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[Federal Register: November 22, 2004 (Volume 69, Number 224)]
[Rules and Regulations]
[Page 68007-68048]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr22no04-19]
[[Page 68007]]
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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
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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://frwebgate.access.gpo.gov/cgi-bin/leaving.cgi?from=leavingFR.html&log=linklog&to=http://ruleforum.llnl.gov. For information about
the interactive rulemaking Web site contact Ms. Carol Gallagher, (301)
415-5905 (e-mail: CAG@nrc.gov).
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: tar@nrc.gov.
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,
[[Page 68009]]
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://frwebgate.access.gpo.gov/cgi-bin/leaving.cgi?from=leavingFR.html&log=linklog&to=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 a | 
