[Federal Register Volume 68, Number 95 (Friday, May 16, 2003)]
[Proposed Rules]
[Pages 26511-26551]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-11696]


 ========================================================================
 Proposed Rules
                                                 Federal Register
 ________________________________________________________________________
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 This section of the FEDERAL REGISTER contains notices to the public of 
 the proposed issuance of rules and regulations. The purpose of these 
 notices is to give interested persons an opportunity to participate in 
 the rule making prior to the adoption of the final rules.
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  Federal Register / Vol. 68, No. 95 / Friday, May 16, 2003 / Proposed 
Rules  

[[Page 26511]]



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: Proposed rule.

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

SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend 
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 proposed amendment 
would revise 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 proposed amendment would permit 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 
proposing a draft regulatory guide to implement the rule.

DATES: Submit comments by July 30, 2003. Comments received after this 
date will be considered if it is practical to do so, but the Commission 
is able to ensure consideration only for comments received on or before 
this date.

ADDRESSES: You may submit comments by any one of the following methods. 
Please include the following number (RIN 3150-AG42) in the subject line 
of your comments. Comments on rulemakings submitted in writing or in 
electronic form will be made available to the public in their entirety 
on the NRC rulemaking web site. Personal information will not be 
removed from your comments.
    Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
    E-mail comments to: [email protected]. If you do not receive a reply e-
mail confirming that we have received your comments, contact us 
directly at (301) 415-1966. You may also submit comments via the NRC's 
rulemaking web site at http://ruleforum.llnl.gov. Address questions 
about our rulemaking website to Carol Gallagher (301) 415-5905; email 
[email protected].
    Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 
20852, between 7:30 am and 4:15 pm Federal workdays. (Telephone (301) 
415-1966).
    Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 
(301) 415-1101.
    Publicly available documents related to this rulemaking may be 
examined and copied for a fee at the NRC's Public Document Room (PDR), 
Public File Area O1 F21, One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland. Selected documents, including comments, can be 
viewed and downloaded electronically via the NRC rulemaking web site at 
http://ruleforum.llnl.gov.
    Publicly available documents created or received at the NRC after 
November 1, 1999, are available electronically at the NRC's Electronic 
Reading Room at http://www.nrc.gov/NRC/ADAMS/index.html. From this 
site, the public can gain entry into the NRC's Agencywide Document 
Access and Management System (ADAMS), which provides text and image 
files of NRC's public documents. If you do not have access to ADAMS or 
if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [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. Rule Initiation
III. Proposed Regulations
IV. Public Input to the Proposed Rule
V. Section by Section Analysis
VI. Other Topics for Public Comment
VII. Guidance
VIII. Criminal Penalties
IX. Compatibility of Agreement State Regulations
X. Availability of Documents
XI. Plain Language
XII. Voluntary Consensus Standards
XIII. Finding of No Significant Environmental Impact
XIV. Paperwork Reduction Act Statement
XV. Regulatory Analysis
XVI. Regulatory Flexibility Act Certification
XVII. Backfit Analysis

I. 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, 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 were defined as ``safety-related,'' and these SSCs were the 
subject of many regulatory requirements designed to ensure that they 
were of high quality, high reliability, and had capability to perform 
during postulated design basis

[[Page 26512]]

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 of Appendix A to 10 CFR part 50) is not 
explicitly defined.
    These prescriptive requirements as to how licensees were 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 structures, systems, or components 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, 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 nonprobabilistic 
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 makes regulatory 
sense. 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 (PRA) 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 staff developed guidance 
on the use of risk information for reactor license amendments and 
issued Regulatory Guide (RG) 1.174. 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.
    Regulatory Guide 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

[[Page 26513]]

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.

II. 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 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, or seismic continue to impose 
substantial burdens. As a result, the replacement of such a low safety 
significant component needs 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 go 
beyond the staff's 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 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 discussed in Section IV.4).
    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 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 designed to maintain consistency between actual 
performance and the performance considered in the assessment process 
that determines their significance. As a result, both the NRC staff and 
industry should be able to better focus their resources on regulatory 
issues of greater safety significance.
    The Commission directed the 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, referred to as Sec.  50.69, to 
contain these alternative requirements.
    The Commission received more than 200 comments in response to the 
ANPR. The staff sent the Commission SECY-00-194 ``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 below.
    The concept developed for this proposed rule, discussed at length 
in the ANPR, was to apply 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 proposed 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 Commission finds the risk-informed

[[Page 26514]]

approach outlined in RG 1.174 is appropriate for use in this 
rulemaking.
    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, this approach should 
allow an acceptable, though reduced, level of assurance that these SSCs 
will satisfy functional requirements.

III. Proposed Regulations

    The Commission is proposing to establish Sec.  50.69 as an 
alternative set of requirements whereby a licensee may undertake 
categorization of its SSCs using risk insights and adjust treatment 
requirements based upon their resulting significance. Under this 
approach, a licensee would be allowed to reduce special treatment 
requirements for SSCs that are determined to be of low safety 
significance and would enhance requirements for treatment of other SSCs 
that are found to be safety significant. The proposed requirements 
would establish a process by which a licensee would categorize SSCs 
using a risk-informed process, adjust treatment requirements consistent 
with the relative significance of the SSC, and manage the process over 
the lifetime of the plant. To implement these requirements, a risk-
informed categorization process would be employed to determine the 
safety significance of SSCs and place the SSCs into one of four risk-
informed safety class (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 would be performed by an 
integrated decisionmaking process which uses both risk insights and 
traditional engineering insights. The safety functions would 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 would be 
conducted to make adjustments to the categorization and treatment 
processes as needed so that SSCs continue to meet applicable 
requirements. The proposed 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 would define 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 the proposed 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 either 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 are nonsafety-
related, and the risk-informed categorization determines them 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 contributors to 
plant safety. These SSCs are termed RISC-3 SSCs. 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.
    Section 50.69 would define 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 proposed rule would impose greater treatment 
requirements on SSCs that perform safety-significant functions (RISC-1 
and RISC-2 SSCs) to ensure that defense-in-depth and safety margins are 
maintained. The proposed rule would also require that the change in 
risk associated with implementation of proposed Sec.  50.69 be small.

III.2.0 Methodology for Categorization

    The cornerstone of proposed 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 proposed 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 that has the expertise and capabilities for 
making a sound decision regarding the SSC's categorization, and that 
information is considered in a manner that ensures the Commission's 
criteria for risk-informed applications are satisfied (i.e., that 
defense-in-depth is maintained, safety margins are maintained, any risk 
change is small, and a monitoring and performance assessment strategy 
is used). This process enables SSCs to be placed in the correct RISC 
category such that the appropriate treatment requirements will be 
applied commensurate with their safety significance. A safety-
significant SSC is an SSC that performs a safety-significant function. 
The proposed rule would require that SSC safety significance be 
determined using quantitative information from an up-to-date PRA 
reasonably representing the current plant configuration, which as a 
minimum covers internal events at full power, and other available risk 
analyses and traditional engineering information to supplement the 
quantitative PRA results.

[[Page 26515]]

[GRAPHIC][TIFF OMITTED]TP16MY03.031

    Section 50.69 would contain requirements to ensure that the PRA is 
adequate for this application. The proposed rule would require that as 
part of the categorization process defense-in-depth is considered, and 
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 the defense-in-depth philosophy (in terms of both large early and 
large late releases). As part of meeting the defense-in-depth 
principle, 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. Thus, the rule contains requirements for 
the IDP to consider defense-in-depth as part of the categorization 
process.
    The risk insights and other traditional information are required to 
be evaluated by an Integrated Decision-Making Panel (IDP) 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, it is important that the membership 
include a variety of expertise about the plant, how it is operated, and 
the safety analyses (both deterministic and probabilistic), so that all 
pertinent information is considered. Hence the available deterministic 
and probabilistic information pertaining to SSC safety significance is 
considered in the decision process. The information considered must 
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 which was either 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 a period of 
time, and the documentation

[[Page 26516]]

gives both the licensee and the NRC the ability to understand the basis 
for the categorization decision, should questions arise at a later 
date.
    The proposed rule would contain general requirements for 
consideration of SSCs, modes of operation or 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. As noted in the ANPR, the Commission could include more 
requirements in the rule itself, rather than only being in the 
guidance. Public comment is requested on the merits of placing the 
additional detail shown in the guidance and discussed in Section V.4 of 
the Statement of Considerations (SOC) in the rule.
    Implementation of the categorization process relies heavily on the 
skills, knowledge, and experience of the people that implement the 
process, in particular on the qualifications of IDP members. Therefore, 
the Commission concludes that requirements 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 in 
making SSC classifications.
    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 completeness and technical adequacy for determining 
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 assurance that 
for IDP decisions that utilize PRA information that the results of the 
categorization process provide a valid representation of the risk 
importance of SSCs.
    Before implementation of Sec.  50.69, the NRC will approve the 
categorization process, through a license amendment, because of the 
importance of the PRA and categorization process to successful 
implementation of the proposed rule. This review will determine whether 
the licensee's application satisfies the Sec.  50.69 requirements, 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 staff review 
of the PRA is necessary because there are key assumptions and modeling 
parameters that can have a significant enough impact on the results 
such 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)(iv) would require that a licensee or applicant 
provide reasonable confidence that for SSCs categorized as RISC-3, 
sufficient safety margins are maintained and that any potential changes 
in core damage frequency (CDF) and large early release frequency (LERF) 
resulting from the implementation of Sec.  50.69 are small. That is, 
plants with total baseline CDF of 10-4 per year or less 
would be permitted CDF increases of up to 10-5 per year, and 
plants with total baseline CDF greater than 10-4 per year 
would be permitted CDF increases of up to 10-6 per year. 
Plants with total baseline LERFs of 10-5 per year or less 
would be permitted LERF increases of up to 10-6 per year, 
and plants with total baseline LERFs greater than 10-5 per 
year would be permitted LERF increases of up to 10-7 per 
year. However, if there is an indication that the baseline CDF or LERF 
may be considerably higher than these values, the focus of the licensee 
should be on finding ways to reduce risk and the licensee may be 
required to present arguments as to why steps should not be taken to 
reduce risk in order to consider the reduction in special treatment 
requirements. 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. Thus, the allowable 
potential increase in risk must be determined in a cumulative way for 
all the SSCs being recategorized.
    Section 50.69 contains requirements for maintaining the design 
basis of the facility. These requirements, considered in conjunction 
with the requirements to maintain the potential change in risk as small 
(as discussed above), ensure that safety margins are maintained. The 
performance of candidate RISC-3 SSCs should not be significantly 
degraded by the removal of special treatment. This is because the 
licensee is required to implement processes that provide reasonable 
confidence that SSCs remain functional, that is, remain capable of 
performing their function with a reliability that is not significantly 
degraded to such an extent that there will be a significant number of 
failures that can lead to unacceptable increases in CDF or LERF.
    The proposed rule would require applicants and licensees to perform 
evaluations to assess the potential impact on risk from changes to 
treatment. For SSCs modeled in the PRA, this would likely be 
accomplished by sensitivity studies to assess the impact of changes in 
SSC failure probabilities or reliabilities that might occur due to the 
revised treatment. For example, a licensee would be expected to 
increase the failure rates of RISC-3 SSCs by appropriate factors to 
understand the potential effect of applying reduced treatment to these 
SSCs (e.g., reduced maintenance, testing, inspection, and quality 
assurance). For other SSCs, other types evaluations would be used to 
provide the basis for concluding that the potential increase in risk 
would be small. A licensee will need to submit its basis to support 
that the evaluations are bounding estimates of the potential change in 
risk and that programs already in existence or implemented for proposed 
Sec.  50.69 can provide sufficient information that any potential risk 
change remains small over the lifetime of the plant. A licensee is 
required to consider potential effects of common-cause interaction 
susceptibility and potential impacts from known degradation mechanisms. 
To meet this requirement, a licensee would need to: (a) Maintain an 
understanding of common-cause effects and degradation mechanisms and 
their potential impact on RISC-3 SSCs; (b) maintain an understanding of 
the programmatic activities that provide defenses against common cause 
failures (CCFs) and failures resulting from degradation; and (c) factor 
this knowledge into the treatment applied to the RISC-3 SSCs.
    The proposed rule focuses on common-cause effects because 
significant increases in common-cause failures could invalidate the 
evaluations, such as sensitivity studies, performed to show a small 
change in risk due to implementation of Sec.  50.69. With respect to 
known degradation mechanisms, this is an acknowledgment that certain 
treatment requirements have evolved over time to deal with such 
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 
would need to provide a basis to conclude that the small increase in 
risk requirement would still be met in light of potential changes in 
treatment. All of these requirements are included in Sec.  50.69 so 
that a licensee has a basis for

[[Page 26517]]

concluding that the evaluations performed to show a small change in 
risk remain valid.
    In addition, the rule would require that implementation be done 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

    Treatment requirements are applied to SSCs commensurate with SSC 
safety significance and as a function of the RISC category into which 
the SSCs are categorized.
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 would maintain 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. 
Additional requirements are being added to these SSCs to ensure that 
their performance remains consistent with the assumed performance in 
the categorization process (including the PRA) for beyond design basis 
conditions. For example, in developing the PRA model, a licensee will 
make assumptions regarding 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 assumed to be performed may exceed the design-basis 
conditions for the applicable SSCs. In the proposed rule, a licensee 
would be required to ensure that the treatment applied to RISC-1 and 
RISC-2 SSCs is 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 categorization assumptions to reflect actual treatment practices. 
In addition, requirements exist for monitoring and adjustment of 
treatment processes (or categorization decisions) as needed based upon 
performance.
III.3.2 RISC-3 Treatment
    For RISC-3 SSCs, Sec.  50.69 would impose requirements which are 
intended to maintain their 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 it is important to maintain their design 
basis functional capability. Maintenance of RISC-3 design basis 
functionality is important to ensuring that defense-in-depth and safety 
margins are maintained. As a result, Sec.  50.69(d)(2) would require 
licensees or applicants to have processes in place that provide 
reasonable confidence in the capability of RISC-3 SSCs to perform their 
safety-related functions under design basis conditions throughout the 
service life. The proposed rule contains high-level requirements for 
the treatment of RISC-3 SSCs with respect to design control; 
procurement; maintenance, inspection, test, and surveillance; and 
corrective action. 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 proposed rule. For example, the 
alternative treatment requirements for RISC-3 SSCs in proposed 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 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.
    The Commission is proposing to specify four processes that must be 
controlled and accomplished for RISC-3 SSCs: Design Control; 
Procurement; Maintenance, Inspection, Testing, and Surveillance; and 
Corrective Action. The high level RISC-3 requirements are structured to 
address the various key elements of SSC functionality by focusing in 
these areas. When SSCs are replaced, RISC-3 SSCs must remain capable of 
performing design basis functions. Hence, the high level requirements 
focus on maintaining this capability through design control and 
procurement requirements. During the operating life of a RISC-3 SSC, a 
sufficient level of confidence is necessary that the SSC continues to 
be able to perform its design basis function; hence, the inclusion of 
high level requirements for maintenance, inspection, test, and 
surveillance. Finally, when data is collected, it must be fed back into 
the categorization and treatment processes, and when important 
deficiencies are found, they must be corrected; hence, requirements are 
also provided in these areas.
    In devising these requirements, the Commission has focused upon 
those critical aspects of the various processes that must exist to 
provide assurance of performance. Thus, in the design area, for 
instance, the design conditions under which equipment is expected to 
perform, such as environmental conditions or seismic conditions, are 
still to be met. As another example, in the procurement area, procured 
items are to satisfy their design requirements. These steps provide the 
basis for concluding that a newly designed and procured replacement 
item will be capable of meeting its design requirements, even though 
the special treatment requirements that previously existed are no 
longer being required.
    In implementing the processes required by the proposed rule, 
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. These 
requirements are necessary because they require the licensee to obtain 
the data necessary to continue to conclude that RISC-3 SSCs remain 
capable of performing design basis functions, and to enable the 
licensee to take actions to restore equipment performance consistent 
with corrective action requirements included in the proposed rule.
    Effective implementation of the treatment requirements provides 
reasonable confidence in the capability of RISC-3 SSCs to perform their 
safety 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 
their low safety significance.
    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. Section 50.69(d)(2)(i), which focuses 
upon design control, is intended to draw a distinction between 
treatment (managed through Sec.  50.69) and design changes (managed 
through other processes such as Sec.  50.59). As

[[Page 26518]]

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 of SSCs.
III.3.3 RISC-4 Treatment
    Section Sec.  50.69 would not impose treatment requirements on 
RISC-4 SSCs. Instead RISC-4 SSCs are simply removed from the scope of 
any applicable special treatment requirements. This is justified in 
view of their low significance considering both safety-related and risk 
information. 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

    RISC-3 and RISC-4 SSCs, through the application of Sec.  50.69, are 
removed from the scope of specific special treatment requirements 
listed in proposed Sec.  50.69. These requirements were initially 
identified in the ANPR based upon a set of criteria as to whether the 
regulation imposed requirements relating to quality assurance, 
qualification, documentation, testing, etc., that were intended to add 
assurance to performance of SSCs.
    The special treatment requirements were originally imposed to 
provide a very 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, 
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. For those SSCs that this new categorization 
identifies as most safety-significant (RISC-1 and RISC-2), the existing 
special treatment requirements are being maintained because the 
Commission still desires a high level of assurance. However, the 
Commission concluded that for the less significant SSCs, it was no 
longer necessary to have the same high level of assurance that they 
would perform as specified. This is because some increased likelihood 
of failure can be tolerated without significantly impacting safety. 
Thus, the Commission decided to remove the RISC-3 and RISC-4 SSCs from 
those detailed, specific requirements that provided the very 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 its treatment processes to achieve performance 
objectives. The more general requirements that the Commission is 
specifying for the RISC-3 SSCs include steps to procure SSCs suitable 
for the conditions under which they are to perform, to conduct 
performance and/or condition monitoring and to take corrective action, 
as a means of maintaining functionality. As discussed elsewhere in this 
notice, the Commission concludes that the requirements in Sec.  50.69 
maintain adequate protection of public health and safety. 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 totally removed from the scope of specific special 
treatment requirements while in other cases the Commission concluded 
that only partial removal was appropriate. The reduced assurance for 
the RISC-3 SSC would be provided by the alternative requirements being 
added by this proposed rule. Finally, there was a set of requirements 
initially identified as special treatment for which the Commission is 
not proposing to remove RISC-3 and RISC-4 SSCs from their scopes. These 
requirements are discussed at the end of this section (III.4.9).
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 non-compliance 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 
promulgate regulations defining these terms.
    The Commission's regulations implementing Section 206 are set forth 
in 10 CFR Part 21 and Sec.  50.55(e) for license holders and 
construction permit holders, respectively. The definitions of ``basic 
component,'' ``defect,'' and ``substantial safety hazard'' in Part 21 
were established by the Commission based upon the premise that the 
deterministic regulatory paradigm embedded in the Commission's 
regulations would 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 very 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 which 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.


[[Page 26519]]


    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 defect within 
five (5) working days of completion of evaluations for determining 
whether noncompliance or defect 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 would substitute 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 proposed 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; and (2) the appropriate reporting 
obligations of licensees and vendors under proposed 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 
proposed 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 which are 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, and 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. 
Inasmuch as no regulatory purpose would be served by reporting for 
RISC-4 SSCs, the Commission proposes that RISC-4 SSCs should not be 
subject to part 21 or Sec.  50.55(e). 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 reasons discussed below, 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 as to whether such a safety hazard requires that 
immediate NRC regulatory action at one or more nuclear power plants be 
taken 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. Such reporting allows a 
licensee using such 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 which 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 such 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: First, that a non-compliance or defect could 
be incorporated into construction where it could never be detected; and 
second, 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 proposed 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 
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 such 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 such that 
immediate licensee and NRC evaluation of the situation and 
implementation of necessary corrective action is necessary to ensure 
adequate 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.
    Those SSCs that are viewed as being of sufficient safety 
significance to require Part 21 reporting are RISC-1 SSCs. 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 
as being safety-significant, as these

[[Page 26520]]

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 maintain integrity of fission product 
barriers, that provide or support the primary success paths for 
shutdown, or that prevent or mitigate accidents that could lead to 
potential offsite exposures). 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 SSC 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 they: (i) Contribute to plant risk by initiating transients 
that could lead to severe accidents (if multiple failures of other 
mitigating SSCs were to occur), or (ii) they can reduce risk by 
providing backup mitigation to RISC-1 SSCs in response to an event. The 
Commission recognizes that, on its face, 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 SSC being 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 for 
consistency between the categorization assumptions and how they are 
treated, performance should only be enhanced. As discussed in Sections 
III.3 and III.5 of this SOC, the Commission is proposing that 
additional regulatory controls be imposed on RISC-2 SSCs to prevent 
their performance from degrading. In addition, the Commission is 
proposing that licensees evaluate treatment being applied for 
consistency with key categorization assumptions, monitor the 
performance of these SSCs, take corrective actions, and report when a 
loss of a safety-significant function occurs. The requirements of the 
maintenance rule (Sec.  50.65 (a)(1) through (a)(3)) also continue to 
apply to these SSCs. 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 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 such RISC-2 SSCs also 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 
WASH-1400 (Reactor Safety Study), 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 that 
would be categorized as RISC-2 SSC would rise to the level of 
significance that their failure or lack of functionality would 
constitute a substantial safety hazard requiring immediate 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 promulgated as cost-beneficial safety 
improvements. The equipment used for station blackout or anticipated 
transients without scram would generally fall within the RISC-2 
category. The Commission believes its conclusion about the relative 
significance of RISC-2 SSC is also supported by plant-specific risk 
studies, such as the IPE and IPEEE \1\, conducted to identify (and 
correct) any plant-specific vulnerabilities to severe accident risk. 
NRC's review of the responses to 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, 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 were found to be cost-
beneficial (and none were considered necessary to provide adequate 
protection of public health and safety).
---------------------------------------------------------------------------

    \1\ 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 (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.
---------------------------------------------------------------------------

    In sum, the Commission believes that in light of risk assessments 
and actions that have already been implemented, there would be no SSCs 
categorized under 50.69 as RISC-2 whose failure would represent a 
significant and substantial safety concern such that immediate 
notification and 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 believes that the multiple simultaneous failures of 
either RISC-2 or RISC-3 components, in the same or in different 
systems, is not a concern such that Part 21 reporting is necessary. 
Even for components of the same type, it is not likely that the 
installed components are identical in terms of their specific 
characteristics or operating and maintenance history such that a defect 
would lead to simultaneous failure of multiple components at the same 
time. For both RISC categories, there are requirements to collect data 
about performance of the SSCs, to review the data to determine if 
adverse performance is occurring and to take

[[Page 26521]]

appropriate action (e.g., correct failures and adjust treatment 
processes). Thus, it would be expected that degradation or problems 
affecting a component type would be detected and dealt with before 
multiple failures becomes likely. For many RISC-2 SSCs, failures tend 
to be self-revealing (as it is initiation of a transient as a result of 
failure of many RISC-2 SSC that makes them significant). For RISC-3 
SSCs, requirements exist for design and procurement for any replacement 
components to meet their design conditions, thus making it unlikely 
that unsuitable components would be installed. Further, for the RISC-3 
SSCs, evaluations will be performed, assuming significantly increased 
failure rates for large number of components occurring simultaneously 
to show that there is no more than a small (potential) change in risk. 
Therefore, the Commission believes appropriate regulatory attention has 
been given to the potential for multiple simultaneous failures.
    The Commission also considered the question as to whether 
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 
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 part. Further, the suppliers of these components 
would then also 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, proposed 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.
    The Commission does not believe that any changes to Part 21 are 
necessary to accomplish the Commission's proposal, and that this 
proposal 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 determine 
that a RISC-2 SSC does not represent a ``substantial safety hazard'' as 
defined in Sec.  21.3 in the context of a risk-informed regulatory 
approach.
    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 safety 
hazard such that immediate licensee and NRC evaluation of the situation 
and implementation of necessary 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 proposes 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 
should exclude RISC-2 SSCs.
    The Commission also proposes 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, at most, only a small change 
in risk, and should not result in a major degradation of essential 
safety-related equipment (see NUREG-0302, Rev. 1).\2\ As discussed 
above, the body of regulatory requirements (the retained requirements 
and the requirements contained in this proposed rule) are sufficient 
such that simultaneous failures in multiple systems (as would be 
necessary to lead to a substantial safety hazard involving RISC-3 SSCs) 
would not occur. Thus, there is little regulatory need for the NRC to 
be informed of instances of noncompliance and defects with RISC-3 SSCs. 
This is consistent with the NRC's current position that a ``substantial 
safety hazard'' involves a major degradation of essential safety-
related equipment (see NUREG-0302). Accordingly, the Commission 
proposes that RISC-3 SSCs should not be subject to reporting 
requirements of part 21 and Sec.  50.55(e).
---------------------------------------------------------------------------

    \2\ 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.
---------------------------------------------------------------------------

    In sum, the Commission proposes that part 21 reporting requirements 
should extend only to SSCs classified as RISC-1 SSCs, since these SSCs 
are those that are important in ensuring public health and safety and 
minimizing risk. RISC-2 SSCs should not be subject to reporting because 
play a lesser role than RISC-1 SSC in protection of public health and 
safety and no regulatory purpose would be served by part 21 reporting 
(as discussed above). 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 the Commission's proposals 
with respect to RISC-2 and RISC-3 SSCs, and that this proposal is 
consistent with the statutory requirements in Section 206 of the ERA. 
As discussed above, 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. As discussed earlier, Sec.  50.69 embodies a risk-
informed regulatory paradigm which 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

[[Page 26522]]

``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 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: (i) Interpreting ``basic component'' to encompass a risk-informed 
view of what SSCs the term encompasses, and (ii) including a second 
definition of ``basic component'' in Sec.  21.3, which would apply only 
to those portions of a plant which 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 ``safety-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, 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-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 are subject to reporting under 
Section 206, and 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 
which they provided under contract as safety-related are now 
categorized as RISC-3, thereby removing the vendor's reporting 
obligation under either part 21 or Sec.  50.55(e). Failing to inform a 
vendor that a safety-related SSC which 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 proposed 
Sec.  50.69, the Commission considered including a new requirement in 
either proposed 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 which it provided has been 
categorized as RISC-3. After consideration, the Commission believes 
that it is unlikely that such a 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 non-safety-related applications. A vendor would have some 
difficulty in determining whether the problem with the supplied SSC 
potentially affects the SSC recategorized 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 does not propose to adopt 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 AEA 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,

[[Page 26523]]

    (2) The capability to shut-down the facility and maintain it in a 
safe shut-down condition, or
    (3) The capability to prevent or mitigate the consequences of 
accidents which 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 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 would be removed from the scope of the 
requirements of Sec.  50.49 through 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 (see 
Section III.4.0). The requirements from 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), would remove 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 proposed 
rulemaking, RISC-3 SSCs would be removed from the scope of these 
requirements, and instead would be subject to the requirements in Sec.  
50.69(d)(2)(iii). 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 contains 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 BPV Code required by Sec.  50.55a(g) for ASME Class 1 SSCs, 
even if they were 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 3 SSCs that are shown to be of 
low safety-significance if categorized as RISC-3, the additional 
assurance from the specific provisions of the ASME Code is not 
considered necessary.
    Section 50.55a(h) incorporates by reference the requirements in 
either Institute of Electrical and Electronics Engineers (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 consistent with the Commission decision already discussed.
III.4.4 Section 50.65 Monitoring the Effectiveness of Maintenance
    The Commission is proposing to remove 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 includes Section III.4.0 and the following 
discussion.
    Section 50.65, referred to as 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 be monitored against licensee established 
goals, in a manner sufficient to provide confidence that the SSCs are 
capable of fulfilling their intended functions. 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 documentation and 
action requirements; thus, they are special treatment requirements. 
Upon implementation of Sec.  50.69, a licensee would not be required to 
apply maintenance rule monitoring, goal setting, corrective action, 
alternate demonstration, or periodic evaluation treatments required by 
Sec. Sec.  50.65(a)(1), (a)(2), and (a)(3) to RISC-3 and RISC-4

[[Page 26524]]

SSCs. The proposed rule does include in Sec.  50.69(e)(3) provisions 
for a licensee to use performance information to feedback into its 
processes to adjust treatment (or categorization) when results so 
indicate. However, this requirement does not require the specific 
monitoring and goal setting as required in Sec.  50.65, in 
consideration of the lesser safety-significance of these SSCs.
    RISC-1 and RISC-2 SSCs that are currently within the scope of Sec.  
50.65(b) would remain subject to existing maintenance rule 
requirements. Any RISC-1 or RISC-2 function not currently within the 
scope of Sec.  50.65(b) would be added to the scope of the maintenance 
rule (as a result of the requirement in Sec.  50.69(e)(2) that requires 
monitoring, evaluation and appropriate action for these SSCs).
    The proposed removal of RISC-3 and 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 proposed maintenance activities. The requirements in Sec.  
50.65(a)(4) remain in effect. It is noted that Sec.  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 safety-significance.
III.4.5 Sections 50.72 and 50.73 Reporting Requirements
    This proposed rule would remove the requirements in Sec. Sec.  
50.72 and 50.73 for RISC-3 and RISC-4 SSCs. The basis for this removal 
follows.
    Sections 50.72 and 50.73 contain requirements for licensees to 
report events involving certain SSCs. These reporting requirements are 
special treatment requirements . NRC requires event reports in part so 
that it can follow-up on corrective action for these circumstances. 
Through this rulemaking, the Commission proposes to remove RISC-3 and 
RISC-4 SSCs from the scope of these requirements. The low safety-
significance of these SSCs does not warrant the burden associated with 
reporting events or conditions only affecting such SSCs, for the 
reasons already discussed.In particular, under NRC's risk-informed 
inspection process, NRC follow-up of corrective action will be focused 
upon safety-significant situations.
III.4.6 10 CFR Part 50 Appendix B Quality Assurance Requirements
    This proposed rule would remove RISC-3 SSCs from the scope of 
requirements in Appendix B to 10 CFR Part 50. These requirements are 
currently not applicable to RISC-4 SSCs so there is no change for these 
SSCs. Appendix B contains requirements for a quality assurance program 
meeting specified attributes. While many of the general attributes are 
still appropriate for RISC-3 SSCs (and in some instances are included 
within the high-level requirements in Sec.  50.69(d)(2)), it was 
considered simpler to remove RISC-3 SSCs from the scope of the existing 
requirements in Appendix B (with its attendant set of guidance and 
implementing documents), and to add back the minimum set of 
requirements viewed as necessary for RISC-3 SSCs, rather than to 
subdivide the existing Appendix B requirements for this purpose.
    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 apply 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 
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
    The proposed rule would remove a subset of RISC-3 and RISC-4 SSCs 
from the scope of the requirements in Appendix J to Part 50 that 
pertain to containment leakage testing. Specifically, RISC-3 and RISC-4 
SSCs that meet specified criteria would be removed from the scope of 
the requirements for Type B and Type C testing. The basis for the 
removal is described below.
    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 which 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 (a) 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 (b) 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 requirement 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.
    The NRC believes that risk-informing this appendix may lead to less 
testing and therefore would reduce unnecessary regulatory burden on the 
licensees. Although the 1995 revision to Appendix J was characterized 
as risk-informed, the changes were not as extensive as those expected 
in the risk-informed Part 50 effort. The revision primarily decreased 
testing frequencies, whereas risk-informing the scope of SSCs that are 
subject to Appendix J testing would remove some components from testing 
(i.e., to the extent that defense-in-depth is maintained in accordance 
with the risk-informed categorization process).
    The proposed rule would exclude certain identified containment 
isolation valves from Type C testing. For RISC-3 components, which 
includes containment isolation valves, leak testing is not required. 
The reliability

[[Page 26525]]

strategy is to monitor and restore component functions once they are 
identified through the corrective action program or the periodic 
feedback process. Similarly, requirements for Type B testing of certain 
penetrations would not be required. The relief from testing is limited 
to components meeting specified criteria such that acceptable results 
for large early release and defense-in-depth are maintained.
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 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 concludes that Type A testing should 
continue to be required as described in Appendix J.
    Type B Testing: The Commission concludes that Type B testing should 
continue to be required for air lock door seals, including door 
operating mechanism penetrations which are part of the containment 
pressure boundary and doors with resilient seals or gaskets except for 
seal-welded doors. 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 less than 1 inch in equivalent diameter.
    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 pressure rating and that is not 
connected to the reactor coolant pressure boundary.
    4. The valve size is 1-inch nominal pipe size or less.
III.4.7.3 Basis for Reduction of Scope
    The first criterion for Type B testing deals with penetrations that 
are pressurized with the pressures in the penetrations being 
continuously monitored by licensees. The pressurization itself 
establishes a leak tight barrier, for such penetrations. The monitoring 
of the pressures in the penetrations, in conjunction with the proposed 
requirements for RISC 3 SSCs (including taking corrective action when 
an SSC fails) provide sufficient assurance, without the need for Type B 
testing, to ensure that these penetrations are functional.
    The second criterion for reducing the scope of Type B testing 
(i.e., penetrations less than 1 inch in equivalent diameter) is 
essentially the same as the fifth criterion for reducing the scope of 
Type C testing (i.e., valve size is 1-inch or less). By definition 
penetrations of this size do not contribute to large early release.
    The Commission finds that these criteria for reducing the scope of 
the Type C testing requirements are reasonable in that, even without 
Type C testing, the probability of significant leakage during an 
accident (that is, leakage to the extent that public health and safety 
is affected) is small. This is true even though some of the valves that 
satisfy these criteria may be fairly large.
    Appendix J to 10 CFR part 50 deals only with leakage rate testing 
of the primary reactor containment and its penetrations. It assumes 
that containment isolation valves are in their safe position. No 
failure is assumed that would cause the containment isolation valves 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, if a valve is open, it is assumed to be capable of being 
closed. Testing to ensure the capability of containment isolation 
valves 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 this proposed revision 
affecting Appendix J is negligible.
    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 in more detail the effect of containment 
leakage on risk 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 proposed 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 which 
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

[[Page 26526]]

containment penetrations typically have redundant isolation valves. 
While one valve may leak, the other may remain leak-tight. The study 
assumed that failure of one valve in a series failed the penetration; 
however, the probability of failure was that for a single valve.
    3. The analysis assumed possible leakage of all valves subject to 
Type C testing, not just those subject to the proposed revision.
    According to this analysis, the proposed revision would 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 proposed 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))
    The proposed rule would remove RISC-3 and RISC-4 SSCs from the 
requirement in Appendix A to 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 its Appendix A 
addresses seismic and geologic siting criteria used by the Commission 
to evaluate suitability of plant design bases in consideration of 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. The rule change would exclude 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. As discussed in Section III.4.0, because 
of the low safety significance of the RISC-3 and RISC-4 SSCs, the 
additional assurance provided by qualification testing (or engineering 
analyses) is not considered necessary.
    For current operating reactors, Appendix A to part 100 is 
applicable. For new plant applications, the seismic design requirements 
are set forth in Appendix S to Part 50. The NRC has determined that 
Appendix S does not need to be included in the proposed Sec.  50.69 
because the wording of the requirements with respect to 
``qualification'' by testing or specific types of analysis is not 
present in this rule. Therefore, a rule change would not be necessary 
to permit a licensee to implement means other than qualification 
testing or the specified methods to demonstrate SSC capability.
III.4.9 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 presented.
III.4.9.1 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.4.9.2 Section 50.36 Technical Specifications
    Section 50.36 establishes operability, surveillance, limiting 
conditions for operation and other requirements on certain SSCs. To the 
extent that this rule specified testing and related requirements, it 
was considered as a candidate for being ``special treatment''. However, 
the Commission concluded that it was not appropriate to revise Sec.  
50.36 for several reasons. First, risk-informed criteria have already 
been established in Sec.  50.36 for determining which SSCs should have 
TS requirements. Improved standard TS have already resulted in 
relocation of requirements for less important SSCs to other documents. 
Further, other improvement efforts are underway that could be 
implemented by individual licensees to make their plant-specific 
requirements more risk-informed. Thus, no changes to this rule (or its 
implementation) are necessary as part of Sec.  50.69 to make the TS 
risk-informed or to accommodate the revised requirements of this 
proposed rule.
III.4.9.3 Section 50.44 Combustible Gas Control
    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. The Commission notes that a 
separate rulemaking is underway to ``rebaseline'' the requirements in 
Sec.  50.44 using risk insights (see August 2, 2002; 67 FR 50374). 
Therefore, the NRC believes that there is no need to include those 
sections of (existing) Sec.  50.44 as being removed for RISC-3 SSC. If 
portions of Sec.  50.44 that were identified as special treatment 
requirements are retained, and/or relocated to other rules (and they 
are not necessary for RISC-3 SSCs), then there may be a need to 
reference these rules within Sec.  50.69(b)(1) when Sec.  50.69 is 
issued as a final rule.
III.4.9.4 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 rules themselves. 
The Commission has developed a proposed rulemaking (see November 1, 
2002; 67 FR 66578) to allow licensees to voluntarily adopt National 
Fire Protection Association (NFPA)-805 requirements in lieu of other 
fire protection requirements. NFPA-805 would permit a licensee to 
implement a risk-informed fire protection program as a voluntary 
alternative to compliance with Sec.  50.48 and 10 CFR Part 50, Appendix 
R. Accordingly, changes to these regulations were not included in the 
scope of the Sec.  50.69 rulemaking.
III.4.9.5 Section 50.59 Changes, Tests and Experiments
    The Commission does not believe that a Sec.  50.59 evaluation need 
be performed when a licensee implements Sec.  50.69 by changing the 
special treatment requirement for RISC-3 and RISC-4 SSCs. Accordingly, 
Sec.  50.69(f)(iii) contains language that removes the requirement for 
a Sec.  50.59 evaluation of the changes in special treatment as part of 
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

[[Page 26527]]

for recategorization and determination of revision to requirements 
resulting from the categorization. Thus, subjecting the implementation 
steps as they relate to changes to treatment from what was described in 
the final safety analysis report (FSAR), to determine if NRC approval 
is needed of those changes, is an unnecessary step. Since 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 would have any effect. As required by 
Sec.  50.69(f)(ii), the licensee/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 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. Section 50.69(d)(2)(i), which 
focuses upon design control, is intended to draw 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 of SSCs.
III.4.9.6 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 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.4.0).
III.4.9.7 10 CFR Part 52 Early Site Permits, Standard Design 
Certifications and Combined Operating Licenses
    Part 52 contains, by cross-reference, 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 the rulemaking 
effort. However, with the proposed ``applicability'' paragraph (Sec.  
50.69(b)) extending to applicants for a facility license or design 
certification under Part 52, the Commission presently sees no need for 
revisions to Part 52 itself.
III.4.9.8 10 CFR Part 54 License Renewal
    In SECY-99-256, 10 CFR part 54, which provides license renewal 
requirements, was identified as a candidate regulation for removal from 
scope of applicability to low significance SSCs. The aging management 
requirements could be viewed as being 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 
prior to 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 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 use of risk in establishing the scoping criteria within part 54 
was addressed by the Commission on May 8, 1995 (60 FR 22461), when 
amending part 54. In the 1995 amendment, the Commission stated that the 
current licensing basis for current operating plants is largely based 
on deterministic engineering criteria. Consequently, there was 
considerable logic in establishing license renewal scoping criteria 
that recognized the deterministic nature of a plant's licensing basis. 
Without the necessary regulatory requirements and appropriate controls 
for plant-specific PRAs, the Commission concluded that it was 
inappropriate to establish a license renewal scoping criterion that 
relied on plant-specific probabilistic analyses. Therefore, the 
Commission concluded further that within the construct of the final 
rule, PRA techniques were of very limited use for license renewal 
scoping (60 FR 22468).
    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,'' (60 FR 22471). Although Sec.  50.69 would remove 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 the 
proposed 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 (60 FR 22468). 
The NRC staff has 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.
III.4.9.9 Other Requirements
    In the ANPR and related documents, the 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 as it is being proposed.

III.5.0 Evaluation and Feedback, Corrective Action and Reporting 
Requirements

    The validity of the categorization process relies on ensuring that 
the performance and condition of SSCs continues to be maintained 
consistent with applicable assumptions. Changes in the level of 
treatment applied to an SSC might result in changes in the

[[Page 26528]]

reliability of the SSCs which are used in the categorization process. 
Additionally, plant changes, changes to operational practices, and 
industry operational experience may impact the categorization 
assumptions. Consequently, the proposed 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 proposed rule would require licensees to review in 
a timely manner but no longer than every 36 months, the changes to the 
plant, operational practices, applicable industry operational 
experience, and, as appropriate, update the PRA and SSC categorization. 
In addition, licensees would be 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 would 
need to ensure that sufficient information is obtained. For RISC-3 
SSCs, there is a relaxation of requirements for obtaining information 
when compared to the applicable special treatment requirements; however 
sufficient information would need to be obtained, and rule requirements 
are being proposed to consider performance data, see if adverse changes 
in performance might occur, and to make necessary adjustments such 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 proposed Sec.  50.69. For safety-significant SSCs, 
all current requirements would continue to apply and, as a consequence, 
Appendix B corrective action requirements would be applied to RISC-1 
SSCs to ensure that conditions adverse to quality are corrected. For 
both RISC-1 and RISC-2 SSCs, requirements would be 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 proposed rule would require that a licensee 
perform timely corrective action (Sec.  50.69(d)(2)(iv)). Further, as 
part of the feedback process, review of operational data may reveal 
inappropriate assumptions 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, and would also restore the 
capability of SSCs to perform their functions.
    Finally, the proposed rule would require reports of events or 
conditions that would have prevented RISC-1 and RISC-2 SSCs from being 
able to perform their safety-significant functions. A new reporting 
requirement would be added in Sec.  50.69(g) for events or conditions 
that would prevent RISC-2 SSCs from performing their safety-significant 
functions (if not otherwise reportable). Because the categorization 
process has determined that RISC-2 SSCs are of safety significance, NRC 
is interested in reports about circumstances where the safety-
significant function 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, 
such that NRC can take any actions deemed appropriate.
    Properly implemented, these requirements would ensure that validity 
of the categorization process and results are maintained throughout the 
operational life of the plant.

III.6.0 Implementation Process Requirements

    The proposed rule would also contain requirements specifying how a 
licensee (or applicant) would be able to use the alternative 
requirements in lieu of the existing requirements. The rule would 
specify applicability requirements as well as requirements on the 
Commission approval process for implementation.
    The Commission is making the provisions of Sec.  50.69 available to 
both applicants for licenses or design certification rules and to 
holders of facility licenses for light-water reactors. The proposed 
rule would be limited to light-water reactors because it was developed 
to risk-inform the scope of special treatment requirements which are 
applied to light-water reactors. Consequently, the technical aspects of 
the rule (e.g., providing reasonable confidence that risk increases 
(e.g., changes in CDF and LERF are small) including the implementation 
guidance, are specific to light-water reactor designs.
    Proposed Sec.  50.69 would rely on robust categorization to provide 
high confidence that the safety significance of SSCs is correctly 
determined. To ensure a robust categorization is employed, proposed 
Sec.  50.69 would require the categorization process to be reviewed and 
approved prior to implementation of Sec.  50.69 either by following the 
license amendment 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 proposed 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 peer review results. The NRC staff 
would also review other evaluations and approaches to be used such as 
margins-type analyses. Additionally, this review would examine any 
aspects of the proposed categorization guidance that are not consistent 
with the staff's regulatory guidance for implementing Sec.  50.69. 
Thus, the proposed rule would require 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 peer review process employed. An applicant would submit this 
information as part of its license application. The Commission 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. Commission 
action on an applicant's request would be part of the Commission 
decision on the license application.
    The Commission is proposing that the approval for a licensee to 
implement Sec.  50.69 be by license amendment. As discussed above, 
prior NRC review and approval of the licensee's proposed PRA, basis for 
sensitivity studies and evaluations, and results of PRA review process 
is required. This review will involve substantial professional 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 review results and sensitivity studies. Consistent 
with the

[[Page 26529]]

Commission's decision in Cleveland Electric Illuminating Co. (Perry 
Nuclear Power Plant, Unit 1), CLI-96-13, 44 NRC 315 (1996), the 
proposed rule would require NRC approval to be provided by issuance of 
a license amendment. The Nuclear Energy Institute (NEI) submitted a 
paper, ``License Amendments: Analysis of Statutory and Legal 
Requirements'' (NEI Analysis) in a July 10, 2002, letter to the 
Director of 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 
proposed 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 proposed Sec.  50.69 would permit 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 
proposed Sec.  50.69 is analogous to the review and approval process in 
Perry, 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 American Society for Testing and Materials (ASTM) standard 
set forth objective, non-discretionary criteria for changes to the 
withdrawal schedule, Sec.  50.69 does not contain such criteria for 
assessing the adequacy of the PRA process, PRA review results, and the 
sensitivity studies. Hence, the NRC's approval of a request to 
implement Sec.  50.69 will involve substantial professional judgment 
and discretion. In sum, 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 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 proposed Sec.  50.69 to contain requirements that ensure the 
categorization is robust to provide high 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 low 
safety significance; and requirements for obtaining sufficient 
information concerning the performance of these SSCs to 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 recognizes that this proposed rule may have 
implications with respect to NRC's reactor oversight process including 
the inspection program, significance determination process, and 
enforcement approach. In its final decision on this rulemaking, the 
Commission proposes to document its conclusions as to whether new or 
revised inspection or enforcement guidance is necessary.
    The Commission included requirements in the proposed rule for 
documenting categorization decisions to facilitate NRC oversight of a 
licensee's or applicant's implementation of the alternative 
requirements. The proposed rule would also include 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 SSC, 
in accordance with the rule requirements contained in Sec.  50.69, the 
Commission concludes that requiring further review as to whether NRC 
approval might be required for such changes is unnecessary burden. If a 
licensee is satisfying the rule requirements, as applied to RISC-3 SSC, 
the Commission could not postulate circumstances under which NRC 
approval of such changes would be required. Thus, a licensee would be 
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 design basis of SSCs.

III.7.0 Adequate Protection

    The Commission believes that reasonable assurance of adequate 
protection of public health and safety will be provided by applying the 
following principles in the development and implementation of proposed 
Sec.  50.69:
    (1) The net increase in plant risk is small;
    (2) Defense-in-depth is maintained;
    (3) Safety margins are maintained; and
    (4) Monitoring and performance assessment strategies are used.
    As described previously, 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. Proposed Sec.  50.69 was developed to incorporate these 
principles, both to ensure consistency with Commission policy, and to 
ensure that the proposed rule maintains adequate protection of public 
health and safety.
    The following discusses how proposed 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
    Proposed Sec.  50.69 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 review and approval of the 
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 safety significance, 
and that these requirements continue to be satisfied for SSCs of safety 
significance. The proposed rule requires that the

[[Page 26530]]

potential net increase in risk from implementation of proposed Sec.  
50.69 be assessed, and that this risk change is small. These 
requirements to provide reasonable confidence that the net change in 
risk is small as part of the categorization decision, in conjunction 
with the proposed 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 Commission has determined to be appropriate. 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 would require that the defense-in-depth philosophy 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 are important to 
defense-in-depth, as outlined in the implementation guidance, will be 
categorized as safety-significant (and will retain their treatment 
requirements). For safety-significant SSCs (i.e., RISC-1 and RISC-2 
SSCs), all current special treatment requirements would remain (i.e., 
the proposed rule does not remove any of these requirements) to provide 
high confidence that they can perform design basis functions, and 
additionally requires sufficient treatment be applied to support the 
credit taken for these SSCs for beyond design basis events. For RISC-3 
SSCs, proposed Sec.  50.69 would impose 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 the defense-
in-depth will continue to be available. The proposed rule does not 
change the design basis of the facility, which was established based 
upon defense-in-depth considerations. Accordingly, the Commission 
concludes that the proposed rule maintains defense-in-depth.
III.7.3. Safety Margins Are Maintained
    Proposed Sec.  50.69 maintains sufficient safety margins by a 
combination of:
    (1) Maintaining all existing functional and treatment requirements 
on RISC-1 and RISC-2 SSCs and additionally ensuring that any credit for 
these SSCs for beyond design basis conditions is valid and maintained; 
(2) maintaining the design basis of the facility for all SSCs, 
including RISC-3 SSCs as described above; and (3) requiring a licensee 
to have reasonable confidence that the overall increase in risk that 
may result due to implementation of proposed Sec.  50.69 is small.
    Maintaining current requirements on RISC-1 and RISC-2 SSCs, and 
ensuring that credit taken for these SSCs in the PRA for beyond design 
basis events is maintained, provides assurance that the safety-
significant SSCs continue to perform as assumed in the categorization 
process. Maintaining the design basis ensures that SSCs continue to be 
designed to criteria that ensure the SSCs perform their design basis 
functions, and therefore are nominally capable of performing their 
design basis functions. Because the only requirements that are relaxed 
are those related to treatment, existing safety margins for SSCs 
arising from the design technical and functional requirements would 
remain. The proposed rule would also require (through monitoring 
requirements) that the SSCs must be maintained such that they continue 
to be capable of performing their design basis functions. The reduction 
in treatment applied to RISC-3 SSCs may result in an increase in RISC-3 
failure rates (i.e., a reduction in RISC-3 reliability). To address how 
this relates to safety margin, proposed Sec.  50.69 would require that 
there be reasonable confidence that any potential increases in CDF and 
LERF be small from assumed changes in reliability resulting from the 
treatment changes permitted by the proposed rule. As a result, 
individual SSCs continue to be capable of performing their design basis 
functions, as well as to perform any beyond design basis functions 
consistent with the categorization process and results. Therefore, the 
Commission concludes that the proposed rule preserves sufficient safety 
margins.
III.7.4 Monitoring and Performance Assessment Strategies Are Used
    Proposed Sec.  50.69(e) would contain requirements that ensure that 
the risk-informed categorization and treatment processes are 
maintained, and reflect operational practices, the facility 
configuration, and SSC performance. In addition, proposed Sec.  
50.69(g) would contain requirements that reports are made to NRC of 
conditions preventing SSCs from performing their safety-significant 
functions. Together, these requirements maintain the validity of the 
risk-informed categorization and treatment processes such that the 
above criteria will continue to be satisfied over the life of the 
facility.
III.7.5 Summary and Conclusions
    Proposed Sec.  50.69 would contain requirements such that the net 
risk increase from implementation of its requirements is small; 
defense-in-depth is maintained; safety margins are maintained; and 
monitoring and performance assessment strategies are used. Together, 
these requirements result in a proposed Sec.  50.69 that is consistent 
with Commission policy on the use of PRA, and that maintains adequate 
protection of public health and safety.

IV. Public Input to the Proposed Rule

IV.1.0 Advance Notice of Proposed Rulemaking (ANPR) Comments

    The Commission published an ANPR (March 3, 2000; 65 FR 11488) to 
solicit public input on the direction and scope of this rulemaking. A 
number of comments were received. The NRC staff provided its 
preliminary responses to the issues raised by the commenters in SECY-
00-194, dated September 7, 2000. The Commission has considered these 
issues in developing the proposed rule. More detailed discussion of the 
comments and the Commission's preliminary positions are contained in a 
separate document (see Section X, Availability of Documents). A summary 
of some of the more substantive issues follows.
IV.1.1 Need for Prior NRC Review and PRA ``Quality''
    As originally envisioned in the ANPR, with development of a 
detailed Appendix T to contain the categorization process requirements, 
implementation of Sec.  50.69 could be undertaken without a prior NRC 
review and approval. As the rulemaking, guidance development, and pilot 
reviews progressed, it became apparent that some degree of NRC review 
would be necessary to determine that the PRA was technically adequate 
to support its use in the categorization process. While the completion 
of documents such as the ASME Standard for Probabilistic Risk 
Assessments for Nuclear Power Plant Applications and completion of peer 
reviews can lead to improved PRAs, there is still some lack of 
definitive guidance on preparation of PRAs that would allow use of PRA 
results in the manner anticipated without some degree of NRC review of 
the PRA itself. Concerns were also raised that excessive detail in the 
rule might be problematic and require exemptions. Thus, the approach 
that has been developed is for a rule with the minimum elements of the 
categorization process defined in the rule, a

[[Page 26531]]

requirement for NRC review and approval of the categorization process 
(including PRA peer review information) to be used, and detailed 
implementation guidance (in the form of a regulatory guide).
IV.1.2 Treatment Attributes
    Many of the ANPR comments focused on what treatment requirements 
should be established for various RISC categories of SSC. For example, 
there were comments that the requirements should not be ``added-on'' to 
existing requirements, but should reflect the significance of the SSCs. 
The Statement of Considerations of this rulemaking provides details 
about the decisions the Commission has made concerning the appropriate 
treatment requirements to include for the various categories of SSCs.
IV.1.3 Selective Implementation
    The Commission received a number of comments on selective 
implementation, both during the ANPR process and later. The Commission 
concludes that selective implementation of Sec.  50.69 should be 
allowed to permit a licensee/applicant to depart from compliance with a 
limited set of the special treatment rules delineated in Sec.  
50.69(b)(1). This topic is discussed further in Section V.5.1. Because 
of the existing requirements that would remain in place, a licensee 
could choose not to revise requirements for all of the rules within the 
scope of Sec.  50.69(b). However, there is no selective implementation 
for the overall requirements in Sec.  50.69. Thus for example, a 
licensee could not elect to adopt paragraph (b)(1) and not (d)(2).
    The other question was whether selective implementation with 
respect to the scope of SSCs to be categorized should be allowed. The 
Commission has determined that selective implementation on a system 
basis should be allowed, but not for components within a system. The 
rule includes specific language about this limitation. 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. As further discussed in 
Section III.2, the implementation, including the categorization process 
must address an entire system or structure, not selected components 
within a system.
    With respect to the question about categorizing only some systems, 
because the process of categorization of individual components within 
the systems can be time-consuming, categorization will occur over a 
period of time. In theory, certain systems might not be categorized at 
all. Initially there was some reservation that a licensee might only 
choose to categorize in systems where they anticipated relief from 
requirements (i.e., with a large set of RISC-3 SSCs) and would not 
categorize a system that would have RISC-2 SSCs. The Commission notes 
that requirements remain for RISC-3 SSCs until they are recategorized, 
and both sets of requirements are intended to maintain the design basis 
functions of RISC-3 SSCs. However, in categorizing any SSC, the 
categorization process may result in making assumptions about other 
SSCs in the plant (through the PRA modeling and in the IDP). In other 
words, for some SSCs to be of low safety significance, it is necessary 
for other SSCs to be safety-significant. For example, a RISC-2 SSC may 
be credited in the categorization process and subsequently another SSC 
becomes RISC-3 (low safety-significant). If a licensee wants to 
selectively implement Sec.  50.69 just for the system in which a 
particular RISC-3 SSC resides, then the licensee would also have to 
assure that the credit for the RISC-2 SSC is maintained also. To ensure 
that the categorization process is valid, such assumptions and credit 
must be retained over time, as determined by the PRA update process. 
Because the NRC will be reviewing the categorization process before 
implementation, NRC can determine if the categorization process is 
compatible with this approach.

IV.2.0 Draft Rule Comments

    On November 29, 2001 (66 FR 59546), the NRC staff released draft 
rule language for proposed Sec.  50.69, in response to guidance from 
the Commission dated August 2, 2001. The draft rule language was 
released to stakeholders as a means of obtaining early input from 
stakeholders about the rulemaking and how it would be implemented. The 
NRC staff received ten sets of comments from stakeholders in response 
to the FR notice. The NRC staff revised the draft rule and re-issued 
the revised language on April 5, 2002, taking into account the issues 
raised by the stakeholders. A third draft of the rule was made publicly 
available on August 2, 2002. Some revisions to the rule resulted from 
the input provided by the stakeholders and others were taken into 
account in the development of the SOC. The remaining discussion 
identifies the significant comments which resulted in changes to the 
draft rule.
    Many of the comments received related to the way in which the high-
level treatment requirements for RISC-3 SSCs were organized and worded. 
Based upon these comments, the NRC reduced the number of separate 
subsections (from 8 to 4), and simplified the wording by removing 
duplication of phrases. Suggested simplifications that were accepted 
were the deletion of details of the types of maintenance (corrective, 
predictive), and deletion of the words ``design inputs.'' Some 
stakeholders, such as the NEI, stated that the requirements were overly 
prescriptive and were not consistent with the concept of removing SSCs 
from the scope of NRC special treatment requirements. The issue about 
level of detail is the topic that drew the most comment during the 
draft rule language process. At the same time, comments and input from 
other stakeholders (including the Advisory Committee on Reactor 
Safeguards (ACRS), were resulting in strengthening of the 
categorization process such that any individual SSC categorized as 
RISC-3 is of very low safety significance. Specific consideration was 
also added in the rule requirements to deal with potential common-cause 
failures. Based upon this evolution, concerns about prescriptiveness as 
stated in these comments led the Commission to simplify the 
requirements on treatment for RISC-3 SSCs.
    Another part of the draft rule that drew comment was the 
requirement for monitoring of RISC-3 SSCs. Some of the comments 
indicated that this was not necessary for low safety-significant SSCs, 
and was inconsistent with the removal of maintenance rule monitoring 
(by removing Sec.  50.65(a)(1) through (3) as requirements). In the 
proposed rule, the Commission has clarified that the type of monitoring 
of availability and failures under the maintenance rule is not 
necessary and that the type of monitoring appropriate for RISC-3 SSCs 
is the performance monitoring specified in Sec.  50.69(d)(2)(iii) and 
the feedback specified in Sec.  50.69(e)(3).
    Other comments proposed that the scope of rules being removed 
should be expanded to include the requirements in Sec.  50.55a (ASME 
code requirements), and Appendix A to Part 100. Rule language was added 
to accomplish this by listing specific subsections of Sec.  50.55a and 
Appendix A to Part 100 in the list of requirements removed, and through 
other changes to the rule designed to maintain the necessary 
reliability of SSCs. The ASME provided comments on the draft rule 
language

[[Page 26532]]

stating that the risk-informed Code Cases and Standards developed by 
ASME should not be directly referenced in the rule, but that there 
should be a framework developed to ensure that the Code Cases are used, 
and that partial use does not occur. The proposed rule permits, but 
does not require, use of the Code Cases for purposes of meeting rule 
requirements. The Commission notes that these Code Cases cover both 
categorization and treatment requirements in the areas of inservice 
inspection, inservice testing, and repair/replacement. The Commission 
expects licensees will utilize the ASME Code Cases as part of their 
implementation of Sec.  50.69.
    Another commenter stated that the rule should be made applicable to 
applicants as well as license holders, and NRC agreed that this was 
appropriate and made revisions to the rule language to accommodate 
this. Another commenter stated that the wording of the requirement to 
``assure risk is small from changes to treatment'' set an impossible 
standard, and that the rule wording should be revised to allow use of 
sensitivity studies to provide confidence that the risk is small. The 
NRC agreed with this comment and revised the rule wording in the manner 
suggested that the licensee provide reasonable confidence that the 
increase in risk is small through performance of appropriate 
evaluations, such as sensitivity studies for SSCs modeled in the PRA.
    A commenter thought it was unnecessary to require that a schedule 
or scope of systems to be categorized be part of the submittal. It was 
noted that implementation of the rule would of necessity occur over 
time, and that existing requirements would remain in effect until SSCs 
were categorized. Thus, the commenter believes that a licensee should 
not be held to any particular schedule for implementation. The NRC's 
intent in requesting a schedule and scope was for informational 
purposes to know what requirements would be in effect, but agrees that 
a firm commitment to a schedule is not required. This part of the rule 
was removed, and instead there is a requirement to update the FSAR, in 
accordance with Sec.  50.71(e), to reflect implementation as it occurs 
for particular systems.

IV.3.0 Pilot Plants

    To aid in the development of the proposed 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. The 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 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 integrated decision-making panel (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 staff developed and 
issued a trip report containing the staff's observations.
    The following points set forth the principal lessons learned and 
key feedback from the NRC staff's observations of the pilot activities.
    [sbull] Potential treatment changes and their potential effects 
need to be understood by the IDP as part of the deliberations on 
categorization.
    [sbull] The pilots showed the importance of documentation of the 
IDP decisions and the basis. The rule contains a requirement for the 
categorization basis to be documented (and records retained) in Sec.  
50.69(f).
    [sbull] 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. Based upon the pilot experience, NRC adjusted its 
approach to 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.
    [sbull] The need for a number of improvements to the implementation 
guidance in NEI 00-04 were noted, for instance, improvement in a 
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.
    [sbull] 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 because of 
such considerations 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.
    [sbull] 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.
    [sbull] 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.
    [sbull] 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 and developed draft revision C of the implementation guidance 
(discussed in Section VI).

IV.4.0 South Texas Exemption as Proof-of-Concept

    A major element of the rulemaking plan described in SECY-99-256 was 
the review of the South Texas Project Nuclear Operating Company 
(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

[[Page 26533]]

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 not risk-significant. The scope of the 
exemptions requested included only those safety-related SSCs that have 
been categorized as low safety-significant or as nonrisk 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 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 NRC 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), 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, to inform the Commission of the staff's finding 
regarding the STPNOC exemption 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 proposed 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 proposed Sec.  50.69, the NRC continues to require a robust 
categorization with a detailed staff review.
    The proposed rule specifies the requirement that the licensee 
provide reasonable confidence in functionality and further specifies 
some high-level requirements for SSC treatment. Under proposed Sec.  
50.69, the NRC does not plan to review each licensee's plan for SSC 
treatment in detail. Licensees will have to establish appropriate 
performance-based SSC treatment processes to maintain the validity of 
the categorization process and its results. The proposed rule would 
require 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 proposed 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 
XIII 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.'' As discussed in Section II of the SOC, RISC-1 SSCs are 
those SSCs that are safety-related (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 which are 
safety-significant. RISC-3 SSCs are safety-related but are low safety-
significant. Finally, RISC-4 SSCs are not safety-related and are 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 are of more importance 
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, including 
risk significance, and not just ``risk-significant.''
    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 DG-1121, ``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 Sec.  50.69(b) provides that the rule may be voluntarily 
implemented by:
    (1) Holders of Sec.  50.21(b) or Sec.  50.22 light water reactor 
(LWR) operating licenses;
    (2) Holders of Part 54 renewed LWR licenses;
    (3) A person seeking a design certification under Part 52 of this 
chapter; or
    (4) Applicants for a LWR license under Sec.  50.22 or under Part 
52.
    For current licensees, implementation will be through a license 
amendment as set forth in Sec.  50.90. Until the request is approved, a 
licensee would continue to follow existing requirements. Upon approval 
of the categorization process (and review of the supporting PRA), the 
licensee can begin implementation by performing categorization of SSCs 
and revising treatment requirements accordingly.

[[Page 26534]]

    Applicants would be permitted to implement the treatment 
requirements, although the process involved for them would likely be 
different, depending upon the stage at which they seek approval. An 
applicant would have to categorize its SSCs into the four RISC 
categories, which would first require the applicant to design the 
facility to meet the Part 50 requirements including classifying SSCs 
according to the safety-related definition of Part 50. The applicant 
could then use the provisions of Sec.  50.69 (upon NRC approval) to 
categorize SSCs into the four RISC categories, and this in turn would 
enable the applicant to initially procure these SSCs to meet the 
applicable Sec.  50.69 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. In the development of Sec.  
50.69, questions have been received regarding what would be the impact 
to licensees that implement the proposed Sec.  50.69 and then 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 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.
    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 considers 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.
    Adopting the proposed Sec.  50.69 requirements for an applicant 
that has not obtained a Sec.  50.21(b) or Sec.  50.22 operating license 
(e.g. for a construction permit holder), is not as straightforward, and 
requires that the applicant first design the facility to meet the 
current Part 50 requirements. Specifically, to use the proposed Sec.  
50.69 requirements requires that SSCs first be classified into the 
traditional safety-related and nonsafety-related classifications. This 
establishes the design basis for the facility, which as previously 
stated, the proposed Sec.  50.69 is not changing. Once the SSC 
categorization has been done consistent with the safety-related 
definition in Sec.  50.2, then proposed Sec.  50.69 can be used to re-
categorize SSCs into RISC-1, RISC-2, RISC-3, and RISC-4, and the 
alternative treatment requirements of proposed Sec.  50.69 implemented. 
A new applicant who chooses to adopt these proposed 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 proposed Sec.  50.69 requirements before 
initial plant operation. An applicant who references a certified design 
and wishes to implement Sec.  50.69 would include the specified 
information as part of its application for a license. This does not 
mean that an applicant would actually construct the facility per all 
Part 50, and 100 requirements first, before applying Sec.  50.69. 
Instead, the facility needs to be designed per these requirements, but 
following approval of application of Sec.  50.69, RISC-3 SSCs could be 
procured per the requirements of Sec.  50.69(d).
    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 only applicable to light-water reactor designs.
    An applicant for a design certification could request to implement 
Sec.  50.69 with respect to categorizing SSCs. Because the rule 
requirements in Sec.  50.69 include elements of procurement and 
installation, as well as inservice activities, implementation of the 
rule by a holder of a manufacturing license or by a design 
certification applicant would have implications for 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. However, applicability of this proposed rule is not excluded 
for manufacturing licenses or design certificate applicants.
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) of the proposed rule 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 of the rule requirements (or portions thereof) that are 
being removed by this rulemaking are listed in a separate item, 
numbered from Sec.  50.69(b)(1)(i) through (ix). The basis for removal 
of these requirements was discussed earlier. 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 
requirements in Sec.  50.69(d)(2), which when effectively implemented 
by licensees to provide a sufficient level of confidence that RISC-3 
SSCs continue to be capable of performing their safety-related 
functions under design basis conditions. Note that special treatment 
requirements are not removed from any SSCs until 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
    Proposed Sec.  50.69(b)(2) would require a licensee who voluntarily 
seeks to implement Sec.  50.69 to submit an application for a license 
amendment pursuant to 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

[[Page 26535]]

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 shall 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 draft regulatory 
guide DG-1121 which endorses NEI 00-04, with some conditions and 
exceptions. If a licensee or applicant wishes to use a different 
approach, the submittal would need to provide 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 proposed Sec.  50.69. The measures 
described would 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 and this is described below in Section 
VI.2. The licensee/applicant would 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) would be required to include 
information about the evaluations they intend to conduct to provide 
reasonable confidence that the increase in risk would be small. This 
would include any sensitivity studies for RISC-3 SSCs, including the 
basis for whatever change in reliability being assumed for these 
analyses. A licensee would need to provide sufficient information for 
the NRC describing the sensitivity studies and other evaluations, and 
the basis for their acceptability as appropriately representing the 
potential increase in risk from implementation of the revised 
requirements in this proposed rule.
    As discussed elsewhere, the RISC-3 SSCs have low safety 
significance under Sec.  50.69. The Commission expects licensees and 
applicants to implement effective treatment processes to maintain RISC-
3 functionality that comply with Sec.  50.69(d). Those processes do not 
need to be described to the NRC as part of the proposed Sec.  50.69 
submittal under Sec.  50.69(b)(2).
V.3.3 Section 50.69(b)(3) Approval for Licensees
    Section 50.69(b)(3) would further provide 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) would require that an applicant for a license 
(or for a design certification) that chooses to implement proposed 
Sec.  50.69 must submit the information listed in Sec.  50.69(b)(2) as 
part of its application for a license. As previously discussed, 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. The above-cited 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 would be met, as part of its action on the application for 
a license or the design certification rule. As noted in Section V.3.0, 
an applicant referencing a certified design that implemented Sec.  
50.69 would need to adopt the remaining provisions of Sec.  50.69 or 
apply the other requirements in Part 50 to its processes.

V.4.0 Section 50.69(c) Categorization Process Requirements

    Section 50.69(c) would establish 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)) that discusses how their proposed 
categorization process, supporting PRA, and evaluations meet the Sec.  
50.69(c) requirements. As described above in Section III.2.0, these 
requirements are intended to ensure that the risk-informed Sec.  50.69 
categorization process determines the safety significance of SSCs with 
a high level of confidence. The introductory paragraph states that SSCs 
must be categorized as RISC-1, 2, 3, or 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. How this is to be accomplished is discussed 
in Section V.4.1.1. 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, which would include low power and shutdown modes, must 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 in order 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

[[Page 26536]]

it represents a coherent, integrated model, and have sufficient detail 
to support the initial 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 also prepared a draft regulatory guide (DG-1122) 
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 staff would use 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. As discussed in Section VI, NRC has developed 
review guidelines for considering the sufficiency of a PRA that was 
subjected to the NEI peer review process, as it would be used in 
implementation of Sec.  50.69. The submittal requirements listed in 
Sec.  50.69(b)(2) include a requirement to provide information about 
the quality of the PRA analysis and about the peer review results.
V.4.1.2 Risk Categorization Process Based on PRA Information
    For SSCs modeled in the PRA, the 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 tools to identify safety significance 
of SSCs, for example, in the implementation of Sec.  50.65.) In 
addition to being a useful tool to help prioritize NRC staff and 
licensee resources, use of importance measures can provide a systematic 
means to identify improvements to current plant practices. 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 must be determined 
based on both CDF and LERF. Importance measures should be chosen so 
that results can provide the IDP 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 criteria 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. IDP 
consideration of Sec. 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 importance 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 calculation of SSC importance should take into account the 
combined 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) event probabilities.
    Another concern that arises because importance measures are 
typically evaluated on the basis of individual events is that 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 
out 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 assumed 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, the importance 
measures should be evaluated for each analysis separately, and the 
results of the analyses should be provided to the IDP.
    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. In IDP deliberations on the 
sensitivity study results, the following should be considered:
    (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

[[Page 26537]]

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, as discussed below. 
Because importance measures are not available for use as screening, 
other criteria or considerations must be used by the IDP to determine 
the significance. To provide the necessary structure, the Commission is 
setting forth guidance on how these deliberations should be conducted; 
this information will also be included in the regulatory guidance for 
this proposed rule. These considerations were selected based upon NRC 
experience about what functions are important to prevention of core 
damage or large early release.
    The proposed rule would also include requirements 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 of ``reasonably reflect'' was 
selected to allow for appropriate PRA modeling and also to make clear 
that the PRA and 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.4.2.1 Initiating Events and Plant Operating Modes Not Modeled in the 
PRA
    When initiating events with frequencies of greater than 
10-\6\ per year are not modeled in the PRA, or when the low 
power and shutdown plant operating modes are not modeled in the PRA, 
other means are needed to determine the safety significance to meet 
Sec.  50.69(c)(1). The proposed implementation guidance contains 
information about how this can be accomplished by the IDP assessments. 
The licensee should demonstrate that the relaxation of regulatory 
requirements will not unacceptably degrade plant response capability 
and will not introduce risk vulnerabilities for the unmodeled 
initiating events or plant operating modes. For these unmodeled events, 
the IDP assessment should consider whether an SSC has an impact on the 
plant's capability to:
    (1) Prevent or mitigate accident conditions,
    (2) Reach and/or maintain safe shutdown conditions,
    (3) Preserve the reactor coolant system pressure boundary 
integrity,
    (4) Maintain containment integrity, or
    (5) Allow monitoring of post-accident conditions.
    In determining the importance of SSCs for each of these functions, 
the following factors should be considered:
    [sbull] Safety function being satisfied by SSC operation
    [sbull] Level of redundancy existing at the plant to fulfill the 
SSC's function
    [sbull] Ability to recover from a failure of the SSC
    [sbull] Performance history of the SSC
    [sbull] Use of the SSC in the Emergency Operating Procedures or 
Severe Accident Management Guidelines
    The licensee or applicant (through the IDP) must document the basis 
for the assignment of an SSC as RISC-3 based on the above 
considerations. Insights and results from risk assessment and risk 
management methodologies (for example the fire and external events 
screening methodologies, the seismic margins analyses, or the shutdown 
safety management models) may be used to help form this basis.
V.4.2.2 SSCs Not Modeled in the PRA
    In addition to being safety-significant in terms of their 
contribution to CDF or LERF, SSCs can also be safety-significant in 
terms of other risk metrics or conditions. Therefore, for SSCs not 
modeled explicitly in the PRA, the IDP should verify low safety 
significance based on traditional engineering analyses and insights, 
operational experience, and information from licensing basis documents 
and design basis accident analyses. The IDP should assess the safety 
significance of these SSCs by determining if:
    (1) Failure of the SSC will significantly increase the frequency of 
an initiating event, including those initiating events originally 
screened out in the PRA.
    (2) Failure of the SSC will compromise the integrity of the reactor 
coolant pressure boundary. It is expected that a sufficiently robust 
categorization process would result in the reactor coolant pressure 
boundary being categorized as RISC-1.
    (3) Failure of the SSC will fail a safety-significant function, 
including SSCs that are assumed to be inherently reliable in the PRA 
(e.g., piping and tanks) and those that may not be explicitly modeled 
(e.g., room cooling systems, and instrumentation and control systems). 
For example, it is expected for PWRs that a sufficiently robust 
categorization process would categorize high energy ASME Section III 
Class 2 piping of the main steam and feedwater systems as RISC-1.
    (4) The SSC supports important operator actions required to 
mitigate an accident, including the operator actions taken credit for 
in the PRA.
    (5) Failure of the SSC will result in failure of safety-significant 
SSCs (e.g., through spatial interactions or through functional reliance 
on another SSC).
    (6) Failure of the SSC will impact the plant's capability to reach 
and/or maintain safe shutdown conditions.
    (7) The SSC is one of a redundant set that can be justifiably 
identified as a common cause failure group.
    (8) The SSC is a part of a system that acts as a barrier to fission 
product release during severe accidents. It is expected that a 
sufficiently robust categorization process would result in fission 
product barriers (e.g., the containment shell or liner) being 
categorized as RISC-1.
    (9) The SSC is depended upon in the Emergency Operating Procedures 
or the Severe Accident Management Guidelines.
    (10) Failure of the SSC will result in unintentional releases of 
radioactive

[[Page 26538]]

material in excess of 10 CFR part 100 guidelines even in the absence of 
severe accident conditions.
    (11) The SSC is relied upon to control or to mitigate the 
consequences of transients and accidents.
    If any of these conditions is true, the IDP should use a 
qualitative evaluation process to determine the impact of relaxing 
requirements on SSC reliability and performance. This evaluation should 
include identifying the functions being supported by SSC operation, the 
relationship between the SSC's failure modes and the functions being 
supported, the SSC failure modes for which the failure rate may 
increase, and the SSC failure modes for which detection could become or 
are more difficult. The IDP can then justify low safety significance of 
the SSC by demonstrating the following:
    [sbull] The categorization is consistent with the defense-in-depth 
philosophy (per Section V.4.3 below).
    [sbull] Operating experience indicates that degradation mechanisms 
(e.g., for piping flow accelerated corrosion or microbiologically-
induced corrosion), for passive and active SSCs are not present, 
relaxing the requirements will have minimal impact on the failure rate 
increase, and degradation in the ability of the SSC to perform its 
safety function can be detected in a timely fashion.
    [sbull] Relaxing the requirements will have a minimal impact on the 
expected onsite occupational or offsite doses from transients and 
accidents that do not contribute to CDF or LERF.
V.4.3 Section 50.69(c)(1)(iii) Maintaining Defense-in-Depth Philosophy
    Section 50.69(c)(1)(iii) requires that the categorization process 
maintain the defense-in-depth philosophy. To satisfy this requirement, 
when categorizing SSCs as low safety-significant, the IDP must 
demonstrate that the defense-in-depth philosophy is maintained. 
Defense-in-depth is considered adequate if the overall redundancy and 
diversity among the plant's systems and barriers is sufficient to 
ensure the risk acceptance guidelines discussed below in Section V.4.4 
are met, and that:
    [sbull] Reasonable balance is preserved among prevention of core 
damage, prevention of containment failure or bypass, and mitigation of 
consequences of an offsite release.
    [sbull] 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.
    [sbull] There is no over-reliance on programmatic activities and 
operator actions to compensate for weaknesses in the plant design, and
    [sbull] 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 the defense-in-depth philosophy 
(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. One way to do this would be to show that these SSCs are not 
relied on to prevent late containment failure during core damage 
accidents. An alternative method would be to demonstrate that a 
potential decrease in reliability of RISC-3 SSCs that support the 
containment function does not have significant impact on the estimate 
of late containment failure probability. In essence, what the NRC 
expects is for a plant-specific understanding of the effects of 
containment systems on large late releases and an understanding of the 
credit given to these systems in maintaining the conditional 
probability for these releases. A licensee or applicant can 
qualitatively argue that an SSC is not relied upon to prevent large 
late containment failure and is thus low safety-significant from this 
standpoint. If an SSC plays a role in supporting the containment 
function in terms of large late releases, and if the licensee wants to 
categorize these SSCs as low safety-significant (for example, because 
of available redundant systems or trains or because failure is 
dominated by factors not related to the SSC), NRC would find acceptable 
the use of sensitivity studies to show that the effects on (i.e., 
change in) the late containment failure probability is small (i.e., 
less than a 10 percent increase from the base value) and that factors 
such as common cause failures or other dependencies are not important. 
Where a licensee categorizes containment isolation valves or 
penetrations as RISC-3, the licensee will need to address the impact of 
the proposed change in treatment on a case-by-case basis to ensure that 
the defense-in-depth principle continues to be satisfied.
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 SSC, 
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 maintenance, 
inspection, tests, and surveillance activities are adequate to prevent, 
detect and correct significant SSC performance and reliability 
degradation. Later sections of this SOC provide discussion on the 
proposed treatment processes the licensee will implement to provide 
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 small are discussed below.
    As part of their submittal, a licensee (or applicant) is to 
describe the evaluations to be conducted for purposes of meeting the 
requirement that there would be no more than a small (potential) 
increase in risk. For SSCs included in the PRA, the Commission expects 
that sensitivity studies (evaluations) would be done to provide a basis 
for concluding that even if reliability of these SSC should degrade 
because of the changes in treatment, the potential risk increase would 
be small. Satisfying the rule requirement that the risk increase is 
small presumes that the increase in failure rates assumed in the PRA 
sensitivity study bounds any reasonable estimate of the increase that 
may be expected as a result of the proposed changes in treatment.
    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 include potential impacts from known degradation 
mechanisms on both active and passive functions. It is

[[Page 26539]]

necessary for a licensee to consider the impact that a change in 
treatment (as a result of removal of special treatment requirements) 
might have on the ability of the SSC to perform its design basis 
function and on reliability of SSCs. The purpose is to provide an 
understanding of the new treatment requirements and their effects on 
RISC-3 SSCs due to removal of special treatment requirements. This will 
help form the basis for the change-in-risk evaluations and will support 
developing a technical basis for concluding that SSC performance is 
consistent with the categorization process and its results and with 
those evaluations performed to show that there is a no more than a 
small increase in risk associated with implementation of Sec.  50.69. 
The basis supporting the evaluations that examine potential SSC 
reliability changes due to treatment changes may be either qualitative 
or quantitative.
    One mechanism that could lead to large increases in CDF/LERF is 
extensive, across system common cause failures. However, for such 
extensive CCFs to occur would require that the mechanisms that lead to 
failure, in the absence of special treatment, were 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.
    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. As an example of how this would 
be implemented, the known existence of certain degradation mechanisms 
affecting pressure boundary SSC integrity might support retaining the 
current requirements on inspections or examinations or use of the risk-
informed ASME Code Cases, as accepted by the NRC regulatory process. An 
alternative might be to relax certain elements of treatment, but retain 
those that were assessed to be the most effective in negating the 
degradation mechanisms. As another example, changing levels of 
treatment on several similar components that might be sensitive to CCF 
potential would require consideration as to whether the planned 
monitoring and corrective action program, or other aspects of 
treatment, would be effective in sufficiently minimizing CCF potential 
such that the evaluations remain bounding.
    The treatment for all RISC-3 SSCs may not need to 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 small and 
acceptable, the above 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 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. The 
Commission regards ``small'' changes for plants with total baseline CDF 
of 10-4 per year or less to be CDF increases of up to 
10-5 per year, and plants with total baseline CDF greater 
than 10-4 per year to be 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 in order for the 
reduction in special treatment requirements to be considered. For 
plants with total baseline LERFs 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 LERFs greater than 
10-5 per year, LERF increases of up to 10-7 per 
year. Similarly, if there is an indication that the LERF may be 
considerably higher than 10-5 per year, the focus of the 
licensee should be on finding ways to decrease rather than increase 
LERF and the licensee may be required to present arguments as to why 
steps should not be taken to reduce LERF in order for the reduction in 
special treatment requirements to be considered. 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.
    The licensee can choose a factor for the increase on unreliability 
such that the corrective action and feedback processes discussed in 
Sec. Sec.  50.69(d)(2) and 50.69(e)(3) would provide sufficient data to 
substantiate that the increased unreliability used in the evaluations 
is not exceeded.
    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 a period of time. 
However, the implementation, including the categorization process, must 
address entire systems or structures; not selected components within a 
system or structure.
V.4.6 Section 50.69(c)(2) Use of Integrated Decision-Making Panel (IDP)
    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, which uses both risk insights and traditional engineering 
insights. In categorizing SSCs as low safety-significant, it should be 
demonstrated that the defense-in-depth philosophy is maintained, that 
sufficient safety margin is maintained, and that increases in risk

[[Page 26540]]

(if any) are small. 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 shall 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 qualification 
of members of the IDP. The IDP should be composed of a group of at 
least five experts 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 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 this philosophy.
    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) Requirements for Structures, Systems, and 
Components

    After SSCs are categorized as either RISC-1, RISC-2, RISC-3, or 
RISC-4, then the Sec.  50.69(d) requirements, which provide the 
treatment requirements applicable to each RISC category, are applied. 
Until a structure or system is categorized using this process, the 
existing requirements on SSCs in that structure or system are retained. 
Section 50.69(d) contains two sub-items. The first contains the 
requirements being imposed on RISC-1 and RISC-2 SSCs. The second 
section contains the ``high-level'' requirements that are being added 
for RISC-3 SSCs to provide necessary confidence that design basis 
capability will be retained for these SSCs. The list of existing 
special treatment requirements that are being removed through this 
rulemaking for RISC-3 and RISC-4 SSCs is contained in Sec.  
50.69(b)(1).
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 
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. To 
meet this, a licensee should first evaluate the treatment being applied 
in light of the credit being taken in the categorization process, with 
appropriate adjustment of treatment or categorization to achieve 
consistency as necessary. For SSCs categorized as RISC-1 or RISC-2, all 
existing applicable requirements continue to apply. This includes any 
applicable special treatment requirements. The rule language notes that 
this evaluation is to focus upon 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 designed or otherwise determined to be 
able to perform as assumed. Other examples might be for the failure 
rates used in the PRA model. As a general matter, for those SSCs 
modeled in the PRA, conformance with industry standards on PRAs would 
also result in such evaluation steps being accomplished in order to 
help assure the PRA represents the facility.
    If a Sec.  50.69 licensee chooses to categorize a selective set of 
SSCs as RISC-3, and the categorization of SSCs as RISC-3 is based on 
credit taken for the performance of other plant SSCs (that would be 
RISC-1 or RISC-2, whether or not these SSCs are within the selective 
implementation set), then the licensee must ensure that consistency of 
performance with what was credited in the categorization. As discussed 
in Section V.4.5, selective implementation of components within a 
system is not permitted. This applies to credit taken in: 1) PRA 
models, inputs and assumptions; 2) screening and margin analyses; and 
3) IDP deliberations. This implies that the licensee must ensure that 
the credited (RISC-2) SSCs perform their functions per Sec.  
50.69(d)(1), and the performance of these SSCs must be monitored per 
Sec.  50.69(e)(2).
V.5.2 Section 50.69(d)(2) RISC-3 Treatment
    Section 50.69(d)(2) contains, as an overall requirement for the 
treatment of RISC-3 SSCs, that licensees shall have processes to 
control the design; procurement; inspection, maintenance, testing, and 
surveillance; and corrective action, for RISC-3 SSCs to provide 
reasonable confidence in the capability of RISC-3 SSCs to perform their 
safety-related functions under design basis conditions throughout their 
service life. In other words, the Commission expects licensees to have 
sufficient treatment controls in place to have reasonable confidence 
that RISC-3 SSCs will be capable of performing their safety functions 
if they were called upon to perform those functions. Licensees may 
decide to apply current practices at their facilities or may establish 
new practices for the treatment of RISC-3 SSCs, provided the 
requirements of Sec.  50.69 are satisfied.
    During its review of the South Texas exemption request, the NRC 
staff identified several instances where the licensee's interpretation 
of the extent to which treatment could be relaxed for low-risk safety-
related SSCs was not consistent with the staff's expectations under 
Option 2 of the NRC's risk-informed rulemaking initiative (i.e., that

[[Page 26541]]

design basis functions be maintained). To ensure more consistent 
implementation of Sec.  50.69, the SOC discusses some of these areas 
for the implementation of proposed Sec.  50.69 about how the treatment 
processes for low-risk safety-related SSCs should be conducted. The 
Commission is also giving examples of what it considers good practice 
to achieve confidence of functionality. The Commission does not believe 
that it is necessary to include these ``expectations'' as specific 
requirements because there may be other means of achieving the 
specified outcome and failure to implement a particular expectation 
would not, by itself, be a regulatory concern. The Commission's intent 
is to place on the licensee the responsibility to determine the 
necessary treatment to maintain functionality without the Commission 
having to establish prescriptive requirements.
    The categorization process assumes that the functionality of SSCs 
in performing their safety functions will be retained, although the 
treatment applied to RISC-3 SSCs may be reduced under proposed Sec.  
50.69. Further, the categorization process may include specific 
reliability assumptions for plant SSCs in performing their intended 
functions. Therefore, when establishing the performance-based treatment 
process for RISC-3 SSCs, the licensee should take these assumptions 
into account to support the evaluations of small increase in risk 
resulting from implementation of the changes in treatment. It is 
important to obtain sufficient information on SSC performance to allow 
the results of the categorization process to remain valid. The 
Commission considers the risk-informed, performance-based ASME Code 
Cases (as endorsed in Sec.  50.55a) to be one acceptable method of 
establishing treatment processes that are consistent with the 
categorization process.
    Proposed Sec.  50.69 identifies four processes that must be 
controlled and accomplished for RISC-3 SSCs: Design Control; 
Procurement; Maintenance, Inspection, Testing, and Surveillance; and 
Corrective Action. The high level RISC-3 requirements are structured to 
address the various key elements of SSC functionality by focusing in 
several areas. When SSCs are replaced, RISC-3 SSCs must remain capable 
of performing design basis functions; hence, the high level 
requirements focus on maintaining this capability through design 
control and procurement requirements. During the operating life of a 
RISC-3 SSC, a sufficient level of confidence is necessary that the SSC 
continues to be able to perform its design basis functions; hence, the 
inclusion of high level requirements for maintenance, inspection, test, 
and surveillance. Finally, when data is collected, it must be fed back 
into the categorization and treatment processes, and when important 
deficiencies are found, they must be corrected; hence, requirements are 
also provided in these areas.
    The Commission notes that use of voluntary consensus standards is 
an effective means of establishing treatment requirements to achieve 
functionality. As an example, ASME risk-informed Code Cases have been 
developed with the purpose of determining appropriate treatment 
requirements for low safety-significant SSCs in their specific 
functional areas. Further, the Commission expects that related 
standards (such as ASME Code Cases N-658 and N-660 on SSC 
categorization and treatment for purposes of repair and replacement) be 
used in conjunction with each other as intended by the accredited 
standards writing body. Where suitable standards do not exist or 
available standards are not sufficient, the Commission expects the 
licensee to establish sufficient controls to provide reasonable 
confidence in the functionality of RISC-3 SSCs, based upon such factors 
as operating experience and vendor recommendations. However, the 
Commission also notes that use of a voluntary consensus standard in and 
of itself might not be sufficient to maintain functionality for 
particular SSCs under certain service conditions, and that the licensee 
might need to supplement its processes to achieve the desired results.
    The proposed rule would require the treatment processes for RISC-3 
SSCs be implemented to provide reasonable confidence in the capability 
of RISC-3 SSCs to perform their safety-related functions under design 
basis conditions. That is to say, the pertinent requirements identified 
in Sec.  50.69 for each process must be satisfied for RISC-3 SSCs 
unless the requirements are clearly not applicable or are not necessary 
in the particular circumstance to achieve functionality of the SSC. As 
an example, a licensee might determine that it is more efficient and 
effective to replace a particular component before the end of its 
design life rather than conducting maintenance to repair the component. 
Further, a licensee might determine that some maintenance activities 
are within the skill of the craft (such as replacing missing bolts on 
motor-operated valve switch compartments), such that detailed work 
orders would not be necessary. On the other hand, an activity to 
procure a replacement component with active functions that is not the 
same as the one being replaced would necessitate use of most of the 
specified processes, with a greater need for documentation and 
independent review to achieve the expected result.
    As part of the high level requirement that RISC-3 SSCs be capable 
of performing their safety-related functions under design basis 
conditions, the Commission emphasizes that implementation of the 
processes must provide reasonable confidence of the future capability 
of the SSC (i.e., not just confidence that the SSC works at a certain 
point in time but rather provides confidence that the component will 
work when called upon). The level of confidence can be less than was 
provided by the special treatment requirements listed in Sec.  
50.69(b)(1). As an example, exercising of a valve or simply starting a 
pump does not provide reasonable confidence in design basis capability, 
will not detect service-induced aging or degradation that could prevent 
the component from performing its design basis functions in the future, 
and is insufficient by itself to satisfy the intent of the rule.
    A licensee implementing Sec.  50.69 is responsible for implementing 
the treatment requirements for RISC-3 SSCs in an effective manner to 
maintain the capability to perform the safety functions under design 
basis conditions. A licensee should address the potential impact on the 
functionality of RISC-3 SSCs as a result of the changes to testing 
programs, such that the categorization process assumptions and results 
remain valid. To provide a basis to conclude that the potential 
increase in risk would be small, a licensee is required to conduct 
evaluations that assume failure rates that might occur as a result of 
the revisions to treatment. These evaluations would need to consider, 
for instance, any planned alteration in a licensee's program for 
diagnostic testing of motor-operated valves. If a likely result of a 
contemplated change in treatment is an increase in failure rate, 
outside the bounds of the evaluations, that change in treatment would 
not be acceptable under proposed Sec.  50.69 because the criterion in 
Sec.  50.69(c)(i)(iv) about providing reasonable confidence of a small 
increase in risk would not be met.
V.5.2.1 Section 50.69(d)(2)(i) Design Control Process
    Section 50.69(d)(2)(i) specifies that the functional requirements 
and bases for RISC-3 SSCs be maintained and controlled. The functional 
requirements

[[Page 26542]]

and bases continue to apply unless they are specifically changed in 
accordance with the appropriate regulatory change control process 
(e.g., Sec.  50.59). The rule further states that RISC-3 SSCs must be 
capable of performing their safety-related functions under design basis 
conditions including (any applicable) design requirements for 
environmental conditions (temperature, pressure, humidity, chemical 
effects, radiation, and submergence), effects (aging and synergisms ), 
and seismic conditions (design load combinations of normal and accident 
conditions with earthquake motions).
    It is recognized that the level of confidence in the design basis 
capability of RISC-3 SSCs may be less than the confidence provided in 
the capability of RISC-1 SSCs to perform their safety functions. The 
proposed treatment requirements for the control of the design of RISC-3 
SSCs are included, in part, to provide a basis for the assumption in 
the categorization process that these SSCs will continue to be capable 
of performing their safety-related functions under design basis 
conditions throughout their service life. The implementation of an 
effective design control process is crucial to the maintenance of the 
functionality of safety-related SSCs because many SSCs cannot be 
monitored or tested to demonstrate design basis capability or to 
identify potential degradation as part of normal plant operations. For 
instance, if the SSC were modified or replaced, the design control 
processes are important means by which the required capability is 
installed and maintained over the life of the component. Further, 
because it is not possible to test or monitor some SSCs under the 
conditions that they might experience in service, other means, such as 
control of design and procurement of SSCs, and condition monitoring, 
are used such that the SSCs are capable of performing their functions. 
The proposed rule would require that licensees have a design control 
process that maintains and applies design requirements to ensure that 
RISC-3 SSCs will be capable of performing their safety-related 
functions under design basis conditions. To meet this performance 
objective, the licensee's design control process would be expected to 
specify appropriate quality standards; select suitable materials, 
parts, and equipment; control design interfaces; coordinate 
participation of design organizations; verify design adequacy; and 
control design changes. The manner in which the design control 
requirements for RISC-3 SSCs are accomplished would be the 
responsibility of the licensees adopting Sec.  50.69. The proposed rule 
would allow flexibility for licensees to focus their resources on the 
SSCs that are most safety-significant while implementing an effective 
design control process for RISC-3 SSCs. For example, licensees might 
provide design control for RISC-3 SSCs through application of (1) the 
process established under Criterion III of 10 CFR Part 50, Appendix B; 
(2) an augmented quality assurance program such as might have been 
established in response to regulatory guidance issued in conjunction 
with Sec.  50.62 (for SSCs used to comply with anticipated transients 
without a plant scram; or (3) a plant-specific process currently in 
place or established to satisfy the treatment requirements of Sec.  
50.69.
    The design control process under Sec.  50.69 is intended to provide 
assurance that the proposed rule is satisfying the principle that the 
design requirements of RISC-3 SSCs would not be changed under Sec.  
50.69. For example, the design provisions of Section III of the ASME 
Boiler and Pressure Vessel Code (BPV Code) required by Sec.  50.55a(c), 
(d), and (e) for RISC-3 SSCs are not affected by the proposed rule. 
Another example is a requirement for fracture toughness of particular 
materials that is part of a licensee's design requirements; such a 
requirement would continue to apply when repair or replacement of 
affected components is undertaken. Licensees would continue to be 
required by Sec.  50.59 to evaluate proposed modifications to design 
requirements for safety-related SSCs, including those categorized as 
RISC-3.
    For RISC-3 SSCs, the proposed rule would remove the requirements 
for a program for environmental qualification of electric equipment 
specified in Sec.  50.49, ``Environmental Qualification of Electric 
Equipment Important to Safety for Nuclear Power Plants.'' However, the 
proposed rule would 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, Criterion 4 of 10 CFR part 50, Appendix A, ``General Design 
Criteria for Nuclear Power Plants,'' requires that SSCs important to 
safety be designed to accommodate the effects of and to be compatible 
with the environmental conditions associated with normal operation, 
maintenance, testing, and postulated accidents. In accordance with 
Sec.  50.69(d)(2), the licensee is required to design, procure, 
install, maintain, and monitor electric equipment important to safety 
such that they are capable of performing their intended functions under 
the environmental conditions listed in Sec.  50.69(d)(2)(i) throughout 
their service life. Further, if RISC-3 electrical equipment is relied 
on to perform its safety-related function beyond its design life, 
licensees should have a basis justifying the continued capability of 
the equipment under adverse environmental conditions.
    RISC-3 and RISC-4 SSCs would continue to be required to function 
under design basis seismic conditions, 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 who 
adopts the proposed 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 Earthquake. The proposed 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. These continue to be part of the design basis 
requirements or procurement requirement for replacement SSCs. The 
proposed rule would 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 proposed rule would permit 
licensees to select a technically defensible method to show that RISC-3 
SSCs will remain functional when subject to design earthquake loads. 
The level of confidence for the design basis capability of RISC-3 SSCs, 
including seismic capability, may be less than the confidence in the 
design basis capability of RISC-1 SSCs. The use of earthquake 
experience data has been mentioned as a potential method to demonstrate 
SSCs will remain functional during earthquakes. However, it would be 
difficult to rely on earthquake experience alone to demonstrate

[[Page 26543]]

functionality of SSCs if the design basis includes multiple earthquake 
events or combinations of loadings unless these specific conditions 
were enveloped by the experience data. 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. Qualification testing of an SSC would be 
necessary if no suitable alternative method is available for showing 
that the SSC will perform its design basis function during an 
earthquake.
    Licensees are responsible for proper installation and post-
installation testing of RISC-3 SSCs as part of design control and other 
treatment processes to provide reasonable confidence in the capability 
of SSCs to perform their functions. The Commission also expects 
licensees to control special processes associated with installation, 
such as welding, to provide reasonable confidence in the design basis 
capability of RISC-3 SSCs. Licensees would be expected to perform 
sufficient post-installation testing to verify that the installed SSC 
is operating within expected parameters and is capable of performing 
its safety functions under design basis conditions. In performing post-
installation testing, licensees may apply engineering analyses to 
extrapolate the test data to demonstrate design basis capability.
V.5.2.2 Section 50.69(d)(2)(ii) Procurement Process
    Section 50.69(d)(2)(ii) specifies that procured RISC-3 SSCs satisfy 
their design requirements. In order to obtain components that meet the 
requirements, the licensee would be expected to specify the technical 
requirements (including applicable design basis environmental and 
seismic conditions) for items to be procured. Further, the Commission 
expects licensees to use established methods (e.g., vendor 
documentation, equivalency evaluation, technical evaluation, technical 
analysis, or testing) to develop a technical basis for the 
determination that the procured item can perform its safety-related 
function under design basis conditions, including applicable design 
basis environmental conditions (temperature, pressure, humidity, 
chemical effects, radiation, and submergence), and effects (aging and 
synergisms), and seismic conditions (design load combinations of normal 
and accident conditions with earthquake motions). In addition to 
appropriately specifying in the procurement the desired component, the 
licensee/applicant would also be expected to conduct activities upon 
receipt to confirm that the received component is what was ordered.
    The proposed rule would allow more flexibility in the 
implementation of the procurement process for RISC-3 SSCs than 
currently provided by 10 CFR part 50, Appendix B. Nevertheless, 
licensees will continue to be responsible for implementing an effective 
procurement process for RISC-3 SSCs. Differences constituting a design 
change are expected to be documented and addressed under the licensee's 
design control process. As an example of one acceptable procurement 
process, a licensee might use an approach similar to that described 
below:
    Vendor Documentation--Vendor documentation could be used when the 
performance characteristics for the SSC, as specified in vendor 
documentation (e.g., catalog information, certificate of conformance), 
satisfy the SSC's design requirements. If the vendor documentation does 
not contain this level of detail, the design requirements could be 
provided in the procurement specifications. The vendor's acceptance of 
the stated design specifications provides sufficient confidence that 
the RISC-3 SSC would be capable of performing its safety-related 
functions under design basis conditions. Equivalency Evaluation--An 
equivalency evaluation could be used when it is sufficient to determine 
that the procured SSC is equivalent to the SSC being replaced (e.g., a 
like-for-like replacement).
    Engineering Evaluation--For minor differences, a technical 
evaluation could be performed to compare the differences between the 
procured SSC and the design requirements of the SSC being replaced and 
determines that differences in areas such as material, size, shape, 
stressors, aging mechanisms, and functional capabilities would not 
adversely affect the ability to perform the safety-related functions of 
the SSC under design basis conditions.
    Engineering Analysis--In cases involving substantial differences 
between the procured SSC and the design requirements of the SSC being 
replaced, a technical analysis could be conducted to determine that the 
procured SSC can perform its safety-related function under design basis 
conditions. The technical analysis would be based on one or more 
engineering methods that include, as necessary, calculations, analyses 
and evaluations by multiple disciplines, test data, or operating 
experience to support functionality of the SSC over its expected life.
    Testing--Testing under simulated design basis conditions could be 
performed on the SSC.
V.5.2.3 Section 50.69(d)(2)(iii) Maintenance, Inspection, Test, and 
Surveillance Process
    Section 50.69(d)(2)(iii) specifies that periodic maintenance, 
inspections, tests, and surveillance activities be established and 
conducted, and their results evaluated using prescribed acceptance 
criteria to determine that the RISC-3 SSCs will remain capable of 
performing their safety-related functions under design basis conditions 
until their next scheduled activity.
    To meet this requirement, licensees are expected to establish the 
scope, frequency, and detail of predictive, preventive, and corrective 
maintenance activities (including post-maintenance testing) to support 
the determination that RISC-3 SSCs will remain capable of performing 
their safety-related functions under design basis conditions throughout 
their service life. For a RISC-3 SSC in service beyond its design life, 
the Commission expects licensees to have a basis to determine that the 
SSC will remain capable of performing its safety-related function. 
Following maintenance activities that affect the capability of an SSC 
to perform its safety-related function, licensees would be expected to 
perform post-maintenance testing to verify that the SSC is performing 
within expected parameters and is capable of performing its safety 
function under design basis conditions. Licensees may apply engineering 
analyses to extrapolate the test data to demonstrate design basis 
capability as part of post-maintenance testing. The Commission expects 
licensees to identify the preventive maintenance needed to preserve the 
capability of RISC-3 SSCs to perform their safety-related functions 
under applicable design basis environmental and seismic conditions for 
their expected service life.
    To have reasonable confidence that SSCs can perform their 
functions, licensees must implement effective processes for inspection, 
testing, and surveillance of RISC-3 SSCs; they may apply their own 
individual approaches such that the requirements of Sec.  50.69 are 
satisfied. As an example, the provisions for risk-informed inspection 
and testing in applicable ASME Code Cases would constitute one 
effective approach in satisfying the Sec.  50.69 requirements. To 
prevent the occurrence of common-

[[Page 26544]]

cause problems that might invalidate the categorization process 
assumptions and results, effective implementation would include a 
determination of the functionality of safety-related SSCs checked using 
measuring and test equipment that was later found to be in error or 
defective.
    With respect to RISC-3 pumps and valves, the Commission expects 
licensees to implement periodic testing or inspection, and evaluation 
of performance data, sufficient to provide reasonable confidence that 
these pumps and valves will be capable of performing their safety 
function under design basis conditions. To determine that SSC will 
remain capable until the next scheduled activity, a licensee would have 
to obtain sufficient operational information or performance data to 
provide reasonable confidence that the RISC-3 pumps and valves will be 
capable of performing their safety function if called upon to function 
under operational or design basis conditions over the interval between 
periodic testing or inspections. A licensee may develop the type and 
frequency of the test or inspection for RISC-3 pumps and valves where 
sufficient to conclude that the pump or valve will perform its safety 
function. These tests or inspections may be less rigorous and less 
frequent than those performed on RISC-1 pumps and valves. For example, 
a licensee might establish more relaxed criteria for grouping of 
similar RISC-3 components, or might apply less stringent test 
acceptance criteria for RISC-3 pumps and valves, than specified for 
RISC-1 components. The licensee could apply staggered test intervals 
for the RISC-3 components to provide confidence that the relaxed 
grouping or acceptance criteria had not resulted in SSC performance 
that is inconsistent with the categorization process or its 
assumptions. Licensees should note that performance data obtained for 
pumps and valves operating under normal conditions may not be capable 
of predicting their capability to perform safety functions under design 
basis conditions without additional evaluation or analysis. This does 
not mean that pumps and valves must be tested or inspected under design 
basis conditions. Methods exist for collecting performance data at 
conditions different than design basis conditions that can be used to 
reach conclusions regarding the design basis capability of components. 
Examples of such methods are described in Regulatory Guide 1.175, An 
Approach for Plant-Specific, Risk-Informed Decision making: Inservice 
Testing, and applicable risk-informed ASME Code Cases (e.g., OMN-1, 
OMN-4, OMN-7, OMN-12) as accepted by 10 CFR 50.55a.
V.5.2.4 Section 50.69(d)(2)(iv) Corrective Action Process
    Section 50.69(d)(2)(iv) would specify that conditions that could 
prevent a RISC-3 SSC from performing its safety-related functions under 
design basis conditions be identified, documented, and corrected in a 
timely manner. A licensee may obtain information from the inspection, 
test and surveillance activities discussed above, or from other 
sources, such as operating experience, that indicates that an SSC is 
not capable of performing its required functions and thus identifies 
that corrective action is needed.
    In meeting proposed Sec.  50.69, licensees may implement a 
corrective action process for RISC-3 SSCs that is different than the 
process established to satisfy 10 CFR Part 50, Appendix B. This more 
general requirement would allow a graded approach, as well as less 
stringent timeliness requirements. The Commission believes an effective 
corrective action process is crucial to maintaining the capability of 
RISC-3 SSCs to perform their safety-related functions because of the 
reduction in requirements for other processes for design control; 
procurement; and maintenance, inspection, test, and surveillance. For 
example, 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.

V.6.0 Section 50.69(e) Feedback and Process Adjustment

    Section 50.69(e)(1) requires the updating of the PRA. The PRA 
configuration control program must incorporate a feedback process to 
update the PRA model. The program must require that plant data, design, 
and procedure changes that affect the PRA models or input parameters be 
incorporated into the model. This update is to account for plant-
specific operating experience as well as general industry experience. 
In particular, the proposed rule would require the licensee to review 
changes to the plant, operational practices, applicable industry 
operational experience, and, as appropriate, update the PRA and SSC 
categorization in a timely manner but no longer than every 36 months 
for RISC-1, RISC-2, RISC-3 and RISC-4 SSCs. Changes must be evaluated 
with respect to the impact on CDF and LERF. If the change would result 
in a significant increase in the CDF or LERF or might change the 
categorization of SSCs, the PRA must be updated in a timely manner; in 
this context it would clearly not be timely to wait to update the PRA 
if there would be a significant change in risk. Other changes are to be 
incorporated within 36 months. The results of the updated PRA and the 
associated risk categorizations based on the updated PRA information 
should be used as part of the feedback and corrective action process, 
and SSCs must be re-categorized as needed.
    Section 50.69(e)(2) and (e)(3) contains the requirements for 
feeding back into the categorization process SSC performance 
information and data, and for adjusting the categorization and 
treatment processes as appropriate, with the goal that the validity of 
the categorization process and its results are maintained. Further, the 
proposed rule would require the licensee to monitor the performance of 
RISC-1 and RISC-2 SSCs and make adjustments as necessary to either the 
categorization or treatment processes. To meet this requirement, the 
Commission expects licensees to monitor all functional failures (i.e., 
not just maintenance preventable unavailabilities and failures as is 
currently required by Sec.  50.65) so that they can determine when 
adjustments are needed. Licensee monitoring programs will also need to 
include the monitoring of SSCs that support beyond design basis 
functions (if applicable) that are not necessarily included in the 
scope of an existing maintenance rule monitoring program.
    If a licensee chooses to categorize a selective set of SSCs as 
RISC-3, and the categorization of SSCs as RISC-3 is based on credit 
taken for the performance of other plant SSCs (whether or not these 
SSCs are within the selective implementation set), then the licensee 
must maintain the credited performance. This applies to credit taken 
in: (1) PRA models, inputs and assumptions; (2) screening and margin 
analyses; and (3) IDP deliberations. This implies that the licensee 
must ensure that the credited SSCs perform their functions per Sec.  
50.69(d)(1), and the performance of these SSCs must be monitored per 
Sec.  50.69(e)(2).
    For RISC-3 SSCs, the proposed rule would require the licensee to 
consider the performance data required by Sec.  50.69(d)(2)(iii) 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 make 
adjustments as necessary to either

[[Page 26545]]

the categorization or treatment processes, to maintain categorization 
process results valid. Section 50.69(d)2)(iii) requires periodic 
maintenance, testing and surveillance activities for RISC-3 SSCs. Based 
upon review of this information, if SSC reliability degrades to the 
point that the evaluations done to show that the potential risk was 
small are no longer bounding, action is necessary to either adjust the 
treatment (to improve reliability) or to perform the categorization 
process again (to determine if any changes in categorization of SSC are 
necessary).

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, for handling planned changes to programs and 
processes and for records. Each subparagraph 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 is expected 
to address why a component was determined to be either safety-
significant or low safety-significant based upon the requirements in 
Sec.  50.69(c).
    The Commission is not, except in limited instances, specifying 
particular records to retain. Since the licensee is responsible for 
compliance with the requirements, subject to NRC oversight and 
inspection, the licensee (or applicant) would need to be able to show 
that they have established the processes required by the rules and 
conducted activities sufficient to provide reasonable confidence in 
functionality of SSCs under design basis conditions.
    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, and thus, may have revised treatment applied 
to the structures and components within that system. This provision is 
included to maintain clear information, at a minimum level of detail, 
about which requirements a licensee is satisfying; detailed information 
about particular SSCs is not required to be submitted. For an 
applicant, this updating would be expected to be either part of the 
original application or as a supplement to the FSAR under Sec.  50.34. 
For licensees, the updating must be in accordance with the provisions 
of Sec.  50.71(e) for licensees.
    Once the NRC has completed its review of a licensee's Sec.  50.69 
submittal as it relates to categorization, the licensee or applicant 
would be able to adjust its treatment processes provided that the rule 
requirements 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 would be made as a result of 
implementation of Sec.  50.69. Any revisions to these documents are to 
be submitted in accordance with the existing requirements of Sec.  
50.54(a)(2) and Sec.  50.71(e) respectively. For instance, Sec.  
50.71(e) states that the FSAR is to contain the latest information 
developed and is to reflect information submitted to the Commission 
since the last update. The regulations further state in the cited 
sections how a licensee is to submit to the NRC revisions to the QA 
plan or to the FSAR. Information in these documents that would no 
longer be accurate upon implementation of Sec.  50.69 must be updated. 
Details of the processes would be expected to be contained in plant 
procedures, procurement documents, surveillance records, etc.
    Section 50.69(f)(3) specifies that for initial implementation of 
the rule, changes to the FSAR for implementation of this proposed 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 for implementation of 
this proposed rule need not include a supporting Sec.  50.54(a) review 
of changes directly related to implementation. 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 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). Because the NRC is 
reviewing and approving a submittal containing the licensee or 
applicant's commitments for categorization, changes that would 
invalidate their submittal would also invalidate the approval. However, 
provided any revised process continues to conform with what was 
submitted or committed to (such as through a commitment to follow a 
particular RG), NRC review would not be needed of lower-tier changes 
(such as to implementing procedures) that might arise.
    No explicit requirements are included in Sec.  50.69 for the period 
for retention of records. The proposed rule would specify 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 would have prevented a RISC-1 or RISC-2 SSCs 
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 assumptions made 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 would no longer apply to RISC-3 (and RISC-4) SSCs under the 
proposed rule.

VI. Other Topics for Public Comment

VI.1.0 Additional Potential Requirements for Public Comment

    The cornerstone of proposed Sec.  50.69 is a robust, risk-informed 
categorization process that provides high confidence that the safety 
significance of SSCs is correctly determined considering all relevant 
information. The categorization requirements incorporated into the 
proposed rule achieve this objective. The Commission proposes to remove

[[Page 26546]]

the RISC-3 and RISC-4 SSCs from the scope of special treatment 
requirements delineated in Sec.  50.69(b)(1), and instead require the 
licensee to comply with more general, high level requirements for 
maintaining functionality. The proposed rule would allow appropriate 
flexibility for implementation while continuing to provide reasonable 
confidence that the SSCs will remain functional. As discussed elsewhere 
in this notice, the Commission concludes that the requirements in 
proposed Sec.  50.69 would maintain adequate protection of public 
health and safety. Previous drafts of this proposed rule posted to the 
NRC web site, contained more detailed requirements in Sec.  50.69(d)(2) 
for RISC-3 SSCs. The Commission believes that this level of detail is 
beyond what is necessary to provide reasonable confidence in RISC-3 
design basis capability in light of the robust categorization 
requirements incorporated into proposed Sec.  50.69. However, the 
Commission recognizes that some stakeholders may disagree and invites 
public comment on this matter. To facilitate public comment, example 
language is provided below that identifies (in quotations and brackets) 
those requirements that were considered for inclusion in Sec.  50.69 
(as well as where they would have appeared in the rule).
    (2) RISC-3 SSCs. The licensee or applicant shall develop and 
implement processes to control the design; procurement; inspection, 
maintenance, testing, and surveillance; and corrective action for RISC-
3 SSCs to provide reasonable confidence in the capability of RISC-3 
SSCs to perform their safety-related functions under design basis 
conditions throughout their service life. [``These processes must meet 
voluntary consensus standards which are generally accepted in 
industrial practice, and address applicable vendor recommendations and 
operational experience. The implementation of these processes and the 
assessment of their effectiveness must be controlled and accomplished 
through documented procedures and guidelines. The treatment processes 
must be consistent with the assumptions credited in the categorization 
process.''] The processes must meet the following requirements, as 
applicable: (i)Design Control. Design functional requirements and bases 
for RISC-3 SSCs must be maintained and controlled, [``including 
selection of suitable materials, methods, and standards; verification 
of design adequacy; control of installation and post-installation 
testing; and control of design changes'']. RISC-3 SSCs must be [``have 
a documented basis to demonstrate that they are''] capable of 
performing their safety-related functions including design requirements 
for environmental conditions (i.e., temperature and pressure, humidity, 
chemical effects, radiation, and submergence) and effects (i.e., aging 
and synergism); and seismic conditions (design load combinations of 
normal and accident conditions with earthquake motions). 
[``Replacements for ASME Class 2 and Class 3 SSCs or parts must meet 
either: (1) The requirements of the ASME Boiler & Pressure Vessel (BPV) 
Code; or (2) the technical and administrative requirements, in their 
entirety, of a voluntary consensus standard that is generally accepted 
in industrial practice applicable to replacement. ASME Class 2 and 
Class 3 SSCs and parts shall meet the fracture toughness requirements 
of the SSC or part being replaced.'']
    (ii) Procurement. Procured RISC-3 SSCs must satisfy their design 
requirements. [``Upon receipt, the licensee shall verify that the item 
received is the item that was ordered.'']
    (iii) Maintenance, Inspection, Testing, and Surveillance. Periodic 
maintenance, inspection, testing, and surveillance activities must be 
established and conducted using prescribed acceptance criteria, and 
their results evaluated to determine that RISC-3 SSCs will remain 
capable of performing their safety-related functions under design basis 
conditions until the next scheduled activity.
    (iv) Corrective Action. Conditions that could prevent a RISC-3 SSC 
from performing its safety-related functions under design basis 
conditions must be identified, documented, and corrected in a timely 
manner. [``In the case of significant conditions adverse to quality, 
measures shall assure that the cause of the condition is determined and 
corrective action taken to preclude repetition.''] The Commission is 
requesting comment as to whether any of these requirements (or other 
requirements) are necessary to provide reasonable confidence of SSC 
functionality commensurate with the safety significance of the RISC-3 
SSC, i.e., whether the requirements on categorization are sufficiently 
robust that the level of detail contained in the proposed rule on 
treatment is appropriate.

VI.2.0 Questions for Public Input

    In addition to seeking comment on the proposed rule and its 
supporting documents, the Commission is also specifically seeking 
public comment on the following questions. Comments should be submitted 
as noted in the ADDRESSES section of this notice.
VI.2.1 PRA Requirements
    The proposed rule requires as a minimum, a PRA that includes 
internal events, at power, which has been subjected to a peer review 
process. The PRA (for that scope) must be capable of determining both 
CDF and LERF (i.e., provide level 2-type results). Proposed Sec.  50.69 
allows licensees to use non-PRA methods to address other modes and 
hazards in the categorization process (see in particular NEI 00-04 and 
DG-1121). The proposed rule requires the licensee to submit information 
about its PRA and these other methods, including information about the 
quality and level of detail about all of the methods to be used.
    The Commission is seeking comment as to 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. Thus, instead of employing 
other methods to account for the contribution from modes and events not 
modeled in the PRA, this more comprehensive PRA would allow for 
quantification of the contribution from these scenarios. This approach 
would involve substantive changes in the implementing guidance as well. 
The Commission is interested in both the benefits of this approach as 
well as any implications for this specific application of risk 
insights. The Commission is also seeking comment on whether a different 
set of PRA requirements, from either of the alternatives described 
above, should be required for this application.
VI.2.2 Review and Approval of Treatment for RISC-3 SSCs
    In the proposed rule, the Commission is proposing to review and 
approve the categorization process to be used by the licensee. For 
treatment requirements, the proposed rule sets forth high-level 
requirements, and does not require NRC review and approval of specific 
processes a licensee would implement to meet these requirements. 
Another way to structure the rule would be to require NRC review and 
approval of the licensee's proposed treatment program for RISC-3 SSCs. 
The Commission is interested in any benefits of this approach as well 
as any implications for this rulemaking and its associated guidance.
VI.2.3 Inspection and Enforcement
    As discussed above, the Commission recognizes that the final rule 
may have implications with respect to NRC's

[[Page 26547]]

reactor oversight process including the inspection program, and 
enforcement. In its final decision on this rulemaking, the Commission 
proposes to document its conclusions as to whether or not new or 
revised inspection or enforcement guidance is necessary. Public comment 
is requested on whether or not changes are needed in our inspection and 
enforcement programs to enable NRC to exercise the appropriate degree 
of regulatory oversight of these aspects of the facility operation.
VI.2.4 Operating Experience
    One of the areas of uncertainty associated with this rulemaking has 
been the potential effects of changes in treatment on SSC reliability 
and common-cause failure potential. This is reflected in the 
requirement for evaluations (sensitivity studies) to provide reasonable 
confidence that any potential increase in risk would be small, with a 
basis provided for the factors to be assumed in these evaluations. 
Further, the rule requires the licensee to consider performance 
information to determine whether there are any adverse changes such 
that SSC unreliability values approach the values used in these 
evaluations, and to make necessary adjustments to the categorization 
and treatment processes. As discussed in Section VII.2, below, draft RG 
(DG-1121) provides some discussion about techniques that might be used 
in determining the factors for these evaluations. The Commission is 
interested in the role that relevant operational experience could play 
in reducing the uncertainty associated with the effects of treatment on 
performance and specifically seeks public comment as to what 
information might be available and how it could be used to support 
implementation of this rulemaking.

VII. Guidance

VII.1 Regulatory Guide and Implementation Guidance for Sec.  50.69
    The Nuclear Energy Institute (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 proposed rule, the NRC staff reviewed drafts of this 
document and in addition, NEI 00-04 was used in the pilot program 
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 
version of NEI 00-04, dated June 28, 2002, forms the basis for the 
draft regulatory guide.
    The NRC staff's review of NEI 00-04 resulted in several areas where 
the staff would find it necessary to identify exceptions to NEI 
guidance or to include further guidance to supplement the document, as 
it is currently written. These areas are discussed in an attachment to 
the draft regulatory guide, DG-1121, ``Guidelines for Categorizing 
Structures, Systems and Components in Nuclear Power Plants According to 
Their Safety Significance.'' Through this document, the Commission is 
also seeking public comment on the DG and the identified issues. 
Comments should be submitted as discussed under the ADDRESSES section. 
Availability of this document is noted in Section X.
VII.2 Review Guidance Concerning PRA Quality and Peer Review
    The NRC has prepared a draft regulatory guide DG-1122, ``An 
Approach for Determining the Technical Adequacy of Probabilistic Risk 
Assessment Results for Risk-Informed Activities.'' This guide provides 
guidance on the NRC position on voluntary consensus standards for PRA 
(in particular on the ASME standard for internal events PRAs) and 
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 whether a PRA providing 
results being used in a decision is technically adequate.
    In a letter dated April 24, 2000, NEI requested 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 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. In order to 
reach the conclusion that the PRA results support the proposed 
categorization, the review guidance is structured to lead the staff 
reviewer to either 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.

VIII. Criminal Penalties

    For the purposes of Section 223 of the Atomic Energy Act, as 
amended, the Commission is issuing the proposed rule to add Sec.  50.69 
under one or more of sections 161b, 161i, or 161o of the AEA. Willful 
violations of the rule would be subject to criminal enforcement. 
Criminal penalties, as they apply to regulations in Part 50 are 
discussed in Sec.  50.111.

IX. 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.

X. 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 Website (Web). The NRC's interactive rulemaking Website 
is located at http://ruleforum.llnl.gov. These documents may be viewed 
and downloaded electronically via this Website.
    NRC's Public Electronic Reading Room (PERR). The NRC's public 
electronic reading room is located at www.nrc.gov/reading-rm.html.

[[Page 26548]]



----------------------------------------------------------------------------------------------------------------
                   Document                         PDR           Web                       PERR
----------------------------------------------------------------------------------------------------------------
Comments on the ANPR.........................            X             X   Available.
Comments on the draft rule language..........            X             X   Available.
ANPR Comment Resolution......................            X             X   ML022630030.
Environmental Assessment.....................            X             X   ML022630050.
Regulatory Analysis..........................            X             X   ML022630028.
OMB Supporting Statement.....................            X             X   ML031000685.
Industry Implementation Guidance.............            X             X   ML021910534.
Draft Regulatory Guide.......................            X             X   ML022630041.
----------------------------------------------------------------------------------------------------------------

XI. Plain Language

    The Presidential memorandum dated June 1, 1998, entitled ``Plain 
Language in Government Writing'' directed that the Government's writing 
be in plain language. This memorandum was published on June 10, 1998 
(63 FR 31883). The NRC requests comments on the proposed rule 
specifically with respect to the clarity and reflectiveness of the 
language used. Comments should be sent to the address listed under the 
ADDRESSES caption of the preamble.

XII. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 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 proposed rule, the 
NRC proposes to use the following Government-unique standard (Draft NRC 
Regulatory Guide DG-1121, August 2002). 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. DG-1121 and DG-1122 (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. DG-1121 explicitly notes such 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 would allow 
use of these voluntary consensus standards, but would not require 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 recategorized. PRA requirements for 
all initiating events and modes of operation, nor other parts of the 
approach laid out such as determining the basis for the evaluations to 
show a small increase in risk. The NRC is not aware of any voluntary 
consensus standard that could be used instead of the proposed 
Government-unique standards. The NRC will consider using a voluntary 
consensus standard if an appropriate standard is identified. If a 
voluntary consensus standard is identified for consideration, the 
submittal should explain how the voluntary consensus standard is 
comparable and why it should be used instead of the proposed standard.

XIII. 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, if adopted, would not be a 
major Federal action significantly affecting the quality of the human 
environment and, therefore, an environmental impact statement is not 
required.
    The determination of this environmental assessment is that there 
will be no significant offsite impact to the public from this action. 
However, the general public should note that the NRC is seeking public 
participation; availability of the environmental assessment is provided 
in Section X. Comments on any aspect of the environmental assessment 
may be submitted to the NRC as indicated under the ADDRESSES heading.
    The NRC has sent a copy of the environmental assessment and this 
proposed rule to every State Liaison Officer and requested their 
comments on the environmental assessment.

XIV. Paperwork Reduction Act Statement

    This proposed rule contains information collection requirements 
that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 
3501, et seq.). This rule has been submitted to the Office of 
Management and Budget for review and approval of the information 
collection requirements.
    The burden to the public for these information collections is 
estimated to average 1032 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. The U.S. Nuclear Regulatory Commission is 
seeking public comment on the potential impact of the information 
collections contained in the proposed rule and on the following issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC, including whether the 
information will have practical utility?
    2. Is the estimate of burden accurate?
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be submitted?
    4. How can the burden of the information collection be minimized, 
including the use of automated collection techniques?
    Send comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden, to the 
Records Management Branch (T-6 E6), U. S. Nuclear Regulatory 
Commission, Washington DC 20555-0001, or by Internet electronic mail to 
[email protected]; and to the Desk Officer, Office of Information 
and Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management 
and Budget, Washington DC 20503.
    Comments to OMB on the information collections or on the above 
issues should be submitted by June 16, 2003. Comments received after 
this date will be considered if it is practical to do so, but assurance 
of consideration cannot be given to comments received after this date.

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

[[Page 26549]]

information collection requirement unless the requesting document 
displays a currently valid OMB control number.

XV. Regulatory Analysis

    The Commission has prepared a draft regulatory analysis on this 
proposed regulation. The analysis examines the costs and benefits of 
the alternatives considered by the Commission. The Commission requests 
public comment on the draft regulatory analysis. Availability of the 
regulatory analysis is provided in Section X. Comments on the draft 
analysis may be submitted to the NRC as indicated under the ADDRESSES 
heading.

XVI. Regulatory Flexibility Certification

    In accordance with the Regulatory Flexibility Act (5 U.S.C. 
605(b)), the Commission certifies that this rule will not, if 
promulgated, have a significant economic impact on a substantial number 
of small entities. This proposed 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).

XVII. Backfit Analysis

    The NRC has determined that the backfit rule does not apply to this 
proposed rule; therefore, a backfit analysis is not required for this 
proposed rule. As a voluntary alternative to existing requirements, 
these amendments do not impose more stringent safety requirements on 10 
CFR Part 50 licensees or applicants and thus do not constitute a 
backfit pursuant to Sec.  50.109.

List of Subjects in 10 CFR Part 50

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

    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. 553, the NRC is proposing to 
adopt the following amendments to 10 CFR Part 50.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    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, 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, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat.1242, as 
amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
    Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951, as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123 (42 
U.S.C. 5851). Sections 50.10 also issued under secs. 101, 185, 68 
Stat. 936, 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 Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C. 2239). Sections 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).

    2. Section 50.8(b) is revised to read as follows:


Sec.  50.8  Information collection requirements: OMB approval.

    (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, 50.71, 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.
    3. Part 50 is amended by adding a new Sec.  50.69 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 Sec. Sec.  50.21(b) or 50.22, a holder of a 
renewed LWR license under Part 54 of this chapter; a person seeking a 
design certification under Part 52 of this chapter, or an applicant for 
a LWR license under Sec.  50.22 or under Part 52, 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) 10 CFR 50.49.
    (iii) 10 CFR 50.55(e).
    (iv) The inservice testing requirements in 10 CFR 50.55a(f); the 
inservice inspection, and repair and replacement, 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).
    (v) 10 CFR 50.65, except for paragraph (a)(4).
    (vi) 10 CFR 50.72.
    (vii) 10 CFR 50.73.
    (viii) Appendix B to 10 CFR part 50.
    (ix) 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 that 
is not connected to the reactor coolant pressure boundary; or
    (4) The valve is 1-inch nominal size or less.
    (x) 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

[[Page 26550]]

Shutdown Earthquake and Operating Basis Earthquake.
    (2) A licensee voluntarily choosing to implement this section shall 
submit an application for license amendment pursuant to 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 shall 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 for a license voluntarily choosing to implement 
this section shall include the information in Sec.  50.69(b)(2) as part 
of application for a license. 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 whether 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 the defense-in-depth philosophy.
    (iv) Include evaluations that provide reasonable confidence that 
for SSCs categorized as RISC-3, sufficient safety 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.  50.69(b)(1) and Sec.  
50.69(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 develop and 
implement processes to control the design; procurement; inspection, 
maintenance, testing, and surveillance; and corrective action for RISC-
3 SSCs to provide reasonable confidence in the capability of RISC-3 
SSCs to perform their safety-related functions under design basis 
conditions throughout their service life. The processes must meet the 
following requirements, as applicable:
    (i) Design control. Design functional requirements and bases for 
RISC-3 SSCs must be maintained and controlled. RISC-3 SSCs must be 
capable of performing their safety-related functions including design 
requirements for environmental conditions (i.e., temperature and 
pressure, humidity, chemical effects, radiation and submergence) and 
effects (i.e., aging and synergism); and seismic conditions (design 
load combinations of normal and accident conditions with earthquake 
motions);
    (ii) Procurement. Procured RISC-3 SSCs must satisfy their design 
requirements;
    (iii) Maintenance, Inspection, Testing, and Surveillance. Periodic 
maintenance, inspection, testing, and surveillance activities must be 
established and conducted using prescribed acceptance criteria, and 
their results evaluated to determine that RISC-3 SSCs will remain 
capable of performing their safety-related functions under design basis 
conditions until the next scheduled activity; and
    (iv) Corrective Action. Conditions that could prevent a RISC-3 SSC 
from performing its safety-related functions under design basis 
conditions must be identified, documented, and corrected in a timely 
manner.
    (e) Feedback and process adjustment. (1) RISC-1, RISC-2, RISC-3 and 
RISC-4 SSCs. In a timely manner but no longer than every 36 months, the 
licensee shall review changes to the plant, operational practices, 
applicable industry operational experience, and, as appropriate, update 
the PRA and SSC categorization.
    (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)(iii) 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 satisfy Sec.  50.69 (c)(1)(iv). The licensee shall make adjustments 
as necessary to either 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

[[Page 26551]]

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 would have 
prevented RISC-1 or RISC-2 SSCs from performing a safety-significant 
function.

    Dated at Rockville, Maryland this 6th day of May, 2003.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 03-11696 Filed 5-15-03; 8:45 am]
BILLING CODE 7590-01-U