1Commission papers cited in this trial regulatory guide are available through the NRC’s public Web site
at v/reading-rm/doc-collections/commission/secys/, and the related Federal Register  notices
are available through the Federal Register Web site sponsored by the Government Printing Office (GPO)
at v/fr/index.html .
Th e U.S. N ucle ar R egu latory C om m ission  (NR C) iss ues  regu latory gu ides  to describe and make available to the public methods that the NRC staff considers acceptable for use in implementing specific parts of the agency’s regulations, techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the staff need in reviewing applications for permits and licenses.  Regulatory guides are not substitutes for regulations, and compliance with them is not required.  Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Comm ission.
Th is guide  was  issue d after c ons idera tion of c om m ents  rece ived  from  the pu blic.  The NR C sta
ff enc oura ges  and  welc om es c om m ents  and  sug ges tions in  con nec tion w ith im prov em ents  to published regulatory guides, as well as items for inclusion in regulatory guides that are currently being developed.  The NR C staff will revise existing guides,as appropriate, to accommodate comm ents and to reflect new information or experience.  W ritten comments may be submitted to the Rules and Directives Branch,Office of Administration, U.S. Nuclear Regulatory Commission, W ashington, DC 20555-0001.
Re gula tory guides are issued in 10 broad divisions:  1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities; 4, E nv iron m en tal an d S iting ; 5, M ate rials an d P lan t Pro tec tion ; 6, P rod uc ts; 7, T ran sp orta tion ; 8, O ccu pa tion al H ea lth; 9, An titrus t and  Fina nc ial R ev iew ; and  10, G en era l.
Re que sts for single copies of draft or active regulatory guides (which may be reproduced) should be made to the U.S. Nuclear Regulatory Commission, W ashington, DC 20555,Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to v .  Ele ctro nic c op ies o f this g uid e an d oth er re ce ntly issued guid es  are  ava ilable through the NRC ’s public W eb site under the Regulatory Guides document collection of the NRC ’
s Electronic Reading Room ’
U.S. NUCLEAR REGULATORY COMMISSION May 2006
Revision 1
REGULATORY GUIDE
OFFICE OF NUCLEAR REGULATORY RESEARCH
REGULATORY GUIDE 1.201
(For Trial Use)
(Previously issued for trial use, January 2006)
GUIDELINES FOR CATEGORIZING
STRUCTURES, SYSTEMS, AND COMPONENTS
IN NUCLEAR POWER PLANTS
ACCORDING TO THEIR SAFETY SIGNIFICANCE
A.  INTRODUCTION
The U.S. Nuclear Regulatory Commission (NRC) has promulgated regulations to permit power reactor
licensees and license applicants to implement an alternative regulatory framework with respect to “special treatment ”where special treatment refers to those requirements that provide increased assurance beyond normal industrial
practices that structures, systems, and components (SSCs) perform their design-basis functions.  Under this framework,licensees using a risk-informed process for categorizing SSCs according to their safety significance can remove
SSCs of low safety significance from the scope of certain identified special treatment requirements.
The genesis of this framework stems from 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, 1998.1
In that Commission paper, the NRC staff recommended developing risk-informed approaches to the application
of special treatment requirements to reduce unnecessary regulatory burden related to SSCs of low safety significance
by removing such SSCs from the scope of special treatment requirements.  The Commission subsequently approved the NRC staff’s rulemaking plan and issuance of an Advanced Notice of Proposed Rulemaking (ANPR) as outlined in SECY-99-256, “Rulemaking Plan for Risk-Informing Special Treatment Requirements,” dated October 29, 1999.
The Commission published the ANPR in the Federal Register (65 FR 11488) on March 3, 2000, and subsequently published a proposed rule for public comment (68 FR 26511) on May 16, 2003. Then, on November 22, 2004, the Commission adopted a new section, referred to as §50.69, within
Title 10, Part 50, of the Code of Federal Regulations, on risk-informed categorization and treatment
of SSCs for nuclear power plants (69 FR 68008).
This trial regulatory guide describes a method that the NRC staff considers acceptable for use
in complying with the Commission’s requirements in §50.69 with respect to the categorization of SSCs that are considered in risk-informing special treatment requirements.  This categorization method uses the process that the Nuclear Energy Institute (NEI) described in Revision 0 of its guidance document NEI 00-04, “10 CFR 50.69 SSC Categorization Guideline,” dated July 2005.2  Specifically, this process determines the safety significance of SSCs and categorizes them into one of four risk-informed
safety class (RISC) categories.
The NRC issued a draft of this guide, Draft Regulatory Guide DG-1121, for public revieweditor evaluating revision
and comment as part of the §50.69 rulemaking package in May 2003.  The staff subsequently received and addressed public comments in developing the previous revision of this guide, which the agency published in January 2006, and has since incorporated additional stakeholder comments in preparing
the current revision.  However, since this is a new regulatory approach to categorizing SSCs, and to ensure that the final guidance adequately addresses lessons learned from the initial applications, the NRC decided to issue this guide for trial use.  Therefore, this trial regulatory guide does not establish any final staff positions for purposes of the Backfit Rule, 10 CFR 50.109, and may continue to be revised
in response to experience with its use.  As such, any changes to this trial guide prior to staff adoption
in final form will not be considered to be backfits as defined in 10 CFR 50.109(a)(1).  This will ensure that the final regulatory guide adequately addresses lessons learned from regulatory review of pilot
and follow-on applications, and that the guidance is sufficient to enhance regulatory stability in the review, approval, and implementation of probabilistic risk assessments (PRAs) and their results in the risk-informed categorization process required by §50.69.
The NRC issues regulatory guides to describe to the public methods that the staff considers acceptable for use in implementing specific parts of the agency’s regulations, to explain techniques
that the staff uses in evaluating specific problems or postulated accidents, and to provide guidance
to applicants.  Regulatory guides are not substitutes for regulations, and compliance with regulatory guides is not required.
This regulatory guide contains information collections that are covered by the requirements
of 10 CFR Part 50 which the Office of Management and Budget (OMB) approved under OMB control n
umber 3150-0011.  The NRC may neither conduct nor sponsor,and a person is not required to respond to
an information collection request or requirement unless the requesting document displays a currently valid OMB control number.
2NEI 00-04, “10 CFR 50.69 SSC Categorization Guideline,” is available through the NRC’s Agencywide Documents Access and Management System (ADAMS), v/reading-rm/adams/web-based.html, under
Accession #ML052910035.
3NEI 00-04 uses the term “high-safety-significant (HSS)” to refer to SSCs that perform safety-significant functions.
The NRC understands HSS to have the same meaning as “safety-significant” (i.e., SSCs that are categorized as RISC-1or RISC-2), as used in §50.69.
B.  DISCUSSION
This trial regulatory guide provides interim guidance for complying with the NRC’s requirements in §50.69, by using the process described in Revision 0 of NEI 00-04 to determine the safety significance of SSCs and place them into the appropriate RISC categories.  The safety significance of SSCs is determined using an integrated decision-making process, which incorporates both risk and traditional e
ngineering insights.  The safety functions of SSCs include both the design-basis functions (derived from the safety-related definition) and functions credited for preventing and/or mitigating severe accidents.  Treatment requirements are then commensurately applied for the categorized SSCs to maintain their functionality.
Figure 1 provides a conceptual understanding of the new risk-informed SSC categorization scheme. The figure depicts the current safety-related versus nonsafety-related SSC categorization scheme with an overlay of the new safety-significance categorization.  In the traditional deterministic approach,
SSCs were generally categorized as either “safety-related” (as defined in 10 CFR 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 3 or low-safety-significant (LSS)(as shown by the horizontal line in the figure).  This results in SSCs being grouped into one of four categories, as represented by the four boxes in Figure 1.
Figure 1.  §50.69 RISC Categories
RISC-1 SSCs are safety-related SSCs that the risk-informed categorization process determines to be significant contributors to plant safety.  Licensees must continue to ensure that RISC-1 SSCs perform t
heir safety-significant functions consistent with the categorization process, including those safety-significant functions that go beyond the functions defined as safety-related for which credit is taken
in the categorization process.
RISC-2 SSCs are those that are defined as nonsafety-related, although the risk-informed categorization process determines that they are significant contributors to plant safety on an individual basis. The NRC staff recognizes that some RISC-2 SSCs may not have existing special treatment requirements. As a result, the focus for RISC-2 SSCs is on the safety-significant functions for which credit is taken
in the categorization process.
RISC-3 SSCs are those that are defined as safety-related, although the risk-informed categorization process determines that they are not significant contributors to plant safety.  Special treatment requirements are removed for RISC-3 SSCs and replaced with high-level requirements.  These high-level requirements are intended to provide sufficient regulatory treatment, such that these SSCs are still expected to perform their safety-related functions under design-basis conditions, albeit at a reduced level of assurance compared to the current special treatment requirements.  However, §50.69
does not allow these RISC-3 SSCs to lose their functional capability or be removed from the facility.
Finally, RISC-4 SSCs are those that are defined as nonsafety-related, and that the risk-informed categorization process determines are not significant contributors to plant safety.  Section 50.69 does not impose alternative treatment requirements for these RISC-4 SSCs.  However, as with the RISC-3 SSCs, changes to the design bases of RISC-4 SSCs must be made in accordance with current applicable design change control requirements (if any), such as those set forth in 10 CFR 50.59.
The NRC staff believes that the guidance in NEI 00-04 provides an acceptable approach for use in categorizing SSCs to support the implementation of §50.69.  Section C of this trial regulatory guide provides the NRC staff’s regulatory positions on NEI 00-04.
4The Commission’s Final Policy Statement on Use of PRA Methods in Nuclear Regulatory Activities, SP-95-146,
announced in the Federal Register (60 FR 42622) on August 16, 1995, is available through the NRC’s public Web site at v/reading-rm/doc-collections/commission/policy/60fr42622.pdf .
5
Copies of NUMARC 91-06, “Guidelines for Industry Actions to Assess Shutdown Management,” dated December 1991,may be obtained from the Nuclear Energy Institute, Attention:  Ms. Tonya Cameron, 1776 I Street, NW, Suite 400,Washington, DC  20006-3708 (phone:  202-739-8148).C.  REGULATORY POSITION
This trial regulatory guide provides interim guidance for trial use of the process and criteria
for determining the safety significance of SSCs using the categorization process described in Revision 0of NEI 00-04, “10 CFR 50.69 SSC Categorization Guideline,” dated July 2005.
1.Other Documents Referenced in Revision 0 of NEI 00-04
Revision 0 of NEI 00-04 references numerous other documents, but the NRC’s endorsement
of Revision 0 of NEI 00-04 does not constitute an endorsement of those other referenced documents.
2.Use of Examples in Revision 0 of NEI 00-04
Revision 0 of NEI 00-04 includes examples to supplement the guidance.  However, the NRC’s endorsement of Revision 0 of NEI 00-04 does not constitute a determination that the examples are ap
plicable for all licensees.  A licensee or applicant must ensure that a given example is applicable to its particular circumstances before implementing the guidance as described in that example.
3.Use of Methods Other Than Revision 0 of NEI 00-04
To meet the requirements of §50.69 for categorization of SSCs, licensees may use methods
other than those set forth in Revision 0 of NEI 00-04.  The NRC staff will determine the acceptability of such other methods by evaluating them against the requirements of §50.69.
4.Limitations of Types of Analyses Used in Implementing Revision 0 of NEI 00-04
In its Final Policy Statement on Use of PRA Methods in Nuclear Regulatory Activities , SP-95-146,dated August 16, 1995, the Commission determined that the use of PRA technology should be increased in all
regulatory matters, to the extent supported by state-of-the-art PRA methods and data.4  Implementation
of risk-informed regulation is possible because the development and use of a quantitative PRA require
s a systematic and integrated evaluation.  Development of a technically defensible quantitative PRA also requires sufficient and structured documentation to allow investigations of all aspects of the evaluation. To meet the requirements of §50.69 for categorization of SSCs, licensees must use risk evaluations and insights that cover the full spectrum of potential events (i.e., internal and external initiating events)and the range of plant operating modes (i.e., full-power, low-power, and shutdown operations).
Revision 0 of NEI 00-04 allows the use of non-PRA-type evaluations (e.g., fire-induced
vulnerability evaluation (FIVE), seismic margins analysis (SMA), and NEI guidance in NUMARC 91-06,“Guidelines for Industry Actions to Assess Shutdown Management,”5 to address shutdown operations),when PRAs have not been performed.  Such non-PRA-type evaluations will result in more conservative categorization, in that special treatment requirements will not be allowed to be relaxed for SSCs that are relied upon in such evaluations.  The degree of relief that the NRC will accept under §50.69 (i.e., SSCs subject to relaxation of special treatment requirements) will be commensurate with the assurance provided by the evaluation.

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