisfy other special
requirements during a procurement process. These requirements prevented
STPNOC from realizing the full potential reduction in unnecessary
regulatory burden for SSCs judged to have little or no safety
importance. In an effort to achieve the full benefit of the graded QA
program (and in fact to go beyond the staff's
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previous approval of graded QA), STPNOC submitted a request, dated July
13, 1999, asking for an exemption from the scope of numerous special
treatment regulations (including 10 CFR part 50 appendix B) for SSCs
categorized as low safety significant or as non-risk significant.
STPNOC's exemption was ultimately approved by the staff in August 2001
(further discussion on this exemption request is provided in Section
IV.2).
The experience with graded QA was a principal factor in the NRC's
determination that rule changes would be necessary to proceed with some
activities to risk-inform requirements. The Commission also believes
that the development of PRA technology and decision-making tools for
using risk information together with deterministic information
supported rulemaking activities to allow the NRC to refocus certain
regulatory requirements using this type of information.
Under Option 2 of SECY-98-300, ``Options for Risk-Informed
Revisions to 10 CFR part 50--`Domestic Licensing of Production and
Utilization Facilities,' '' dated December 23, 1998, the NRC staff
recommended that risk-informed approaches to the application of special
treatment requirements be developed as one application of risk-informed
regulatory changes. Option 2 (also referred to as RIP50 Option 2)
addresses the implementation of changes to the scope of SSCs needing
special treatment while still providing assurance that the SSCs will
perform their design functions. Changes to the requirements pertaining
to the design basis functional requirements of the plant or the design
basis accidents are not included in Option 2. These technical risk-
informed changes are addressed under Option 3 of SECY-98-300. The
Commission approved proceeding with Option 2 in a staff requirements
memorandum (SRM) dated June 8, 1999.
The stated purpose of the ``Option 2'' rulemaking was to develop an
alternative regulatory framework that enables licensees, using a risk-
informed process for categorizing SSCs according to their safety
significance (i.e., a decision that considers both traditional
deterministic insights and risk insights), to reduce unnecessary
regulatory burden for SSCs of low safety significance by removing these
SSCs from the scope of special treatment requirements. As part of this
process, those SSCs found to be of risk-significance would be brought
under a greater degree of regulatory control through the requirements
being added to the rule, which are designed to maintain consistency
between actual performance and the performance credited in the
assessment process that determines their significance. As a result,
both the NRC and industry should be able to better focus their
resources on regulatory issues of greater safety significance.
The Commission directed the NRC staff to evaluate strategies to
make the scope of the nuclear power reactor regulations that impose
special treatment risk-informed. SECY-99-256, ``Rulemaking Plan for
Risk-Informing Special Treatment Requirements,'' dated October 29,
1999, was sent to the Commission to obtain approval for a rulemaking
plan and issuance of an Advance Notice of Proposed Rulemaking (ANPR).
By SRM dated January 31, 2000, the Commission approved publication of
the ANPR and approved the rulemaking plan. The ANPR was published in
the Federal Register on March 3, 2000 (65 FR 11488), for a 75-day
comment period, which ended on May 17, 2000. In the rulemaking plan,
the NRC proposed to create a new section within part 50, now identified
as Sec. 50.69, to contain these alternative requirements.
The Commission received more than 200 comments in response to the
ANPR. The NRC staff sent the Commission SECY-00-0194, ``Risk-Informing
Special Treatment Requirements,'' dated September 7, 2000, which
provided the staff's preliminary views on the ANPR comments and
additional thoughts on the preliminary regulatory framework for
implementing a rule to revise the scope of special treatment
requirements for SSCs. The comments from the ANPR are further discussed
in Section IV.1.0 of SECY-02-0176, ``Proposed Rulemaking to Add New
Section 10 CFR 50.69, ``Risk-Informed Categorization and Treatment of
Structures, Systems, and Components,'' dated September 30, 2002 (ADAMS
accession number ML022630007).
The concept developed for this rule, discussed at length in the
ANPR, applies treatment requirements based upon the safety significance
of SSCs, determined through consideration of both risk insights and
deterministic information. Thus, the risk-informed approach discussed
in this rule for establishing an alternative scope of SSCs subject to
special treatment requirements uses both risk and traditional
deterministic methods in a blended ``risk-informed'' approach.
The NRC staff prepared a proposed rule package and provided it to
the Commission in SECY-02-0176. The Commission approved issuance of
proposed 10 CFR 50.69 for public comment in a SRM dated March 28, 2003.
The proposed 10 CFR 50.69 rule was published for public comment in the
Federal Register on May 16, 2003 (68 FR 26511). The Commission received
26 sets of comments in response to the proposed rule. The comments are
discussed in Section II below.
The NRC staff provided the Commission the draft final rule in SECY-
04-0109 dated June 30, 2004. The Commission subsequently approved the
final rule subject to the changes denoted during the session and
documented in SRM dated October 7, 2004 (ADAMS accession number
ML042810516).
I.3 Rule Overview
Section 50.69 represents an alternative set of requirements whereby
a licensee or applicant may voluntarily undertake categorization of its
SSCs consistent with the requirements in Sec. 50.69(c), remove the
special treatment requirements listed in Sec. 50.69(b) for SSCs that
are determined to be of low individual safety significance, and
implement alternative treatment requirements in Sec. 50.69(d). The
regulatory requirements not removed by Sec. 50.69(b) continue to apply
as well as the requirements specified in Sec. 50.69. The rule contains
requirements by which a licensee categorizes SSCs using a risk-informed
process, adjusts treatment requirements consistent with the relative
significance of the SSC, and manages the process over the lifetime of
the plant. To implement these requirements, a risk-informed
categorization process is employed to determine the safety significance
of SSCs and place the SSCs into one of four risk-informed safety class
(RISC) categories. The determination of safety significance is
performed by an integrated decision-making process which uses both risk
insights and traditional engineering insights. The safety functions
include both the design basis functions (derived from the ``safety-
related'' definition, which includes external events), as well as,
functions credited for severe accidents (including external events).
Treatment for the SSCs is required to be applied as necessary to
maintain functionality and reliability, and is a function of the
category into which the SSC is categorized. Finally, assessment
activities are conducted to make adjustments to the categorization and
treatment processes as needed so that SSCs continue to meet applicable
requirements. The rule contains requirements for obtaining prior NRC
review and approval of the categorization process and for
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maintaining certain plant records and reports. For a more detailed
discussion of the rule requirements refer to Sections III and V of this
rule.
It is important to note that this rulemaking effort, while intended
to ensure that the scope of special treatment requirements imposed on
SSCs is risk-informed, is not intended to allow for the elimination of
SSC functional requirements or to allow equipment that is required by
the deterministic design basis to be removed from the facility (i.e.,
changes to the design of the facility must continue to meet the current
requirements governing design change; most notably Sec. 50.59).
Instead, this rulemaking should enable licensees and the staff to focus
their resources on SSCs that make a significant contribution to plant
safety by restructuring the regulations to allow an alternative risk-
informed approach to special treatment. Conversely, for SSCs that do
not significantly contribute to plant safety on an individual basis,
this approach should allow an acceptable, though reduced, level of
confidence (i.e., ``reasonable confidence'') that these SSCs will
satisfy functional requirements. However, continued maintenance of the
health and safety of the public will depend on effective implementation
of Sec. 50.69 by the licensee or applicant applying the rule at its
nuclear power plant.
II. Public Comments
II.1.0 Comments on Proposed Rule
The Commission published proposed Sec. 50.69 for public comment on
May 16, 2003 (68 FR 26511). Twenty-six sets of comments were received
(comments are available at http://ruleforum.llnl.gov/cgi-bin/rulemake?source=SSC_PRULE&st=prule). The Commission requested
feedback on several specific issues in Section VI of the proposed rule
notice. A summary of the public feedback concerning these issues, as
well as a discussion of the more significant comments, follows. A
detailed discussion of the issues raised by all comments is contained
in a separate document (see Section IX, Availability of Documents).
II.1.1 Consideration of More Detailed Language for Sec. 50.69(d)(2)
Regarding RISC-3 SSC Treatment Requirements
As discussed in the proposed rule, the Commission believed that
detailed rule language for the treatment of RISC-3 SSCs (i.e., safety-
related SSCs that are categorized as low safety significant) was not
necessary to provide reasonable confidence in RISC-3 design basis
capability and, as a consequence, constructed proposed Sec. 50.69 to
contain high-level (i.e., less detailed) RISC-3 treatment requirements.
However, the Commission recognized that some stakeholders could
disagree with this approach and invited comment on this issue. For the
most part, industry commenters asserted that there was no need for more
detailed treatment requirements for RISC-3 SSCs in the rule. The state
commenters and public interest groups considered the proposed rule
language to be inadequate to provide reasonable confidence in the
capability of RISC-3 SSCs to perform their safety-related functions
under design basis conditions. In reviewing the public comments, the
Commission found significant divergence in the interpretation of the
proposed rule language by industry commenters from the Commission's
expectations as described in the Statement of Considerations--
preamble--(SOC) for the proposed rule. After consideration of all
stakeholder comments, the Commission revised Sec. 50.69(d)(2) to adopt
a more performance-based approach that provides licensees and
applicants greater flexibility in establishing RISC-3 treatment
consistent with the low safety significance of RISC-3 SSCs.
Accordingly, the Commission has removed the more prescriptive
requirements regarding RISC-3 treatment activities and adopted rule
language that focuses on the performance requirements for RISC-3 SSCs.
II.1.2 PRA Requirements
The Commission requested stakeholder comment on whether the NRC
should amend the requirements in Sec. 50.69(c) to require a level 2
internal and external initiating events, all-mode, peer-reviewed PRA
that must be submitted to, and reviewed by, the NRC. Stakeholder
comments ranged from those supporting such PRA requirements to those
who conclude that the proposed PRA requirements in Sec. 50.69(c) are
sufficient. The industry commenters stated that additional PRA
requirements were not necessary because the other categorization
requirements in Sec. 50.69(c) addressed other modes and events not
addressed by the PRA and as a result, all sources of risk were
addressed. The states and public interest groups supported increased
PRA requirements. The Commission concludes that the Sec. 50.69 PRA
requirements in the proposed rule are sufficient for this application.
The supporting guidance for the rule has been structured such that
licensees will gain more benefit when PRA methods are used (beyond the
minimum PRA requirements in Sec. 50.69(c)), and where non-PRA methods
are used, the requirements and associated implementation guidance
account for this situation by requiring a process that tends to
conservatively categorize SSCs into RISC-1 and RISC-2 (i.e., no special
treatment requirements are removed). There are several other features
to the regulatory framework that also contribute to ensuring sound PRA
is used such as requiring aspects of the categorization process to be
reviewed and approved before implementation, requiring the PRA to be
peer reviewed, Integrated Decision-Making Panel (IDP) requirements,
provisions for addressing all modes and events regardless of whether in
the PRA, feedback and update requirements, and supporting standards.
(Also see the Commission's SRM on PRA quality dated December 18, 2003,
ADAMS Accession No. ML033520457.)
II.1.3 Review and Approval of RISC-3 Treatment
The Commission requested stakeholder comment on whether the NRC
should review and approve the RISC-3 treatment processes being
developed by the licensee or applicant before implementation in
addition to reviewing the categorization process. Public interest
groups and comments from state organizations generally stressed the
need for the NRC to review and approve RISC-3 treatment processes in
advance of implementation to confirm appropriate treatment will be
applied to RISC-3 SSCs given that these SSCs are safety-related. On the
other hand, industry commenters did not consider prior review and
approval of RISC-3 treatment to be necessary in light of the low safety
significance of individual RISC-3 SSCs, other requirements that help
maintain safety, and the availability of inspection and enforcement by
the NRC. The NRC agrees that the individual low safety significance of
RISC-3 SSCs supports allowing licensees to establish treatment for
RISC-3 SSCs without prior NRC review. This conclusion is based on the
rule containing:
(1) Robust categorization and PRA requirements;
(2) Requirements to show that implementation risk is small;
(3) Feedback requirements of paragraph (e) to help maintain the
validity of the categorization process; and
(4) The high-level, performance-based RISC-3 requirements designed
to
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maintain RISC-3 SSC design basis functional capability.
In addition, a provision has been added to the final rule to make
it clear that the treatment applied to RISC-3 SSCs must be consistent
with (i.e., maintain the validity of) the categorization process. To
provide additional assurance, the NRC intends to conduct sample
inspections at nuclear power plants implementing Sec. 50.69 to address
programmatic issues related to the categorization and treatment
processes (see below).
II.1.4 Inspection and Enforcement
The Commission requested stakeholder comment on whether or not
changes are needed in the NRC's reactor oversight process including the
inspection program and enforcement to enable NRC to exercise the
appropriate degree of regulatory oversight of these aspects of facility
operation regarding Sec. 50.69. The public comments on the proposed
rule indicated general support for providing regulatory oversight of
the implementation of processes established under Sec. 50.69 through
the NRC's inspection and enforcement process. Some stakeholders
considered the current inspection and enforcement process to be
sufficient without adjustment. Other stakeholders recommended that the
NRC consider additional training and guidance to inspectors to support
implementation of Sec. 50.69. Some stakeholders provided specific and
constructive suggestions regarding the inspection and enforcement
process under Sec. 50.69 including aspects of treatment processes to
be inspected, and the application of enforcement discretion. Based on
its consideration of this issue, the Commission plans to conduct
inspections of Sec. 50.69 implementation. These inspections will be
performed on a sampling basis (in terms of the number of plants
inspected) and will depend on the number of licensees who decide to
implement Sec. 50.69. These sample inspections are intended to gather
information that will enable the NRC to assess whether modifications
are needed to the ongoing baseline inspection program. The principal
focus of the inspection will be on the safety significant aspects of
Sec. 50.69 implementation such as categorization and treatment of
RISC-1 and RISC-2 SSCs, but the inspection will also consider the
implementation of RISC-3 treatment focusing on programmatic and common
cause issues, which could undermine the categorization process and its
results.
II.1.5 Operating Experience
The Commission requested stakeholder feedback regarding the role
that relevant operational experience could play in reducing the
uncertainty associated with the effects of treatment on performance and
specifically sought public comment as to what information might be
available and how it could be used to support implementation of this
rulemaking. Some stakeholders commented that relevant operating
experience argues against the removal of special treatment requirements
and that regulatory attention should be increased for this equipment.
Other stakeholders suggested that there is a large amount of data that
demonstrates that commercial and safety-related SSCs have comparable
failure rates with the implication that special treatment requirements
can be removed with little impact. The specific study referenced by
those stakeholders was not submitted for formal NRC review. The
Commission concludes that a single unreviewed study does not provide a
sufficient basis to make broad conclusions regarding the performance of
SSCs subject to commercial and industrial practices for fabrication,
installation, and maintenance. Other stakeholders commented that there
are already opportunities for industry to share experience data with
existing industry and regulatory programs implying that a new program
is not necessary. In some instances, however, those referenced programs
will be eliminated for RISC-3 SSCs under Sec. 50.69. To emphasize the
importance of applying operating experience in maintaining plant
safety, the final rule has been revised to clarify that Sec.
50.69(e)(1) requires the feedback of plant operational experience in
addition to the requirements to feed back performance data, plant
changes, operational changes, and industry experience. This plant
operational information may be obtained from the corrective action
program and processes, as well as other sources.
II.1.6 Other Substantive Issues
In addition to the issues addressed in Section II.1.5, stakeholders
provided substantive comments that caused the NRC to re-examine the
Sec. 50.69 framework and make changes. Those issues and comments are
discussed below. Additionally, there were several issues that involved
a significant number of stakeholder comments, and even though the
Commission decided not to revise its approach, those issues and
comments are also discussed in this section.
II.1.6.1 SOC Guidance
Numerous comments were received from the industry regarding the
nature of the information in the proposed rule's SOC supporting both
Sec. 50.69(d)(2) and Sec. 50.69(c). Several industry commenters
stated that the discussion in the SOC was inconsistent with the rule
requirements. For example, some commenters suggested that, contrary to
the SOC discussion, the treatment requirements for RISC-3 SSCs in Sec.
50.69(d)(2) would allow exercising of pumps and valves as a means of
providing reasonable confidence in the design basis capability of those
components. Another commenter claimed that, contrary to the SOC
discussion, Sec. 50.69 would allow the leakage tests required by 10
CFR part 50, Appendix J, for containment isolation valves to be
eliminated without considering the capability of those valves to close
under design basis conditions. Other commenters asserted that the
corrective action process alone would be sufficient to satisfy the
high-level requirements for feedback and monitoring of RISC-3 SSCs in
Sec. 50.69. These industry comments raised concerns regarding the
interpretation of the rule language. The Commission clarified the rule
requirements and simplified the SOC to focus on the meaning of the rule
language (see Sections II.1.6.2 through II.1.6.3, Section V.5.2, and
the responses to comments d-32 and e-4 in Table 3 of ``Response to
Comments on Proposed Sec. 50.69'' as referenced in Section IX of this
document).
II.1.6.2 RISC-3 Treatment Requirements
Numerous stakeholder comments were received concerning the Sec.
50.69(d)(2) requirements for RISC-3 SSCs. Some public stakeholders
provided their view that the RISC-3 treatment requirements were
inadequate in light of previous industry experience (e.g., regarding
the use of substandard parts) and that more detailed RISC-3
requirements were needed to address common cause failures, significant
degradation, and in general to avoid an increase in risk to the health
and safety of the public. Industry stakeholders tended to view the
RISC-3 requirements as too prescriptive and beyond what is necessary to
maintain reasonable confidence of RISC-3 SSC design basis capability.
Some of the industry comments revealed that the rule requirements might
not be implemented consistent with the Commission's expectations
discussed in the SOC. Therefore, the Commission clarified the
[[Page 68013]]
rule and SOC as discussed in the following sections.
II.1.6.2.1 Fracture Toughness
In the SOC for the proposed rule, the Commission noted that design
requirements for fracture toughness would continue to apply for
replacement ASME components categorized as RISC-3 SSCs. One industry
commenter asserted that fracture toughness is not a design issue while
other commenters argued in general that the SOC discussion exceeded the
rule requirements. The Commission emphasizes that the intent of Sec.
50.69 is to remove special treatment requirements while maintaining
design requirements for RISC-3 SSCs. The Commission considers fracture
toughness to be an important design consideration. Fracture toughness
is a property of the material that prevents premature failure of an SSC
at abrupt geometry changes, or at small undetected flaws. Adequate
fracture toughness of SSCs is necessary to prevent common cause
failures due to design basis events, such as earthquakes. To ensure
that this design consideration continues to be applicable to Sec.
50.69 licensees, Sec. 50.69(b)(1)(v) was clarified to exclude fracture
toughness from the scope of Sec. 50.55a repair and replacement
requirements which are removed for RISC-3 SSCs.
II.1.6.2.2 Consistency With the Categorization Process
Several industry comments indicated that licensees might not
consider the impact of changes in treatment on RISC-3 SSCs as part of
the categorization process. For example, one industry commenter
asserted that sensitivity studies eliminate the need to specifically
consider SSC reliability changes that might occur due to treatment
changes. Another industry commenter stated that cross-system common
cause interactions are rarely modeled in PRAs. Similarly, another
industry commenter indicated that degradation mechanisms resulting from
treatment processes are typically not considered in PRAs. The treatment
practices for plant SSCs must support the capability credited in the
categorization process for there to be reasonable confidence that any
increase in risk remains small. Therefore, Sec. 50.69(d)(2) was
clarified to explicitly require the treatment of RISC-3 SSCs to be
consistent with the categorization process.
II.1.6.2.3 Voluntary Consensus Standards
In the SOC for the proposed rule, the Commission discussed the use
of voluntary consensus standards as one effective means to establish
treatment requirements for RISC-3 SSCs. In its comments, the ASME did
not recommend adding a provision on voluntary consensus standards in
the rule itself because it considered the SOC to provide adequate
guidance for RISC-3 treatment. However, several industry commenters
suggested that licensees might only apply general industrial practices
when implementing treatment requirements for RISC-3 SSCs. For example,
some industry commenters believed that exercising a pump or valve would
provide sufficient assurance under Sec. 50.69 of the capability of the
pump or valve to perform its design basis safety functions. Although
exercising a pump or valve might be consistent with general industrial
practices, operating experience has demonstrated that exercising a pump
or valve is not sufficient to ensure with reasonable confidence its
design basis capability. For example, the Commission modified Sec.
50.55a to require licensees implementing the ASME Code for Operation
and Maintenance of Nuclear Power Plants to periodically verify the
design basis capability of motor-operated valves to perform their
safety functions in light of the recognized inadequacies in stroke-time
testing (somewhat more informative than exercising) to assess the
operational readiness of those valves. The NRC issued Regulatory Issue
Summary 00-03 (March 15, 2000), ``Resolution of Generic Safety Issue
158, Performance of Safety-Related Power-Operated Valves Under Design
Basis Conditions,'' to discuss the importance of this issue relative to
safety-related air-operated and other power-operated valves. Further,
the ASME developed comprehensive pump testing provisions to provide
more appropriate testing under significant flow conditions in light of
the weakness of the previous Code testing under minimal loading
conditions. In SECY-00-0194, the NRC noted that a wide variation
existed in industrial practices. Therefore, certain industrial
practices may not be sufficient to satisfy the treatment requirements
for RISC-3 SSCs in Sec. 50.69. To address these concerns, the
Commission clarified the rule requirements to indicate that the
treatment of RISC-3 SSCs must be consistent with the categorization
process. One way to achieve this consistency could be the application
of consensus standards. However, licensees or applicants must recognize
that the application of such standards must meet Sec. 50.69(d)(2)
requirements to be acceptable. The determination of consistency between
treatment and categorization also includes consideration of applicable
operational experience, which may be found from such sources as NRC
information notices, bulletins, and generic letters; and vendor
recommendations.
II.1.6.2.4 Design Control Process
In the SOC for the proposed rule, the Commission listed several
attributes to be considered as part of the design control process for
RISC-3 SSCs in satisfying the high-level treatment requirements in
Sec. 50.69. One industry commenter suggested that a focused list of
design control attributes be substituted in Sec. 50.69 for the
proposed rule language. This list would include selection of suitable
materials; verification of design adequacy, and control of design
changes. After consideration of these comments, the Commission has
decided not to include detailed design control process requirements in
the final rule. The final rule requirements require that licensees and
applicants ensure with reasonable confidence that RISC-3 SSCs remain
capable of performing their safety-related functions under design basis
conditions. With respect to design changes, as noted in several places
in the notice for the final rule, Sec. 50.69 is not changing design
basis functional requirements and Sec. 50.59 remains applicable to all
changes to non-special treatment aspects of RISC-3 SSCs. The Commission
believes that a performance-based requirement will allow licensees who
choose to implement Sec. 50.69 to have greater flexibility to
implement treatment that they have determined is needed, commensurate
with the safety significance of the SSCs in order to ensure with
reasonable confidence that RISC-3 safety-related functional capability
is maintained.
II.1.6.2.5 Design Basis Conditions
Under Sec. 50.69, RISC-3 SSCs will be exempt from special
treatment requirements for qualification methods for environmental
conditions and effects and seismic conditions. Nevertheless, RISC-3
SSCs continue to be required to be capable of performing their safety-
related functions under applicable environmental conditions and effects
and seismic conditions, albeit at a lower level of confidence as
compared to RISC-1 SSCs. Based on industry comments on the proposed
rule, some
[[Page 68014]]
licensees appeared to interpret the proposed rule language as not
requiring evaluation of environmental and seismic capability of RISC-3
SSCs. For example, one industry commenter stated that Sec. 50.69
exempts RISC-3 electrical equipment from aging issues and that the rule
would not require the establishment of design life for RISC-3
electrical equipment. Contrary to the public comment, a licensee
implementing Sec. 50.69 must consider operating life (aging) and
combinations of operating life parameters (synergistic effects) in the
design of RISC-3 electrical equipment. This is particularly important
if the equipment contains materials which are known to be susceptible
to significant degradation due to thermal, radiation, and/or wear
(cyclic) aging including any known synergistic effects that could
impair the ability of the equipment to meet its design basis function.
However, the Commission agrees that the applicable rule language can be
simplified and has revised the final rule to utilize a performance-
based approach to ensuring with reasonable confidence the functional
capabilities of RISC-3 SSCs. Accordingly, the final rule has been
revised by deleting the reference to the specific conditions that were
parenthetically listed in the proposed rule.
II.1.6.2.6 Corrective Action
Some public commenters raised concerns regarding the lack of
requirements for the consideration of common-cause issues for RISC-3
SSCs. An industry commenter also noted this omission in the proposed
rule and provided proposed rule language to resolve this issue.
Therefore, the Commission decided to revise Sec. 50.69(d)(2)(ii) to
require that, for significant conditions adverse to quality associated
with RISC-3 SSCs, measures shall be taken to provide reasonable
confidence that the cause of the condition is determined and corrective
action is taken to preclude repetition. The revised corrective action
requirement is consistent with a proposal by the Nuclear Energy
Institute and uses language that is similar to 10 CFR part 50 Appendix
B Criterion XVI. As such, this should be a well-understood requirement
that minimizes the potential for common cause failures. It is also
consistent with the principle of performance-based regulation that non-
compliance with the performance requirement should provide sufficient
margin such that reasonable assurance of public health and safety
continues to be provided.
II.1.6.2.7 Seismic Experience Data
Several industry commenters stated that the SOC for the proposed
rule might create additional burden on plants licensed before
implementation of Appendix A to 10 CFR Part 100. In establishing Sec.
50.69, the Commission does not intend to alter the existing seismic
design requirements for RISC-3 SSCs in any plant's design basis.
Industry commenters also raised concerns regarding the SOC discussion
on use of seismic experience data. In meeting Sec. 50.69, the licensee
or applicant must have adequate technical bases to conclude that RISC-3
SSCs will perform their safety-related functions under seismic design
basis conditions, which includes the number and magnitude of earthquake
events specified for the SSC design. Some commenters implied that it
would be acceptable to use ``experience data'' alone to have reasonable
confidence that an SSC is capable of functioning during an earthquake
even if there is no actual ``experience data'' for the SSC. While the
use of experience data is not prohibited by the rule, it may be
difficult for a licensee or applicant to show that experience data
alone will satisfy the applicable design requirements of 10 CFR part
100 (which Sec. 50.69 leaves intact). The Commission clarified the SOC
with respect to the use of seismic experience data and to indicate that
Sec. 50.69 will not change the seismic design basis for Unresolved
Safety Issue (USI) A-46\1\ plants or impose additional seismic
requirements for those plants.
---------------------------------------------------------------------------
\1\ In December 1980 the NRC designated ``Seismic Qualification
of Equipment in Operating Plants'' as an unresolved safety issue.
For more information refer to GL 87-02.
---------------------------------------------------------------------------
II.1.6.3 Feedback
Several industry commenters requested adjustments to the feedback
requirements in Sec. 50.69(e)(1) to provide more efficient
implementation of the rule. Upon consideration of those comments, the
Commission revised Sec. 50.69(e)(1) to replace the maximum time
interval for updating the categorization and treatment processes from
36 months to two refueling outages, and to indicate that the licensee
or applicant may adjust either its categorization process or its
treatment processes in satisfying the feedback requirement.
II.1.6.4 Section 50.46a/Appendix B Requirements for High Point Vents
A comment was submitted that the NRC should undertake a review of
the recently revised Sec. 50.44 to determine whether the new rule
contains special treatment requirements that should be within the scope
of Sec. 50.69. The Commission agreed with this comment. The Commission
noted in the proposed rule (Section III.4.9.3) that there may be a need
to scope into Sec. 50.69 certain provisions of the old Sec. 50.44
dependent on the outcome of the effort to risk inform the Sec. 50.44
requirements. The revised Sec. 50.44 has no special treatment
requirements. However, when Sec. 50.44 was revised, a portion of the
old Sec. 50.44 regarding application of Appendix B requirements to
high point vents was moved to Sec. 50.46a. This particular requirement
was not risk-informed as part of the Sec. 50.44 effort and was instead
simply relocated. Because application of Appendix B is a special
treatment requirement, the Appendix B portion of Sec. 50.46a(b) has
been included within the scope of Sec. 50.69 by the inclusion of Sec.
50.69(b)(1)(ii).
II.1.6.5 Basis for RISC-3 SSC Reliability Used in Sec. 50.69(c)(1)(iv)
Evaluation
A number of comments were received regarding the technical basis
for the RISC-3 SSC reliability (failure rates) to be used in the risk
sensitivity study performed to meet Sec. 50.69(c)(1)(iv) requirements
to demonstrate reasonable confidence that any potential risk increase
from implementation of the rule is small. Some commenters suggested
that licensees or applicants that voluntarily implement the rule should
be required to characterize and reasonably bound the specific effects
of eliminating treatment on SSC reliability under design basis and
severe accident conditions. Other commenters suggested that there is
evidence that reductions in treatment (using industry practices) has no
impact on SSC reliability.
The NRC recognizes that the reliability of RISC-3 SSCs could
potentially decrease (RISC-3 SSC failure rates increase) due to the
reduction in treatment applied to these SSCs as a result of Sec. 50.69
implementation. This is the reason why the Commission requires in the
rule that the licensee demonstrate with reasonable confidence that any
potential risk increase due to implementation of the rule will be
small. However, the NRC also recognizes that it is difficult a priori
to relate specific changes in treatment directly to specific changes in
SSC reliability. The rule has been constructed to account for this
difficulty. First, the categorization process that a licensee uses must
comply with the rule's requirements. Second, this categorization
process will be reviewed and approved by the NRC
[[Page 68015]]
before implementation. These steps are to have high confidence that
SSCs are appropriately categorized so that RISC-3 SSCs are of low
individual safety significance. Third, licensees are required to
provide reasonable confidence that any risk increase due to
implementation is acceptably small and this assessment must be
supported by a supporting technical justification that discusses why
the assessment adequately addresses the potential reliability changes
for RISC-3 SSCs. This basis may include reliance on the capability of
the licensee's data collection, feedback, and corrective action , which
are also addressed by requirements of the rule. Finally, the rule has
been revised to clarify the linkage between treatment and
categorization and specifically to ensure that the treatment process is
consistent with the categorization process, including the risk
sensitivity study (i.e., maintain that any risk increase due to reduced
treatment is acceptably small). Therefore, the rule is structured to
contain:
(1) Robust categorization and PRA requirements;
(2) Requirements to show that implementation risk is small;
(3) A new provision to make it clear that the treatment applied to
RISC-3 SSCs must be consistent with (i.e., maintain the validity of)
the categorization process;
(4) Feedback requirements of Sec. 50.69(e) to maintain the
validity of the categorization process; and,
(5) The high-level RISC-3 requirements designed to maintain RISC-3
SSC design basis functional capability.
Thus, the Commission finds that the rule, as revised, has the
appropriate provisions for addressing the concerns regarding the basis
for RISC-3 SSC reliability used in the risk sensitivity study to be
performed to meet the Sec. 50.69(c)(1)(iv) requirement to demonstrate
with reasonable confidence that any potential risk increase from
implementation of the rule is small.
II.1.6.6 RISC-1 and RISC-2 Treatment Requirements and Crediting SSCs
A number of industry stakeholders commented on the treatment
requirements applicable to RISC-1 and RISC-2 SSCs in Sec. 50.69(d)(1).
These stakeholders commented that this requirement obligated a licensee
implementing Sec. 50.69 to evaluate treatment applied to all safety
significant SSCs to ensure adequacy of treatment and cited this as an
added burden that is neither necessary nor appropriate because RISC-1
SSCs are already subjected to full regulatory requirements. They also
commented that it appeared that this requirements was extending special
treatment requirements (such as Appendix B) to RISC-2 SSCs. In fact
there was a general consensus of comments that any additional treatment
requirements for RISC-1 and RISC-2 SSCs should be removed from the SOC
or that the SOC be clarified to address the specific beyond design
basis scope of additional regulatory controls. First, the Commission
notes that Sec. 50.69(d)(1) does not require licensees or applicants
to evaluate the application of special treatment requirements to RISC-1
SSCs. These requirements are to maintain the design basis functional
requirements with a high level of assurance. The special treatment
requirements remain intact and unchanged, and hence there is no reason
that an evaluation of the application of special treatment requirements
should be required. Secondly, the Commission notes that it is not the
intent of Sec. 50.69(d)(1) to simply extend special treatment
requirements such as Appendix B to RISC-1 and RISC-2 beyond design
basis functions. Instead, the focus of Sec. 50.69(d)(1) is on the PRA
credited performance of RISC-1 and RISC 2 SSCs for beyond design basis
conditions, and specifically for ensuring that there is a valid
technical basis for the credit taken in the PRA (i.e., there must be a
valid technical basis for the failure rate/probability of the SSC
performing the function). The basis for this credit should already be
established and documented in the PRA supporting documentation, so this
should not be an additional burden for licensees to capture and
implement. If an existing technical basis does not exist or is
insufficient to support the credit taken in the PRA, then Sec.
50.69(d)(1) would require that a technical basis be developed for the
credit taken; potentially including the creation of a treatment program
for the SSC that validates the capability credited.
Regarding the issue of ``credited'' SSCs, several commenters stated
that the SOC implied an enormous program would be required if a
licensee decides to selectively implement Sec. 50.69 for a set of
systems. It was commented that this enormous program would result due
to the application of Sec. Sec. 50.69(d)(1) and 50.69(e)(2) to
maintain credited performance within the PRA and thereby enable the
selected set of SSCs to be categorized as low safety significant. As
the Commission has already noted, Sec. 50.69(d)(1) obligates licensees
to have a basis to support the performance of RISC-1 and RISC-2 SSCs
credited in the PRA used in the categorization process, including the
performance credited for beyond design basis conditions. This is an
important aspect of the rule. The categorization process will result in
a number of safety-related SSCs being determined to be of low safety
significance (i.e., RISC-3) and subject to reduced treatment. This
determination of low safety significance will implicitly take credit
for the performance capability of other SSCs in the PRA, some or all of
which may not be included in the scope of the licensee's categorization
process (due to the allowance for licensees to selectively implement
the rule and to phase that implementation over time). To maintain the
validity of the categorization process, and more importantly to
maintain any potential risk increase as small, it is necessary to
maintain the ``credited'' SSCs per Sec. 50.69, and this means the
application of Sec. Sec. 50.69(d)(1) and 50.69(e)(2) requirements as
suggested by the comment.
II.1.6.7 Adequate Protection Comments
The NRC received several comments indicating that the proposed
regulation would not maintain adequate protection of public health and
safety. The Commission disagrees with these comments and concludes that
both the proposed rule requirements and the final rule requirements
maintain adequate protection for the reasons discussed in Section
III.7.0 of this notice.
II.1.6.8 License Amendment
A commenter stated that the requirement to prepare, submit, and
then receive approval of a license amendment to implement Sec. 50.69
is a disincentive to its use. The commenter argued that, in light of
the desire to move to a more performance-based regulatory regime,
voluntary implementation of Sec. 50.69 should be developed by
licensees using the requirements in the rule and any attendant
regulatory guidance, with routine NRC inspection serving to verify
acceptable compliance. The Commission has decided not to revise Sec.
50.69 in response to this comment. The Commission continues to conclude
that (as discussed in Section III.6.0 of this rule) the review of the
license amendment submittal will involve substantial engineering
judgment on the part of NRC reviewers, inasmuch as the rule does not
contain objective, non-discretionary criteria for assessing the
adequacy of the PRA process, PRA
[[Page 68016]]
review results and sensitivity studies. Consistent with the
Commission's decision in Cleveland Electric Illuminating Co. (Perry
Nuclear Power Plant, Unit 1), CLI-96-13, 44 NRC 315 (1996), the final
rule requires NRC approval to be provided by issuance of a license
amendment.
III. Final Rule
The Commission is establishing Sec. 50.69 as an alternative set of
requirements whereby a licensee or applicant may undertake
categorization of its SSCs consistent with the requirements in Sec.
50.69(c) and adjust treatment requirements per Sec. 50.69(d) based
upon the resulting significance. Under this approach, a licensee or
applicant is allowed to remove the special treatment requirements
listed in Sec. 50.69(b) for SSCs that are determined to be of low
safety significance while potentially enhancing requirements for
treatment of other SSCs that are found to be safety significant. The
requirements establish a process by which a licensee categorizes SSCs
using a risk-informed process, adjusts treatment requirements
consistent with the relative significance of the SSC, and manages the
process over the lifetime of the plant. To implement these
requirements, a risk-informed categorization process is employed to
determine the safety significance of SSCs and place the SSCs into one
of four RISC categories. It is important that this categorization
process be robust to enable the Commission to remove requirements for
SSCs determined to be of low safety significance. The determination of
safety significance is performed by an integrated decision-making
process which uses both risk insights and traditional engineering
insights. The safety functions include both the design basis functions
(derived from the ``safety-related'' definition, which includes
external events), as well as functions credited for severe accidents
(including external events). Treatment requirements for the SSCs are
applied as necessary to maintain functionality and reliability and are
a function of the category into which the SSC is categorized. Finally,
assessment activities are conducted to make adjustments to the
categorization and treatment processes as needed so that SSCs continue
to meet applicable requirements. The rule also contains requirements
for obtaining NRC approval of the categorization process and for
maintaining plant records and reports.
III.1.0 Categorization of SSCs
Section 50.69 defines four RISC categories into which SSCs are
categorized. Four categories were chosen because it is the simplest
approach for transitioning between the previous SSC classification
scheme and the new scheme used in Sec. 50.69. The depiction in Figure
1 provides a conceptual understanding of the new RISC categories. The
figure depicts the current safety-related versus nonsafety-related SSC
categorization scheme with an overlay of the new risk-informed
categorization. In the traditional deterministic approach, SSCs were
generally categorized as either ``safety-related'' (as defined in Sec.
50.2) or nonsafety-related. This division is shown by the vertical line
in the figure. Risk insights, including consideration of severe
accidents, can be used to identify SSCs as being safety significant or
low safety significant (shown by the horizontal line). Hence, the
application of a risk-informed categorization results in SSCs being
grouped into one of four categories as represented by the four boxes in
Figure 1.
Box 1 of Figure 1 depicts safety-related SSCs that a risk-informed
categorization process determines are significant contributors to plant
safety. These SSCs are termed RISC-1 SSCs. RISC-2 SSCs, depicted by box
2 in Figure 1, are nonsafety-related SSCs that the risk-informed
categorization determines to be significant contributors to plant
safety. The third category are those SSCs that are safety-related SSCs
and that a risk-informed categorization process determines are not
significant individual contributors to plant safety. These SSCs are
termed RISC-3 SSCs and are depicted by box 3 in Figure 1. Finally,
there are SSCs that are nonsafety-related and that a risk-informed
categorization process determines are not significant contributors to
plant safety. These SSCs are termed RISC-4 SSCs and are depicted by box
4 in Figure 1.
[[Page 68017]]
[GRAPHIC] [TIFF OMITTED] TR22NO04.000
Section 50.69 defines the terminology ``safety significant
function'' as functions whose loss or degradation could have a
significant adverse effect on defense-in-depth, safety margins, or
risk. This definition was chosen to be consistent with the concepts
described in RG 1.174. The rule maintains more treatment requirements
on SSCs that perform safety significant functions (RISC-1 and RISC-2
SSCs) than on SSCs that perform low safety significant functions to
ensure that defense-in-depth and safety margins are maintained. The
rule also requires that the licensee or applicant provide reasonable
confidence that the change in risk associated with implementation of
Sec. 50.69 will be small.
III.2.0 Methodology for Categorization
The cornerstone of Sec. 50.69 is the establishment of a robust,
risk-informed categorization process that provides high confidence that
the safety significance of SSCs is correctly determined considering all
relevant information. As such, all the categorization requirements
incorporated into Sec. 50.69 are to achieve this objective.
Essentially, the process is structured to ensure that all relevant
information pertaining to SSC safety significance is considered by a
panel (referred to as either an expert panel or an integrated decision-
making panel (IDP)) that has the expertise and capabilities for making
a sound decision regarding the SSC's categorization, and that the
assembled information is considered in a manner that ensures the
Commission's criteria for risk-informed applications are satisfied
(i.e., defense-in-depth is maintained, reasonable confidence that
safety margins are maintained, reasonable confidence that any risk
increase is small, and a monitoring and performance assessment
[[Page 68018]]
strategy is used). This process enables SSCs to be placed in the
correct RISC category so that the appropriate treatment requirements
will be applied commensurate with the SSC's safety significance. A
safety significant SSC is an SSC that performs a safety significant
function as defined in Sec. 50.69. The rule requires that SSC safety
significance be determined using quantitative information from a PRA
that reasonably represents the as-built, as-operated, current plant
configuration, and which at a minimum covers internal events at full
power. The categorization process must address both internal events and
external events for all modes of operation and can use other available
risk analyses and traditional engineering information to supplement the
quantitative PRA results to address modes and events not within the
scope of the PRA.
Section 50.69(c)(1)(i) ensures that the PRA is adequate for this
application. Section 50.69(c)(1)(iii) requires that defense-in-depth is
maintained as part of the categorization process. Section
50.69(c)(1)(iv) requires that the revised treatment applied to RISC-3
SSCs be considered for its potential impact on risk. As an example, the
Commission's position is that the containment and its systems are
important in the preservation of defense-in-depth (in terms of both
large early and large late releases). As part of maintaining defense-
in-depth, a licensee must demonstrate that the function of the
containment as a barrier (including fission product retention and
removal) is not significantly degraded when SSCs that support the
functions are moved to RISC-3.
Section 50.69(c)(2) requires the risk insights and other
traditional information to be evaluated by the IDP and this panel must
be comprised of expert, plant-knowledgeable members whose expertise
includes PRA, safety analysis, plant operation, design engineering, and
system engineering. Because the IDP makes the final determination about
the safety significance of an SSC, the Commission concludes that the
requirements in Sec. 50.69(c)(2) are necessary for the composition of
the panel to be experienced personnel who possess diverse knowledge and
insights in plant design and operation and who are capable in the use
of deterministic knowledge and risk insights to categorize SSCs.
As mentioned previously, the Sec. 50.69 categorization process
requires that available deterministic and probabilistic information
pertaining to SSC safety significance be considered in the decision
process. The information considered must reasonably reflect the as-
built and as-operated plant so that the decisions are based upon
correct information, leading to proper categorization. Where
applicable, the information is to come from a PRA that is adequate for
this application (i.e., categorization of SSC safety significance).
From this perspective, the IDP decision process can be viewed as an
extension of the previous process for determining SSC safety
classification (i.e., safety-related or nonsafety-related), in that it
is making use of relevant risk information that was not considered or
not available when the SSCs were initially classified. The IDP makes
the final determination of the safety significance of SSCs using a
process that takes all this information into consideration, in a
structured, documented manner. The structure provides consistency to
decisions that may be made over time and the documentation gives both
the licensee and the NRC the ability to understand the basis for the
categorization decision, should questions arise at a later date.
Section 50.69(c)(1)(ii) contains general requirements for
consideration of SSCs, modes of operation, and initiating events not
modeled in the PRA. As a result, the implementing guidance plays a
significant role in effective implementation and bolsters the need for
NRC review and approval of the categorization process before
implementation.
The PRA used to provide the risk information to the categorization
process is required to be subjected to a peer review. The peer review
focuses on the PRA's completeness and technical adequacy for
determining the importance of particular SSCs, including consideration
of the scope, level of detail, and technical quality of the PRA model,
the assumptions made in the development of the results, and the
uncertainties that impact the analysis. This provides confidence that
for IDP decisions that use PRA information, the results of the
categorization process provide a valid representation of the risk
importance of SSCs.
Before a licensee may implement Sec. 50.69, the NRC must approve
the categorization process through a license amendment. This is
necessary because of the importance of the PRA and categorization
process to successful implementation of the rule. This review and
approval of the categorization process is a one-time, process approval
(i.e., the approval is not restricted to a set of systems or
structures, and can be applied to any system or structure in the plant
and the licensee is not required to come back to the NRC for review of
the categorization process provided that licensee remains within the
scope of the NRC's safety evaluation). The NRC's review of the Sec.
50.69 submittal will determine whether Sec. 50.69 requirements are
satisfied and consider the adequacy of the PRA; focusing on the results
of the peer review and the actions taken by the licensee to address any
peer review findings. The Commission has determined that a focused NRC
review of the PRA is necessary because there are key assumptions and
modeling parameters that can have a significant impact on the results
so that NRC review of their adequacy for this application is considered
necessary to verify that the overall categorization process will yield
acceptable decisions.
Section 50.69(c)(1)(iv) requires reasonable confidence that the
increase in the overall plant core damage frequency (CDF) and large
early release frequency (LERF) resulting from potential decreases in
the reliability of RISC-3 SSCs as a result of the changes in treatment
be small. The rule further requires the licensee (or applicant) to
describe the evaluations to be performed to meet this requirement. As
presented in RG 1.174, the NRC considers small changes to be relative
and to depend on the current plant CDF and LERF (hence we also refer to
``acceptably small'' changes in other portions of this notice since
small can be different for different plants with different baseline
levels of risk). For plants with total baseline CDF of 10-4 per year or
less, small means CDF increases of up to 10-5 per year and for plants
with total baseline CDF greater than 10-4 per year, small means CDF
increases of up to 10-6 per year. However, if there is an indication
that the CDF may be considerably higher than 10-4 per year, the focus
of the licensee should be on finding ways to decrease rather than
increase CDF and the licensee may be required to present arguments as
to why steps should not be taken to reduce CDF for the reduction in
special treatment requirements to be considered. For plants with total
baseline LERF of 10-5 per year or less, small LERF increases
are considered to be up to 10-6 per year, and for plants
with total baseline LERF greater than 10-5 per year, small LERF
increases are considered to be up to 10-7 per year. However, if there
is an indication that the baseline CDF or LERF may be considerably
higher than 10-4 or 10-5, respectively, the
licensee either must find ways to reduce risk and present the arguments
to the NRC staff before implementation of Sec. 50.69, otherwise it
[[Page 68019]]
is likely that the NRC will deny the Sec. 50.69 application. This is
consistent with the guidance in Section 2.2.4 of RG 1.174. It should be
noted that this allowed increase shall be applied to the overall
categorization process, even for those licensees that will implement
Sec. 50.69 in a phased manner. This means that the allowable potential
increase in risk must be determined in a cumulative way for all SSCs
being categorized under Sec. 50.69.
Section 50.69 is structured to maintain the design basis functional
requirements of the plant. These requirements (that maintain design
basis functional requirements) when considered in conjunction with the
requirements to provide reasonable confidence that the potential change
in risk is small (as previously discussed), also provide reasonable
confidence that safety margins are maintained. Specifically, licensees
are required to ensure with reasonable confidence that RISC-3 SSCs
remain capable of performing their design basis functions and these
SSCs must remain capable of performing their design basis function,
e.g., by providing a reliability that is not significantly degraded, to
provide reasonable confidence that any increases in CDF or LERF will be
acceptably small.
Section 50.69(c)(1)(iv) requires applicants and licensees to
perform evaluations to assess the potential impact on risk from changes
to treatment. Further, Sec. 50.69(d)(2) requires that the treatment
applied to RISC-3 SSCs be consistent with the categorization process.
For SSCs modeled in the PRA, the licensee or applicant might conduct a
risk sensitivity study that assesses the impact of changes in SSC
failure probabilities or reliabilities that might occur due to the
revised treatment. For example, a licensee could increase the failure
rates of RISC-3 SSCs by appropriate factors to provide insights into
the potential changes in risk that might result from reduced treatment
(e.g., reduced maintenance, testing, inspection, and quality
assurance). For other SSCs, other types of evaluations would be used to
provide the basis for concluding that the potential increase in risk
would be small. Under Sec. 50.69(b)(2)(iv), a licensee will need to
submit its basis supporting the evaluations that estimate the potential
change in risk. A licensee is required by Sec. 50.69(b)(2)(iv) to
consider potential effects of common-cause interaction susceptibility
and potential impacts from known degradation mechanisms.
The rule focuses on common-cause effects because significant
increases in common-cause failures could invalidate the evaluations
performed to show that any potential change in risk due to
implementation of Sec. 50.69 would be small. With respect to known
degradation mechanisms, this is an acknowledgment that certain
treatment requirements have evolved over time to deal with these
mechanisms (e.g., use of particular inspection techniques or
frequencies), and that when contemplating changes to treatment, the
lessons from this experience are to be taken into account.
For SSCs categorized by means other than PRA models, the licensee
needs to provide a basis to conclude that any potential increase in
risk that might result from reduced treatment would be small. These
requirements are included in Sec. 50.69 so that a licensee has a basis
for concluding that the evaluations performed to provide reasonable
confidence that only a acceptably small change in risk will result
remain valid.
In addition, the rule requires that implementation be performed for
an entire system or structure and not for selected components within a
system or structure. This required scope ensures that all safety
functions associated with a system or structure are properly identified
and evaluated when determining the safety significance of individual
components within a system or structure and that the entire set of
components that comprise a system or structure are considered and
addressed.
III.3.0 Treatment Requirements
The final rule applies treatment requirements to SSCs commensurate
with their safety significance.
III.3.1 RISC-1 and RISC-2 Treatment
For SSCs determined by the IDP to be safety significant (i.e.,
RISC-1 and RISC-2 SSCs), Sec. 50.69 maintains the current regulatory
requirements (i.e., it does not remove any requirements from these
SSCs) for special treatment. These current requirements are adequate
for addressing design basis performance of these SSCs. Additionally,
Sec. 50.69(d)(1) requires that sufficient treatment be applied to
support the credit taken for these SSCs for beyond design basis events.
For example, in developing the PRA model, a licensee must determine the
availability, capability, and reliability of RISC-1 and RISC-2 SSCs in
performing specific functions under various plant conditions. These
functions may be beyond the design basis for individual SSCs. Further,
the conditions under which those functions are to be performed may
exceed the design basis conditions for the applicable SSCs. Section
50.69(d)(1) requires the treatment applied to RISC-1 and RISC-2 SSCs to
be consistent with the performance credited in the categorization
process. This includes credit with respect to prevention and mitigation
of severe accidents. In some cases, licensees might need to enhance the
treatment applied to RISC-1 or RISC-2 SSCs to support the credit taken
in the categorization process, or conversely adjust the credit for
performance of the SSC in the categorization process to reflect actual
treatment practices and/or documented performance capability. In
addition, Sec. 50.69(e) requires monitoring and adjustment of
treatment processes or categorization decisions as needed based upon
operational experience.
III.3.2 RISC-3 Treatment
Section 50.69(d)(2) imposes requirements that are intended to
maintain RISC-3 SSC design basis capability. Although individually
RISC-3 SSCs are not significant contributors to plant safety, they do
perform functions necessary to respond to certain design basis events
of the facility. Thus, collectively, RISC-3 SSCs can be safety
significant and as such, it is important to maintain their design basis
functional capability. Maintenance of RISC-3 design basis functionality
is important to ensure that defense-in-depth and safety margins are
maintained. As a result, Sec. 50.69(d)(2) requires that licensees or
applicants ensure with reasonable confidence that RISC-3 SSCs remain
capable of performing their safety-related functions under design basis
conditions, including seismic conditions and environmental conditions
and effects throughout their service life. To support this requirement,
Sec. 50.69(d)(2) contains inspection, testing, and corrective action
requirements, and in addition requires that the treatment of RISC-3
SSCs be consistent with the categorization process. The requirements
are performance-based and give licensees the flexibility to implement
treatment that they have determined is needed, commensurate with the
low safety significance of the SSCs in order to provide reasonable
confidence that their safety-related functional capability is
maintained. In this context, ``reasonable confidence'' is a somewhat
reduced level of confidence as compared with the relatively high level
of confidence provided by the current special treatment requirements.
These alternative treatment requirements for RISC-3 SSCs represent a
relaxation of those special treatment requirements that are removed for
RISC-3 SSCs by the rule. For example, the alternative
[[Page 68020]]
treatment requirements for RISC-3 SSCs in Sec. 50.69 are less detailed
than provided in the special treatment requirements and allow
significantly more flexibility by licensees in treating RISC-3 SSCs.
The Commission is allowing greater flexibility and a lower level of
assurance to be provided for RISC-3 SSCs in recognition of their low
individual safety significance and this recognition includes a
consideration for the potential change in reliability that might occur
when treatment is reduced from what had previously been required by the
special treatment requirements.
In implementing the rule requirements, licensees will need to
obtain data or information sufficient to make a technical judgement
that RISC-3 SSCs will remain capable of performing their safety-related
functions under design basis conditions, and to enable the licensee to
take actions to restore equipment performance consistent with
corrective action requirements included in the rule.
Effective implementation of the treatment requirements should
result in reasonable confidence that RISC-3 SSCs will perform their
safety-related function under normal and design basis conditions. This
level of confidence is both less than that associated with RISC-1 SSCs,
which are subject to all special treatment requirements, and consistent
with the low individual safety significance of RISC-3 SSCs.
It is noted that changes that affect any non-treatment aspects of
an SSC (e.g., changes to the SSC design basis functional requirements)
are still required to be evaluated in accordance with other regulatory
requirements, such as Sec. 50.59. The Commission, in developing Sec.
50.69, is drawing a distinction between treatment (managed through
Sec. 50.69) and design changes (managed through other processes, such
as Sec. 50.59). As previously noted, this rulemaking is only risk-
informing the scope of special treatment requirements. The process and
requirements established in Sec. 50.69 do not extend to making changes
to the design basis functional requirements of SSCs.
III.3.3 RISC-4 Treatment
Section Sec. 50.69 does not impose any new treatment requirements
on RISC-4 SSCs. Instead, RISC-4 SSCs are simply removed from the scope
of any applicable special treatment requirements identified in Sec.
50.69(b)(1). This is justified in view of their low significance
considering both safety-related and risk information. Requirements
applicable to RISC-4 SSCs not removed by Sec. 50.69(b)(1) continue to
apply. Any changes (beyond changes to special treatment requirements)
must be made per existing design change control requirements including
Sec. 50.59, as applicable.
III.4.0 Removal of RISC-3 and RISC-4 SSCs From the Scope of Special
Treatment Requirements
Through the application of Sec. 50.69, RISC-3 and RISC-4 SSCs are
removed from the scope of the specific special treatment requirements
listed in Sec. 50.69(b)(1). The special treatment requirements were
originally imposed to provide a high level of assurance that safety-
related SSCs would perform when called upon with high reliability. As
previously noted, the requirements include extensive quality assurance
requirements and qualification testing requirements, as well as
inservice inspection and testing requirements. These requirements can
be quite demanding and expensive, as indicated in the data provided in
the regulatory analysis on procurement costs. The Commission concluded
that, in light of the low individual safety significance of RISC-3
SSCs, it is unnecessary to have the same high level of assurance that
they would perform as designed. This is because some increased
likelihood of their individual failure can be tolerated without
significant impact to safety. Thus, the Commission decided to remove
the RISC-3 and RISC-4 SSCs from those detailed, specific requirements
that provided the high level of assurance. However, the functional
requirements for these SSCs remain. As an example, a RISC-3 component
must still be designed to withstand any harsh environment it would
experience under a design basis event, but the NRC will not require
that this capability be demonstrated by a qualification test. Further,
the performance (and treatment) of these RISC-3 SSCs remain under
regulatory control, but in a different way. Instead of the special
treatment requirements, the Commission has set forth more general
requirements by which a licensee is to maintain functionality. These
requirements give the licensee more latitude in applying treatment to
maintain the design basis functional capability of the RISC-3 SSCs. The
more general requirements that the Commission is specifying for the
RISC-3 SSCs include inspection, testing, and corrective action, as a
means of maintaining functionality. As discussed elsewhere in the SOC
of this rule, the Commission concludes that the requirements in Sec.
50.69 will maintain adequate protection of public health and safety if
effectively implemented by licensees. Hence, implementation of Sec.
50.69 should result in a better focus for both the licensee and the
regulator on issues that pertain to plant safety and is consistent with
the Commission's policy statement for the use of PRA.
In some cases, the Commission concluded that the RISC-3 and RISC-4
SSCs could be removed from the scope of specific special treatment
requirements, while in other cases the Commission concluded that only
partial removal was appropriate. Finally, there was a set of
requirements initially identified as special treatment for which the
Commission is not removing RISC-3 and RISC-4 SSCs from their scopes.
These requirements are discussed in Section III.4.10.
III.4.1 Reporting Requirements Under 10 CFR Part 21 and Sec. 50.55(e)
Section 206 of the Energy Reorganization Act of 1974 (ERA) requires
the directors and responsible officers of nuclear power plant licensees
and firms supplying ``components of any facility or activity * * *
licensed or otherwise regulated by the Commission'' to ``immediately
report'' to the Commission if they have information that such facility,
activity, or basic components supplied to such facility or activity
either fails to comply with the AEA, or Commission rule, regulation,
order or license ``relating to substantial safety hazards,'' or
contains a ``defect which could create a substantial safety hazard * *
*'' Id., paragraph (a). Congress adopted Section 206 to ensure that
individuals, and responsible directors and officers of licensees and
firms supplying important components to nuclear power plants notify the
NRC in a timely fashion of potentially significant safety problems or
noncompliance with NRC requirements. The NRC then may assess the
reported information and take any necessary regulatory action in a
timely fashion to protect public health and safety or common defense
and security. Congress did not include definitions for the terms,
``components,'' ``basic components,'' or ``substantial safety hazard,''
in Section 206, but instead directed the Commission to issue
regulations defining these terms.
The Commission's regulations implementing Section 206 appear in 10
CFR part 21 and Sec. 50.55(e) for license holders and construction
permit holders, respectively. The Commission established definitions of
``basic component,'' ``defect,'' and ``substantial safety hazard'' in
part 21 on the premise that the deterministic regulatory paradigm
embedded in the Commission's regulations would
[[Page 68021]]
continue to be the appropriate basis for determining the safety
significance of an SSC, and therefore, the extent of the reporting
obligation under Section 206. This is most evident in the Sec. 21.3
definition of ``basic component,'' which is similar to the definition
of ``safety-related'' SSCs in Sec. 50.2 (originally embodied in Sec.
50.49). Part 21 also recognizes that Congress did not intend that every
potential noncompliance or ``defect'' in a component raises such
significant safety issues that the NRC must be informed of every
identified or potential noncompliance or defect. Instead, Congress
limited the Section 206 reporting requirement to those instances of
noncompliance and defects that represent a ``substantial safety
hazard.'' Thus, part 21 limits the reporting requirement to instances
of noncompliance and defects representing ``substantial safety
hazard,'' which part 21 defines as:
A loss of safety function to the extent there is a major reduction in
the degree of protection afforded to public health and safety for any
facility or activity licensed, other than for export, pursuant to parts
30, 40, 50, 60, 61, 63, 70, 71, or 72 of this chapter.
Finally, Part 21 establishes that a licensee or vendor should
``immediately report'' potential noncompliance or defects to the NRC in
a telephonic ``notification'' (see Sec. 21.3) within two (2) days of
receipt of information identifying a noncompliance or defect in a basic
component (see Sec. 21.21(d)). In addition, part 21 requires that
vendors/suppliers of basic components must make notifications to
purchasers or licensees of a reportable noncompliance or deviation
within five (5) working days of completion of evaluations for
determining whether noncompliance or deviation constitutes a
substantial safety hazard (see Sec. 21.21(b)). Thus, Part 21
establishes a reporting scheme for immediate reporting of the most
safety significant noncompliances and defects, as contemplated by
Section 206 of the ERA.
Section 50.69 substitutes a risk-informed approach for regulating
nuclear power plant SSCs for the current deterministic approach.
Therefore, it is necessary from the standpoint of regulatory coherence
to determine: (1) what categories of SSCs (i.e., RISC-1, RISC-2, RISC-
3, and RISC-4) should be subject to part 21 and Sec. 50.55(e)
reporting under Sec. 50.69 and whether changes to part 21 and/or Sec.
50.55(e) are necessary to ensure proper reporting of substantial safety
hazards caused by these SSCs; and (2) the appropriate reporting
obligations of licensees and vendors under Sec. 50.69, and whether
changes to part 21 and/or Sec. 50.55(e) are necessary to impose the
intended reporting obligations on these entities under Sec. 50.69.
III.4.1.1 RISC-1, RISC-2, RISC-3, and RISC-4 SSCs
After consideration of the underlying purposes of Section 206 and
the risk-informed approach embodied in Sec. 50.69 (which blends both
deterministic and risk information), the Commission believes that RISC-
1 SSCs should be subject to the reporting requirements in part 21 and
Sec. 50.55(e) because of their high safety significance. The NRC
should be informed of any potential defects or noncompliance with
respect to RISC-1 SSCs so that it may evaluate the significance of the
defects or noncompliance and take appropriate action. The fact that
properly-categorized RISC-1 SSCs in all likelihood fall within the
Commission's definition of ``basic components'' and are currently
subject to part 21 and Sec. 50.55(e) provides confirmation that the
Commission's determination is prudent.
Similarly, the Commission believes that SSCs categorized as RISC-4
should continue to be beyond the scope of, and not be subject to, part
21 and Sec. 50.55(e). SSCs properly categorized as RISC-4 have little
or no risk significance. It is highly unlikely that any significant
regulatory action would be taken by the NRC based upon information on
defects or instances of noncompliance in RISC-4 SSCs so reporting them
serves no regulatory purpose. Again, the fact that SSCs properly
categorized as RISC-4 do not otherwise fall within the definition of
``basic component'' and, therefore, are not subject to part 21 and
Sec. 50.55(e) provides some confirmation of the prudence of the
Commission's determination.
Thus, the most problematic issue from the standpoint of regulatory
coherence is determining the appropriate scope of reporting for RISC-2
and RISC-3 SSCs. For the following reasons, the Commission proposes
that neither RISC-2 nor RISC-3 SSCs be subject to part 21 and Sec.
50.55(e) reporting requirements.
The Commission begins by considering the regulatory objective of
Part 21 and Sec. 50.55(e) reporting under Section 206 and believes
that there are two parallel regulatory purposes inherent in these
reporting schemes. The first objective is to ensure that the NRC is
immediately informed of a potentially significant noncompliance or
defect in supplied components (in the broad sense of ``basic
components'' as defined in Sec. 21.3) so that the NRC may make a
determination if such a safety hazard requires that immediate NRC
regulatory action be taken at one or more nuclear power plants to
ensure adequate protection to public health and safety or common
defense and security. The second is to ensure that nuclear power plant
licensees are immediately informed of a potentially significant
noncompliance or defect in supplied components. This reporting allows a
licensee using these components to immediately evaluate the
noncompliance or defect to determine if a safety hazard exists at the
plant and take timely corrective action as necessary. In both cases,
the regulatory objective is limited to components that have the highest
significance with respect to ensuring adequate protection to public
health and safety and common defense and security and whose failure or
lack of proper functioning could create an imminent safety hazard so
that immediate evaluation of the situation and implementation of
necessary corrective action is necessary to ensure adequate protection.
In the context of a construction permit, the safety hazard is two-fold:
(1) That a noncompliance or defect could be incorporated into
construction where it could never be detected; and,
(2) That a noncompliance or defect would, upon initial operation
and without prior indications of failure, create a substantial safety
hazard.
The Commission believes that the regulatory objectives embodied in
part 21 and Sec. 50.55(e) reporting remain the same regardless of
whether the nuclear power plant is operating under the existing,
deterministic regulatory system or the alternative, risk-informed
system embodied in Sec. 50.69. In both cases, the reporting scheme
should focus on immediate reporting to the NRC and licensee of
potentially significant noncompliances and defects that could create a
substantial safety hazard requiring immediate evaluation and corrective
action to ensure continuing adequate protection. Accordingly, in
determining whether RISC-2 and RISC-3 SSCs should be subject to part 21
reporting, the Commission assessed whether failure or malfunction of
these SSCs could reasonably lead to a safety hazard so that immediate
evaluation of the situation and implementation of necessary corrective
action is necessary to ensure adequate protection.
For RISC-2 SSCs, the Commission does not believe their failure or
malfunction could reasonably lead to a safety hazard so that immediate
licensee and NRC evaluation of the situation and implementation of
necessary corrective action is necessary to ensure adequate
[[Page 68022]]
protection. Although a RISC-2 SSC may be of significance for particular
sequences and conditions, for the reasons discussed below, the
Commission believes that no RISC-2 SSC, in and of itself, is of such
significance that its failure or lack of function would necessitate
immediate notification and action by licensees and the NRC.
The categorization process embodied in Sec. 50.69 determines the
relative significance of SSCs, with those in RISC-1 and RISC-2 being
more significant than those in RISC-3 or RISC-4. This does not mean
that any RISC-2 SSC would rise to the level of necessitating immediate
action if defects were identified.
RISC-1 SSCs are viewed as being of sufficient safety significance
to require part 21 reporting. It is the capability provided by these
RISC-1 SSCs for purposes of satisfying safety-related functional
requirements that also leads to RISC-1 SSCs being safety-significant,
as these are key functions in prevention and mitigation of severe
accidents. Thus, RISC-1 SSCs are generally significant for a range of
events and conditions and, as the primary means of accident prevention
and mitigation, the Commission wants to continue to achieve the high
level of quality, reliability, preservation of margins, and assurance
of performance of current regulatory requirements.
By contrast, RISC-2 SSCs are less important than RISC-1 SSCs
because they do not play a role in prevention and mitigation of design
basis events (i.e., the SSCs that assure the integrity of the reactor
coolant boundary, the capability to shut down the reactor and maintain
it in a safe shutdown condition, or the capability to prevent or
mitigate the consequences of accidents that could result in potential
offsite exposures comparable to the applicable exposure guidelines set
forth in Sec. 50.34(a)(1) or Sec. 100.11). For example, they are not
part of the reactor protection system or engineered safety features
that perform critical safety functions such as reactivity control,
inventory control and heat removal. When viewed from a deterministic
standpoint, RISC-2 SSCs are not considered to rise to the level of a
potential substantial safety hazard. From the risk-informed
perspective, SSCs may end up classified as RISC-2 for a number of
reasons. The classification might occur because: (1) they contribute to
plant risk by initiating transients that could lead to severe accidents
(if multiple failures of other mitigating SSCs were to occur); or (2)
they can reduce risk by providing backup mitigation to RISC-1 SSCs in
response to an event.
The Commission recognizes that noncompliance by, or defects in,
RISC-2 SSCs, which could increase risk, such as by more frequent
initiation of a transient, may appear to constitute a ``substantial
safety hazard.'' However, upon closer examination, the Commission
believes otherwise. The risk significance of such ``transient-
initiating'' RISC-2 SSCs depends upon their frequency of initiation,
with resultant consequences depending upon the failure of multiple
other components of varying types in different systems. Further, their
risk significance, as identified by the categorization process, is a
result of the reliability (failure rates) currently being achieved for
these SSCs treated as commercial-grade components, which includes the
possibility of noncompliances and defects. Because requirements on
RISC-2 SSCs are not being reduced, there is no reason to believe that
their performance would degrade as a result of implementation of Sec.
50.69. In fact, by better understanding of their safety significance,
and through the added requirements in this rule for RISC-2 SSCs to
achieve consistency between their categorization and treatment,
performance should, at a minimum, be maintained and in some cases,
enhanced. As discussed in Sections III.3 and III.5 of this rule, the
Commission is imposing additional regulatory controls on RISC-2 SSCs to
prevent their performance from degrading. In addition, the Commission
is requiring: (1) that licensees evaluate treatment being applied for
consistency with the performance credited in the categorization; (2)
monitoring of the performance of these SSCs; (3) corrective actions;
and (4) reporting when a loss of a safety significant function occurs.
Thus, there are requirements for corrective action by the licensee if
noncompliances involving these SSCs are identified. The Commission
concludes that these requirements are sufficient to preclude the need
for Part 21 reporting, because no RISC-2 SSC is so significant as to
necessitate immediate Commission (or licensee) action.
For RISC-2 SSCs that provide backup mitigation to RISC-1 SSCs, the
Commission also finds it prudent and desirable from a risk-informed
standpoint to provide an enhanced level of assurance that RISC-2 SSCs
can perform their safety significant functions, but the failure or
malfunction of these RISC-2 SSCs does not raise a concern about
imminent safety hazards. Moreover, over the last several years, the
current fleet of power reactors have been subjected to a number of risk
studies, including NUREG-1150, and other generic and plant-specific
reviews. While some safety improvements have been identified as a
result of these reviews, none has been of such significance as to
require immediate action. This essentially means that no SSCs
categorized as RISC-2 would rise to the level of significance that
their failure or lack of functionality would constitute a substantial
safety hazard requiring immediate NRC regulatory action. For example,
in the case of two key risk scenarios, Station Blackout and Anticipated
Transient Without Scram, the Commission imposed regulatory requirements
to reduce risk from these events. However, the rules were issued as
cost-beneficial safety improvements. The Commission believes its
conclusion about the relative significance of RISC-2 SSCs is also
supported by plant-specific risk studies, such as the Individual Plant
Examination (IPE) and Individual Plant Examination of External Events
(IPEEE),\2\ conducted to identify (and correct) any plant-specific
vulnerabilities to severe accident risk. NRC's review of the licensee
submittals has not identified any situations requiring immediate action
for protection of public health and safety. In addition, as part of
license renewal reviews, the NRC reviews severe accident mitigation
alternatives (SAMAs), to identify and evaluate plant design changes
with the potential for improving severe accident safety performance. In
the license renewals completed to date, only a few candidate SAMAs have
been found to be cost-beneficial (and none were considered necessary to
provide adequate protection of public health and safety).
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\2\ In Generic letter 88-20, dated November 23, 1988, licensees
were requested to perform individual plant examinations to identify
plant-specific vulnerabilities to severe accidents that might exist
in their facilities and report the results to the Commission. As
part of their review and report, licensees were asked to determine
any cost-beneficial improvements to reduce risk. In supplement 4 to
the generic letter dated June 28, 1991, this request was extended to
include external events (e.g., earthquakes, fires, floods). The NRC
staff reviewed the plant-specific responses and prepared a staff
evaluation report on each submittal. Further, the set of results
were presented in NUREG-1560, IPE Program: Perspectives on Reactor
Safety and Plant Performance. A similar report on IPEEE results was
issued as NUREG-1742. In addition, as discussed in SECY-00-0062, the
staff has conducted IPE follow-up activities with owners groups and
licensees to confirm that identified improvements have been
implemented and if any other actions were warranted.
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In light of risk assessments and actions that have already been
implemented, the Commission believes there would be no SSCs categorized
[[Page 68023]]
under Sec. 50.69 as RISC-2 whose failure would represent a significant
and substantial safety hazard so that immediate notification under Part
21 and NRC regulatory action is required. Accordingly, the results of
these risk assessments provide additional confidence to the Commission
that Part 21 requirements need not be imposed on RISC-2 SSCs.
The Commission also considered if notification of component defects
should be required from the perspective of other potentially-affected
licensees. The set of SSCs that are RISC-2 would vary from site to site
because it depends upon the specifics of plant design and operation,
particularly for the balance-of-plant which typically differs more from
plant to plant than does the nuclear steam supply portion. Further, the
suppliers of these components would vary. Therefore, the specific type
of notifications under Part 21, for the purposes of NRC assessment of
generic implications of component defects and to assure notification of
licensees with the same components in service, would not fulfill a
useful regulatory function. The Commission notes that although Part 21
and Sec. 50.55(e) (component defect) reporting will not be required
for RISC-2 SSCs, Sec. 50.69(g) contains enhanced reporting
requirements applicable to loss of system function attributable to,
inter alia, failure or lack of function of RISC-2 SSCs. This is
discussed in greater detail in Section III.5.
Therefore, because of the more supporting role that the RISC-2 SSCs
play with respect to ensuring critical safety functions, a
noncompliance or defect in a RISC-2 SSC would not result in a
substantial safety hazard such that immediate licensee and NRC
evaluation of the situation and implementation of corrective action is
necessary to ensure adequate protection. Thus, the Commission believes
that a noncompliance or defect in a RISC-2 SSC does not constitute a
substantial safety hazard for which reporting is necessary under Part
21. Accordingly, the Commission concludes that reporting requirements
to comply with Section 206 of the ERA are not necessary for RISC-2 SSCs
and that the scope of Part 21 and Sec. 50.55(e) reporting requirements
exclude RISC-2 SSCs.
The Commission also concludes that RISC-3 SSCs should not be
subject to Part 21 and Sec. 50.55(e) reporting. A failure of a
properly-categorized RISC-3 SSC should result in only a small change in
risk and should not result in a major degradation of essential safety-
related equipment (see NUREG-0302, Rev. 1) \3\. As previously
discussed, the body of regulatory requirements (i.e., the retained
requirements and the requirements contained in this rule) are
sufficient, if effectively implemented, so that simultaneous failures
in multiple systems (as would be necessary to lead to a substantial
safety hazard involving RISC-3 SSCs) would not occur. Further, the
broad applicability of information from a single RISC-3 SSC that would
be provided under Part 21 and Sec. 50.55(e) reporting would be
questionable because of the significant changes in treatment for RISC-3
SSCs allowed under Sec. 50.69. Accordingly, the Commission concludes
that RISC-3 SSCs should not be subject to reporting requirements of
Part 21 and Sec. 50.55(e).
---------------------------------------------------------------------------
\3\ NUREG-0302, ``Remarks Presented (Questions and Answers
Discussed) At Public Regional Meetings to Discuss Regulations (10
CFR Part 21) for Reporting of Defects and Noncompliances.'' Copies
of NUREGs may be purchased from the Superintendent of Documents,
U.S. Government Printing Office, P.O. Box 37082, Washington DC
20013-7082. Copies are also available from the National Technical
Information Service, 5285 Port Royal Road, Springfield, VA 22161. A
copy is also available for inspection and/or copying for a fee at
the NRC Public Document Room, One White Flint North, 11555 Rockville
Pike, Public File Area O1-F21, Rockville, MD.
---------------------------------------------------------------------------
The Commission concludes that Part 21 reporting requirements extend
only to RISC-1 SSCs because they are important in ensuring public
health and safety. RISC-2 SSCs are not subject to reporting because
they play a lesser role than RISC-1 SSCs in protection of public health
and safety and with the significant changes in treatment allowed under
Sec. 50.69, no regulatory purpose would be served by Part 21 reporting
(as previously discussed). Individually, RISC-3 and RISC-4 SSCs have
little or no risk significance and no regulatory purpose would be
served by subjecting RISC-3 and RISC-4 SSCs to Part 21 and Sec.
50.55(e).
The Commission does not believe that any changes to Part 21 or
Sec. 50.55(e) are necessary to accomplish its conclusions with respect
to RISC-2 and RISC-3 SSCs. The Commission believes this is consistent
with the statutory requirements in Section 206 of the ERA. Section 206
does not contain any definition of ``substantial safety hazard,'' but
contains a direction to the Commission to define this term by
regulation. Nothing in the legislative history suggests that Congress
had in mind a fixed and unchanging concept of ``substantial safety
hazard'' or that the term was limited to deterministic regulatory
principles. Hence, the Commission has broad discretion and authority to
determine the appropriate scope of reporting under Section 206. The
Commission believes that the current definition of ``substantial safety
hazard'' in Sec. 21.3 is broadly written to permit the Commission to
interpret it as applying, in the context of a risk-informed regulatory
approach, only to RISC-1 SSCs. Section 50.69 embodies a risk-informed
regulatory paradigm that is different in key respects from the
Commission's historical deterministic approach and applies the risk-
informed approach to classifying a nuclear power plant's SSCs according
to the SSC's risk significance. SSCs that are classified as RISC-1 are
those that represent the most important SSCs from both a risk and
deterministic standpoint: they perform the key functions of preventing,
controlling, and mitigating accidents and controlling risk. Failure of
RISC-1 SSCs represent, from a risk-informed regulatory perspective, the
most important and significant safety concerns (i.e., a ``substantial
safety hazard''). Therefore, the Commission believes that, in the
context of the risk-informed regulatory approach embodied in Sec.
50.69, it is reasonable for the Commission to interpret ``substantial
safety hazard'' as applying only to RISC-1 SSCs and that reporting
under Section 206 may be limited to RISC-1 SSCs.
The Commission considered two alternative approaches for limiting
the reporting requirements in Part 21 and Sec. 50.55(e) to RISC-1
SSCs:
(1) Interpreting ``basic component'' to encompass a risk-informed
view of what SSCs the term encompasses; and,
(2) Including a second definition of ``basic component'' in Sec.
21.3, which would apply only to those portions of a plant that have
been categorized in accordance with Sec. 50.69 and would be defined as
an SSC categorized as RISC-1 under Sec. 50.69.
The Commission does not believe that the Part 21 definition of
``basic component'' may easily be read as simultaneously permitting
both a deterministic concept of basic component and risk-informed
concept, inasmuch as the Part 21 definition was drawn from, and was
intended to be consistent with the definition of ``safety-related SSC''
in Sec. 50.2. The Sec. 50.2 definition of ``afety-related SSC''
refers to the ability of the SSC to remain functional during ``design
basis events.'' The term, ``design basis events'' in Commission
practice has referred to the deterministic approach of defining the
events and conditions (e.g., shutdown, normal operation, and accident)
for which an SSC is expected to function (or not fail). Identification
of design basis events is inherently different conceptually when
compared to a risk-
[[Page 68024]]
informed approach, which attempts to identify all possible outcomes (or
a reasonable surrogate) and assign a probability to each outcome and
consequence before integrating the probability of the total set of
outcomes. The Commission rejected the second approach of adopting an
alternative definition of ``basic component,'' because a change to the
definition in Sec. 21.3 could be misunderstood as a change to the
reporting requirements for licensees who choose not to comply with
Sec. 50.69.
III.4.1.2 Reporting Obligations of Vendors for RISC-3 SSCs
The reporting requirements of Section 206 apply to individuals,
directors, and responsible officers of a firm constructing, owning,
operating or supplying the basic components of any NRC-licensed
facility or activity. Nuclear power plant licensees and nuclear power
plant construction permit holders who are subject to reporting under
Section 206, Part 21, and Sec. 50.55(e) will continue to provide for
such reporting by those entities. Section 206 also imposes a reporting
obligation on ``vendors'' (i.e., firms who supply basic components to
nuclear power plant licensees and construction permit holders). The
Commission does not intend to change the reporting obligations under
Part 21 or Sec. 50.55(e) for licensees, construction permit holders,
or vendors with respect to RISC-1 SSCs and the Commission does not
intend to require reporting under Part 21 and Sec. 50.55(e) for RISC-
2, RISC-3 or RISC-4 SSCs.
Thus, a vendor who supplied a safety-related component to a
licensee that was subsequently classified by the licensee as RISC-3
would no longer be legally obligated to comply with Part 21 or Sec.
50.55(e) reporting requirements. However, as a practical matter that
vendor would likely continue to comply with Part 21 or Sec. 50.55(e).
Vendors are informed of their Part 21 or Sec. 50.55(e) obligations as
part of the contract supplying the basic component to the licensee/
construction permit holder. Vendors supplying basic components that
have been categorized as RISC-3 at the time of contract ratification
would know that they have no Part 21 or Sec. 50.55(e) obligations.
However, vendors that provide (or in the past provided) safety-related
SSCs would not know, absent communication from the licensee or
construction permit holder implementing Sec. 50.69, whether the SSCs
that they provided under contract as safety-related are now categorized
as RISC-3, thereby removing the vendor's reporting obligation under
Part 21 or Sec. 50.55(e). Failing to inform a vendor that a safety-
related SSC that it provided is no longer subject to Part 21 or Sec.
50.55(e) reporting because of its reclassification as a RISC-3 SSC
could result in unnecessary reporting to the licensee and the NRC. It
may also result in unnecessary expenditure of resources by the vendor
in determining whether a problem with a supplied SSC rises to the level
of a reportable defect or noncompliance under the existing provisions
of Part 21 and Sec. 50.55(e).
To address the potential for unnecessary reporting under Sec.
50.69, the Commission considered including a new requirement in either
Sec. 50.69 or Part 21 and Sec. 50.55(e). The new provision would
require the licensee or construction permit holder to inform a vendor
that a safety-related SSC that it provided has been categorized as
RISC-3. After consideration, the Commission believes that it is
unlikely that this provision would result in any great reduction in the
potential scope of reporting by vendors. The NRC does not receive many
Part 21 reports, so the overall reporting burden to be reduced may be
insubstantial. Furthermore, the Commission believes that the proposal
could cause confusion, inasmuch as a vendor may supply many identical
components to a licensee/holder, with some of the items intended for
use in SSCs categorized as RISC-3 and other items intended in
nonsafety-related applications. A vendor would have some difficulty in
determining whether the problem with the supplied SSC potentially
affects the SSC categorized as RISC-3 (as opposed to the supplied SSC
used in nonsafety-related applications). The Commission also believes
there may be some value in notification of the NRC when defects are
identified, as they may reveal issues about the quality processes or
implications for basic components at other facilities. Finally, the NRC
notes that the vendor has already been compensated by the licensee for
the burden associated with Part 21 and Sec. 50.55(e) as part of the
initial procurement process. For these reasons, the Commission is not
adopting a provision in Sec. 50.69, Part 21, or Sec. 50.55(e)
requiring a licensee or construction permit holder to inform a vendor
of safety-related SSCs that its SSCs have been categorized as RISC-3.
III.4.1.3 Criminal Liability Under Section 223.b. of the AEA
As discussed earlier, Section 206 of the ERA authorizes the
imposition of civil penalties for a licensee's and vendor's failure to
report instances of noncompliance or defects in ``basic components''
that create a ``substantial safety hazard.'' However, in addition to
the civil penalties authorized by Section 206, criminal penalties may
be imposed under Section 223.b. of the AEA on an individual director,
officer, or employee of a firm that supplies components to a nuclear
power plant, that knowingly and willfully violate regulations that
results (or could have resulted) in a ``significant impairment of a
basic component * * *.'' Licensees, applicants, and vendors should note
the difference in the definition of ``basic component'' in Part 21
versus the definition set forth in Section 223.b:
For the purposes of this subsection, the term ``basic component''
means a facility structure, system, component or part thereof necessary
to assure--
(1) The integrity of the reactor coolant pressure boundary;
(2) The capability to shutdown the facility and maintain it in a
safe shutdown condition; or
(3) The capability to prevent or mitigate the consequences of
accidents that could result in an unplanned offsite release of
quantities of fission products in excess of the limits established by
the Commission.
The U.S. Department of Justice is responsible for prosecutorial
decisions involving violations of Section 223.b.
III.4.1.4 Posting Requirements
Both AEA Section 223.b and ERA Section 206 require posting of their
statutory requirements at the premises of all licensed facilities. This
is implemented through 10 CFR parts 19 and 21.
As a result of implementation of Sec. 50.69, rights and
responsibilities of licensee workers would be slightly different. For
instance, SSCs categorized as RISC-3 would no longer be subject to Part
21. However, RISC-1 SSCs (and ``safety-related'' SSCs not yet
categorized per Sec. 50.69) are subject to the Part 21 requirements.
No additional responsibilities for identification or notification are
involved. The supporting information, such as procedures to be made
available to workers, would need to reflect the reduction in scope of
requirements. For the reasons already mentioned, the Commission
concludes that there would be no impact on vendors with respect to
posting requirements in that these changes in categorization would be
``transparent'' to them as suppliers.
III.4.2 Section 50.49 Environmental Qualification of Electrical
Equipment
The general requirement that certain SSCs be designed to be
compatible with environmental conditions associated
[[Page 68025]]
with postulated accidents is contained in GDC-4. Section 50.49 was
written to provide specific programmatic requirements for a
qualification program and documentation for electrical equipment, and
thus, is a special treatment requirement.
Section 50.49(b) imposes requirements on licensees to have an
environmental qualification program that meets the requirements
contained therein. It defines the scope of electrical equipment
important to safety that must be included under the environmental
qualification program. Further, this regulation specifies methods to be
used for qualification of the equipment for identified environmental
conditions and documentation requirements.
RISC-3 and RISC-4 SSCs are removed from the scope of the
requirements of Sec. 50.49 by Sec. 50.69(b)(2)(ii). For SSCs
categorized as RISC-3 or RISC-4, the Commission has concluded that for
low safety significant SSCs, additional assurance, such as that
provided by the detailed provisions in Sec. 50.49 for testing,
documentation files and application of margins, are not necessary (for
the reasons stated in Section III.4.0). The requirements in GDC-4 as
they relate to RISC-3 and RISC-4 SSCs, and the design basis
requirements for these SSCs, including the environmental conditions
such as temperature and pressure, remain in effect. Thus, these SSCs
must continue to remain capable of performing their safety-related
functions under design basis environmental conditions.
III.4.3 Section 50.55a(f), (g), and (h) Codes and Standards
Section 50.69(b)(2)(iv) removes RISC-3 SSCs from the scope of
certain provisions of Sec. 50.55a, relating to Codes and Standards.
The provisions being removed are those that relate to ``treatment''
aspects, such as inspection and testing, but not those pertaining to
design requirements established in Sec. 50.55a. Each of the
subsections being removed is discussed in the paragraphs below.
Section 50.55a(f) incorporates by reference provisions of the ASME
Code, as endorsed by NRC, that contains inservice testing requirements.
These are special treatment requirements. Through this rulemaking,
RISC-3 SSCs are removed from the scope of these requirements and
instead are subject to the requirements in Sec. 50.69(d)(2). For the
reasons discussed in Section III.4.0, the Commission has determined
that for low safety significant SSCs, it is not necessary to impose the
specific detailed provisions of the Code, as endorsed by NRC, and these
requirements can be replaced by the more ``high-level'' alternative
treatment requirements, which allow greater flexibility to licensees in
implementation.
Section 50.55a(g) incorporates by reference provisions of the ASME
Code, as endorsed by NRC, that contain the inservice inspection, and
repair and replacement requirements for ASME Class 2 and Class 3 SSCs.
The Commission will not remove the repair and replacement provisions of
the ASME Code required by Sec. 50.55a(g) for ASME Class 1 SSCs, even
if they are categorized as RISC-3, because those SSCs constitute
principal fission product barriers as part of the reactor coolant
system or containment. For Class 2 and Class 3 SSCs that are shown to
be of low safety significance and categorized as RISC-3, the additional
assurance obtained from the specific provisions of the ASME Code is not
considered necessary. However, the Commission has not removed the
requirements for fracture toughness specified for ASME Class 2 and
Class 3 SSCs because fracture toughness is a significant design
parameter for the material used to construct the SSC. Fracture
toughness is a property of the material that prevents premature failure
of an SSC at abrupt geometry changes, or at small undetected flaws.
Adequate fracture toughness of SSCs is necessary to prevent common
cause failures due to design basis events, such as earthquakes.
Section 50.55a(h) incorporates by reference the requirements in
either IEEE 279, ``Criteria for Protection Systems for Nuclear Power
Generating Stations,'' or IEEE 603-1991, ``IEEE Standard Criteria for
Safety Systems for Nuclear Power Generating Stations.'' Within these
IEEE standards are special treatment requirements. Specifically,
Sections 4.3 and 4.4 of IEEE 279 and Sections 5.3 and 5.4 of IEEE 603-
1991 contain quality and environmental qualification requirements.
RISC-3 SSCs are being removed from the scope of this special treatment
requirement.
III.4.4 Section 50.65 Monitoring the Effectiveness of Maintenance
The Commission is removing RISC-3 and RISC-4 SSCs from the scope of
the requirements of Sec. 50.65 (except for paragraph (a)(4)). The
basis for this removal is provided in Section III.4.0 and the following
discussion.
Section 50.65, the Maintenance Rule, imposes requirements for
licensees to monitor the effectiveness of maintenance activities for
safety significant plant equipment to minimize the likelihood of
failures and events caused by the lack of effective maintenance.
Specifically, Sec. 50.65 requires the performance of SSCs defined in
Sec. 50.65(b) to either be monitored against licensee established
goals in a manner sufficient to provide confidence that the SSCs are
capable of fulfilling their intended functions, or demonstrated to be
effectively controlled through the performance of appropriate
preventative maintenance. The rule further requires that where
performance does not match the goals, appropriate corrective action
shall be taken. Included within the scope of Sec. 50.65(b) are SSCs
that are relied upon to remain functional during design basis events or
in emergency operating procedures and nonsafety-related SSCs whose
failure could result in the failure of a safety function or cause a
reactor scram or activation of a safety-related system.
Sections 50.65(a)(1), (a)(2), and (a)(3) impose action
requirements; thus, they are special treatment requirements. Upon
implementation of Sec. 50.69, a licensee is not required to apply
maintenance rule monitoring, goal setting, corrective action, alternate
demonstration, or periodic evaluation treatments required by Sec.
50.65(a)(1), (a)(2), and (a)(3) to RISC-3 and RISC-4 SSCs. The rule
includes provisions for a licensee to use performance information to
feedback into its processes to adjust treatment (or categorization)
when results so indicate in Sec. 50.69(e)(3). However, this
requirement does not require the specific monitoring and goal setting
as required in Sec. 50.65, in consideration of the lower safety
significance of these SSCs.
RISC-1 and RISC-2 SSCs that are currently within the scope of Sec.
50.65(b) remain subject to existing maintenance rule requirements.
Furthermore, Sec. 50.69(e)(2) requires additional monitoring,
evaluation and appropriate action for these SSCs.
The removal of RISC-3 and RISC-4 SSCs from the scope of
requirements does not include Sec. 50.65(a)(4), which contains
requirements to assess and manage the increase in risk that may result
from maintenance activities. The requirements in Sec. 50.65(a)(4)
remain in effect. Section 50.65(a)(4) already includes provisions by
which a licensee can limit the scope of the assessment required to SSCs
that a risk-informed evaluation process has shown to be significant to
public health and safety. Thus, there is no need to revise the
requirements to permit a licensee to apply requirements commensurate
with SSC safety-significance.
[[Page 68026]]
III.4.5 Sections 50.72 and 50.73 Reporting Requirements
This rule removes the requirements in Sec. 50.72 and Sec. 50.73
for RISC-3 and RISC-4 SSCs. Sections 50.72 and 50.73 contain
requirements for licensees to report events involving certain SSCs.
These reporting requirements are special treatment requirements. The
NRC requires event reports in part so that it can follow-up on
corrective action for these circumstances. Through this rulemaking, the
Commission is removing RISC-3 and RISC-4 SSCs from the scope of these
requirements. The broad applicability of information obtained under
Sec. 50.72 and Sec. 50.73 for RISC-3 SSCs would be questionable
because of the significant changes in treatment allowed under Sec.
50.69 (see the similar discussion for Part 21 in Section III.4.1.1).
Therefore, the Commission does not consider the burden associated with
reporting events or conditions only affecting these SSCs to be
warranted.
III.4.6 10 CFR Part 50, Appendix B Quality Assurance Requirements
This rule removes RISC-3 and RISC-4 SSCs from the scope of
requirements in Appendix B to 10 CFR part 50. Appendix B contains
requirements for a quality assurance program meeting specified
attributes. The intent of Appendix B to 10 CFR part 50, and the
complementary regulations, is to provide quality assurance requirements
for the design, construction, and operation of nuclear power plants.
The quality assurance requirements of Appendix B are to provide
adequate confidence that an SSC will perform satisfactorily in service.
These requirements were developed to be applied to safety-related SSCs.
In the implementation of Appendix B, a licensee is bound to detailed
and prescriptive quality requirements to apply to activities affecting
those SSCs. As such, these requirements meet the Commission's
definition of special treatment requirements. These requirements are
removed from application to RISC-3 and RISC-4 SSCs because their low
individual safety significance does not warrant the level of quality
requirements that currently exist with Appendix B.
III.4.7 10 CFR Part 50, Appendix J Containment Leakage Testing
Section 50.69(b)(1)(x) removes a subset of RISC-3 and RISC-4 SSCs
from the scope of the requirements in Appendix J to 10 CFR part 50 that
pertain to containment leakage testing. Specifically, RISC-3 and RISC-4
SSCs that meet specified criteria in Sec. 50.69(b)(1)(x) are removed
from the scope of the requirements for Type B and Type C testing. It is
important to note that this removes only the Appendix J leakage testing
requirements from these SSCs. These SSCs must still be capable of
performing their design basis functions (i.e., to close or isolate
containment). The basis for the removal of the Appendix J leakage
testing requirements follows.
One of the conditions of all operating licenses for water-cooled
power reactors as specified in Sec. 50.54(o), is that primary reactor
containments shall meet the containment leakage test requirements set
forth in Appendix J to 10 CFR part 50. These test requirements provide
for preoperational and periodic verification by tests of the leak-tight
integrity of the primary reactor containment, and systems and
components that penetrate containment of water-cooled power reactors
and establish the acceptance criteria for these tests. As such, these
tests are special treatment requirements. The purposes of the tests are
to assure that:
(1) Leakage through the primary reactor containment, or through
systems and components penetrating primary containment, shall not
exceed allowable leakage rate values as specified in the technical
specifications; and
(2) Periodic surveillance of reactor containment penetrations and
isolation valves is performed so that proper maintenance and repairs
are made during the service life of the containment, and systems and
components penetrating primary containment.
Appendix J includes two options; Option A and Option B. Option A
includes prescriptive requirements while Option B identifies
performance-based requirements and criteria for preoperational and
subsequent periodic leakage rate testing. A licensee may choose either
option for meeting the requirements of Appendix J.
The discussion contained in Appendix J to 10 CFR part 50 can be
divided into two categories. Parts of Appendix J contain testing
requirements. Other parts contain information, such as definitions or
clarifications, necessary to explain the testing requirements. A review
of Appendix J did not identify any technical requirements other than
those describing the methods of the required testing. Therefore,
Appendix J was considered to be, in its entirety, a special treatment
requirement.
Although the 1995 revision to Appendix J was characterized as risk-
informed, the changes were not as extensive as those expected by
inclusion of Appendix J within the scope of Sec. 50.69. The 1995
revision to Appendix J primarily decreased testing frequencies, whereas
risk-informing the scope of SSCs that are subject to Appendix J testing
removes some components from testing (i.e., to the extent that defense-
in-depth is maintained in accordance with the risk-informed
categorization process).
III.4.7.1 Types of Tests Required by Appendix J
Appendix J testing is divided into three types: Type A, Type B, and
Type C. Type A tests are intended to measure the primary reactor
containment overall integrated leakage rate after the containment has
been completed and is ready for operation and at periodic intervals
thereafter. Type B tests are intended to detect local leaks and to
measure leakage across each pressure-containing or leakage-limiting
boundary. Primary reactor containment penetrations required to be Type
B tested are identified in Appendix J. Type C tests are intended to
measure containment isolation valve (CIV) leakage rates. The
containment isolation valves required to be Type C tested are
identified in Appendix J.
III.4.7.2 Reduction in Scope for Appendix J Testing
Type A Testing: The Commission is not changing the Type A testing
requirements of Appendix J.
Type B Testing: The Commission is not changing the Type B testing
requirements for air lock door seals, including door operating
mechanism penetrations that are part of the containment pressure
boundary and doors with resilient seals or gaskets, except for seal-
welded doors. However, the Commission concludes that Type B testing is
not necessary for other penetrations that are determined to be of low
safety significance and that meet one or both of the following
criteria:
1. Penetrations pressurized with the pressure being continuously
monitored.
2. Penetrations are 1 inch nominal size or less.
Type C Testing: The Commission concludes that Type C testing is not
necessary for valves that are determined to be of low safety
significance and that meet one or more of the following criteria:
1. The valve is required to be open under accident conditions to
prevent or mitigate core damage events.
2. The valve is normally closed and in a physically closed, water-
filled system.
3. The valve is in a physically closed system whose piping pressure
rating exceeds the containment design
[[Page 68027]]
pressure rating and is not connected to the reactor coolant pressure
boundary.
4. The valve size is 1-inch nominal pipe size or less.
The Commission has made a determination that the size specified in
Sec. 50.69(b)(x) and identified above is acceptable. At this time, the
NRC has not determined that a larger size is acceptable for application
to Sec. 50.69, nor has the NRC received such a proposal. At this time,
for the Commission to entertain a larger penetration/CIV size, and
subsequently revise the rule language to reflect any such review
(assuming that such a size is acceptable) would likely cause the NRC to
re-notice Sec. 50.69 for stakeholder comment. Licensees and applicants
are free to pursue exemptions (to Sec. 50.69(b)(x)) to this criteria
if they conclude a larger penetration opening can be justified for
their containment design. If such a proposal is ultimately reviewed and
accepted, and can be applied generically, the NRC will consider a
revision to Sec. 50.69 to reflect the new criteria.
III.4.7.3 Basis for Reduction of Scope
The first category of penetrations which are excluded from Type B
testing are penetrations that are pressurized with the pressures in the
penetrations being continuously monitored by licensees. This monitoring
would detect significant leakage from the penetrations. The monitoring
of the pressures in the penetrations, in conjunction with the
requirements for RISC-3 SSCs (including taking corrective action when
an SSC fails), ensures with reasonable confidence, without the need for
Type B testing, that these penetrations are functional.
The second category of penetrations excluded from Type B testing
are penetrations that are 1 inch nominal size or less. These
penetrations do not contribute to large early releases. Accordingly,
the failure of such penetrations does not contribute in a significant
way to safety or increased risk. The Commission concludes that such
penetrations will not be subject to Type B testing.
Regarding Type C containment leakage testing, the Commission finds
that for the four categories of containment isolation valves identified
in Sec. 50.69(b)(1)(x), the removal of Type C testing requirements is
reasonable because even without Type C testing, the probability of
significant leakage during an accident (i.e., leakage to the extent
that public health and safety is affected) is small.
Appendix J to 10 CFR part 50 deals only with leakage rate testing
of the primary reactor containment and its penetrations. It assumes
that CIVs are in their safe position. No failure is assumed that causes
the CIVs to be open when they are supposed to be closed. The valve
would be open if needed to transmit fluid into or out of containment to
mitigate an accident or closed if not needed for this purpose. For
purposes of this evaluation, it is assumed that an open valve is
capable of being closed. The licensee or applicant implementing Sec.
50.69 must apply treatment to RISC-3 CIVs that ensures with reasonable
confidence that those valves are capable of performing their safety-
related function to close under design basis conditions. Testing to
ensure the capability of CIVs to reach their safe position is not
within the scope of Appendix J and as such is not within the scope of
this evaluation. Therefore, the valves addressed by this evaluation are
considered to be closed, but may be leaking. The increase in risk due
to these SSCs being removed from the scope of Appendix J requirements
is negligible.
The acceptability of the removal of Appendix J leakage testing for
the RISC-3 CIVs is based on the assumption that those valves are
capable of achieving the full seated position by means of the actuator.
Therefore, even though a RISC-3 CIV might be exempt from Appendix J
leakage testing, the RISC-3 CIV must meet the treatment requirements in
Sec. 50.69(d) to provide reasonable confidence that the CIV can
perform its safety function (e.g., to close) under design basis
conditions. Because it is likely that most CIVs will be categorized as
RISC-3, the licensee or applicant must evaluate the proposed change in
the treatment of RISC-3 CIVs to ensure that defense-in-depth is
maintained by ensuring with reasonable confidence that the RISC-3 CIVs
are capable of performing their safety-related functions under design
basis conditions. Although the licensee or applicant is allowed
flexibility in addressing this issue, the rule requires that the
licensee or applicant ensure with reasonable confidence the capability
of RISC-3 CIVs to perform their safety functions to maintain defense-
in-depth as discussed in RG 1.174.
Past studies (e.g., NUREG-1150, ``Severe Accident Risks: An
Assessment for Five U.S. Nuclear Power Plants; Final Summary Report,''
dated December 1990) show that the overall reactor accident risks are
not sensitive to variations in containment leakage rate. This is
because reactor accident risk is dominated by accident scenarios in
which the containment either fails or is bypassed. These very low
probability scenarios dominate predicted accident risks due to their
high consequences.
The Commission examined the effect of containment leakage on risk
in more detail as part of the Appendix J to 10 CFR part 50, Option B,
rulemaking. The results of these studies are applicable to this
evaluation. NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' dated September 1995, calculated the containment leakage
necessary to cause a significant increase in risk and found that the
leakage rate must typically be approximately 100 times the Technical
Specification leak rate, La. It is improbable that even the
leakage of multiple valves in the categories under consideration would
exceed this amount. Operating experience shows that most measured leaks
are much less than 100 times La. A more direct estimate of
the increase in risk for the revision to Appendix J can be obtained
from the Electric Power Research Institute (EPRI) report TR-104285,
``Risk Impact Assessment of Revised Containment Leak Rate Testing
Intervals,'' dated August 1994. This report examined the change in the
baseline risk (as determined by a plant's IPE risk assessment) due to
extending the leakage rate test intervals. For the pressurized water
reactor (PWR) large dry containment examined in the EPRI report, for
example, the percent increase in baseline risk from extending the Type
C test interval from 2 years to 10 years was less than 0.1 percent.
While this result was for a test interval of 10 years vs. the current
proposal to do no more Type C testing of the subject valves for the
life of a plant, the analysis may reasonably apply to this situation
because it contains several conservative assumptions that offset the
10-year time interval. These assumptions include the following:
1. The study used leakage rate data from operating plants. Any
leakage over the plant's administrative leakage limit was considered a
leakage failure. An administrative limit is a utility's internal limit
and does not imply violation of any Appendix J limits. Therefore, the
probability of a leakage failure is overestimated.
2. Failure of one valve to meet the administrative limit does not
imply that the penetration would leak because containment penetrations
typically have redundant isolation valves. While one valve may leak,
the other valve may remain leak-tight. The study assumed that failure
of one valve in a series failed the penetration. Therefore, the
probability of a penetration leak is overestimated.
[[Page 68028]]
3. The analysis assumed possible leakage of all valves subject to
Type C testing, not just those subject to the relief per Sec. 50.69.
According to this analysis, the removal of SSCs from the scope of
Appendix J requirements does not have a significant effect on risk. The
NUREG-1493 analysis shows that the amount of leakage necessary to
significantly increase risk is two orders of magnitude greater than a
typical Technical Specification leakage rate limit. Therefore, the risk
to the public will not significantly increase due to the relief from
the requirements of Appendix J to 10 CFR part 50.
III.4.8 Appendix A to 10 CFR Part 100 (and Appendix S to 10 CFR Part 50
(Seismic Requirements))
Section 50.69(b)(1)(xi) removes RISC-3 and RISC-4 SSCs from the
requirement in Appendix A to 10 CFR part 100 to demonstrate that SSCs
are designed to withstand the safe shutdown earthquake (SSE) by
qualification testing or specific engineering methods. GDC-2 requires
that SSCs ``important to safety'' be capable of withstanding the
effects of natural phenomena, such as earthquakes. The requirements of
10 CFR part 100 pertain to reactor site criteria and Appendix A
addresses seismic and geologic siting criteria used by the Commission
to evaluate the suitability of plant design bases considering these
characteristics. Sections VI(a)(1) and (2) of Appendix A to 10 CFR part
100 address the engineering design for the SSE and operating basis
earthquake (OBE), respectively. Section 50.69 excludes RISC-3 and RISC-
4 SSCs from the scope of the requirements of Sections VI(a)(1) and (2)
of Appendix A to 10 CFR part 100, only to the extent that the rule
requires testing and specific types of analyses to demonstrate that
safety-related SSCs are designed to withstand the SSE and OBE. It is
only these aspects of Appendix A to 10 CFR part 100 that are considered
special treatment requirements. As discussed in Section III.4.0 of this
rulemaking, because of the low individual safety significance of the
RISC-3 and RISC-4 SSCs, the additional assurance provided by
qualification testing (or specific methods of analysis) is not
considered necessary.
Appendix A to part 100 is applicable for current operating
reactors. The seismic design requirements are set forth in Appendix S
to part 50 for new plant applications. The NRC has determined that
Appendix S does not need to be included within the scope of Sec. 50.69
because the wording of the requirements with respect to
``qualification'' by testing or specific types of analysis is not
present in Appendix S. Therefore, a revision to the regulations is not
necessary to permit a licensee to implement means other than
qualification testing or the specified methods to demonstrate SSC
capability.
III.4.9 Section 50.46a(b) Appendix B Requirements for Reactor Coolant
System Vents
The Commission established new requirements for combustible gas
control in Sec. 50.44 using risk insights and issued the revised rule
on September 16, 2003 (68 FR 54123). As part of the Sec. 50.44
rulemaking, portions of the old Sec. 50.44 were relocated to more
appropriate regulations. In particular, requirements formerly located
in Sec. 50.44 were relocated to Sec. 50.46a(b) concerning the design
of vents and associated controls, instruments, and power sources and
the need for these components to conform to 10 CFR part 50 Appendix B.
This rule removes RISC-3 SSCs from the scope of Appendix B quality
assurance requirements, as discussed in Section III.4.6. These same
arguments apply to the requirements in Sec. 50.46a(b) where Appendix B
is being imposed on a specific set of components. As such, this rule
removes the RISC-3 and RISC-4 SSCs from the scope of Appendix B
requirements contained in Sec. 50.46a(b). This applies only to the
requirements relating to Appendix B in Sec. 50.46a(b); the remaining
requirements of Sec. 50.46a remain unchanged.
III.4.10 Requirements Not Removed by Sec. 50.69(b)(1)
In the following paragraphs, the Commission discusses certain rules
that were considered as candidates for removal as requirements for
RISC-3 and RISC-4 SSCs during development of this rulemaking. These
rules were identified as candidate rules in SECY-99-256. They are not
part of this rulemaking for the reasons stated.
III.4.10.1 Section 50.34 Contents of Applications
Section 50.34 identifies the required information that applicants
must provide in preliminary and final safety analysis reports. Because
Sec. 50.69 contains the documentation requirements for licensees and
applicants who choose to implement Sec. 50.69, and these requirements
do not conflict with Sec. 50.34, it is not necessary to revise Sec.
50.34 to implement Sec. 50.69.
III.4.10.2 Section 50.36 Technical Specifications
Section 50.36 establishes operability, surveillance, limiting
conditions for operation and other requirements on certain SSCs.
Because this rule specifies testing and related requirements, it was
considered as a candidate special treatment rule. However, the
Commission concluded that it was not appropriate to revise Sec. 50.36
for several reasons.
Currently, the NRC staff and the industry are developing risk-
informed improvements to technical specifications. These improvements,
or initiatives, are intended to maintain or improve safety while
reducing unnecessary burden, and to bring technical specifications into
congruence with the Commission's other risk-informed regulatory
requirements, in particular risk management requirements of the
Maintenance Rule in 10 CFR 50.65(a)(4). Eight initiatives for
fundamental improvements to the Standard Technical Specifications (TS)
have been proposed. Two of the initiatives have been approved and
offered to licensees for adoption, and six are being developed by the
industry and NRC staff. All of the initiatives involve, to some
prescribed degree, assessing and managing plant risk using a
configuration risk management program consistent with and in some cases
exceeding the requirements of the Maintenance Rule in 10 CFR 50.65. The
two approved initiatives involve: permitting the extension of up to one
surveillance interval of an inadvertently missed surveillance; and,
permitting plant mode transitions with inoperable equipment,
anticipating the imminent return of the equipment to operability. The
six initiatives under development involve: shutting down to hot
shutdown rather than cold shutdown to repair equipment; permitting the
temporary extension of allowed outage times; permitting the
determination of surveillance frequencies through the use of an
approved methodology; permitting time to restore equipment operability
rather than immediately shutting down; providing extended time to
restore support systems to operability; and, revising the scope of
technical specifications to include only on risk significant systems,
which would require rulemaking.
Improved standard TSs have already resulted in the relocation of
requirements for less important SSCs to other documents. Given the
ongoing regulatory efforts to risk-inform the TSs, it was not
considered necessary to scope
[[Page 68029]]
Sec. 50.36 into Sec. 50.69 as a special treatment requirement.
III.4.10.3 Section 50.44 Combustible Gas Control
During the effort to identify candidate special treatment rules
(refer to SECY-99-256), certain provisions within Sec. 50.44 were
identified as containing special treatment requirements in that they
specified conformance with Appendix B for particular design features,
specified requirements for qualification, and related statements. For
proposed Sec. 50.69, the Commission elected not to identify Sec.
50.44 as a special treatment rule, and instead decided to wait on the
outcome of the effort to risk inform Sec. 50.44. The Commission
subsequently rebaselined the requirements in Sec. 50.44 using risk
insights and issued the revised rule on September 16, 2003 (68 FR
54123). As a result, the NRC concludes that there is no need to include
Sec. 50.44 within the scope of Sec. 50.69. However, as part of the
September 16, 2003, rulemaking, portions of the old Sec. 50.44 were
relocated to more appropriate regulations. In particular, requirements
were relocated to Sec. 50.46a(b) concerning the design of vents and
associated controls, instruments, and power sources and the need for
these components to conform to 10 CFR Part 50 Appendix B. Because this
aspect of the relocated requirements is a special treatment requirement
(and this same requirement was also identified in the old Sec. 50.44
as being a special treatment requirement) it is now captured within the
scope of Sec. 50.69(b)(1) as discussed in Section III.4.9.
III.4.10.4 Section 50.48 (Appendix R and GDC 3) Fire Protection
Initially, fire protection requirements were considered to be
within the scope of this rulemaking effort. There are augmented quality
provisions applied to fire protection systems and these augmented
quality provisions are considered special treatment requirements.
However, these provisions are not contained in the Commission's
regulations and therefore a revision to the rules (i.e., to scope them
into Sec. 50.69) is not required to support a change (i.e., changes to
these requirements can be made without a revision to the rules).
Additionally, the Commission has issued a final rule that would allow
licensees to voluntarily adopt National Fire Protection Association
(NFPA)-805 requirements in lieu of other fire protection requirements.
NFPA-805 sets forth requirements for establishing and implementing a
risk-informed fire protection program. Inasmuch as the NRC has
addressed fire protection in another rulemaking, fire protection
requirements were not included in the scope of the Sec. 50.69
rulemaking.
III.4.10.5 Section 50.59 Changes, Tests, and Experiments
There is no change is being made to Sec. 50.59 as a result of
Sec. 50.69, however, the Commission does not believe that a Sec.
50.59 evaluation need be performed when a licensee implements Sec.
50.69 and thereby changes the special treatment requirements applied to
RISC-3 and RISC-4 SSCs. Accordingly, Sec. 50.69(f) contains language
that removes the requirement for licensees to perform Sec. 50.59
evaluations for the changes in special treatment that stem from Sec.
50.69 implementation. The process of adjusting treatment for RISC-3 and
RISC-4 SSCs does not need to be subject to Sec. 50.59 because the
rulemaking already provides the decision process for categorization and
determination of revision to requirements resulting from the
categorization. Because it is only in the area of treatment for RISC-3
and RISC-4 SSCs that might be viewed as involving a reduction in
requirements, these are the only aspects for which this rule provision
applies. As required by Sec. 50.69(f), the licensee or applicant will
be required to update the FSAR appropriately to reflect incorporation
of its treatment processes into the FSAR. However, it is important to
recognize that changes that may affect any non-treatment aspects of an
SSC (e.g., changes to the SSC design basis functional requirements) are
required to be evaluated in accordance with the requirements of Sec.
50.59. The Commission, in developing Sec. 50.69, is drawing a
distinction between treatment (managed through Sec. 50.69) and design
changes (managed through other processes such as Sec. 50.59). As
previously noted, this rulemaking is only risk-informing the scope of
special treatment requirements. The process and requirements
established in Sec. 50.69 do not extend to making changes to the non-
treatment portion of the design basis of SSCs.
III.4.10.6 Appendix A to 10 CFR Part 50 General Design Criteria (GDC)
The NRC has concluded that the GDC of Appendix A to 10 CFR Part 50
do not need to be revised because they specify design requirements and
do not specify special treatment requirements. Because this rulemaking
is not revising the non-treatment portion of the design basis of the
facility, the GDC should remain intact and are not within the scope of
Sec. 50.69. This subject is discussed in more detail in the NRC's
action on the South Texas exemption request, in which their request for
exemption from certain GDCs was denied as being unnecessary to
accomplish what was proposed (see Section IV.2.0).
III.4.10.7 10 CFR Part 52 Early Site Permits, Standard Design
Certifications and Combined Operating Licenses
Part 52 cross-references regulations from other parts of Chapter 10
of the Code of Federal Regulations, most notably Part 50. Therefore, it
was initially considered for inclusion in this rulemaking effort.
However, the ``applicability'' paragraph (Sec. 50.69(b)) makes clear
that Sec. 50.69 is available to applicants for, and holders of a
facility license. Accordingly, there is no need to revise Part 52 to
assure the availability of Sec. 50.69. There are issues associated
with Part 52 design certifications and these are currently excluded
from the group of entities who may adopt the provisions of 50.69 as
discussed in Section V.3.0.
II.4.10.8 10 CFR Part 54 License Renewal
10 CFR part 54, which sets forth the license renewal requirements
for nuclear power reactors, was identified as a candidate special
treatment requirement in SECY-99-256. The Part 54 aging management
requirements are special treatment requirements in that they provide
assurance that SSCs will continue to meet their licensing basis
requirements during the renewed license period. Section 54.4 explicitly
defines the scope of the license renewal rule using the traditional
deterministic approach. Part 54 imposes aging management requirements
in Sec. 54.21 on the scope of SSCs meeting Sec. 54.4.
In SECY-00-0194, the NRC staff provided its preliminary view that
RISC-3 SSCs should not be removed from the scope of Part 54 and that
licensees can renew their licenses in accordance with Part 54 by
demonstrating that the Sec. 50.69 treatment provides adequate aging
management in accordance with Sec. 54.21. The NRC staff suggested that
no changes are necessary to Part 54 to implement Sec. 50.69 either
before renewing a licensing or after license renewal.
The goal of the license renewal program is to establish a stable,
predictable, and efficient license renewal process. The Commission
believes that a revision of Part 54 at this time could have a
significant effect on the stability and consistency of the processes
established for preparation of license renewal applications and for NRC
staff review. Further, as discussed
[[Page 68030]]
below, the Commission believes that the requirements in Part 54 are
compatible with the Sec. 50.69 approach, including use of risk
information in establishing treatment (aging management) requirements.
Refer to Section V.3.0 for additional discussion regarding the
implementation of Sec. 50.69 for a facility that has already received
a renewed license. Thus, Part 54 requires no changes at this time.
However, in the future, the Commission will consider whether revisions
to the scope of Part 54 are appropriate.
The 1995 amendment to Part 54 excluded active components to
``reflect a greater reliance on existing licensee programs that manage
the detrimental effects of aging on functionality, including those
activities implemented to meet the requirements of the maintenance
rule'' (May 8, 1995; 60 FR 22471). Although Sec. 50.69 removes RISC-3
components from the scope of the maintenance rule requirements in Sec.
50.65(a)(1), (a)(2), and (a)(3), a licensee is required under Sec.
50.69(d)(2) to provide confidence in the capability of RISC-3 SSCs to
perform their safety-related functions under design-basis conditions
when challenged. The SOC for Part 54 also indicated the Commission's
recognition that risk insights could be used in evaluating the
robustness of an aging management program (May 8, 1995; 60 FR 22468).
III.4.10.9 Other Requirements
In the ANPR and related documents, the NRC staff and stakeholders
suggested a number of other regulatory requirements that might be
candidates for inclusion in Sec. 50.69. These included Sec. 50.12
(exemptions), Sec. 50.54(a), (p), and (q) (plan change control), and
Sec. 50.71(e) (FSAR updates). As the rulemaking progressed, the
Commission concluded that these requirements did not need to be changed
to allow a licensee to adopt Sec. 50.69.
III.5.0 Feedback, Documentation, and Reporting Requirements
The validity of the categorization process relies on ensuring that
the performance and condition of SSCs continue to be maintained
consistent with applicable assumptions. Changes in the level of
treatment applied to an SSC might result in changes in the reliability
of the SSCs credited in the categorization process. Additionally, plant
changes, changes to operational practices, and plant and industry
operational experience may impact categorization process results.
Consequently, the rule contains requirements for updating the
categorization and treatment processes when conditions warrant to
assure that continued SSC performance is consistent with the
categorization process and results.
Specifically, the rule requires licensees to review the changes to
the plant, operational practices, applicable plant and industry
operational experience, and, as appropriate, update the PRA and SSC
categorization. The review must be performed in a timely manner but no
longer than once every two refueling outages. In addition, licensees
are required to obtain sufficient information on SSC performance to
verify that the categorization process and its results remain valid.
For RISC-1 SSCs, much of this information may be obtained from present
programs for inspection, testing, surveillance, and maintenance.
However, for RISC-2 SSCs and for RISC-1 SSCs credited for beyond design
basis accidents, licensees need to ensure that sufficient information
is obtained. For RISC-3 SSCs, there is a relaxation of the requirements
for obtaining information when compared to the applicable special
treatment requirements. However, sufficient information still needs to
be obtained. The rule requires considering performance data,
determining if adverse changes in performance have occurred, and making
the necessary adjustments so that desired performance is achieved so
that the evaluations conducted to meet Sec. 50.69(c)(1)(iv) remain
valid. The feedback and adjustment process is crucial to ensuring that
the SSC performance is maintained consistent with the categorization
process and its results.
Taking timely corrective action is an essential element for
maintaining the validity of the categorization and treatment processes
used to implement Sec. 50.69. For safety significant SSCs, all current
requirements continue to apply and, as a consequence, Appendix B
corrective action requirements are applied to the design basis aspects
of RISC-1 SSCs to ensure that conditions adverse to quality are
corrected. For both RISC-1 and RISC-2 SSCs, requirements are included
in Sec. 50.69(e)(2) for monitoring and for taking action when SSC
performance degrades.
When a licensee or applicant determines that a RISC-3 SSC does not
meet its established acceptance criteria for performance of design
basis functions, the rule requires that a licensee perform timely
corrective action (Sec. 50.69(d)(2)(ii)). Further, as part of the
feedback process, the review of operational data may reveal
inappropriate credit for reliability or performance and a licensee
would need to re-visit the findings made in the categorization process
or modify the treatment for the applicable SSCs (Sec. 50.69(e)(3)).
These provisions would then restore the facility to the conditions that
were considered in the categorization process and would also restore
the capability of the SSCs to perform their functions.
Section 50.69(f) requires the licensee or applicant to document the
basis for its categorization of SSCs before removing special treatment
requirements. Section 50.69(f) also requires the licensee or applicant
to update the final safety analysis report to reflect which systems
have been categorized.
Finally, Sec. 50.69(g) requires reporting of events or conditions
that prevented, or would have prevented, a RISC-1 or RISC-2 SSC from
performing a safety significant function. Because the categorization
process has determined that RISC-2 SSCs are of safety significance, NRC
is interested in reports about circumstances where a safety significant
function was, or would have been, prevented because of events or
conditions. This reporting will enable NRC to be aware of situations
impacting those functions found to be significant under Sec. 50.69, so
that NRC can take any actions deemed appropriate.
Properly implemented, these requirements ensure that the validity
of the categorization process and results are maintained throughout the
operational life of the plant.
III.6.0 Implementation Process Requirements
The Commission is making the provisions of Sec. 50.69 available to
both applicants for licenses and to holders of facility licenses for
light-water reactors. The rule is limited to light-water reactors
because the Commission does not yet have substantial experience or
information sufficient to develop risk-informed requirements applicable
to non-light water reactors. Consequently, the technical aspects of the
rule (e.g., providing reasonable confidence that risk increases are
small), including the implementation guidance, are specific to light-
water reactor designs.
Section 50.69 relies on a robust categorization process to provide
reasonable confidence that the safety significance of SSCs is correctly
determined. To ensure a robust categorization is employed, Sec. 50.69
requires the categorization process to be reviewed and approved by the
NRC before implementation of Sec. 50.69 by following the license
amendment
[[Page 68031]]
process of Sec. 50.90 or as part of the license application review.
While detailed regulatory guidance has been developed to provide
guidance for implementing categorization consistent with the rule
requirements, the Commission concluded that a prior review and approval
was still necessary to enable the NRC staff to review the scope and
quality of the plant-specific PRA; taking into account industry peer
review results. The NRC staff will also review other evaluations and
approaches that may be used, such as margins-type analyses, as well as
examine any aspects of the proposed categorization process that are not
consistent with the NRC's regulatory guidance for implementing Sec.
50.69. Thus, the rule requires that a licensee who wishes to implement
Sec. 50.69 submit an application for license amendment to the NRC
containing information about the categorization process and about the
industry peer review process employed. An applicant would submit this
information as part of its license application. The NRC will approve,
by license amendment, a request to allow a licensee to implement Sec.
50.69 if it is satisfied that the categorization process to be used
meets the requirements in Sec. 50.69.
NEI submitted a paper, ``License Amendments: Analysis of Statutory
and Legal Requirements'' (NEI Analysis) in a July 10, 2002, letter to
the Director of the Office of Nuclear Reactor Regulation (NRR). In this
analysis, NEI contends that approval of a licensee's/applicant's
request to implement Sec. 50.69 need not be accomplished by a license
amendment. NEI essentially argues that the rule does not increase the
licensee's operating authority, but merely provides a ``different means
of complying with the existing regulations * * *'' Id., p.8. The
Commission disagrees with this position, inasmuch as Sec. 50.69
permits the licensee/applicant, once having obtained approval from the
NRC, to depart from compliance with the ``special treatment''
requirements set forth in those regulations delineated in Sec. 50.69.
NEI also argues that the NRC's review and approval of the SSC
categorization process under Sec. 50.69 is analogous to the review and
approval process in Cleveland Electric Illuminating Co. (Perry Nuclear
Power Plant, Unit 1), CLI-96-13, 44 NRC 315 (1996), which the
Commission determined did not require a license amendment. Unlike the
Perry case, where the license already provided for the possibility of
material withdrawal schedule changes and the governing ASTM standard
set forth objective, non-discretionary criteria for changes to the
withdrawal schedule, Sec. 50.69 does not contain these criteria for
assessing the adequacy of the categorization process, PRA peer review
results, and the basis for sensitivity studies. Hence, the NRC's
approval of a request to implement Sec. 50.69 will involve substantial
professional judgment and discretion. The Commission does not agree
with NEI's assertion that the NRC's approval of a request to implement
Sec. 50.69 may be made without a license amendment in accordance with
the Perry decision.
The Commission does not believe it is necessary to perform a prior
review of the treatment processes to be implemented for RISC-3 SSCs in
lieu of the special treatment requirements. Instead, the NRC has
developed Sec. 50.69 to contain requirements that ensure the
categorization process is sufficiently robust to provide reasonable
confidence that SSC safety significance is correctly determined;
sufficient requirements on RISC-3 SSCs to provide a level of assurance
that these SSCs remain capable of performing their design basis
functions commensurate with their individual low safety significance;
and requirements for obtaining information concerning the performance
of these SSCs to help enable corrective actions to be taken before
RISC-3 SSC reliability degrades beyond the values used in the
evaluations conducted to satisfy Sec. 50.69(c)(1)(iv). The NRC
concludes that compliance with these requirements, in conjunction with
inspection of Sec. 50.69 licensees, is a sufficient level of
regulatory oversight for these SSCs.
The Commission included requirements in the rule for documenting
categorization decisions to facilitate NRC oversight of a licensee's or
applicant's implementation of the alternative requirements. The rule
also includes provisions to have the FSAR and other documents updated
to reflect the revised requirements and progress in implementation.
These requirements will allow the NRC and other stakeholders to remain
knowledgeable about how a licensee is implementing its regulatory
obligations as it transitions from past requirements to the revised
requirements in Sec. 50.69. As part of these provisions, the
Commission has concluded that requiring evaluations under Sec. 50.59
(for changes to the facility or procedures as described in the FSAR) or
under Sec. 50.54(a) (for changes to the quality assurance plan) is not
necessary for those changes directly related to implementation of Sec.
50.69. For implementation of treatment processes for low safety
significant SSCs, in accordance with the rule requirements contained in
Sec. 50.69, the Commission concludes that requiring further review if
NRC approval might be required for these changes is an unnecessary
burden. Thus, a licensee is permitted to make changes concerning
treatment requirements that might be contained in these documents. The
Commission is limiting this relief to changes directly related to
implementation (with respect to treatment processes). Changes that
affect any non-treatment aspects of an SSC (e.g., changes to the SSC
design basis functional requirements) are still required to be
evaluated in accordance with other regulatory requirements such as
Sec. 50.59. This rulemaking is only risk-informing the scope of
special treatment requirements. The process and requirements
established in Sec. 50.69 do not extend to making changes to the non-
treatment portion of the design basis.
III.7.0 Adequate Protection
The Commission concludes that Sec. 50.69 provides reasonable
assurance of adequate protection of public health and safety because
the principles listed below were used in the development of Sec. 50.69
and because these principles will continue to be employed in the NRC's
continuing regulatory oversight of Sec. 50.69 implementation. Those
principles are:
(a) Reasonable confidence that the net increase in plant risk is
small;
(b) Defense-in-depth is maintained;
(c) Reasonable confidence that safety margins are maintained; and
(d) Monitoring and performance assessment strategies are used.
These principles were established in RG 1.174, which provided
guidance on an acceptable approach to risk-informed decision-making
consistent with the 1995 Commission policy on the use of PRA. Section
50.69 was developed to incorporate these principles, both to ensure
consistency with Commission policy, and to ensure that the rule
maintains adequate protection of public health and safety.
The following discusses how Sec. 50.69 meets the four criteria,
and as a result, maintains adequate protection of public health and
safety.
III.7.1 Net Increase in Risk Is Small
Section 50.69(c) requires the use of a robust, risk-informed
categorization process that ensures that all relevant information
concerning the safety significance of an SSC is considered by a
competent and knowledgeable panel who makes the final determination of
the safety significance of SSCs. The NRC review and approval of the
licensee's
[[Page 68032]]
categorization process ensures that it meets the requirements of Sec.
50.69(c) and that, as a result, the correct SSC safety significance is
determined with high confidence. Correctly determining safety
significance of an SSC provides confidence that special treatment
requirements are only removed from SSCs with low individual safety
significance and that these requirements continue to be satisfied for
SSCs of safety significance. The rule requires that the potential net
increase in risk from implementation of Sec. 50.69 be assessed and
that reasonable confidence be provided that this risk change is small.
These requirements to provide reasonable confidence that the net change
in risk is acceptably small as part of the categorization decision, in
conjunction with the rule requirements for maintaining design basis
functions and the processes noted below for feedback and adjustment
over time, all contribute to preventing risk from increasing beyond the
ranges that the NRC has considered to be appropriate as discussed in
the RG 1.174 acceptance guidelines. As a result, these requirements are
a contributing element for maintaining adequate protection of public
health and safety.
III.7.2 Defense-in-Depth Is Maintained
Section 50.69 (c)(1)(iii) requires that defense-in-depth be
maintained as part of the categorization requirements of Sec.
50.69(c)(1) and as a result, defense-in-depth is considered explicitly
in the categorization process. Thus, SSCs that otherwise might be
considered low safety significant, but are important to defense-in-
depth as discussed in the implementation guidance, will be categorized
as safety significant (and will remain subject to special treatment
requirements). For safety significant SSCs (i.e., RISC-1 and RISC-2
SSCs), all current special treatment requirements remain (i.e., the
rule does not remove any of these requirements) to provide high
confidence that they can perform design basis functions. Additionally,
Sec. 50.69(d)(1) requires sufficient treatment be applied to support
the credit taken for these SSCs for beyond design basis events. For
RISC-3 SSCs, Sec. 50.69 imposes high-level treatment requirements that
when effectively implemented, maintain the capability of RISC-3 SSCs to
perform their design basis functions. Thus, the complement of SSCs
installed at the facility that provide defense-in-depth will continue
to be available and capable of performing the functions necessary to
support defense-in-depth. The rule does not change the design basis
functional requirements of the facility, which were established based
upon defense-in-depth considerations. Accordingly, the Commission
concludes that Sec. 50.69 maintains defense-in-depth.
III.7.3 Safety Margins Are Maintained
Section 50.69(c)(1)(iv) requires that evaluations be performed that
provide reasonable confidence that sufficient safety margins are
maintained. This is provided by a combination of:
(1) Maintaining all existing functional and treatment requirements
on RISC-1 and RISC-2 SSCs and additionally ensuring, through the
application of sufficient treatment and feedback requirements, that any
credit for these SSCs for beyond design basis conditions is valid and
maintained;
(2) Maintaining the design basis functional requirements of the
facility for all SSCs, including RISC-3 SSCs as described in Section
III.7.2; and
(3) Requiring a licensee to have reasonable confidence that the
overall increase in risk that may result due to implementation of Sec.
50.69 is small.
Maintaining all current requirements on RISC-1 and RISC-2 SSCs and
requiring sufficient treatment be applied to support the credit taken
for these SSCs for beyond design basis events provides assurance that
the safety significant SSCs continue to perform as credited in the
categorization process. Maintaining design basis functional
requirements for RISC-3 SSCs ensures that these SSCs continue to be
designed to criteria that enable them to perform their design basis
functions. The reduction in treatment applied to RISC-3 SSCs results in
an increased level of uncertainty concerning the functionality of RISC-
3 SSCs. This reduction in treatment may result in an increase in RISC-3
SSC failure rates (i.e., a reduction in RISC-3 SSC reliability). To
address this possibility and its relationship to safety margin, Sec.
50.69 requires that there be reasonable confidence that any potential
increases in CDF and LERF that might stem from changes in RISC-3 SSC
reliability due to reduced treatment permitted by Sec. 50.69, be
small. As discussed in Section III.7.4, the rule requires (through
monitoring requirements) that the SSCs must be maintained so that they
continue to be capable of performing their design basis functions. For
these reasons, the Commission concludes that Sec. 50.69 maintains
sufficient safety margins.
III.7.4 Monitoring and Performance Assessment Strategies Are Used
Section 50.69(e) contains requirements that ensure that the risk-
informed categorization and treatment processes are updated and
maintained over time. Data that reflect operational practices, the
facility configuration, plant and industry experience, and SSC
performance are required to be fed back into the PRA and the
categorization process on a periodic basis and when appropriate,
adjustments to the categorization and/or treatment processes are
required to maintain the validity of these processes. In addition,
Sec. 50.69(g) contains requirements that reports are made to NRC of
conditions preventing RISC-1 and RISC-2 SSCs from performing their
safety significant functions. Together, these requirements maintain the
validity of the risk-informed categorization and treatment processes so
that the above criteria will continue to be satisfied over the life of
the facility.
III.7.5 Summary and Conclusions
Section Sec. 50.69 contains requirements that:
1. Provide reasonable confidence that any net risk increase from
implementation of its requirements is small;
2. Maintain defense-in-depth;
3. Provide reasonable confidence that safety margins are
maintained; and
4. Require the use of monitoring and performance assessment
strategies.
Together, these requirements result in a rule that is consistent
with the Commission's policy on the use of PRA and, more importantly,
maintains adequate protection of public health and safety.
IV. Pilot Activities
IV.1.0 Pilot plants
To aid in the development of the rule and associated implementation
guidance, several plants volunteered to conduct pilot activities with
the objective of exercising the proposed NEI implementation guidance
and using the feedback and lessons-learned to improve both the
implementation guidance and the governing regulatory framework. There
were two separate pilot efforts. The first pilot effort focused on the
categorization guidance and IDP performance. This effort is discussed
in Section IV.1.1 Categorization Pilot. The second pilot effort is
ongoing and is focused on the Sec. 50.69 submittal and its review.
This pilot effort is discussed in Section IV.1.2 Submittal Pilot.
IV.1.1 Categorization Pilot
The categorization pilot effort was supported by three of the
industry owners groups who identified pilots for their reactor types
and participated by piloting sample systems using the draft
[[Page 68033]]
NEI implementation guidance. Supporting the pilot effort were the
Westinghouse Owners Group with lead plants Wolf Creek and Surry, the
BWR Owners Group with lead plant Quad Cities, and the CE Owners Group
with lead plant Palo Verde. The B&W Owners Group did not participate,
but did follow the pilot activities.
The NRC staff's participation and principal point of interaction in
the pilot effort was primarily in observation of the deliberations of
the IDP. By observing the IDP, the NRC staff was able to view the
culmination of the categorization effort and gain good insights
regarding both the robustness of the categorization process in general
and the IDP decision-making process specifically. Following each of the
pilot IDPs, the NRC staff developed and issued a trip report containing
its observations.
The following points set forth the principal lessons learned and
key feedback from the NRC staff's observations of the pilot activities:
Potential treatment changes and their potential effects
need to be understood by the IDP as part of the deliberations on
categorization.
The pilots showed the importance of documenting IDP
decisions and the basis for them. The rule contains a requirement for
the categorization basis to be documented (and records retained) in
Sec. 50.69(f).
The pilots experienced difficulty in explicit
consideration about safety margins, especially in view of the fact that
functionality must be retained. In the first draft rule language
posted, requirements were included for the IDP to consider safety
margins in its deliberations. On the basis of the pilot experience, NRC
adjusted its approach to safety margins to include this in the section
of the rule that requires consideration of effects of changes in
treatment and the use of evaluations as the means of providing
reasonable confidence safety margins are maintained.
The need for a number of improvements to the industry
implementation guidance provided in NEI 00-04 were noted. For example,
two areas for improvement were the defense-in-depth matrix presented
therein and the need for more specific guidance on making decisions
where quantitative information is not available. These lessons learned
were factored into the revised version of NEI 00-04.
During the pilot activity, pressure boundary (``passive'')
functions were also categorized using the draft version of an ASME Code
Case on categorization available at that time. A separate
categorization process was used for these passive functions because it
was recognized by pilot participants that the approach for these SSCs
must be somewhat different than for ``active'' functions due to
considerations such as spatial interaction. Specifically, if a pressure
boundary SSC failed, the resulting high-energy release or flooding
might impact other equipment in physical proximity, so the process
needed to account for those effects in addition to the significance of
the SSC that initially failed. Improvements to the ASME Code Case for
categorization of piping (and related components) were identified and
fed back into the code development process.
The pilot experiences also revealed the intricacies of the
relationship between ``functions'' (which play a role in decisions on
safety significance) and ``components'' (importance measures are
associated with components and treatment is also generally applied on a
component basis). Because a particular component may support more than
one function, the categorization of the component needs to correspond
with the most significant function and means must be provided for a
licensee to ``map'' the components to the functions they support.
At each pilot, the NRC noted that the IDP needed to
include consideration of long term containment heat removal in
characterizing SSCs. The NRC considers retention of long term
containment heat removal capability important to defense-in-depth for
light water reactors.
Finally, a number of lessons were learned about how to
conduct the IDP process, such as training needs, materials to be
provided to the panel, etc. As a result of this feedback, NEI revised
NEI 00-04 (discussed in Section VI).
IV.1.2 Submittal Pilot
The submittal pilot effort is a currently ongoing effort that
focuses on the Sec. 50.69 submittal and the NRC staff's review and
approval of that submittal. This pilot effort is supported by the
Westinghouse Owners Group with lead plants Wolf Creek and Surry. The
objectives of this pilot effort are to:
Enable the staff to develop reviewer guidance for review
and approval of the Sec. 50.69 submittal.
To acquire experience with the use of RG 1.201 and use
this experience to improve the guidance and address the technical
interpretation/implementation issues identified in RG 1.201.
Enable industry to develop (beyond RG 1.201/NEI 00-04) the
specific information that will be required for a license amendment
submittal that will be submitted for prior staff review and approval
for implementing Sec. 50.69.
The NRC staff will use the results of this pilot effort to improve
RG 1.201 and to develop the reviewer guidance for Sec. 50.69
submittals. Industry expects to use the results of the pilot to develop
a template for a Sec. 50.69 license amendment submittal.
IV.2.0 South Texas Exemption as Proof of Concept
A major element of the rulemaking plan described in SECY-99-256 was
the review of the STPNOC exemption request. The review of the STPNOC
exemption request was viewed as a proof-of-concept prototype for this
rulemaking rather than a pilot because it preceded development of draft
rule language or related implementation guidance.
By letter dated July 13, 1999, STPNOC requested approval of
exemption requests to enable implementation of processes for
categorizing the safety significance of SSCs and treatment of those
SSCs consistent with its categorization process. The STPNOC process
included many similar elements to that described in this rulemaking,
but with some differences. Their process identified SSCs as being
either high, medium, low or non-risk significant. The scope of the
exemptions requested included only those safety-related SSCs that have
been categorized as low safety significant or as non-risk significant
using STPNOC's categorization process. The licensee indicated that the
categorization and treatment processes would be implemented over the
remaining licensed period of the facility. Thus, the basis for the
exemptions granted was the NRC staff's approval of the licensee's
categorization process and alternative treatment elements, rather than
a comprehensive review of the final categorization and treatment of
each SSC (review of the process rather than the results is also the
approach planned under the rulemaking). As a result of discussions with
the staff on a number of topics, STPNOC submitted a revised exemption
request on August 31, 2000.
On November 15, 2000, the NRC staff issued a draft safety
evaluation (SE)(ADAMS accession number ML003761558), based on the
revised exemption requests. Following the licensee's response to the
draft SE, the staff prepared SECY-01-0103 dated June 12, 2001 (ADAMS
accession number ML011560317), to inform the Commission of the staff's
finding regarding the STPNOC exemption
[[Page 68034]]
review. The staff approved the STPNOC exemption requests by letter
dated August 3, 2001 (ADAMS accession number ML011990368).
The NRC has applied lessons learned from the review of the STPNOC
exemption request in developing Sec. 50.69 and the description of
intended implementation of the rule in this SOC. For example, in the
STPNOC review, the NRC staff reviewed the categorization process
proposed by the licensee in detail. With respect to Sec. 50.69, the
NRC continues to require a robust categorization with a detailed staff
review.
The rule specifies the requirement that the licensee shall ensure
with reasonable confidence functionality and further specifies some
high-level requirements for RISC-3 SSC treatment. Under Sec. 50.69,
the NRC will not review and approve licensee's RISC-3 treatment
programs. Licensees will have to establish appropriate performance-
based SSC treatment to maintain the validity of the categorization
process and its results. The rule requires that licensees adjust the
categorization or treatment processes, as appropriate, in response to
the SSC performance information obtained as part of the treatment
process.
V. Section-by-Section Analysis
V.1.0 Section 50.8 Information Collection
This rule includes a revision to Sec. 50.8(b). This section
pertains to approval by the Office of Management and Budget (OMB) of
information collection requirements associated with particular NRC
requirements. Because the new Sec. 50.69 includes information
collection requirements, a conforming change to Sec. 50.8(b) is
necessary to list Sec. 50.69 as one of these rules. See also Section
XII of the SOC for discussion about information collection requirements
of Sec. 50.69.
V.2.0 Section 50.69(a) Definitions
Section 50.69(a) provides the definition for the four RISC
categories and the definition of the term ``safety significant
function.'' RISC-1 SSCs are safety-related SSCs (as defined in Sec.
50.2) and that are found to be safety significant (using the risk-
informed categorization process being established by this rule). RISC-2
SSCs are SSCs that do not meet the safety-related definition, but
determined to be safety significant. RISC-3 SSCs are safety-related
SSCs that are determined to be low safety significant on an individual
basis. Finally, RISC-4 SSCs are SSCs that are not safety-related and
that are determined to be low safety significant. The NRC selected the
terms ``safety significant'' and ``low safety significant'' as the best
representations of their meaning. Every component (if categorized) is
either safety significant or low safety significant. The ``low''
category could include those SSCs that have no safety significance, as
well as some SSCs that individually are not safety significant, but
collectively can have a significant impact on plant safety (and hence
the need for maintaining the design basis capability of these SSCs).
Similarly, within the category of ``safety significant,'' some SSCs
have more safety significance than others; so it did not appear
appropriate to call them all ``high safety significant.'' The RISC
definitions of paragraph (a) are used in subsequent paragraphs of Sec.
50.69 where the treatment requirements are applied to SSCs as a
function of RISC category.
The definitions provided in paragraph (a) are written in terms of
SSCs that perform functions. In the categorization process, it is the
various functions performed by systems that are assessed to determine
their safety significance. For those functions of significance, the
structures and components that support that function are then
designated as being of that RISC category. Then, the treatment
requirements are specified for the SSCs that perform those functions.
Where an SSC performs functions that fall in more than one category,
the treatment requirements derive from the more safety significant
function (i.e., if a component has both a RISC-1 and a RISC-3 function,
it is treated as RISC-1).
The rule also contains a definition of ``safety-significant''
function. NRC selected the term ``safety-significant'' instead of
``risk-significant'' because the categorization process employed in
Sec. 50.69 considers both probabilistic and deterministic information
in the decision process. Thus, it is more accurate to represent the
outcome as a determination of overall safety significance, that
includes the consideration of risk, as opposed to characterizing the
outcome as purely ``risk-significance.''
Those functions that are not determined to be safety significant
are considered to be low safety-significant. The determination as to
which functions are safety significant is done by following the
categorization process outlined in paragraph (c), as implemented
following the guidance in RG 1.201, ``Guidelines for Categorizing
Structures, Systems, and Components in Nuclear Power Plants According
to their Safety Significance.''
V.3.0 Section 50.69(b) Applicability
Section 50.69(b) may be voluntarily implemented by:
(1) A holder of a license to operate a light water reactor (LWR)
nuclear power plant under this part;
(2) Holders of Part 54 renewed LWR licenses;
(3) An applicant for a construction permit or operating license
under this part; and
(4) An applicant for a design approval, a combined license, or
manufacturing license under Part 52 of this chapter.
For current licensees, implementation will be through a license
amendment as set forth in Sec. 50.90. This review and approval of the
categorization process is a one-time process approval (i.e., the
approval is not restricted to a set of systems or structures, and
instead can be applied to any system or structure in the plant). The
licensee is not required to come back to the NRC for review of the
categorization process provided they remain within the scope of the
NRC's safety evaluation. Until the request is approved, a licensee is
free to develop (at their own risk) the Sec. 50.69 processes and
perform the Sec. 50.69 categorization. However, they must continue to
follow existing requirements until approval. Upon approval of the
categorization process, the licensee can implement the results of the
categorization process including the revised Sec. 50.69 treatment
requirements.
For part 54 license holders, implementation is the same as that for
a holder of an operating license under part 50, that is, to apply for
an amendment to the (renewed) license. For the case where a licensee
renewed its license first and then implemented Sec. 50.69, a licensee
might revise some aging management programs for RISC-3 SSCs, consistent
with the requirements of Sec. 50.69. The Commission believes that
there should be little or no impediment for doing so because the
categorization process that allows for the reduction in the special
treatment requirements for RISC-3 components is expected to provide an
appropriate level of safety for the respective structures, systems and
components.
In the development of Sec. 50.69, questions were considered
regarding the impact to licensees that implement Sec. 50.69 and
subsequently apply to renew their license. Because part 54 includes
scoping criteria that bring safety-related components within its scope,
these components could not be exempted without amending part 54 to
allow for their exclusion. However, there are still options available
to
[[Page 68035]]
applicants for renewal that have implemented Sec. 50.69 first. Because
Sec. 50.69 includes alternative treatment requirements for RISC-3
components, an applicant may be able to provide an evaluation that
justifies why these alternative treatment criteria (Sec. 50.69(d)(2))
provide a sufficient demonstration that aging management of the
components will be achieved during the renewal period to ensure the
functionality of the structure, system, or component. In addition, in
the 1995 amendment to part 54, the Commission recognized that risk
insights could be used in evaluating the robustness of an aging
management program. The NRC staff has already received and accepted one
proposal (Arkansas Unit 1) for a risk-informed program for small-bore
piping which demonstrates that risk arguments can be used to a degree.
Adopting Sec. 50.69 requirements for an applicant for a
construction permit or operating license under this part requires that
the applicant first design the facility to meet the current part 50
requirements. Specifically, to use the Sec. 50.69 requirements
requires that SSCs first be classified into the traditional safety-
related and nonsafety-related classifications. This establishes the
design basis functional requirements for the facility, which as
previously stated, Sec. 50.69 is not changing. Once the SSC
categorization has been done consistent with the safety-related
definition in Sec. 50.2, then Sec. 50.69 can be used to categorize
SSCs into RISC-1, RISC-2, RISC-3, and RISC-4 and the alternative
treatment requirements of Sec. 50.69 implemented. A new applicant who
chooses to adopt the Sec. 50.69 requirements, must seek approval of
the categorization process as part of its license application and,
following NRC approval, would be able to procure RISC-3 SSCs to Sec.
50.69 requirements before initial plant operation.
An applicant for a design approval, a combined license, or
manufacturing license under part 52 of this chapter may adopt Sec.
50.69 requirements. An applicant for a design approval, or
manufacturing license would follow a process very similar (from the
standpoint of Sec. 50.69) to that described above for an applicant for
a construction permit or operating license under part 50 (i.e., SSCs
must first be classified into the traditional safety-related and
nonsafety-related classifications which establishes the design basis
functional requirements for the facility and then Sec. 50.69 can be
used to categorize SSCs into RISC-1, RISC-2, RISC-3, and RISC-4).
Because Sec. 50.69 includes elements of procurement and installation,
as well as inservice activities, implementation of the rule by a holder
of a manufacturing license or by a part 52 applicant that references
such a design would place restrictions on the eventual operator of the
facility. The entity that actually constructs and operates the facility
would also have to implement Sec. 50.69 to maintain consistency with
the categorization process and feedback requirements. Otherwise, the
operator would be required to meet other part 50 requirements, such as
Appendix B or Sec. 50.55a, which may not be compatible with the
facility as manufactured by the manufacturing licensee.
An applicant for a part 52 combined license can apply Sec. 50.69
to a referenced design certification that did not comply with Sec.
50.69 provided the design is a LWR design that used the safety-related
definition in Sec. 50.2. An applicant who references a certified
design and wishes to implement Sec. 50.69 would include the specified
information in Sec. 50.69(b)(2) as part of its application for a
license. This does not mean that an applicant would actually construct
the facility per all parts 50 and 100 requirements first, before
applying Sec. 50.69. Instead, the facility needs to be designed per
these requirements, but following approval of the application request
under Sec. 50.69(b)(4), RISC-3 SSCs could be procured per the
requirements of Sec. 50.69(d).
The final rule excludes applicants for standard design
certifications from the group of entities who may take advantage of the
provisions of Sec. 50.69. In considering whether to extend the
applicability of Sec. 50.69 to design certifications, the Commission
identified a number of difficult issues which would have to be resolved
to support such an extension. For example, it is unclear whether the
dynamic process of recategorizing SSCs under Sec. 50.69 would be
inconsistent with the special change restrictions in Sec. 52.63(a),
thereby requiring the inclusion of a special change provision in the
individual design certification rule. Inasmuch as the proposed rule did
not include a provision that would have allowed design certification
applicants to use Sec. 50.69, the NRC has not had the benefit of the
views of the industry and the public on these issues. Moreover, the
industry has not expressed any interest in submitting a design
certification using the principles of Sec. 50.69. Accordingly, the
final rule does not address the issue of applying Sec. 50.69 to new
design certifications; issues associated with the application of Sec.
50.69 to design certification rulemaking can be addressed on a case-by-
case basis as necessary. In the future, the Commission could initiate
rulemaking to extend Sec. 50.69 to new design certifications after the
NRC has had some experience in this area. For much the same reasons,
the rule does not provide a process for changing an existing design
certification rule to voluntarily comply with Sec. 50.69. In addition,
a rulemaking would be necessary to change an existing certified design
(see Section VIII of Appendix A to 10 CFR part 52), and it is unlikely
that such a change would satisfy the requirements of Sec. 52.63(a)(1).
A request for a generic change to adopt Sec. 50.69 would not meet the
special backfit requirements of Section VIII. Therefore, the NRC would
not review the request. Additionally, the NRC would not want to expend
resources reviewing changes to designs that may not be referenced.
However, applicants for COLs that reference a certified design could
adopt Sec. 50.69 and the rule provides for that approach.
The rule provisions were devised to provide means for licensees and
applicants for light water reactors to implement Sec. 50.69. In view
of some of the specific provisions of the rule, for example, ``safety-
related'' definition and use of CDF/LERF metrics, the Commission is
making this rule applicable only to light-water reactor designs.
V.3.1 Section 50.69(b)(1) Removal of RISC-3 and RISC-4 SSCs From the
Scope of Treatment Requirements
Section 50.69 (b)(1) lists the specific special treatment
requirements from whose scope the RISC-3 and RISC-4 SSCs are being
removed through the application of Sec. 50.69. In this paragraph, each
regulatory requirement (or portions thereof) removed by this rulemaking
is listed in a separate item, numbered from Sec. 50.69(b)(1)(i)
through (b)(1)(xi). The basis for removal of these requirements was
discussed in Section III.4. These requirements are being removed due to
the low safety significance of RISC-3 and RISC-4 SSCs as determined by
an approved risk-informed categorization process meeting the
requirements of Sec. 50.69(c). The special treatment requirements for
RISC-3 SSCs are replaced with the high-level, performance-based
requirements in Sec. 50.69(d)(2) that require the licensee to provide
reasonable confidence that RISC-3 SSCs will continue to be capable of
performing their safety-related functions under design basis
conditions. These performance-based RISC-3 requirements in paragraph
(d)(2) are discussed below in greater detail. Note that special
treatment requirements are not removed from any
[[Page 68036]]
SSCs until the NRC approves the categorization process and a licensee
(or applicant) has categorized those SSCs using the requirements of
Sec. 50.69(c) to provide the documented basis for the decision that
they are of low safety significance.
V.3.2 Section 50.69 (b)(2) Application Process
Section 50.69(b)(2) requires a licensee who voluntarily seeks to
implement Sec. 50.69 to submit an application for a license amendment
under Sec. 50.90 that contains the following information:
(i) A description of the categorization process that meets the
requirements of Sec. 50.69(c).
(ii) A description of the measures taken to assure that the quality
and level of detail of the systematic processes that evaluate the plant
for internal and external events during normal operation, low power,
and shutdown (including the plant-specific PRA, margins-type
approaches, or other systematic evaluation techniques used to evaluate
severe accident vulnerabilities) are adequate for the categorization of
SSCs.
(iii) Results of the PRA review process to be conducted to meet
Sec. 50.69(c)(1)(i).
(iv) A description of, and basis for acceptability of, the
evaluations to be conducted to satisfy Sec. 50.69(c)(1)(iv). The
evaluations must include the effects of common cause interaction
susceptibility, and the potential impacts from known degradation
mechanisms for both active and passive functions, and address
internally and externally initiated events and plant operating modes
(e.g., full power and shutdown conditions).
Regarding the categorization process description, the NRC expects
that most licensees and applicants will commit to RG 1.201 which
endorses NEI 00-04, with some conditions and exceptions. If a licensee
or applicant wishes to use a different approach, the submittal must
provide a sufficient description of how the categorization would be
conducted. As part of the submittal, a licensee or applicant is to
describe the measures they have taken to assure that the plant-specific
PRA, as well as other methods used, are adequate for application to
Sec. 50.69. The measures described include such items as any peer
reviews performed, any actions taken to address peer review findings
that are important to categorization, and any efforts to compare the
plant-specific PRA to the ASME PRA standard. The NRC has developed
reviewer guidance applicable to these submittals that is described in
Section VI. The licensee or applicant must also describe what measures
they have used for the methods other than a PRA to determine their
adequacy for this application.
Further, the licensee or applicant is required to include
information about the evaluations they intend to conduct to provide
reasonable confidence that the potential increase in risk would be
small. This includes any risk sensitivity study for RISC-3 SSCs,
including the basis for whatever change in reliability is being assumed
for these analyses. A licensee must provide sufficient information to
the NRC, describing the risk sensitivity study and other evaluations
and the basis for their acceptability as appropriately representing the
potential increase in risk from implementation of the requirements in
this rule.
RISC-3 SSCs are defined as having low individual safety
significance under Sec. 50.69. Licensees and applicants must implement
effective treatment, consisting of, at a minimum, inspection, testing,
and corrective action, to maintain RISC-3 SSC functionality as required
by Sec. 50.69(d)(2). This treatment need not be described to the NRC
as part of the Sec. 50.69 submittal as provided in Sec. 50.69(b)(2).
V.3.3 Section 50.69 (b)(3) Approval for Licensees
Section 50.69(b)(3) provides that the Commission will approve a
licensee's implementation of this section by license amendment if it
determines that the proposed process for categorization of RISC-1,
RISC-2, RISC-3, and RISC-4 SSCs satisfies the requirements of Sec.
50.69(c).
The NRC will review the description of the categorization process
set forth in the application to confirm that it contains the elements
required by the rule. The NRC will also review the information provided
about the plant-specific PRA, including the peer review process to
which it was subjected, and methods other than a PRA relied upon in the
categorization process. The NRC intends to use review guidance
(discussed in more detail in Section VI) for this purpose. The NRC will
approve the licensee's use of Sec. 50.69 by issuing a license
amendment.
V.3.4 Section 50.69(b)(4) Process for Applicants
Section 50.69(b)(4) requires that an applicant for a license,
standard design approval, or manufacturing license that chooses to
implement Sec. 50.69 must submit the information listed in Sec.
50.69(b)(2) as part of its application. The rule is structured to
transition from the ``safety-related'' classification (and related
treatment requirements) to a ``safety significant'' classification.
Thus, an applicant would first need to design the facility to meet
applicable Part 50 design requirements and then apply the requirements
of Sec. 50.69. This information must be submitted in addition to other
technical information necessary to meet Sec. 50.34. The NRC will
provide its approval of implementation of Sec. 50.69, if it concludes
that the rule requirements are met, as part of its action on the
application.
V.4.0 Section 50.69(c) Categorization Process Requirements
Section 50.69(c) establishes the requirements for the risk-informed
categorization process including requirements for the supporting PRA.
Licensees or applicants who wish to adopt the requirements of Sec.
50.69 will need to make a submittal (per Sec. 50.69(b)(2) or Sec.
50.69(b)(4) respectively) that discusses how their proposed
categorization process, supporting PRA, and evaluations meet the Sec.
50.69(c) requirements. As described in Section III.2.0, these
requirements are intended to ensure that the risk-informed Sec. 50.69
categorization process determines the appropriate safety significance
of SSCs with high confidence. The introductory paragraph of Sec.
50.69(c) states that SSCs must be categorized as RISC-1, RISC-2, RISC-
3, or RISC-4 by a process that determines whether the SSC performs one
or more safety significant functions and identifies those functions.
V.4.1 Section 50.69(c)(1)(i) Results and Insights From a Plant-Specific
Probabilistic Risk Assessment
Section 50.69(c)(1)(i) contains the requirements for the PRA
itself, and how it is to be used in the categorization process. The PRA
must have sufficient capability and quality to support the
categorization of the SSCs. Section V.4.1.1 discusses these
requirements in more detail. The PRA and associated sensitivity studies
are used primarily in the categorization of the SSCs as to their safety
significance as discussed in Section V.4.1.2, and the PRA is also used
to perform evaluations to assess the potential risk impact of the
proposed change in treatment of the RISC-3 SSCs, as discussed in
Section V.4.4.
V.4.1.1 Scope, Capability, and Quality of the PRA To Support the
Categorization Process
As required in Sec. 50.69(c)(1)(ii), initiating events from
sources both internal and external to the plant and for all modes of
operation, including low power and shutdown modes, must
[[Page 68037]]
be considered when performing the categorization of SSCs. It is
recognized that few licensees have fully developed PRA models that
cover such a scope. However, as a minimum, the PRA to be used to
support categorization under Sec. 50.69(c)(1) must model internal
initiating events occurring at full power operations. The PRA will have
to be able to calculate both core damage frequency and large early
release frequency to meet the requirement in Sec. 50.69(c)(iv). The
PRA must reasonably represent the current configuration and operating
practices at the plant to meet Sec. 50.69(c)(1)(ii). The PRA model
should be of sufficient technical quality and level of detail to
support the categorization process. This means that it represents a
coherent, integrated model, and has sufficient detail to support the
categorization of SSCs into the safety significant and low safety
significant categories.
The quality and scope of the plant-specific PRA will be assessed by
the NRC taking into account appropriate standards and peer review
results. The NRC has prepared a regulatory guide (RG 1.200) on
determining the technical adequacy of PRA results for risk-informed
activities. As one step in the assurance of technical quality, the PRA
must have been subjected to a peer review process assessed against a
standard or set of acceptance criteria that is endorsed by the NRC.
Thus, the NRC will rely on the NEI Peer Review Process, as modified in
the NRC's approval, or the ASME/ANS Peer Review Process, as modified in
the NRC's approval both of which are (or will be) documented in RG
1.200. As discussed in Section VI, NRC has also developed review
guidelines for considering the sufficiency of a PRA that was subjected
to the NEI peer review process for this application in Sec. 50.69.
This guidance was developed based on an earlier draft version of NEI
00-04 and could be useful in ensuring the adequacy of the PRA for this
application. The submittal requirements listed in Sec. 50.69(b)(2)
include a requirement to provide information about the quality of the
PRA analysis and other supporting analyses and about the peer review
results.
V.4.1.2 Risk Categorization Process Based on PRA Information
For SSCs modeled in the PRA, the typical categorization process
relies on the use of importance measures as a screening method to
assign the preliminary safety significance of SSCs. (Other
methodologies such as success path identification methodologies can
also be used, however, this discussion will focus on the use of
importance measures because these are the most commonly used methods to
identify safety significance of SSCs using a PRA, for example, in the
implementation of Sec. 50.65). The determination of the safety
significance of SSCs by importance measures is also important because
it can identify potential risk outliers and therefore, changes that
exacerbate these outliers can be avoided; and it can facilitate IDP
deliberations of SSCs that are not modeled in the PRA, for example,
events from the ranked list can be used as surrogates for those SSCs
that are not modeled or are only implicitly modeled in the PRA.
For SSCs modeled in the PRA, SSC importance is effectively
determined (see Sec. 50.69(c)(1)(iv)) based on both CDF and LERF.
Importance measures should be chosen so that the IDP can be provided
with information on the relative contribution of an SSC to total risk.
Examples of importance measures that can accomplish this are: the
Fussell-Vesely (F-V) importance and the Risk Reduction Worth (RRW)
importance. Importance measures should also be used to provide the IDP
with information on the margin available should an SSC fail to
function. The Risk Achievement Worth (RAW) importance and the Birnbaum
importance are example measures that are suitable for this purpose.
In choosing screening criteria to be used with the PRA importance
measures, it should be noted that importance measures do not directly
relate to changes in the absolute value of risk. Therefore, the final
criteria for categorizing SSCs into the safety significant and the low
safety significant categories must be based on an assessment of the
potential overall impact of SSC categorization and a comparison of this
potential impact to the acceptance guidelines for changes in CDF and
LERF. However, typically in the initial screening stages, an SSC with
F-V < 0.005 based on CDF and LERF, and RAW < 2 based on CDF and LERF
can be considered as potentially low safety-significant. In addition,
the appropriateness of the importance measures in specifically
addressing SSC CCF contributions and associated screening criteria
should be considered. IDP consideration of Sec. 50.69(c)(1)(ii),
(c)(1)(iii), and (c)(1)(iv) should be carried out to confirm the low
safety significance of these SSCs.
In determining the safety significance of SSCs, consideration
should be given to the potential for the multiple failure modes for the
SSC. PRA basic events represent specific failure events and failure
modes of SSCs. The determination of SSC safety significance should take
into account the effects of all associated basic PRA events (such as
failure to start and failure to run), including indirect contributions
through associated common cause failure (CCF) events.
Because importance measures are typically evaluated on the basis of
individual events, single-event importance measures have the potential
to dismiss all elements of a system or group despite the system or
group having a high importance when taken as a whole. Conversely, there
may be grounds for screening groups of SSCs, owing to the unimportance
of the systems of which they are elements. One approach around this
problem is to first determine the importance of system functions
performed by the selected plant systems. If necessary, each component
in a system is then evaluated to identify the system function(s)
supported by that component. SSCs may be initially assigned the same
category as the most limiting system function they support. System
operating configuration, reliability history, recovery time available,
and other factors can then be considered when evaluating the effect on
categorization from an SSC's redundancy or diversity. The primary
consideration in the process is whether the failure of an SSC will fail
or severely degrade the safety function. If the answer is no, then a
licensee may factor into the categorization the SSC's redundancy, as
long as the SSC's reliability credited in the categorization process
and that of its redundant counterpart(s) have been taken into account.
When the PRA used in the importance analyses includes models for
external initiating events and/or plant operating modes other than full
power, caution should be used when considering the results of the
importance calculations. The PRA models for external initiating events
(e.g., events initiated by fires or earthquakes) and for low power and
shutdown plant operating modes may be more conservative and have a
greater degree of uncertainty than for internal initiating events. Use
of conservative models can influence the calculation of importance
measures by moving more SSCs into the low safety significance category.
Therefore, when PRA models for external event initiators and for the
low power and shutdown modes of operation are available and used, the
importance measures should be evaluated for each analysis separately
and collectively, and the results of these evaluations should be
provided to the IDP.
[[Page 68038]]
As part of the demonstration of PRA adequacy, the sensitivity of
SSC importance to uncertainties in the parameter values for component
availability/reliability, human error probabilities, and CCF
probabilities should be evaluated. Results of these sensitivity
analyses should be provided to the IDP. The following should be
considered in IDP deliberations on the sensitivity study results:
(1) The change in event importance when the parameter value is
varied over its uncertainty range for the event probability can in some
cases provide SSC categorization results that are different. Therefore,
in considering the sensitivity of component categorization to
uncertainties in the parameter values, the IDP should ensure that SSC
categorization is not affected by data uncertainties.
(2) PRAs typically model recovery actions, especially for dominant
accident sequences. Estimating the success probability for the recovery
actions involves a certain degree of subjectivity. The concerns in this
case stem from situations where very high success probabilities are
assigned to a sequence, resulting in related components being ranked as
low risk contributors. Furthermore, it is not desirable for the
categorization of SSCs to be impacted by recovery actions that
sometimes are only modeled for the dominant scenarios. Sensitivity
analyses should be used to show how the SSC categorization would change
if recovery actions were removed. The IDP should ensure that the
categorization is not unduly impacted by the modeling of recovery
actions.
(3) CCFs are modeled in PRAs to account for dependent failures of
redundant components within a system. CCF probabilities can impact PRA
results by enhancing or obscuring the importance of components. A
component may be ranked as a high risk contributor mainly because of
its contribution to CCFs or a component may be ranked as a low risk
contributor mainly because it has negligible or no contribution to
CCFs. The IDP should ensure that the categorization is not unduly
impacted by the modeling of CCFs. The IDP should also be aware that
removing or relaxing requirements may increase the CCF contribution,
thereby changing the risk impact of an SSC.
V.4.2 Section 50.69(c)(1)(ii) Integrated Assessment of SSC Function
Importance
Section 50.69(c)(1)(ii) contains requirements for an integrated,
systematic process to address events including those not modeled in the
PRA, including both design basis and severe accident functions. For
various reasons, many SSCs in the plant will not be modeled explicitly
in the PRA. Therefore, the categorization process must determine the
safety significance of these SSCs by other means. Because importance
measures are not available for use as screening, other criteria or
considerations must be used by the IDP to determine the significance.
Guidance on how these deliberations should be conducted is included in
the NRC regulatory guidance associated with this rule, and in the
industry guidance.
Section 50.69(c)(1)(ii) requires that all aspects of the processes
used to categorize SSC must ``reasonably reflect'' the current plant
configuration, operating practices, and applicable operating
experience. The terminology, ``reasonably reflect,'' was selected to
allow for appropriate PRA modeling and also to make clear that the PRA
and categorization processes do not need to be instantaneously revised
when a plant change occurs (see also requirements in Sec. 50.69(e)(1)
on PRA updating).
V.4.3 Section 50.69(c)(1)(iii) Maintaining Defense-in-Depth
Section 50.69(c)(1)(iii) requires that the categorization process
maintain defense-in-depth. To satisfy this requirement, when
categorizing SSCs as low safety significant, the IDP must demonstrate
that defense-in-depth is maintained. Defense-in-depth is adequate if
the overall redundancy and diversity among the plant's systems and
barriers is sufficient to ensure the risk acceptance guidelines
discussed in Section V.4.4 are met, and that:
Reasonable balance is preserved among prevention of core
damage, prevention of containment failure or bypass, and mitigation of
consequences of an offsite release.
System redundancy, independence, and diversity is
preserved commensurate with the expected frequency of challenges,
consequences of failure of the system, and associated uncertainties in
determining these parameters.
There is no over-reliance on programmatic activities and
operator actions to compensate for weaknesses in the plant design.
Potential for common cause failures is taken into account.
The Commission's position is that the containment and its systems
are important in the preservation of defense-in-depth (in terms of both
large early and large late releases). Therefore, as part of meeting the
defense-in-depth principle, a licensee should demonstrate that the
function of the containment as a barrier (including fission product
retention and removal) is not significantly degraded when SSCs that
support the functions are moved to RISC-3 (e.g., containment isolation
or containment heat removal systems). The concepts used to address
defense-in-depth for functions required to prevent core damage may also
be useful in addressing issues related to those SSCs that are required
to preserve long-term containment integrity. Where a licensee
categorizes containment isolation valves or penetrations as RISC-3, the
licensee should address the impact of the change in treatment to ensure
that defense-in-depth continues to be satisfied. Where the impact of
changes in treatment does not support the reliability assumptions in
the categorization process, the licensee should resolve this situation
by adjustments to the categorization process assumptions or treatment
of the component.
V.4.4 Section 50.69(c)(1)(iv) Include Evaluations To Provide Reasonable
Confidence That Sufficient Safety Margins Are Maintained and That Any
Potential Increases in CDF and LERF Resulting From Changes in Treatment
Permitted by Implementation of Sec. 50.69(b)(1) and Sec. 50.69(d)(2)
Are Small
Section 50.69(c)(1)(iv) specifies that the categorization process
include evaluations to provide reasonable confidence that as a result
of implementation of revised treatment permitted for RISC-3 SSCs,
sufficient safety margins are maintained and any potential increases in
CDF and LERF are small. Safety margins can be maintained if the
licensee maintains the functionality of the SSCs following
implementation of the revised requirements and if periodic inspection,
testing, and corrective action activities are adequate to prevent,
detect and correct significant SSC performance and reliability
degradation. Later sections of this SOC provide discussion on the
treatment the licensee will implement to ensure with reasonable
confidence that RISC-3 SSCs remain capable of performing their safety
functions under design basis conditions. The requirements of the rule
to show that sufficient safety margins are maintained and that
potential increases in risk are acceptably small are discussed below.
As part of their submittal, a licensee or applicant is to describe
the evaluations to be conducted for purposes of providing reasonable
confidence that there would be no more than an acceptably small
(potential) increase in risk. For SSCs included in the PRA, the
Commission expects a risk sensitivity study (evaluation) to be
[[Page 68039]]
performed to provide a basis for concluding that if the reliability of
these RISC-3 SSCs should collectively degrade because of the changes in
treatment, the potential risk increase would be small. Satisfying the
rule requirement that the risk increase is acceptably small presumes
that the increase in failure rates credited in the PRA risk sensitivity
study bounds any reasonable estimate of the increase that may be
expected as a result of the changes in treatment; also considering the
feedback and corrective action aspects of the rule.
The categorization process encompasses both active and passive
functions of SSCs. Section 50.69(b)(2)(iv) includes the requirement
that the change-in-risk evaluations performed to satisfy Sec.
50.69(c)(1)(iv) must address potential impacts from known degradation
mechanisms on both active and passive functions. The manner of
addressing these potential impacts may be either qualitative or
quantitative and may rely on the maintenance of current programs that
address these degradation mechanisms (e.g., microbiologically-induced
corrosion, flow-assisted corrosion) and/or may incorporate existing
risk-informed approaches (e.g., risk-informed inservice inspection).
One mechanism that could lead to large increases in CDF/LERF is
extensive, across system common cause failures. These CCFs could occur
where the mechanisms that lead to failure, in the absence of special
treatment, are sufficiently rapidly developing or are not self-
revealing that there would be few opportunities for early detection and
corrective action. Thus, when deciding how much to assume that SSC
reliability might change, the applicant or licensee is expected to
consider potential effects of common-cause interaction susceptibility,
including cross-system interactions and potential impacts from known
degradation mechanisms; while also considering the feedback and
corrective actions aspects of the rule.
Those aspects of treatment that are necessary to prevent SSC
degradation or failure from known degradation mechanisms, to the extent
that the results of the evaluations are invalidated, must be retained.
Identifying those aspects will involve an understanding of what the
degradation mechanisms are and what elements of treatment are
sufficient to prevent the degradation.
The treatment for all RISC-3 SSCs may not be the same. As an
example, motor operated valves (MOVs) operating in a severe environment
(e.g., in the steam tunnel) would be more susceptible to failure
because of grease degradation if they were not regularly maintained and
tested. However, not all MOVs, even if they have the same design and
are identical in other respects, will be exposed to the same
environment. Therefore, the other MOVs may not be as susceptible to
failure as those in the steam tunnel and less frequent maintenance and
testing would be acceptable. While it may be simpler to increase the
unreliability or unavailability of all the RISC-3 SSCs by a certain
bounding factor to demonstrate that the change in risk is acceptably
small, this example suggests that it may also be appropriate to use
different factors for different groups of SSCs depending on the impact
of reducing treatment on those SSCs.
Section 50.69(c)(1)(iv) requires reasonable confidence that the
increase in the overall plant CDF and LERF resulting from potential
decreases in the reliability of RISC-3 SSCs as a result of the changes
in treatment, be small. The rule further requires the licensee or
applicant to describe the evaluations to be performed to meet this
requirement. As presented in RG 1.174, the NRC considers small changes
to be relative and to depend on the current plant CDF and LERF (hence
we also refer to ``acceptably small'' changes in other portions of this
notice since small can be different for different plants with different
baseline levels of risk). For plants with total baseline CDF of
10-4 per year or less, small means CDF increases of up to
10-5 per year and for plants with total baseline CDF greater
than 10-4 per year, small means CDF increases of up to
10-6 per year. However, if there is an indication that the
CDF may be considerably higher than 10-4 per year, the focus
of the licensee should be on finding ways to decrease rather than
increase CDF and the licensee may be required to present arguments as
to why steps should not be taken to reduce CDF for the reduction in
special treatment requirements to be considered. For plants with total
baseline LERF of 10-5 per year or less, small LERF increases
are considered to be up to 10-6 per year, and for plants
with total baseline LERF greater than 10-5 per year, small
LERF increases are considered to be up to 10-7 per year.
However, if there is an indication that the baseline CDF or LERF may be
considerably higher than 10-4 or 10-5,
respectively, the licensee either must find ways to reduce risk and
present the arguments to the staff before implementation of Sec.
50.69, otherwise it is likely that the staff will reject the Sec.
50.69 application. This is consistent with the guidance in Section
2.2.4 of RG 1.174. It should be noted that this allowed increase shall
be applied to the overall categorization process, even for those
licensees that will implement Sec. 50.69 in a phased manner.
If a PRA model does not exist for the external initiating events or
the low power and shutdown operating modes, justification should be
provided, on the basis of bounding analyses or qualitative
considerations, that the effect on risk (from the unmodeled events or
modes of operation) is not significant and that the total effect on
risk from modeled and unmodeled events and modes of operation is small,
consistent with Section 2.2.4 of RG 1.174.
V.4.5 Section 50.69(c)(1)(v) System or Structure Level Review
Section 50.69(c)(1)(v) specifies that the categorization be done at
the system or structure level; not for selected components within a
system. A licensee or applicant is allowed to implement Sec. 50.69 for
a subset of the plant systems and structures (i.e., partial
implementation) and to phase in implementation over time. However, the
implementation, including the categorization process, must address
entire systems or structures; not selected components within a system
or structure. Note that this requirement should be understood to
exclude entire support systems (e.g., if system A is categorized as
RISC-3, but is dependent on system B components which in turn have been
categorized as RISC-1, then system A is understood not to include the
system B components and is not to be categorized as RISC-1). This
required scope ensures that all safety functions associated with a
system or structure are properly identified and evaluated when
determining the safety significance of individual components within a
system or structure and that the entire set of components that comprise
a system or structure are considered and addressed.
V.4.6 Section 50.69(c)(2) Use of Integrated Decision-Making Panel
Section 50.69(c)(2) sets forth the requirements for using an IDP to
make the determination of safety significance, and for the composition
of the IDP. The fundamental requirement for the categorization process
(as stated in Sec. 50.69(c)(1)(ii)) is that it include use of an
integrated systematic process. The determination of safety significance
of SSCs is to be performed as part of an integrated decision-making
process. By ``integrated decision-making process,'' the Commission
means a process that integrates both risk insights and
[[Page 68040]]
traditional engineering insights. In categorizing SSCs as low safety-
significant, defense-in-depth must be maintained (per Sec.
50.69(c)(1)(iii)) and there must be reasonable confidence that
sufficient safety margin is maintained by showing that any increases in
risk are small per Sec. 50.69(c)(1)(iv). To account for each of these
factors and to account for risk insights not found in the plant-
specific PRA, Sec. 50.69(c)(2) requires that the final categorization
of each SSC be performed using an integrated decision-making panel
(IDP). A structured and systematic process using documented criteria
must be used to guide the decision-making process. Categorization is an
iterative process based on expert judgment to integrate the qualitative
and quantitative elements that impact SSC safety significance. The
insights and varied experience of IDP members are relied on to ensure
that the final result reflects a comprehensive and justifiable
judgment.
The panel must be composed of experienced personnel who possess
diverse knowledge and insights in plant design and operation and who
are capable in the use of deterministic knowledge and risk insights in
making SSC classifications. The NRC places significant reliance on the
capability of a licensee to implement a robust categorization process
that relies heavily on the skills, knowledge, and experience of the
people that implement the process, in particular on the qualifications
of the members of the IDP. The IDP must be composed of a group of
individuals who collectively have expertise in plant operation, design
(mechanical and electrical) engineering, system engineering, safety
analysis, and probabilistic risk assessment. At least three members of
the IDP should have a minimum of five years experience at the plant,
and there should be at least one member of the IDP who has worked on
the modeling and updating of the plant-specific PRA for a minimum of
three years.
The IDP should be trained in the specific technical aspects and
requirements related to the categorization process. Training should
address at a minimum the purpose of the categorization; present
treatment requirements for SSCs including requirements for design basis
events; PRA fundamentals; details of the plant-specific PRA including
the modeling, scope, and key assumptions, the interpretation of risk
importance measures, and the role of sensitivity studies and the
change-in-risk evaluations; and the defense-in-depth philosophy and
requirements to maintain defense-in-depth.
The licensee or applicant (through the IDP) shall document its
decision criteria for categorizing SSCs as safety significant or low
safety significant pursuant to Sec. 50.69(f)(1). Decisions of the IDP
should be arrived at by consensus. Differing opinions should be
documented and resolved, if possible. If a resolution cannot be
achieved concerning the safety significance of an SSC, then the SSC
should be classified as safety-significant. SSC categorization shall be
revisited by the licensee or applicant (through the IDP) when the PRA
is updated or when the other criteria used by the IDP are affected by
changes in plant operational data or changes in plant design or plant
procedures. Requirements for PRA updating are contained in Sec.
50.69(e)(1).
V.5.0 Section 50.69(d) Treatment Requirements for Structures, Systems,
and Components
Treatment requirements applicable to RISC-1, RISC-2, and RISC-3
SSCs are specified in Sec. 50.69(d). Any regulatory requirements
applicable to RISC-1, RISC-2, RISC-3, and RISC-4 SSCs not removed by
Sec. 50.69(b)(1) continue to apply.
V.5.1 Section 50.69(d)(1) RISC-1 and RISC-2 Treatment
Section 50.69(d)(1) requires that a licensee or applicant ensure
that RISC-1 and RISC-2 SSCs perform their functions consistent with the
categorization process assumptions by evaluating treatment being
applied to these SSCs to ensure that it supports the key assumptions in
the categorization process that relate to their assumed performance.
This rule language means that the licensee or applicant must evaluate
the treatment associated with those key assumptions in the PRA that
relate to performance of particular SSCs. For example, if a relief
valve was being credited with capability to relieve water (as opposed
to its design condition of steam), such an evaluation would look at
whether the component has been determined to be able to perform as
assumed.
Because RISC-1 and RISC-2 SSCs are the safety significant SSCs and
their performance as credited in the PRA is important to maintaining an
acceptable level of plant risk, given that special treatment
requirements are being removed from RISC-3 SSCs, it is a key and
necessary part of Sec. 50.69 to ensure these SSCs can perform as
credited in the PRA. However, the requirements in Sec. 50.69(d)(1) do
not extend special treatment requirements to RISC-1 beyond design basis
functions and to RISC-2 SSCs.
The performance conditions for beyond design basis capabilities of
RISC-1 SSCs credited in the PRA are not subject to the requirements of
10 CFR Part 50, Appendix B. However, plant SSCs credited for beyond
design basis capabilities must have a valid technical basis for the
credit (i.e., the failure rate/probability of the SSC performing the
beyond design basis function) given in the PRA. Further, the basis for
this credit should already be established and documented in the PRA
supporting documentation so this should not be an additional burden for
licensees to capture and implement. If an existing technical basis does
not exist or is insufficient to support the credit taken for beyond
design basis capability (e.g., the supporting test program does not
test the SSC at the beyond design basis conditions), the licensee or
applicant is required by Sec. 50.69(d)(1) to develop a technical basis
for the credit taken in the PRA potentially including a treatment
program for the SSC that validates the capability credited.
For SSCs categorized as RISC-1 or RISC-2, all existing applicable
requirements continue to apply (i.e., no special treatment requirements
are removed by Sec. 50.69). This rule does not require licensees to
evaluate the effectiveness of special treatment requirements for RISC-1
SSCs to ensure that they are capable of performing their design basis
functions. The special treatment requirements in other NRC regulations
address the design basis capability of RISC-1 SSCs.
The categorization process will result in a number of safety-
related SSCs being determined to be of low safety significance (i.e.,
RISC-3) and subject to reduced treatment. This determination of low
safety significance will implicitly take credit for the performance
capability of other SSCs in the PRA, some, or all of which, may not be
included in the scope of the licensee's categorization process (due to
the allowance for licensees to selectively implement the rule and to
phase that implementation over time). To maintain the validity of the
categorization process, and more importantly to maintain any potential
risk increase as small, it is necessary to maintain the ``credited''
SSCs per Sec. 50.69, and this means the application of Sec.
50.69(d)(1) and Sec. 50.69(e)(2) requirements.
V.5.2 Section 50.69(d)(2) RISC-3 Treatment
Section 50.69(d)(2) requires that the licensee or applicant must
ensure with reasonable confidence that RISC-3 SSCs
[[Page 68041]]
remain capable of performing their safety-related functions under
design basis conditions, including seismic conditions and environmental
conditions and effects throughout their service life. By ``reasonable
confidence'', the Commission means that the licensee or applicant is
required to provide a ``reasonable confidence'' level with regard to
maintaining the capability of RISC-3 safety-related functions. As
indicated previously in this notice, ``reasonable confidence'' is a
level of confidence that is both less than that associated with RISC-1
SSCs which are subject to all the special treatment requirements, and
consistent with their individual low safety significance. The term
``ensure'' is intended to convey the Commission's determination that
the licensee is under a legally-binding regulatory requirement to
provide the requisite ``reasonable confidence.''
Although Sec. 50.69(b)(1) removes for RISC-3 SSCs the
environmental qualification requirements of Sec. 50.49, it does not
eliminate the requirements in 10 CFR part 50, Appendix A, ``General
Design Criteria for Nuclear Power Plants,'' that electric equipment
important to safety be capable of performing their intended functions
under the applicable environmental conditions. For example, GDC-4 of 10
CFR part 50, Appendix A, requires that SSCs important to safety be
designed to accommodate the effects of, and to be compatible with, the
environmental conditions and effects associated with normal operation,
maintenance, testing, and postulated accidents. To satisfy the
provisions of GDC-4 of 10 CFR part 50, Appendix A, the licensee or
applicant must address environmental conditions such as temperature,
pressure, humidity, chemical effects, radiation, and submergence; and
environmental effects such as aging and synergisms. In accordance with
Sec. 50.69(d)(2), the licensee or applicant must design electric
equipment important to safety so they are capable of performing their
intended functions under applicable environmental conditions and
effects throughout their service life. If RISC-3 electrical equipment
is relied on to perform its safety-related function beyond its design
life, Sec. 50.69(d)(2) requires the licensee or applicant to have a
basis for the continued capability of the equipment under adverse
environmental conditions and effects.
Under Sec. 50.69, RISC-3 SSCs would continue to be required to
function under design basis seismic conditions (such as design load
combinations of normal and accident conditions with earthquake
motions), but would not be required to be qualified by testing or
specific engineering methods in accordance with the requirements stated
in 10 CFR part 100, Appendix A. A licensee or applicant who adopts the
rule would no longer be required to meet certain requirements in
Appendix A to part 100, Sections VI(a)(1) and VI(a)(2), to the extent
that those requirements have been interpreted as mandating
qualification testing and specific engineering methods to demonstrate
that RISC-3 SSCs are designed to withstand the Safe Shutdown Earthquake
and Operating Basis Earthquakes. The rule does not remove the design
requirements related to the capability of RISC-3 SSCs to remain
functional considering Safe Shutdown Earthquake and Operating Basis
Earthquake seismic loads, including applicable concurrent loads. The
rule does not change the design input earthquake loads (magnitude of
the loads and number of events) or the required load combinations used
in the design of RISC-3 SSCs. For example, for the replacement of an
existing safety-related SSC that is subsequently categorized as RISC-3,
the same seismic design loads and load combinations would still apply.
The rule would permit the licensee or applicant to select a technically
defensible method to show that RISC-3 SSCs will remain functional when
subject to design earthquake loads. Several public comments on the
proposed rule supported the use of earthquake experience data as a
method to demonstrate SSCs will remain functional during earthquakes.
If the licensee or applicant chooses to use only earthquake experience
data to demonstrate that the SSC will perform its safety-related
function, with no further engineering evaluation, then the earthquake
experience data must envelope the SSC design basis, including the
number of earthquake events and the design load combinations.
Additionally, if the SSC is required to function during or after the
earthquake, the experience data would need to contain explicit
information that the SSC actually functioned during or after the design
basis earthquake events as required by the SSC design basis. The
successful performance of an SSC after the earthquake event does not
demonstrate it would have functioned during the event. Implementation
of Sec. 50.69 does not change the seismic design basis for USI A-46
facilities and, therefore, does not impose additional requirements on
these facilities.
Section 50.69(d)(2) should not be interpreted to extend or expand
design basis conditions to SSCs where such conditions were not
previously part of its design basis.
Section 50.69(d)(2) requires that the treatment of RISC-3 SSCs be
consistent with the categorization process. This rule language means
that, when establishing the treatment for RISC-3 SSCs, the licensee or
applicant must take into account the assumptions in the categorization
process regarding the design basis capability and reliability of RISC-3
SSCs to perform their safety-related functions throughout their service
life. The evaluation by the licensee or applicant of the consistency of
the treatment of RISC-3 SSCs with the categorization process may be
qualitative so long as it provides reasonable confidence of the design
basis capability of RISC-3 SSCs, based on plant-specific and industry-
wide operational experience and vendor information. In establishing
treatment for RISC-3 SSCs, the licensee or applicant is responsible for
addressing applicable vendor recommendations and operational experience
such that the treatment established for RISC-3 SSCs provides reasonable
confidence for design basis capability. For example, operational
experience might be described in NRC information notices or identified
in responses to NRC bulletins, generic letters, or other licensee
commitment documents. The treatment applied to RISC-3 SSCs must also
support the assumptions used in justifying the removal of requirements
applicable to those SSCs. For example, where a licensee or applicant
intends as part of implementing Sec. 50.69 to eliminate leakage
testing required in 10 CFR Part 50, Appendix J, for containment
isolation valves, the treatment applied to those valves must support
the assumption that they are capable of closing under design basis
conditions.
Some public comments on the proposed rule suggested that a
reference to general industrial practices would be sufficient to
satisfy the requirements for the treatment for RISC-3 SSCs. However, as
described in NUREG/CR-6752, ``A Comparative Analysis of Special
Treatment Requirements for Systems, Structures, and Components (SSCs)
of Nuclear Power Plants with Commercial Requirements of Non-Nuclear
Power Plants,'' significant variation exists in the application of
industrial practices at nuclear power plants. Hence, a simple reference
to these practices does not provide a basis to satisfy the rule's
requirements. To satisfy the requirement that the treatment of RISC-3
SSCs be consistent with the categorization process, the licensee or
applicant must establish
[[Page 68042]]
treatment that provides reasonable confidence SSCs perform their
safety-related functions under design basis conditions and is
consistent with the assumptions in the categorization process (e.g.,
reliability levels). The licensee or applicant must either establish
treatment that provides this level of reliability or use consensus
standards that provide a proven level of reliability based on
experience. In using consensus standards, the licensee or applicant
must note that combining or omitting provisions of standards might
result in ineffective implementation of Sec. 50.69 by causing RISC-3
SSCs to be incapable of performing their design basis safety functions.
The NRC considers the ASME code cases endorsed in Sec. 50.55a and
listed in RG 1.84, 1.147, and 1.192 to be one acceptable method of
establishing treatment of RISC-3 SSCs, where applicable, in that those
applicable endorsed code cases adjust treatment based on the safety
significance of the components.
Under Sec. 50.69, most special treatment requirements will be
removed from RISC-3 SSCs, which will typically comprise a large
percentage of safety-related SSCs in a nuclear power plant. These
special treatment requirements will be replaced with the high-level
treatment requirements in Sec. 50.69(d)(2) that will allow significant
reduction in the treatment applied to RISC-3 SSCs. This reduction in
treatment can introduce common-cause concerns and weaken defenses
against them. Therefore, Sec. 50.69(d)(2) requires that inspection,
testing and corrective action be provided for RISC-3 SSCs. The
inspection and testing requirement in Sec. 50.69(d)(2)(i) is to
provide sufficient performance data for RISC-3 SSCs to determine if the
reduction in treatment has adversely affected their design basis
capability and to provide reasonable confidence that the SSC can
perform its safety function throughout their service life. The
corrective action requirement in Sec. 50.69(d)(2)(ii) is to address
SSC failures and provide reasonable confidence in avoiding future
problems. These requirements are necessary to provide reasonable
confidence that RISC-3 safety related functional capability is
maintained and thereby avoid adverse impacts on the reliability and
availability of multiple RISC-3 SSCs, which could reduce plant safety
beyond the categorization process assumptions or results and invalidate
the risk sensitivity results.
A licensee or applicant may not simply assume that a sensitivity
study that increases the failure probability for all RISC-3 SSCs
simultaneously, with no additional basis to support it, would
necessarily bound the potential change in risk that could result due to
implementation of Sec. 50.69. There is a potential that risk due to
implementation of Sec. 50.69 could increase as a result of the
reduction in treatment due to common-cause interactions or degradation,
and this impact might not be uniform across the population of RISC-3
SSCs. For example, if a licensee were to simply eliminate maintenance,
testing, or lubrication of pumps or valves, it could significantly
impact performance of those specific components and the impact might
exceed the cumulative impact of individually reducing the reliability
of all RISC-3 SSCs by a few percent or less. Public comments on the
proposed rule indicated that cross-system common-cause interactions and
degradation mechanisms are typically addressed through the treatment
processes applied to plant equipment, rather than being addressed in
the categorization process. In satisfying the rule, the licensee or
applicant must consider potential common-cause interactions and
degradation mechanisms in establishing treatment for RISC-3 SSCs so
there is a reasonable basis to support the assumptions made for the
risk sensitivity study.
V.5.2.1 Section 50.69(d)(2)(i) Inspection and Testing
Section 50.69(d)(2)(i) requires the licensee to conduct periodic
inspection and testing activities to determine whether RISC-3 SSCs will
remain capable of performing their safety-related functions under
design basis conditions.
The prescriptive special treatment requirements in Sec. Sec.
50.55a and 50.65 for inspection, testing, and surveillance have been
removed for RISC-3 SSCs. In lieu of those prescriptive requirements,
the final rule requires the licensee or applicant to implement
inspection and testing of RISC-3 SSCs sufficient to provide reasonable
confidence that RISC-3 SSCs remain capable of performing their safety-
related functions under design basis conditions throughout their
service life. The licensee or applicant may apply industrial practices
for the treatment of RISC-3 SSCs if those practices maintain the
capability of the RISC-3 SSCs to perform their design-basis safety
functions.
With respect to RISC-3 pumps and valves, the rule language in Sec.
50.69(d)(2)(i) means that the licensee or applicant must implement
periodic testing or inspection sufficient to provide reasonable
confidence that these pumps and valves will be capable of performing
their safety-related functions under design basis conditions. To
determine that the pump or valve will remain capable of performing its
safety-related function, the licensee or applicant will need to obtain
sufficient operational information or performance data to provide with
reasonable confidence that the RISC-3 pumps and valves will be capable
of performing their safety-related functions if called upon to function
under operational or design basis conditions over the interval between
periodic testing or inspections. In addition, the operational
information and performance data must be sufficient to satisfy the
requirements of Sec. 50.69(d)(2)(i) for use in identifying the need
for corrective action under Sec. 50.69(d)(2)(ii) and in providing
information for feedback to the categorization and treatment processes
under Sec. 50.69(e)(3).
In some cases, a licensee or applicant implementing Sec. 50.69
might apply more rigorous test methods than previously applied to
satisfy the ASME Code inservice testing provisions because Sec. 50.69
does not specify restrictive time limits on test intervals that were
provided in the ASME Code. As a result, Sec. 50.69 allows significant
flexibility by the licensee or applicant in verifying the design basis
capability of their safety-related SSCs categorized as RISC-3. However,
the licensee or applicant needs to consider the lessons learned over
the last 20 years regarding SSC performance in establishing the
treatment for RISC-3 SSCs. Contrary to suggestions in some public
comments on the proposed rule, operating experience and research does
not support an assumption that exercising a valve or pump will provide
reasonable confidence of design-basis capability in that such
exercising will not detect service-induced aging or degradation that
could prevent the component from performing its design basis functions
in the future, and therefore is insufficient by itself to satisfy Sec.
50.69(d)(2)(i). The licensee or applicant may develop the type and
frequency of tests or inspections for RISC-3 pumps and valves provided
they are sufficient to conclude that the pump or valve will perform its
safety-related function throughout the service life. The provisions for
risk-informed inspection and testing in applicable ASME code cases (as
incorporated in Sec. 50.55a) would constitute one effective approach
for satisfying the Sec. 50.69 requirements.
[[Page 68043]]
V.5.2.2 Section 50.69(d)(2)(ii) Corrective Action Process
Section 50.69(d)(2)(ii) requires that conditions that would prevent
a RISC-3 SSC from performing its safety-related functions under design
basis conditions must be corrected in a timely manner. In the case of
significant conditions adverse to quality, the rule requires that
measures be taken to provide reasonable confidence that the cause of
the condition is determined and corrective action taken to preclude
repetition. Significant conditions adverse to quality include common-
cause concerns for multiple RISC-3 SSCs or concerns related to the
validity of the categorization process or its results. For example, if
measuring and test equipment is found to be in error or defective, the
licensee or applicant will be responsible for determining the
functionality of safety-related SSCs checked using that equipment to
prevent the occurrence of common-cause problems that might invalidate
the categorization process assumptions and results. Effective
implementation of the corrective action process would include timely
response to information from plant SSCs, overall plant operations, and
industry generic activities that might reveal performance concerns for
RISC-3 SSCs on both an individual and common-cause basis. Contrary to
some public comments on the proposed rule, the corrective action
process alone is insufficient to monitor the effects of reduced
treatment on RISC-3 SSCs, and therefore the Commission has incorporated
feedback requirements into Sec. 50.69.
V.6.0 Section 50.69(e) Feedback and Process Adjustment
Section 50.69(e)(1) requires the licensee or applicant to review
changes to the plant, operational practices, applicable plant and
industry operational experience and, as appropriate, update the PRA and
SSC categorization and treatment processes, in a timely manner, but no
longer than every two refueling outages for RISC-1, RISC-2, RISC-3, and
RISC-4 SSCs. The date the NRC grants the license amendment to implement
10 CFR 50.69 begins the updating interval and provides a recognizable
date for the periodic updating of the categorization and treatment
processes. Depending on the timing of license amendment issuance (for
example, just before a refueling outage), the licensee or applicant
might have minimal plant changes, operational practices, or operational
experience to review in updating the categorization and treatment
processes in the early phases of implementing the rule. If plant
changes, operational practices, or operational experience would result
in a significant adverse impact on plant safety or public health and
safety, the licensee or applicant must update the categorization or
treatment processes in a timely manner without waiting for the two
refueling outage schedule. The information collected under Sec.
50.69(e)(2) and (e)(3) would be among the information used to determine
the need for updating the categorization or treatment processes in a
timely manner required under Sec. 50.69(e)(1). The plant and industry
operational experience referred to in Sec. 50.69(e)(1) includes the
data collected under Sec. 50.69(e)(3) for RISC-3 SSCs. In addition to
the periodic updating of the quantitative reliability information, the
feedback of plant operational experience is intended to include
qualitative information on the performance of plant SSCs obtained
through the corrective action program and processes as well as from
applicable vendor recommendations and operational experience. For
example, lessons learned from operational experience might be described
in NRC information notices or implemented in response to NRC bulletins
or generic letters. The evaluation of the categorization process
includes verifying the continued validity of the risk sensitivity study
and the associated SSC performance assumptions.
Section 50.69(e)(2) requires the licensee or applicant to monitor
the performance of RISC-1 and RISC-2 SSCs and make adjustments as
necessary to either the categorization (i.e., by moving other RISC-3 or
RISC-4 SSCs back into RISC-1 or RISC-2 until the change in risk is
acceptably small) or treatment processes so the categorization process
and results are maintained valid. To meet this requirement, the
licensee or applicant must monitor all unavailabilities and functional
failures so they can determine when adjustments to the categorization
or treatment processes are needed. The licensee or applicant will also
need to monitor SSCs that are credited in the PRA for performing beyond
design basis functions (if applicable) that are not necessarily
included in the scope of an existing maintenance rule program.
The categorization process will result in a number of safety-
related SSCs being determined to be of low safety significance (i.e.,
RISC-3) and subject to reduced treatment. This determination of low
safety significance will implicitly take credit for the performance
capability of other SSCs in the PRA, some, or all of which, may not be
included in the scope of the licensee's categorization process (due to
the allowance for licensees to selectively implement the rule and to
phase that implementation over time). To maintain the validity of the
categorization process, and more importantly to maintain any potential
risk increase as small, it is necessary to maintain the ``credited''
SSCs per Sec. 50.69.
In Sec. 50.69(e)(3) the rule requires the licensee or applicant to
consider the performance data collected in Sec. 50.69(d)(2)(i) for
RISC-3 SSCs to determine whether there are any adverse changes in
performance such that the SSC unreliability values approach or exceed
the values used in the evaluations conducted to meet Sec. 50.69(c)(iv)
and to make adjustments as necessary to either the categorization or
treatment processes so the categorization process and results are
maintained valid. Based on the review of this information, if SSC
reliability degrades so as not to support the categorization process
assumptions, the licensee or applicant must adjust the treatment to
improve SSC reliability or make appropriate changes to the
categorization of SSCs.
V.7.0 Section 50.69(f) Program Documentation and Change Control and
Records
Section 50.69(f) contains administrative requirements for keeping
information current, handling planned changes to programs and
processes, and records. Each requirement is discussed below.
Section 50.69(f)(1) states that the licensee or applicant shall
document the basis for categorization of SSCs in accordance with this
section before removing any requirements. The documentation must
address why a component was determined to be either safety significant
or low safety significant based upon the requirements in Sec.
50.69(c).
Section 50.69(f)(2) specifies that the licensee must update its
FSAR to reflect which systems have been categorized using the
provisions of Sec. 50.69. Systems that are categorized by Sec. 50.69
will have their treatment revised consistent with the RISC category
into which the SSC is categorized and the associated treatment
requirements of Sec. 50.69(d). This provision is included to maintain
clear information, at a minimum level of detail, about which
requirements a licensee is satisfying. However, detailed information
about particular SSCs is not required to be submitted to the NRC. For
an applicant, this updating would be expected to be either part of the
original
[[Page 68044]]
application or as a supplement to the FSAR under Sec. 50.34(b). For
licensees, the updating must be in accordance with the provisions of
Sec. 50.71(e).
Once the NRC has completed its review of a Sec. 50.69 application,
the licensee can adjust its treatment processes provided that the
requirements of Sec. 50.69 are met. NRC does not plan to perform a
pre-implementation review of the revised treatment requirements under
Sec. 50.69(d). However, the Commission recognizes that existing
information in the quality assurance (QA) plan or in the FSAR may need
to be revised to reflect the changes to treatment that are made as a
result of implementation of Sec. 50.69. Any revisions to these
documents are to be submitted to NRC in accordance with the existing
requirements of Sec. 50.54(a)(2) and Sec. 50.71(e), respectively.
Section 50.69(f)(3) specifies that for initial implementation of
the rule, changes to the FSAR for implementation of this rule need not
include a supporting Sec. 50.59 evaluation of changes directly related
to implementation. Future changes to the treatment processes and
procedures for Sec. 50.69 implementation may be made, provided the
requirements of the rule and Sec. 50.59 continue to be met. While the
licensee is to update its programs to reflect implementation of Sec.
50.69, the Commission concluded that no additional review under Sec.
50.59 is necessary for such changes to these parts of the FSAR that
might occur.
Section 50.69(f)(4) specifies that for initial implementation of
the rule, changes to the quality assurance plan directly related to
implementation of this rule need not be considered a reduction in
commitment for the purposes of Sec. 50.54(a). Future changes to the
treatment processes and procedures for Sec. 50.69 implementation may
also be made, provided the requirements of the rule and Sec. 50.54(a)
continue to be met. While the licensee is to update its programs to
reflect implementation of Sec. 50.69, the Commission concluded that no
additional NRC staff review under Sec. 50.54(a) is necessary for
changes to these parts of the QA plan.
No specific change control process is being established for the
categorization process outlined by Sec. 50.69(c). At this time, the
NRC is unable to determine generic criteria for the control of changes
to the categorization process during its implementation that could be
included in Sec. 50.69. As a result, the NRC will review and approve a
license amendment submittal containing the licensee or applicant's
categorization process and intends to impose a license condition upon
which the categorization process approval is based to control
categorization process changes. The license condition will require the
licensee to notify the NRC in advance of implementing changes with
respect to specific aspects of the categorization process. With
experience in the application of Sec. 50.69, the NRC might modify the
rule to specify generic criteria for the control of changes to the
categorization process during implementation of the rule.
No explicit requirements are included in Sec. 50.69 for the period
for retention of records. The rule specifies only a few specific types
of records that must be prepared (e.g., those for the basis for
categorization in Sec. 50.69(f)(1)). In accordance with Sec.
50.71(c), these records are to be maintained until the Commission
terminates the facility license.
V.8.0 Section 50.69(g) Reporting
Section 50.69(g) provides a new reporting requirement applicable to
events or conditions that prevented, or would have prevented, a RISC-1
or RISC-2 SSC from performing a safety significant function. Most
events involving these SSCs will meet existing Sec. 50.72 and Sec.
50.73 reporting criteria. However, it is possible for events and
conditions to arise that impact whether RISC-1 or RISC-2 SSCs would
perform beyond design basis functions consistent with the performance
capability credited in the categorization process. This reporting
requirement is intended to capture these situations. The reporting
requirement is contained in Sec. 50.69, rather than as a revision of
Sec. 50.73, so that its applicability only to those facilities that
have implemented Sec. 50.69 is clear. The existing reporting
requirements in Sec. 50.72 and Sec. 50.73 are removed for RISC-3 (and
RISC-4) SSCs under Sec. 50.69(b)(vii) and (viii).
V.9.0 Inspection of 10 CFR 50.69 Implementation
The NRC will review and update, as appropriate, the current
inspection procedures under the NRC Reactor Oversight Process to
incorporate inspection guidance for monitoring the implementation of
Sec. 50.69 at nuclear power plants. The NRC intends to conduct sample
inspections of plants implementing Sec. 50.69 in a manner that is
sensitive to conditions that could significantly increase risk. These
sample inspections are intended to gather information that will enable
the NRC to assess whether modifications are needed to the ongoing
baseline inspection program. The sample inspections will focus on the
implementation of the categorization process approved as part of the
NRC review of the Sec. 50.69 license amendment request. The sample
inspections will also evaluate the treatment established under Sec.
50.69 with primary attention directed to programmatic and common-cause
issues; including those associated with known degradation mechanisms.
The inspections might help provide operating experience information on
RISC-3 SSCs that can also be provided to other licensees.
VI. Guidance
VI.1 Regulatory Guide and Implementation Guidance for Sec. 50.69
NEI submitted a proposed implementation guide for this rulemaking
in the form of NEI 00-04, ``10 CFR 50.69 SSC Categorization
Guideline''. As part of the effort to develop the rule, the NRC staff
reviewed drafts of this document and in addition, NEI 00-04 was used in
the pilot programs discussed earlier. The objective of the staff's
review was to determine the acceptability of the proposed implementing
guidance, with the intent that the NEI guidance could be endorsed in an
NRC regulatory guide. The revision of NEI 00-04 submitted on April 14,
2004 forms the basis for the NRC RG ``Guidelines for Categorizing
Structures, Systems and Components in Nuclear Power Plants According to
Their Safety Significance.'' Availability of this document is noted in
Section IX.
The NRC staff's review of NEI 00-04 resulted in several areas where
the staff finds it necessary to identify clarifications, limitations,
and conditions to the NEI guidance or to include further guidance to
supplement the document, as it is currently written. These
clarifications, limitations, and conditions, and the reasons therefore,
are set forth in Section C of RG 1.201. These issues are best resolved
by testing the guide against actual applications. Therefore, this RG is
being issued for trial use. This RG does not establish any final staff
positions, and may be revised in response to experience with its use.
As such, this trial regulatory guide does not establish a staff
position for purposes of the Backfit Rule, 10 CFR 50.109, and any
changes to this RG prior to staff adoption in final form will not be
considered to be backfits as defined in 10 CFR 50.109(a)(1). This will
ensure that the lessons learned from regulatory review of pilot and
follow-on applications are adequately addressed in this document and
that the guidance is sufficient to enhance regulatory
[[Page 68045]]
stability in the review, approval, and implementation in the use of
PRAs and their results in the risk informed categorization process
required by 10 CFR 50.69.
The NRC staff and NEI continue to interact on the implementation
guidance. Consequently, it is expected that NEI will submit an improved
revision to NEI 00-04 that will enable the NRC to issue a RG with fewer
clarifications, limitations, and conditions, and as a consequence, the
NRC is delaying issuance of the RG.
VI.2 Review Guidance Concerning PRA Quality and Peer Review
RG 1.200, ``An Approach for Determining the Technical Adequacy of
Probabilistic Risk Assessment Results for Risk-Informed Activities,''
provides guidance on the NRC position on voluntary consensus standards
for PRA (in particular on the ASME standard for internal events PRAs)
and associated industry PRA documents (e.g., NEI 00-02, ``Probabilistic
Risk Assessment Peer Review Process Guideline''). Further, this guide
will be modified to address PRA standards on fire, external events, and
low power and shutdown modes, as they become available. The NRC has
also developed a draft supporting Standard Review Plan, SRP 19.1, to
provide guidance to the staff on how to determine if a PRA providing
results being used in a decision is technically acceptable.
In a letter dated April 24, 2000, NEI requested that the NRC staff
review the suitability of the peer review process described in NEI 00-
02 to address PRA quality issues for this application. NRC issued a
request for additional information on September 19, 2000, to which NEI
responded by letter dated January 18, 2001. By letter dated April 2,
2002 (ADAMS accession number ML020930632), the NRC staff sent to NEI,
draft staff review guidance that was developed as a result of its
review of NEI 00-02, for intended use for Sec. 50.69 applications.
The draft staff review guidance is for a focused review of the
plant-specific PRA based on a review of NEI 00-02 and NEI 00-04. To
reach the conclusion that the PRA results support the proposed
categorization, the review guidance is structured to lead the staff
reviewer to look for evidence that the impact of a given peer review
issue on PRA results has been adequately addressed in the peer review
report and, when necessary, has been identified for consideration by
the IDP, or to request further information from the licensee.
VII. Criminal Penalties
For the purposes of Section 223 of the Atomic Energy Act, as
amended, the Commission is issuing a rule to add Sec. 50.69 under one
or more of Sections 161b, 161i, or 161o of the AEA. Willful violations
of the rule are subject to criminal enforcement. Criminal penalties, as
they apply to regulations in Part 50, are discussed in Sec. 50.111.
VIII. Compatibility of Agreement State Regulations
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement States Programs,'' approved by the Commission on June 20,
1997, and published in the Federal Register (62 FR 46517, September 3,
1997), this rule is classified as compatibility ``NRC.'' Compatibility
is not required for Category ``NRC'' regulations. The NRC program
elements in this category are those that relate directly to areas of
regulation reserved to the NRC by the AEA or the provisions of Title 10
of the Code of Federal Regulations and, although an Agreement State may
not adopt program elements reserved to NRC, it may wish to inform its
licensees of certain requirements via a mechanism that is consistent
with the particular State's administrative procedure laws, but does not
confer regulatory authority on the State.
IX. Availability of Documents
The NRC is making the documents identified below available to
interested persons through one or more of the following methods as
indicated.
Public Document Room (PDR). The NRC Public Document Room is located
at 11555 Rockville Pike, Rockville, Maryland.
Rulemaking Web site (Web). The NRC's interactive rulemaking Web
site is located at http://ruleforum.llnl.gov. These documents may be
viewed and downloaded electronically via this Web site.
NRC's Public Electronic Reading Room (PERR). The NRC's public
electronic reading room is located at http://www.nrc.gov/reading-rm.html.
Note: Public access to documents, including access via ADAMS and
the PDR, has been temporarily suspended so that security reviews of
publicly available documents may be performed and potentially
sensitive information removed. However, access to the documents
identified in this rule continues to be available through the
rulemaking web site at http://ruleforum.llnl.gov, which was not
affected by the ADAMS shutdown. Please check with the listed NRC
contact concerning any issues related to document availability.
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Document PDR Web PERR
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Response to Public Comments............... X X ML042990011
Environmental Assessment.................. X X ML041040236
Regulatory Analysis....................... X X ML041000474
Industry Implementation Guidance.......... X X ML041120253
Regulatory Guide.......................... X X ML041340087
Final Rule SRM............................ X X ML042810516
SRM on PRA Quality........................ X X ML033520457
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X. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995, Pub.
L. 104-113, requires that Federal agencies use technical standards that
are developed or adopted by voluntary consensus standards bodies unless
using such a standard is inconsistent with applicable law or is
otherwise impractical. In this rule, the NRC is using the following
Government-unique standard (RG 1.201, June 2004). The Commission notes
the development of voluntary consensus standards on PRAs, such as an
ASME Standard on Probabilistic Risk Assessment for Nuclear Power Plant
Applications. RG 1.201 and RG 1.200 (PRA Technical Adequacy) discuss
how this standard could be used for the purpose of the internal events,
full-power PRA.
In addition, the Commission acknowledges development of risk-
informed code cases by the ASME on categorization of certain
components, particularly with respect to pressure boundary
considerations. RG 1.201 explicitly notes these code cases and that
they could be proposed by a licensee or applicant as part of the means
for satisfying the rule requirements. The government standards allow
use of these voluntary consensus standards, but do not require
[[Page 68046]]
their use. The Commission does not believe that these other standards
are sufficient to provide the overall construct for the alternative
approach to categorization and treatment of SSCs that is the goal of
this rulemaking. For example, the current standards do not address all
types of components that might be categorized, nor do standards
currently exist for addressing the PRA requirements for all initiating
events and modes of operation. Additionally, there are no voluntary
consensus standards that can address other parts of the approach laid
out such as determining the basis for the evaluations to show an
acceptably small increase in risk. The NRC is not aware of any
voluntary consensus standard that could be used instead of the
Government-unique standards.
XI. Finding of No Significant Environmental Impact: Environmental
Assessment: Availability
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
Subpart A of 10 CFR part 51, that this rule is not a major Federal
action significantly affecting the quality of the human environment
and, therefore, an environmental impact statement is not required. As
set forth in the final environmental assessment, this action will not
have a significant environmental impact principally because it is
structured to maintain the design basis functional requirements for the
SSCs in the facility, because the rule contains feedback and process
adjustment requirements to maintain the validity of the categorization
process over time, and because the standards and requirements
applicable to radiological releases and effluents are not affected by
this rulemaking.
The NRC requested public comments on any aspect of the
environmental assessment. No public comments were received. The NRC
requested the views of the States on the environmental assessment for
this rule. No State comments were received. Availability of the final
environmental assessment is provided in Section IX.
XII. Paperwork Reduction Act Statement
This rule contains information collection requirements that are
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). These requirements were approved by the Office of Management and
Budget, approval number 3150-0011.
The burden to the public for these information collections is
estimated to average 1,032 hours per response, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
information collection. Send comments on any aspect of these
information collections, including suggestions for reducing the burden,
to the Records and FOIA/Privacy Services Branch (T-5 F52), U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, or by Internet
electronic mail to [email protected]; and to the Desk Officer, John
A. Asalone, Office of Information and Regulatory Affairs, NEOB-10202,
(3150-0011), Office of Management and Budget, Washington, DC 20503. You
may also e-mail comments to [email protected] or comment by
telephone at (202) 395-4650.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
XIII. Regulatory Analysis
The Commission has prepared a regulatory analysis on this
regulation. The analysis examines the costs and benefits of the
alternatives considered by the Commission. Availability of the
regulatory analysis is provided in Section IX.
XIV. Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act (5 U.S.C.
605(b)), the Commission certifies that this rule does not have a
significant economic impact on a substantial number of small entities.
This rule affects only the licensing and operation of nuclear power
plants. The companies that own these plants do not fall within the
scope of the definition of ``small entities'' set forth in the
Regulatory Flexibility Act or the size standards established by the NRC
(10 CFR 2.810).
XV. Backfit Analysis
The NRC has determined that the Backfit Rule does not apply to this
rule; therefore, a backfit analysis is not required for this rule. As a
voluntary alternative to existing requirements, the final rule does not
impose different or new requirements on 10 CFR part 50 licensees or
applicants and thus does not constitute a backfit pursuant to Sec.
50.109.
XVI. Small Business Regulatory Enforcement Fairness Act
In accordance with the Small Business Regulatory Enforcement
Fairness Act of 1996, the NRC has determined that this action is not a
major rule and has verified this determination with the Office of
Information and Regulatory Affairs of OMB.
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
0
For the reasons set out in the preamble and under the authority of the
Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of
1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting the
following amendments to 10 CFR part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
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1. The authority citation for Part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat.1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); See
1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
(42 U.S.C. 5841). Sections 50.10 also issued under secs. 101, 185,
68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd),
and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42
U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,
50.55a, and Appendix Q also issued under sec. 102, Pub. L. 91-190,
83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued
under Sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58,
50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42
U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939
(42 U.S.C. 2152). Sections 50.80, 50.81 also issued under sec. 184,
68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued
under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
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2. In Sec. 50.8 paragraph (b) is revised to read as follows:
Sec. 50.8 Information collection requirements: OMB approval.
* * * * *
[[Page 68047]]
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 50.30, 50.33, 50.33a, 50.34, 50.34a,
50.35, 50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54,
50.55, 50.55a, 50.59, 50.60, 50.61, 50.62, 50.63, 50.64, 50.65, 50.66,
50.68, 50.69, 50.70, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90,
50.91, 50.120, and appendices A, B, E, G, H, I, J, K, M, N,O, Q, R, and
S to this part.
* * * * *
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3. A new Sec. 50.69 is added under center heading ``Issuance,
Limitations, and Conditions of Licenses and Construction Permits'' to
read as follows:
Sec. 50.69 Risk-informed categorization and treatment of structures,
systems and components for nuclear power reactors.
(a) Definitions.
Risk-Informed Safety Class (RISC)-1 structures, systems, and
components (SSCs) means safety-related SSCs that perform safety
significant functions.
Risk-Informed Safety Class (RISC)-2 structures, systems and
components (SSCs) means nonsafety-related SSCs that perform safety
significant functions.
Risk-Informed Safety Class (RISC)-3 structures, systems and
components (SSCs) means safety-related SSCs that perform low safety
significant functions.
Risk-Informed Safety Class (RISC)-4 structures, systems and
components (SSCs) means nonsafety-related SSCs that perform low safety
significant functions.
Safety significant function means a function whose degradation or
loss could result in a significant adverse effect on defense-in-depth,
safety margin, or risk.
(b) Applicability and scope of risk-informed treatment of SSCs and
submittal/approval process. (1) A holder of a license to operate a
light water reactor (LWR) nuclear power plant under this part; a holder
of a renewed LWR license under part 54 of this chapter; an applicant
for a construction permit or operating license under this part; or an
applicant for a design approval, a combined license, or manufacturing
license under part 52 of this chapter; may voluntarily comply with the
requirements in this section as an alternative to compliance with the
following requirements for RISC-3 and RISC-4 SSCs:
(i) 10 CFR part 21.
(ii) The portion of 10 CFR 50.46a(b) that imposes requirements to
conform to Appendix B to 10 CFR part 50.
(iii) 10 CFR 50.49.
(iv) 10 CFR 50.55(e).
(v) The inservice testing requirements in 10 CFR 50.55a(f); the
inservice inspection, and repair and replacement (with the exception of
fracture toughness), requirements for ASME Class 2 and Class 3 SSCs in
10 CFR 50.55a(g); and the electrical component quality and
qualification requirements in Section 4.3 and 4.4 of IEEE 279, and
Sections 5.3 and 5.4 of IEEE 603-1991, as incorporated by reference in
10 CFR 50.55a(h).
(vi) 10 CFR 50.65, except for paragraph (a)(4).
(vii) 10 CFR 50.72.
(viii) 10 CFR 50.73.
(ix) Appendix B to 10 CFR part 50.
(x) The Type B and Type C leakage testing requirements in both
Options A and B of Appendix J to 10 CFR part 50, for penetrations and
valves meeting the following criteria:
(A) Containment penetrations that are either 1-inch nominal size or
less, or continuously pressurized.
(B) Containment isolation valves that meet one or more of the
following criteria:
(1) The valve is required to be open under accident conditions to
prevent or mitigate core damage events;
(2) The valve is normally closed and in a physically closed, water-
filled system;
(3) The valve is in a physically closed system whose piping
pressure rating exceeds the containment design pressure rating and is
not connected to the reactor coolant pressure boundary; or
(4) The valve is 1-inch nominal size or less.
(xi) Appendix A to part 100, Sections VI(a)(1) and VI(a)(2), to the
extent that these regulations require qualification testing and
specific engineering methods to demonstrate that SSCs are designed to
withstand the Safe Shutdown Earthquake and Operating Basis Earthquake.
(2) A licensee voluntarily choosing to implement this section shall
submit an application for license amendment under Sec. 50.90 that
contains the following information:
(i) A description of the process for categorization of RISC-1,
RISC-2, RISC-3 and RISC-4 SSCs.
(ii) A description of the measures taken to assure that the quality
and level of detail of the systematic processes that evaluate the plant
for internal and external events during normal operation, low power,
and shutdown (including the plant-specific probabilistic risk
assessment (PRA), margins-type approaches, or other systematic
evaluation techniques used to evaluate severe accident vulnerabilities)
are adequate for the categorization of SSCs.
(iii) Results of the PRA review process conducted to meet Sec.
50.69(c)(1)(i).
(iv) A description of, and basis for acceptability of, the
evaluations to be conducted to satisfy Sec. 50.69(c)(1)(iv). The
evaluations must include the effects of common cause interaction
susceptibility, and the potential impacts from known degradation
mechanisms for both active and passive functions, and address
internally and externally initiated events and plant operating modes
(e.g., full power and shutdown conditions).
(3) The Commission will approve a licensee's implementation of this
section if it determines that the process for categorization of RISC-1,
RISC-2, RISC-3, and RISC-4 SSCs satisfies the requirements of Sec.
50.69(c) by issuing a license amendment approving the licensee's use of
this section.
(4) An applicant choosing to implement this section shall include
the information in Sec. 50.69(b)(2) as part of application. The
Commission will approve an applicant's implementation of this section
if it determines that the process for categorization of RISC-1, RISC-2,
RISC-3, and RISC-4 SSCs satisfies the requirements of Sec. 50.69(c).
(c) SSC Categorization Process. (1) SSCs must be categorized as
RISC-1, RISC-2, RISC-3, or RISC-4 SSCs using a categorization process
that determines if an SSC performs one or more safety significant
functions and identifies those functions. The process must:
(i) Consider results and insights from the plant-specific PRA. This
PRA must at a minimum model severe accident scenarios resulting from
internal initiating events occurring at full power operation. The PRA
must be of sufficient quality and level of detail to support the
categorization process, and must be subjected to a peer review process
assessed against a standard or set of acceptance criteria that is
endorsed by the NRC.
(ii) Determine SSC functional importance using an integrated,
systematic process for addressing initiating events (internal and
external), SSCs, and plant operating modes, including those not modeled
in the plant-specific PRA. The functions to be identified and
considered include design bases functions and functions credited for
mitigation and prevention of severe accidents. All aspects of the
integrated, systematic process used to characterize SSC importance must
reasonably reflect the current plant configuration and operating
practices, and applicable plant and industry operational experience.
(iii) Maintain defense-in-depth.
(iv) Include evaluations that provide reasonable confidence that
for SSCs categorized as RISC-3, sufficient safety
[[Page 68048]]
margins are maintained and that any potential increases in core damage
frequency (CDF) and large early release frequency (LERF) resulting from
changes in treatment permitted by implementation of Sec. Sec.
50.69(b)(1) and (d)(2) are small.
(v) Be performed for entire systems and structures, not for
selected components within a system or structure.
(2) The SSCs must be categorized by an Integrated Decision-Making
Panel (IDP) staffed with expert, plant-knowledgeable members whose
expertise includes, at a minimum, PRA, safety analysis, plant
operation, design engineering, and system engineering.
(d) Alternative treatment requirements.--(1) RISC-1 and RISC 2
SSCs. The licensee or applicant shall ensure that RISC-1 and RISC-2
SSCs perform their functions consistent with the categorization process
assumptions by evaluating treatment being applied to these SSCs to
ensure that it supports the key assumptions in the categorization
process that relate to their assumed performance.
(2) RISC-3 SSCs. The licensee or applicant shall ensure, with
reasonable confidence, that RISC-3 SSCs remain capable of performing
their safety-related functions under design basis conditions, including
seismic conditions and environmental conditions and effects throughout
their service life. The treatment of RISC-3 SSCs must be consistent
with the categorization process. Inspection and testing, and corrective
action shall be provided for RISC-3 SSCs.
(i) Inspection and testing. Periodic inspection and testing
activities must be conducted to determine that RISC-3 SSCs will remain
capable of performing their safety-related functions under design basis
conditions; and
(ii) Corrective action. Conditions that would prevent a RISC-3 SSC
from performing its safety-related functions under design basis
conditions must be corrected in a timely manner. For significant
conditions adverse to quality, measures must be taken to provide
reasonable confidence that the cause of the condition is determined and
corrective action taken to preclude repetition.
(e) Feedback and process adjustment.--(1) RISC-1, RISC-2, RISC-3
and RISC-4 SSCs. The licensee shall review changes to the plant,
operational practices, applicable plant and industry operational
experience, and, as appropriate, update the PRA and SSC categorization
and treatment processes. The licensee shall perform this review in a
timely manner but no longer than once every two refueling outages.
(2) RISC-1 and RISC-2 SSCs. The licensee shall monitor the
performance of RISC-1 and RISC-2 SSCs. The licensee shall make
adjustments as necessary to either the categorization or treatment
processes so that the categorization process and results are maintained
valid.
(3) RISC-3 SSCs. The licensee shall consider data collected in
Sec. 50.69(d)(2)(i) for RISC-3 SSCs to determine if there are any
adverse changes in performance such that the SSC unreliability values
approach or exceed the values used in the evaluations conducted to
satisfy Sec. 50.69(c)(1)(iv). The licensee shall make adjustments as
necessary to the categorization or treatment processes so that the
categorization process and results are maintained valid.
(f) Program documentation, change control and records. (1) The
licensee or applicant shall document the basis for its categorization
of any SSC under paragraph (c) of this section before removing any
requirements under Sec. 50.69(b)(1) for those SSCs.
(2) Following implementation of this section, licensees and
applicants shall update their final safety analysis report (FSAR) to
reflect which systems have been categorized, in accordance with Sec.
50.71(e).
(3) When a licensee first implements this section for a SSC,
changes to the FSAR for the implementation of the changes in accordance
with Sec. 50.69(d) need not include a supporting Sec. 50.59
evaluation of the changes directly related to implementation.
Thereafter, changes to the programs and procedures for implementation
of Sec. 50.69(d), as described in the FSAR, may be made if the
requirements of this section and Sec. 50.59 continue to be met.
(4) When a licensee first implements this section for a SSC,
changes to the quality assurance plan for the implementation of the
changes in accordance with Sec. 50.69(d) need not include a supporting
Sec. 50.54(a) review of the changes directly related to
implementation. Thereafter, changes to the programs and procedures for
implementation of Sec. 50.69(d), as described in the quality assurance
plan may be made if the requirements of this section and Sec. 50.54(a)
continue to be met.
(g) Reporting. The licensee shall submit a licensee event report
under Sec. 50.73(b) for any event or condition that prevented, or
would have prevented, a RISC-1 or RISC-2 SSC from performing a safety
significant function.
Dated in Rockville, Maryland this 15th day of November 2004.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 04-25665 Filed 11-19-04; 8:45 am]
BILLING CODE 7590-01-P