68008 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 50 

RIN 3150–AG42 

Risk-Informed Categorization and 

Treatment of Structures, Systems and 

Components for Nuclear Power 

Reactors

AGENCY: Nuclear Regulatory 

Commission.

ACTION: Final rule.

SUMMARY: The Nuclear Regulatory 

Commission (NRC) is amending its 

regulations to provide an alternative 

approach for establishing the 

requirements for treatment of structures, 

systems and components (SSCs) for 

nuclear power reactors using a riskinformed method of categorizing SSCs 

according to their safety significance. 

The amendment revises requirements 

with respect to ‘‘special treatment,’’ that 

is, those requirements that provide 

increased assurance (beyond normal 

industrial practices) that SSCs perform 

their design basis functions. This 

amendment permits licensees (and 

applicants for licenses) to remove SSCs 

of low safety significance from the scope 

of certain identified special treatment 

requirements and revise requirements 

for SSCs of greater safety significance. In 

addition to the rulemaking and its 

associated analyses, the Commission is 

also issuing a regulatory guide (RG) to 

implement the rule.

EFFECTIVE DATE: December 22, 2004.

ADDRESSES: The final rule and related 

documents are available on NRC’s 

rulemaking Web site at http://

ruleforum.llnl.gov. For information 

about the interactive rulemaking Web 

site contact Ms. Carol Gallagher, (301) 

415–5905 (e-mail: CAG@nrc.gov).

FOR FURTHER INFORMATION CONTACT: Mr. 

Timothy Reed, Office of Nuclear Reactor 

Regulation, U.S. Nuclear Regulatory 

Commission, Washington DC 20555–

0001; telephone (301) 415–1462; e-mail: 

tar@nrc.gov.

SUPPLEMENTARY INFORMATION:

Table of Contents 

I. Background 

II. Comments on Proposed Rule 

III. Final Rule 

IV. Pilot Activities 

V. Section by Section Analysis 

VI. Guidance 

VII. Criminal Penalties 

VIII. Compatibility of Agreement State 

Regulations 

IX. Availability of Documents 

X. Voluntary Consensus Standards (Public 

Law 104–113) 

XI. Finding of No Significant Environmental 

Impact 

XII. Paperwork Reduction Act Statement 

XIII. Regulatory Analysis 

XIV. Regulatory Flexibility Act Certification 

XV. Backfit Analysis 

XVI. Small Business Regulatory Enforcement 

Fairness Act

I. Background 

I.1 History and General Background 

The NRC has established a set of 

regulatory requirements for commercial 

nuclear reactors to ensure that a reactor 

facility does not impose an undue risk 

to the health and safety of the public, 

thereby providing reasonable assurance 

of adequate protection to public health 

and safety. The current body of NRC 

regulations and their implementation 

are largely based on a ‘‘deterministic’’ 

approach. 

This deterministic approach 

establishes requirements for engineering 

margin and quality assurance in design, 

manufacture, and construction. In 

addition, it assumes that adverse 

conditions can exist (e.g., equipment 

failures and human errors) and 

establishes a specific set of design basis 

events (DBEs). The deterministic 

approach contains implied elements of 

probability (qualitative risk 

considerations), from the selection of 

accidents to be analyzed (e.g., reactor 

vessel rupture is considered too 

improbable to be included) to the 

system level requirements for 

emergency core cooling (e.g., safety train 

redundancy and protection against 

single failure). The deterministic 

approach then requires that the licensed 

facility include safety systems capable 

of preventing and/or mitigating the 

consequences of those DBEs to protect 

public health and safety. Those SSCs 

necessary to defend against the DBEs are 

defined as ‘‘safety-related,’’ and these 

SSCs are the subject of many regulatory 

requirements designed to ensure that 

they are of high quality and high 

reliability, and have the capability to 

perform during postulated design basis 

conditions. Typically, the regulations 

establish the scope of SSCs that receive 

special treatment using one of three 

different terms: ‘‘safety-related,’’ 

‘‘important to safety,’’ or ‘‘basic 

component.’’ The terms ‘‘safety-related 

‘‘and ‘‘basic component’’ are defined in 

the regulations, while ‘‘important to 

safety,’’ used principally in the general 

design criteria (GDC) of Appendix A to 

10 CFR part 50, is not explicitly defined.

These prescriptive requirements as to 

how licensees are to treat SSCs, 

especially those that are defined as 

‘‘safety-related,’’ are referred to in the 

rulemaking as ‘‘special treatment 

requirements.’’ These requirements 

were developed to provide greater 

assurance that these SSCs would 

perform their functions under particular 

conditions (e.g., seismic events or harsh 

environments), with high quality and 

reliability, for as long as they are part of 

the plant. These include particular 

examination techniques, testing 

strategies, documentation requirements, 

personnel qualification requirements, 

independent oversight, etc. In many 

instances, these ‘‘special treatment’’ 

requirements were developed as a 

means to gain assurance when more 

direct measures (e.g., testing under 

design basis conditions or routine 

operation) could not show that SSCs 

were functionally capable. 

Special treatment requirements are 

imposed on nuclear reactor applicants 

and licensees through numerous 

regulations that have been issued since 

the 1960’s. These requirements specify 

different scopes of equipment for 

different special treatment requirements 

depending on the specific regulatory 

concern, but are derived from 

consideration of the deterministic DBEs. 

Treatment for an SSC, as a general 

term and as it will be used in this 

rulemaking, refers to activities, 

processes, and/or controls that are 

performed or used in the design, 

installation, maintenance, and operation 

of SSCs as a means of: 

(1) Specifying and procuring SSCs 

that satisfy performance requirements; 

(2) Verifying over time that 

performance is maintained; 

(3) Controlling activities that could 

impact performance; and 

(4) Providing assessment and 

feedback of results to adjust activities as 

needed to meet desired outcomes. 

Treatment includes, but is not limited 

to, quality assurance, testing, 

inspection, condition monitoring, 

assessment, evaluation, and resolution 

of deviations. The distinction between 

‘‘treatment’’ and ‘‘special treatment’’ is 

the degree of NRC specification as to 

what must be implemented for 

particular SSCs or for particular 

conditions. 

Defense-in-depth is an element of the 

NRC’s safety philosophy that employs 

successive measures to prevent 

accidents or mitigate damage if a 

malfunction, accident, or naturally 

caused event occurs at a nuclear facility. 

Defense-in-depth is a philosophy used 

by the NRC to provide redundancy as 

well as the philosophy of a multiplebarrier 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, 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68009

maintenance, or operation of a nuclear 

facility. The net effect of incorporating 

defense-in-depth into design, 

construction, maintenance, and 

operation is that the facility or system 

in question tends to be more tolerant of 

failures and external challenges. 

A probabilistic approach to regulation 

enhances and extends the traditional 

deterministic approach by allowing 

consideration of a broader set of 

potential challenges to safety, providing 

a logical means for prioritizing these 

challenges based on safety significance, 

and allowing consideration of a broader 

set of resources to defend against these 

challenges. Until the accident at Three 

Mile Island (TMI), the NRC only used 

probabilistic criteria in specialized 

areas, such as for certain man-made 

hazards and for natural hazards (with 

respect to initiating event frequency). 

The major investigations of the TMI 

accident recommended that 

probabilistic risk assessment (PRA) 

techniques be used more widely to 

augment traditional non-probabilistic 

methods of analyzing plant safety. 

In contrast to the deterministic 

approach, PRAs address credible 

initiating events by assessing the event 

frequency. Mitigating system reliability 

is then assessed, including the potential 

for common cause failures. The 

probabilistic treatment goes beyond the 

single failure requirements used in the 

deterministic approach. The 

probabilistic approach to regulation is 

therefore considered an extension and 

enhancement of traditional regulation 

by considering risk in a more coherent 

and complete manner. 

The primary need for improving the 

implementation of defense-in-depth in a 

risk-informed regulatory system is 

guidance to determine how many 

measures are appropriate and how good 

these should be. Instead of merely 

relying on bottom-line risk estimates, 

defense-in-depth is invoked as a strategy 

to ensure public safety given there exists 

both unquantified and unquantifiable 

uncertainty in engineering analyses 

(both deterministic and risk 

assessments). 

Risk insights can make the elements 

of defense-in-depth clearer by 

quantifying them to the extent 

practicable. Although the uncertainties 

associated with the importance of some 

elements of defense may be substantial, 

the fact that these elements and 

uncertainties have been quantified can 

aid in determining how much defense is 

appropriate from a regulatory 

perspective. Decisions on the adequacy 

of, or the necessity for, elements of 

defense should reflect risk insights 

gained through identification of the 

individual performance of each defense 

system in relation to overall 

performance. 

The Commission published a Policy 

Statement on the ‘‘Use of Probabilistic 

Risk Assessment’’ on August 16, 1995 

(60 FR 42622). In the policy statement, 

the Commission stated that the use of 

PRA technology should be increased in 

all regulatory matters to the extent 

supported by the state of the art in PRA 

methods and data, and in a manner that 

supports the NRC’s traditional defensein-depth philosophy. The policy 

statement also stated that, in making 

regulatory judgments, the Commission’s 

safety goals for nuclear power reactors 

and subsidiary numerical objectives (on 

core damage frequency and containment 

performance) should be used with 

appropriate consideration of 

uncertainties. 

To implement this Commission 

policy, the NRC staff developed 

guidance on the use of risk information 

for reactor license amendments and 

issued Regulatory Guide (RG) 1.174, 

‘‘An Approach for Using Probabilistic 

Risk Assessment in Risk-Informed 

Decisions on Plant-Specific Changes to 

the Licensing Basis.’’ This RG provided 

guidance on an acceptable approach to 

risk-informed decision-making 

consistent with the Commission’s 

policy, including a set of key principles. 

These principles include: 

(1) Be consistent with the defense-indepth philosophy; 

(2) Maintain sufficient safety margins; 

(3) Any changes allowed must result 

in only a small increase in core damage 

frequency or risk, consistent with the 

intent of the Commission’s Safety Goal 

Policy Statement; and, 

(4) Incorporate monitoring and 

performance measurement strategies. 

RG 1.174 states that consistency with 

the defense-in-depth philosophy will be 

preserved by ensuring that: 

(1) A reasonable balance is preserved 

among prevention of accidents, 

prevention of barrier failure, and 

mitigation of consequences; 

(2) An over-reliance on programmatic 

activities to compensate for weaknesses 

in equipment or device design is 

avoided; 

(3) System redundancy, 

independence, and diversity are 

preserved commensurate with the 

expected frequency, consequences of 

challenges to the system, and 

uncertainties (e.g., no risk outliers);

(4) Defenses against potential 

common cause failures are preserved, 

and the potential for the introduction of 

new common cause failure mechanisms 

is assessed; 

(5) The independence of barriers is 

not degraded; and, 

(6) Defenses against human errors are 

preserved. 

I.2 Rule Initiation 

In addition to RG 1.174, the NRC also 

issued other regulatory guides on riskinformed 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, Riskinformed Inservice Inspection. In this 

respect, the Commission has been 

successful in developing and 

implementing a regulatory means for 

considering risk insights into the 

current regulatory framework. One such 

risk-informed application, the South 

Texas Project (STP) submittal on graded 

quality assurance, is particularly 

noteworthy. 

In March 1996, STP Nuclear 

Operating Company (STPNOC) 

requested that the NRC approve a 

revised Operations Quality Assurance 

Program (OQAP) that incorporated the 

methodology for grading quality 

assurance (QA) based on PRA insights. 

The STP graded QA proposal was an 

extension of the existing regulatory 

framework. Specifically, the STP 

approach continued to use the 

traditional safety-related categorization, 

but allowed for gradation of safety 

significance within the ‘‘safety-related’’

categorization (consistent with 10 CFR 

part 50 appendix B) through use of a 

risk-informed process. Following 

extensive discussions with the licensee 

and substantial review, the NRC staff 

approved the proposed revision to the 

OQAP on November 6, 1997. 

Subsequent to NRC’s approval, STPNOC 

identified implementation difficulties 

associated with the graded QA program. 

Despite the reduced QA requirement 

applied for a large number of SSCs in 

which the licensee judged to be of low 

safety significance, other regulatory 

requirements such as environmental 

qualification, the American Society of 

Mechanical Engineers (ASME) Boiler 

and Pressure Vessel Code (BPV), or 

seismic requirements, continued to 

impose substantial burdens. As a result, 

the replacement of a low safety 

significant component needed to satisfy 

other special requirements during a 

procurement process. These 

requirements prevented STPNOC from 

realizing the full potential reduction in 

unnecessary regulatory burden for SSCs 

judged to have little or no safety 

importance. In an effort to achieve the 

full benefit of the graded QA program 

(and in fact to go beyond the staff’s 

68010 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

previous approval of graded QA), 

STPNOC submitted a request, dated July 

13, 1999, asking for an exemption from 

the scope of numerous special treatment 

regulations (including 10 CFR part 50 

appendix B) for SSCs categorized as low 

safety significant or as non-risk 

significant. STPNOC’s exemption was 

ultimately approved by the staff in 

August 2001 (further discussion on this 

exemption request is provided in 

Section IV.2). 

The experience with graded QA was 

a principal factor in the NRC’s 

determination that rule changes would 

be necessary to proceed with some 

activities to risk-inform requirements. 

The Commission also believes that the 

development of PRA technology and 

decision-making tools for using risk 

information together with deterministic 

information supported rulemaking 

activities to allow the NRC to refocus 

certain regulatory requirements using 

this type of information. 

Under Option 2 of SECY–98–300, 

‘‘Options for Risk-Informed Revisions to 

10 CFR part 50—‘Domestic Licensing of 

Production and Utilization Facilities,’ ’’

dated December 23, 1998, the NRC staff 

recommended that risk-informed 

approaches to the application of special 

treatment requirements be developed as 

one application of risk-informed 

regulatory changes. Option 2 (also 

referred to as RIP50 Option 2) addresses 

the implementation of changes to the 

scope of SSCs needing special treatment 

while still providing assurance that the 

SSCs will perform their design 

functions. Changes to the requirements 

pertaining to the design basis functional 

requirements of the plant or the design 

basis accidents are not included in 

Option 2. These technical risk-informed 

changes are addressed under Option 3 

of SECY–98–300. The Commission 

approved proceeding with Option 2 in 

a staff requirements memorandum 

(SRM) dated June 8, 1999. 

The stated purpose of the ‘‘Option 2’’

rulemaking was to develop an 

alternative regulatory framework that 

enables licensees, using a risk-informed 

process for categorizing SSCs according 

to their safety significance (i.e., a 

decision that considers both traditional 

deterministic insights and risk insights), 

to reduce unnecessary regulatory 

burden for SSCs of low safety 

significance by removing these SSCs 

from the scope of special treatment 

requirements. As part of this process, 

those SSCs found to be of risksignificance would be brought under a 

greater degree of regulatory control 

through the requirements being added 

to the rule, which are designed to 

maintain consistency between actual 

performance and the performance 

credited in the assessment process that 

determines their significance. As a 

result, both the NRC and industry 

should be able to better focus their 

resources on regulatory issues of greater 

safety significance. 

The Commission directed the NRC 

staff to evaluate strategies to make the 

scope of the nuclear power reactor 

regulations that impose special 

treatment risk-informed. SECY–99–256, 

‘‘Rulemaking Plan for Risk-Informing 

Special Treatment Requirements,’’ dated 

October 29, 1999, was sent to the 

Commission to obtain approval for a 

rulemaking plan and issuance of an 

Advance Notice of Proposed 

Rulemaking (ANPR). By SRM dated 

January 31, 2000, the Commission 

approved publication of the ANPR and 

approved the rulemaking plan. The 

ANPR was published in the Federal 

Register on March 3, 2000 (65 FR 

11488), for a 75-day comment period, 

which ended on May 17, 2000. In the 

rulemaking plan, the NRC proposed to 

create a new section within part 50, now 

identified as § 50.69, to contain these 

alternative requirements. 

The Commission received more than 

200 comments in response to the ANPR. 

The NRC staff sent the Commission 

SECY–00–0194, ‘‘Risk-Informing 

Special Treatment Requirements,’’ dated 

September 7, 2000, which provided the 

staff’s preliminary views on the ANPR 

comments and additional thoughts on 

the preliminary regulatory framework 

for implementing a rule to revise the 

scope of special treatment requirements 

for SSCs. The comments from the ANPR 

are further discussed in Section IV.1.0 

of SECY–02–0176, ‘‘Proposed 

Rulemaking to Add New Section 10 CFR 

50.69, ‘‘Risk-Informed Categorization 

and Treatment of Structures, Systems, 

and Components,’’ dated September 30, 

2002 (ADAMS accession number 

ML022630007). 

The concept developed for this rule, 

discussed at length in the ANPR, 

applies treatment requirements based 

upon the safety significance of SSCs, 

determined through consideration of 

both risk insights and deterministic 

information. Thus, the risk-informed 

approach discussed in this rule for 

establishing an alternative scope of 

SSCs subject to special treatment 

requirements uses both risk and 

traditional deterministic methods in a 

blended ‘‘risk-informed’’ approach. 

The NRC staff prepared a proposed 

rule package and provided it to the 

Commission in SECY–02–0176. The 

Commission approved issuance of 

proposed 10 CFR 50.69 for public 

comment in a SRM dated March 28, 

2003. The proposed 10 CFR 50.69 rule 

was published for public comment in 

the Federal Register on May 16, 2003 

(68 FR 26511). The Commission 

received 26 sets of comments in 

response to the proposed rule. The 

comments are discussed in Section II 

below.

The NRC staff provided the 

Commission the draft final rule in 

SECY–04–0109 dated June 30, 2004. 

The Commission subsequently 

approved the final rule subject to the 

changes denoted during the session and 

documented in SRM dated October 7, 

2004 (ADAMS accession number 

ML042810516). 

I.3 Rule Overview 

Section 50.69 represents an 

alternative set of requirements whereby 

a licensee or applicant may voluntarily 

undertake categorization of its SSCs 

consistent with the requirements in 

§ 50.69(c), remove the special treatment 

requirements listed in § 50.69(b) for 

SSCs that are determined to be of low 

individual safety significance, and 

implement alternative treatment 

requirements in § 50.69(d). The 

regulatory requirements not removed by 

§ 50.69(b) continue to apply as well as 

the requirements specified in § 50.69. 

The rule contains requirements by 

which a licensee categorizes SSCs using 

a risk-informed process, adjusts 

treatment requirements consistent with 

the relative significance of the SSC, and 

manages the process over the lifetime of 

the plant. To implement these 

requirements, a risk-informed 

categorization process is employed to 

determine the safety significance of 

SSCs and place the SSCs into one of 

four risk-informed safety class (RISC) 

categories. The determination of safety 

significance is performed by an 

integrated decision-making process 

which uses both risk insights and 

traditional engineering insights. The 

safety functions include both the design 

basis functions (derived from the 

‘‘safety-related’’ definition, which 

includes external events), as well as, 

functions credited for severe accidents 

(including external events). Treatment 

for the SSCs is required to be applied as 

necessary to maintain functionality and 

reliability, and is a function of the 

category into which the SSC is 

categorized. Finally, assessment 

activities are conducted to make 

adjustments to the categorization and 

treatment processes as needed so that 

SSCs continue to meet applicable 

requirements. The rule contains 

requirements for obtaining prior NRC 

review and approval of the 

categorization process and for 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68011

maintaining certain plant records and 

reports. For a more detailed discussion 

of the rule requirements refer to 

Sections III and V of this rule. 

It is important to note that this 

rulemaking effort, while intended to 

ensure that the scope of special 

treatment requirements imposed on 

SSCs is risk-informed, is not intended to 

allow for the elimination of SSC 

functional requirements or to allow 

equipment that is required by the 

deterministic design basis to be 

removed from the facility (i.e., changes 

to the design of the facility must 

continue to meet the current 

requirements governing design change; 

most notably § 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 riskinformed approach to special treatment. 

Conversely, for SSCs that do not 

significantly contribute to plant safety 

on an individual basis, this approach 

should allow an acceptable, though 

reduced, level of confidence (i.e., 

‘‘reasonable confidence’’) that these 

SSCs will satisfy functional 

requirements. However, continued 

maintenance of the health and safety of 

the public will depend on effective 

implementation of § 50.69 by the 

licensee or applicant applying the rule 

at its nuclear power plant. 

II. Public Comments 

II.1.0 Comments on Proposed Rule 

The Commission published proposed 

§ 50.69 for public comment on May 16, 

2003 (68 FR 26511). Twenty-six sets of 

comments were received (comments are 

available at http://ruleforum.llnl.gov/

cgi-bin/rulemake?source=SSC_

PRULE&st=prule). The Commission 

requested feedback on several specific 

issues in Section VI of the proposed rule 

notice. A summary of the public 

feedback concerning these issues, as 

well as a discussion of the more 

significant comments, follows. A 

detailed discussion of the issues raised 

by all comments is contained in a 

separate document (see Section IX, 

Availability of Documents). 

II.1.1 Consideration of More Detailed 

Language for § 50.69(d)(2) Regarding 

RISC–3 SSC Treatment Requirements 

As discussed in the proposed rule, the 

Commission believed that detailed rule 

language for the treatment of RISC–3 

SSCs (i.e., safety-related SSCs that are 

categorized as low safety significant) 

was not necessary to provide reasonable 

confidence in RISC–3 design basis 

capability and, as a consequence, 

constructed proposed § 50.69 to contain 

high-level (i.e., less detailed) RISC–3 

treatment requirements. However, the 

Commission recognized that some 

stakeholders could disagree with this 

approach and invited comment on this 

issue. For the most part, industry 

commenters asserted that there was no 

need for more detailed treatment 

requirements for RISC–3 SSCs in the 

rule. The state commenters and public 

interest groups considered the proposed 

rule language to be inadequate to 

provide reasonable confidence in the 

capability of RISC–3 SSCs to perform 

their safety-related functions under 

design basis conditions. In reviewing 

the public comments, the Commission 

found significant divergence in the 

interpretation of the proposed rule 

language by industry commenters from 

the Commission’s expectations as 

described in the Statement of 

Considerations—preamble—(SOC) for 

the proposed rule. After consideration 

of all stakeholder comments, the 

Commission revised § 50.69(d)(2) to 

adopt a more performance-based 

approach that provides licensees and 

applicants greater flexibility in 

establishing RISC–3 treatment 

consistent with the low safety 

significance of RISC–3 SSCs. 

Accordingly, the Commission has 

removed the more prescriptive 

requirements regarding RISC–3 

treatment activities and adopted rule 

language that focuses on the 

performance requirements for RISC–3 

SSCs. 

II.1.2 PRA Requirements 

The Commission requested 

stakeholder comment on whether the 

NRC should amend the requirements in 

§ 50.69(c) to require a level 2 internal 

and external initiating events, all-mode, 

peer-reviewed PRA that must be 

submitted to, and reviewed by, the NRC. 

Stakeholder comments ranged from 

those supporting such PRA 

requirements to those who conclude 

that the proposed PRA requirements in 

§ 50.69(c) are sufficient. The industry 

commenters stated that additional PRA 

requirements were not necessary 

because the other categorization 

requirements in § 50.69(c) addressed 

other modes and events not addressed 

by the PRA and as a result, all sources 

of risk were addressed. The states and 

public interest groups supported 

increased PRA requirements. The 

Commission concludes that the § 50.69 

PRA requirements in the proposed rule 

are sufficient for this application. The 

supporting guidance for the rule has 

been structured such that licensees will 

gain more benefit when PRA methods 

are used (beyond the minimum PRA 

requirements in § 50.69(c)), and where 

non-PRA methods are used, the 

requirements and associated 

implementation guidance account for 

this situation by requiring a process that 

tends to conservatively categorize SSCs 

into RISC–1 and RISC–2 (i.e., no special 

treatment requirements are removed). 

There are several other features to the 

regulatory framework that also 

contribute to ensuring sound PRA is 

used such as requiring aspects of the 

categorization process to be reviewed 

and approved before implementation, 

requiring the PRA to be peer reviewed, 

Integrated Decision-Making Panel (IDP) 

requirements, provisions for addressing 

all modes and events regardless of 

whether in the PRA, feedback and 

update requirements, and supporting 

standards. (Also see the Commission’s 

SRM on PRA quality dated December 

18, 2003, ADAMS Accession No. 

ML033520457.)

II.1.3 Review and Approval of RISC–3 

Treatment 

The Commission requested 

stakeholder comment on whether the 

NRC should review and approve the 

RISC–3 treatment processes being 

developed by the licensee or applicant 

before implementation in addition to 

reviewing the categorization process. 

Public interest groups and comments 

from state organizations generally 

stressed the need for the NRC to review 

and approve RISC–3 treatment 

processes in advance of implementation 

to confirm appropriate treatment will be 

applied to RISC–3 SSCs given that these 

SSCs are safety-related. On the other 

hand, industry commenters did not 

consider prior review and approval of 

RISC–3 treatment to be necessary in 

light of the low safety significance of 

individual RISC–3 SSCs, other 

requirements that help maintain safety, 

and the availability of inspection and 

enforcement by the NRC. The NRC 

agrees that the individual low safety 

significance of RISC–3 SSCs supports 

allowing licensees to establish treatment 

for RISC–3 SSCs without prior NRC 

review. This conclusion is based on the 

rule containing: 

(1) Robust categorization and PRA 

requirements; 

(2) Requirements to show that 

implementation risk is small; 

(3) Feedback requirements of 

paragraph (e) to help maintain the 

validity of the categorization process; 

and 

(4) The high-level, performance-based 

RISC–3 requirements designed to 

68012 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

maintain RISC–3 SSC design basis 

functional capability. 

In addition, a provision has been 

added to the final rule to make it clear 

that the treatment applied to RISC–3 

SSCs must be consistent with (i.e., 

maintain the validity of) the 

categorization process. To provide 

additional assurance, the NRC intends 

to conduct sample inspections at 

nuclear power plants implementing 

§ 50.69 to address programmatic issues 

related to the categorization and 

treatment processes (see below). 

II.1.4 Inspection and Enforcement 

The Commission requested 

stakeholder comment on whether or not 

changes are needed in the NRC’s reactor 

oversight process including the 

inspection program and enforcement to 

enable NRC to exercise the appropriate 

degree of regulatory oversight of these 

aspects of facility operation regarding 

§ 50.69. The public comments on the 

proposed rule indicated general support 

for providing regulatory oversight of the 

implementation of processes established 

under § 50.69 through the NRC’s 

inspection and enforcement process. 

Some stakeholders considered the 

current inspection and enforcement 

process to be sufficient without 

adjustment. Other stakeholders 

recommended that the NRC consider 

additional training and guidance to 

inspectors to support implementation of 

§ 50.69. Some stakeholders provided 

specific and constructive suggestions 

regarding the inspection and 

enforcement process under § 50.69 

including aspects of treatment processes 

to be inspected, and the application of 

enforcement discretion. Based on its 

consideration of this issue, the 

Commission plans to conduct 

inspections of § 50.69 implementation. 

These inspections will be performed on 

a sampling basis (in terms of the number 

of plants inspected) and will depend on 

the number of licensees who decide to 

implement § 50.69. These sample 

inspections are intended to gather 

information that will enable the NRC to 

assess whether modifications are 

needed to the ongoing baseline 

inspection program. The principal focus 

of the inspection will be on the safety 

significant aspects of § 50.69 

implementation such as categorization 

and treatment of RISC–1 and RISC–2 

SSCs, but the inspection will also 

consider the implementation of RISC–3 

treatment focusing on programmatic and 

common cause issues, which could 

undermine the categorization process 

and its results. 

II.1.5 Operating Experience 

The Commission requested 

stakeholder feedback regarding the role 

that relevant operational experience 

could play in reducing the uncertainty 

associated with the effects of treatment 

on performance and specifically sought 

public comment as to what information 

might be available and how it could be 

used to support implementation of this 

rulemaking. Some stakeholders 

commented that relevant operating 

experience argues against the removal of 

special treatment requirements and that 

regulatory attention should be increased 

for this equipment. Other stakeholders 

suggested that there is a large amount of 

data that demonstrates that commercial 

and safety-related SSCs have 

comparable failure rates with the 

implication that special treatment 

requirements can be removed with little 

impact. The specific study referenced by 

those stakeholders was not submitted 

for formal NRC review. The Commission 

concludes that a single unreviewed 

study does not provide a sufficient basis 

to make broad conclusions regarding the 

performance of SSCs subject to 

commercial and industrial practices for 

fabrication, installation, and 

maintenance. Other stakeholders 

commented that there are already 

opportunities for industry to share 

experience data with existing industry 

and regulatory programs implying that a 

new program is not necessary. In some 

instances, however, those referenced 

programs will be eliminated for RISC–

3 SSCs under § 50.69. To emphasize the 

importance of applying operating 

experience in maintaining plant safety, 

the final rule has been revised to clarify 

that § 50.69(e)(1) requires the feedback 

of plant operational experience in 

addition to the requirements to feed 

back performance data, plant changes, 

operational changes, and industry 

experience. This plant operational 

information may be obtained from the 

corrective action program and 

processes, as well as other sources. 

II.1.6 Other Substantive Issues 

In addition to the issues addressed in 

Section II.1.5, stakeholders provided 

substantive comments that caused the 

NRC to re-examine the § 50.69 

framework and make changes. Those 

issues and comments are discussed 

below. Additionally, there were several 

issues that involved a significant 

number of stakeholder comments, and 

even though the Commission decided 

not to revise its approach, those issues 

and comments are also discussed in this 

section. 

II.1.6.1 SOC Guidance 

Numerous comments were received 

from the industry regarding the nature 

of the information in the proposed rule’s 

SOC supporting both § 50.69(d)(2) and 

§ 50.69(c). Several industry commenters 

stated that the discussion in the SOC 

was inconsistent with the rule 

requirements. For example, some 

commenters suggested that, contrary to 

the SOC discussion, the treatment 

requirements for RISC–3 SSCs in 

§ 50.69(d)(2) would allow exercising of 

pumps and valves as a means of 

providing reasonable confidence in the 

design basis capability of those 

components. Another commenter 

claimed that, contrary to the SOC 

discussion, § 50.69 would allow the 

leakage tests required by 10 CFR part 50, 

Appendix J, for containment isolation 

valves to be eliminated without 

considering the capability of those 

valves to close under design basis 

conditions. Other commenters asserted 

that the corrective action process alone 

would be sufficient to satisfy the highlevel requirements for feedback and 

monitoring of RISC–3 SSCs in § 50.69. 

These industry comments raised 

concerns regarding the interpretation of 

the rule language. The Commission 

clarified the rule requirements and 

simplified the SOC to focus on the 

meaning of the rule language (see 

Sections II.1.6.2 through II.1.6.3, Section 

V.5.2, and the responses to comments 

d–32 and e–4 in Table 3 of ‘‘Response 

to Comments on Proposed § 50.69’’ as 

referenced in Section IX of this 

document). 

II.1.6.2 RISC–3 Treatment 

Requirements 

Numerous stakeholder comments 

were received concerning the 

§ 50.69(d)(2) requirements for RISC–3 

SSCs. Some public stakeholders 

provided their view that the RISC–3 

treatment requirements were inadequate 

in light of previous industry experience 

(e.g., regarding the use of substandard 

parts) and that more detailed RISC–3 

requirements were needed to address 

common cause failures, significant 

degradation, and in general to avoid an 

increase in risk to the health and safety 

of the public. Industry stakeholders 

tended to view the RISC–3 requirements 

as too prescriptive and beyond what is 

necessary to maintain reasonable 

confidence of RISC–3 SSC design basis 

capability. Some of the industry 

comments revealed that the rule 

requirements might not be implemented 

consistent with the Commission’s 

expectations discussed in the SOC. 

Therefore, the Commission clarified the 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68013

rule and SOC as discussed in the 

following sections. 

II.1.6.2.1 Fracture Toughness 

In the SOC for the proposed rule, the 

Commission noted that design 

requirements for fracture toughness 

would continue to apply for 

replacement ASME components 

categorized as RISC–3 SSCs. One 

industry commenter asserted that 

fracture toughness is not a design issue 

while other commenters argued in 

general that the SOC discussion 

exceeded the rule requirements. The 

Commission emphasizes that the intent 

of § 50.69 is to remove special treatment 

requirements while maintaining design 

requirements for RISC–3 SSCs. The 

Commission considers fracture 

toughness to be an important design 

consideration. Fracture toughness is a 

property of the material that prevents 

premature failure of an SSC at abrupt 

geometry changes, or at small 

undetected flaws. Adequate fracture 

toughness of SSCs is necessary to 

prevent common cause failures due to 

design basis events, such as 

earthquakes. To ensure that this design 

consideration continues to be applicable 

to § 50.69 licensees, § 50.69(b)(1)(v) was 

clarified to exclude fracture toughness 

from the scope of § 50.55a repair and 

replacement requirements which are 

removed for RISC–3 SSCs. 

II.1.6.2.2 Consistency With the 

Categorization Process 

Several industry comments indicated 

that licensees might not consider the 

impact of changes in treatment on 

RISC–3 SSCs as part of the 

categorization process. For example, one 

industry commenter asserted that 

sensitivity studies eliminate the need to 

specifically consider SSC reliability 

changes that might occur due to 

treatment changes. Another industry 

commenter stated that cross-system 

common cause interactions are rarely 

modeled in PRAs. Similarly, another 

industry commenter indicated that 

degradation mechanisms resulting from 

treatment processes are typically not 

considered in PRAs. The treatment 

practices for plant SSCs must support 

the capability credited in the 

categorization process for there to be 

reasonable confidence that any increase 

in risk remains small. Therefore, 

§ 50.69(d)(2) was clarified to explicitly 

require the treatment of RISC–3 SSCs to 

be consistent with the categorization 

process.

II.1.6.2.3 Voluntary Consensus 

Standards 

In the SOC for the proposed rule, the 

Commission discussed the use of 

voluntary consensus standards as one 

effective means to establish treatment 

requirements for RISC–3 SSCs. In its 

comments, the ASME did not 

recommend adding a provision on 

voluntary consensus standards in the 

rule itself because it considered the SOC 

to provide adequate guidance for RISC–

3 treatment. However, several industry 

commenters suggested that licensees 

might only apply general industrial 

practices when implementing treatment 

requirements for RISC–3 SSCs. For 

example, some industry commenters 

believed that exercising a pump or valve 

would provide sufficient assurance 

under § 50.69 of the capability of the 

pump or valve to perform its design 

basis safety functions. Although 

exercising a pump or valve might be 

consistent with general industrial 

practices, operating experience has 

demonstrated that exercising a pump or 

valve is not sufficient to ensure with 

reasonable confidence its design basis 

capability. For example, the 

Commission modified § 50.55a to 

require licensees implementing the 

ASME Code for Operation and 

Maintenance of Nuclear Power Plants to 

periodically verify the design basis 

capability of motor-operated valves to 

perform their safety functions in light of 

the recognized inadequacies in stroketime testing (somewhat more 

informative than exercising) to assess 

the operational readiness of those 

valves. The NRC issued Regulatory Issue 

Summary 00–03 (March 15, 2000), 

‘‘Resolution of Generic Safety Issue 158, 

Performance of Safety-Related PowerOperated Valves Under Design Basis 

Conditions,’’ to discuss the importance 

of this issue relative to safety-related airoperated and other power-operated 

valves. Further, the ASME developed 

comprehensive pump testing provisions 

to provide more appropriate testing 

under significant flow conditions in 

light of the weakness of the previous 

Code testing under minimal loading 

conditions. In SECY–00–0194, the NRC 

noted that a wide variation existed in 

industrial practices. Therefore, certain 

industrial practices may not be 

sufficient to satisfy the treatment 

requirements for RISC–3 SSCs in 

§ 50.69. To address these concerns, the 

Commission clarified the rule 

requirements to indicate that the 

treatment of RISC–3 SSCs must be 

consistent with the categorization 

process. One way to achieve this 

consistency could be the application of 

consensus standards. However, 

licensees or applicants must recognize 

that the application of such standards 

must meet § 50.69(d)(2) requirements to 

be acceptable. The determination of 

consistency between treatment and 

categorization also includes 

consideration of applicable operational 

experience, which may be found from 

such sources as NRC information 

notices, bulletins, and generic letters; 

and vendor recommendations. 

II.1.6.2.4 Design Control Process 

In the SOC for the proposed rule, the 

Commission listed several attributes to 

be considered as part of the design 

control process for RISC–3 SSCs in 

satisfying the high-level treatment 

requirements in § 50.69. One industry 

commenter suggested that a focused list 

of design control attributes be 

substituted in § 50.69 for the proposed 

rule language. This list would include 

selection of suitable materials; 

verification of design adequacy, and 

control of design changes. After 

consideration of these comments, the 

Commission has decided not to include 

detailed design control process 

requirements in the final rule. The final 

rule requirements require that licensees 

and applicants ensure with reasonable 

confidence that RISC–3 SSCs remain 

capable of performing their safetyrelated functions under design basis 

conditions. With respect to design 

changes, as noted in several places in 

the notice for the final rule, § 50.69 is 

not changing design basis functional 

requirements and § 50.59 remains 

applicable to all changes to non-special 

treatment aspects of RISC–3 SSCs. The 

Commission believes that a 

performance-based requirement will 

allow licensees who choose to 

implement § 50.69 to have greater 

flexibility to implement treatment that 

they have determined is needed, 

commensurate with the safety 

significance of the SSCs in order to 

ensure with reasonable confidence that 

RISC–3 safety-related functional 

capability is maintained. 

II.1.6.2.5 Design Basis Conditions 

Under § 50.69, RISC–3 SSCs will be 

exempt from special treatment 

requirements for qualification methods 

for environmental conditions and effects 

and seismic conditions. Nevertheless, 

RISC–3 SSCs continue to be required to 

be capable of performing their safetyrelated functions under applicable 

environmental conditions and effects 

and seismic conditions, albeit at a lower 

level of confidence as compared to 

RISC–1 SSCs. Based on industry 

comments on the proposed rule, some 

68014 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

1 In December 1980 the NRC designated ‘‘Seismic 

Qualification of Equipment in Operating Plants’’ as 

an unresolved safety issue. For more information 

refer to GL 87–02.

licensees appeared to interpret the 

proposed rule language as not requiring 

evaluation of environmental and 

seismic capability of RISC–3 SSCs. For 

example, one industry commenter 

stated that § 50.69 exempts RISC–3 

electrical equipment from aging issues 

and that the rule would not require the 

establishment of design life for RISC–3 

electrical equipment. Contrary to the 

public comment, a licensee 

implementing § 50.69 must consider 

operating life (aging) and combinations 

of operating life parameters (synergistic 

effects) in the design of RISC–3 

electrical equipment. This is 

particularly important if the equipment 

contains materials which are known to 

be susceptible to significant degradation 

due to thermal, radiation, and/or wear 

(cyclic) aging including any known 

synergistic effects that could impair the 

ability of the equipment to meet its 

design basis function. However, the 

Commission agrees that the applicable 

rule language can be simplified and has 

revised the final rule to utilize a 

performance-based approach to 

ensuring with reasonable confidence the 

functional capabilities of RISC–3 SSCs. 

Accordingly, the final rule has been 

revised by deleting the reference to the 

specific conditions that were 

parenthetically listed in the proposed 

rule. 

II.1.6.2.6 Corrective Action 

Some public commenters raised 

concerns regarding the lack of 

requirements for the consideration of 

common-cause issues for RISC–3 SSCs. 

An industry commenter also noted this 

omission in the proposed rule and 

provided proposed rule language to 

resolve this issue. Therefore, the 

Commission decided to revise 

§ 50.69(d)(2)(ii) to require that, for 

significant conditions adverse to quality 

associated with RISC–3 SSCs, measures 

shall be taken to provide reasonable 

confidence that the cause of the 

condition is determined and corrective 

action is taken to preclude repetition. 

The revised corrective action 

requirement is consistent with a 

proposal by the Nuclear Energy Institute 

and uses language that is similar to 10 

CFR part 50 Appendix B Criterion XVI. 

As such, this should be a wellunderstood requirement that minimizes 

the potential for common cause failures. 

It is also consistent with the principle 

of performance-based regulation that 

non-compliance with the performance 

requirement should provide sufficient 

margin such that reasonable assurance 

of public health and safety continues to 

be provided. 

II.1.6.2.7 Seismic Experience Data 

Several industry commenters stated 

that the SOC for the proposed rule 

might create additional burden on 

plants licensed before implementation 

of Appendix A to 10 CFR Part 100. In 

establishing § 50.69, the Commission 

does not intend to alter the existing 

seismic design requirements for RISC–3 

SSCs in any plant’s design basis. 

Industry commenters also raised 

concerns regarding the SOC discussion 

on use of seismic experience data. In 

meeting § 50.69, the licensee or 

applicant must have adequate technical 

bases to conclude that RISC–3 SSCs will 

perform their safety-related functions 

under seismic design basis conditions, 

which includes the number and 

magnitude of earthquake events 

specified for the SSC design. Some 

commenters implied that it would be 

acceptable to use ‘‘experience data’’

alone to have reasonable confidence that 

an SSC is capable of functioning during 

an earthquake even if there is no actual 

‘‘experience data’’ for the SSC. While 

the use of experience data is not 

prohibited by the rule, it may be 

difficult for a licensee or applicant to 

show that experience data alone will 

satisfy the applicable design 

requirements of 10 CFR part 100 (which 

§ 50.69 leaves intact). The Commission 

clarified the SOC with respect to the use 

of seismic experience data and to 

indicate that § 50.69 will not change the 

seismic design basis for Unresolved 

Safety Issue (USI) A–461 plants or 

impose additional seismic requirements 

for those plants.

II.1.6.3 Feedback 

Several industry commenters 

requested adjustments to the feedback 

requirements in § 50.69(e)(1) to provide 

more efficient implementation of the 

rule. Upon consideration of those 

comments, the Commission revised 

§ 50.69(e)(1) to replace the maximum 

time interval for updating the 

categorization and treatment processes 

from 36 months to two refueling 

outages, and to indicate that the licensee 

or applicant may adjust either its 

categorization process or its treatment 

processes in satisfying the feedback 

requirement. 

II.1.6.4 Section 50.46a/Appendix B 

Requirements for High Point Vents 

A comment was submitted that the 

NRC should undertake a review of the 

recently revised § 50.44 to determine 

whether the new rule contains special 

treatment requirements that should be 

within the scope of § 50.69. The 

Commission agreed with this comment. 

The Commission noted in the proposed 

rule (Section III.4.9.3) that there may be 

a need to scope into § 50.69 certain 

provisions of the old § 50.44 dependent 

on the outcome of the effort to risk 

inform the § 50.44 requirements. The 

revised § 50.44 has no special treatment 

requirements. However, when § 50.44 

was revised, a portion of the old § 50.44 

regarding application of Appendix B 

requirements to high point vents was 

moved to § 50.46a. This particular 

requirement was not risk-informed as 

part of the § 50.44 effort and was instead 

simply relocated. Because application of 

Appendix B is a special treatment 

requirement, the Appendix B portion of 

§ 50.46a(b) has been included within the 

scope of § 50.69 by the inclusion of 

§ 50.69(b)(1)(ii).

II.1.6.5 Basis for RISC–3 SSC 

Reliability Used in § 50.69(c)(1)(iv) 

Evaluation 

A number of comments were received 

regarding the technical basis for the 

RISC–3 SSC reliability (failure rates) to 

be used in the risk sensitivity study 

performed to meet § 50.69(c)(1)(iv) 

requirements to demonstrate reasonable 

confidence that any potential risk 

increase from implementation of the 

rule is small. Some commenters 

suggested that licensees or applicants 

that voluntarily implement the rule 

should be required to characterize and 

reasonably bound the specific effects of 

eliminating treatment on SSC reliability 

under design basis and severe accident 

conditions. Other commenters suggested 

that there is evidence that reductions in 

treatment (using industry practices) has 

no impact on SSC reliability. 

The NRC recognizes that the 

reliability of RISC–3 SSCs could 

potentially decrease (RISC–3 SSC failure 

rates increase) due to the reduction in 

treatment applied to these SSCs as a 

result of § 50.69 implementation. This is 

the reason why the Commission 

requires in the rule that the licensee 

demonstrate with reasonable confidence 

that any potential risk increase due to 

implementation of the rule will be 

small. However, the NRC also 

recognizes that it is difficult a priori to 

relate specific changes in treatment 

directly to specific changes in SSC 

reliability. The rule has been 

constructed to account for this 

difficulty. First, the categorization 

process that a licensee uses must 

comply with the rule’s requirements. 

Second, this categorization process will 

be reviewed and approved by the NRC 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68015

before implementation. These steps are 

to have high confidence that SSCs are 

appropriately categorized so that RISC–

3 SSCs are of low individual safety 

significance. Third, licensees are 

required to provide reasonable 

confidence that any risk increase due to 

implementation is acceptably small and 

this assessment must be supported by a 

supporting technical justification that 

discusses why the assessment 

adequately addresses the potential 

reliability changes for RISC–3 SSCs. 

This basis may include reliance on the 

capability of the licensee’s data 

collection, feedback, and corrective 

action , which are also addressed by 

requirements of the rule. Finally, the 

rule has been revised to clarify the 

linkage between treatment and 

categorization and specifically to ensure 

that the treatment process is consistent 

with the categorization process, 

including the risk sensitivity study (i.e., 

maintain that any risk increase due to 

reduced treatment is acceptably small). 

Therefore, the rule is structured to 

contain: 

(1) Robust categorization and PRA 

requirements; 

(2) Requirements to show that 

implementation risk is small; 

(3) A new provision to make it clear 

that the treatment applied to RISC–3 

SSCs must be consistent with (i.e., 

maintain the validity of) the 

categorization process; 

(4) Feedback requirements of 

§ 50.69(e) to maintain the validity of the 

categorization process; and, 

(5) The high-level RISC–3 

requirements designed to maintain 

RISC–3 SSC design basis functional 

capability. 

Thus, the Commission finds that the 

rule, as revised, has the appropriate 

provisions for addressing the concerns 

regarding the basis for RISC–3 SSC 

reliability used in the risk sensitivity 

study to be performed to meet the 

§ 50.69(c)(1)(iv) requirement to 

demonstrate with reasonable confidence 

that any potential risk increase from 

implementation of the rule is small. 

II.1.6.6 RISC–1 and RISC–2 Treatment 

Requirements and Crediting SSCs 

A number of industry stakeholders 

commented on the treatment 

requirements applicable to RISC–1 and 

RISC–2 SSCs in § 50.69(d)(1). These 

stakeholders commented that this 

requirement obligated a licensee 

implementing § 50.69 to evaluate 

treatment applied to all safety 

significant SSCs to ensure adequacy of 

treatment and cited this as an added 

burden that is neither necessary nor 

appropriate because RISC–1 SSCs are 

already subjected to full regulatory 

requirements. They also commented 

that it appeared that this requirements 

was extending special treatment 

requirements (such as Appendix B) to 

RISC–2 SSCs. In fact there was a general 

consensus of comments that any 

additional treatment requirements for 

RISC–1 and RISC–2 SSCs should be 

removed from the SOC or that the SOC 

be clarified to address the specific 

beyond design basis scope of additional 

regulatory controls. First, the 

Commission notes that § 50.69(d)(1) 

does not require licensees or applicants 

to evaluate the application of special 

treatment requirements to RISC–1 SSCs. 

These requirements are to maintain the 

design basis functional requirements 

with a high level of assurance. The 

special treatment requirements remain 

intact and unchanged, and hence there 

is no reason that an evaluation of the 

application of special treatment 

requirements should be required. 

Secondly, the Commission notes that it 

is not the intent of § 50.69(d)(1) to 

simply extend special treatment 

requirements such as Appendix B to 

RISC–1 and RISC–2 beyond design basis 

functions. Instead, the focus of 

§ 50.69(d)(1) is on the PRA credited 

performance of RISC–1 and RISC 2 SSCs 

for beyond design basis conditions, and 

specifically for ensuring that there is a 

valid technical basis for the credit taken 

in the PRA (i.e., there must be a valid 

technical basis for the failure rate/

probability of the SSC performing the 

function). The basis for this credit 

should already be established and 

documented in the PRA supporting 

documentation, so this should not be an 

additional burden for licensees to 

capture and implement. If an existing 

technical basis does not exist or is 

insufficient to support the credit taken 

in the PRA, then § 50.69(d)(1) would 

require that a technical basis be 

developed for the credit taken; 

potentially including the creation of a 

treatment program for the SSC that 

validates the capability credited. 

Regarding the issue of ‘‘credited’’

SSCs, several commenters stated that 

the SOC implied an enormous program 

would be required if a licensee decides 

to selectively implement § 50.69 for a 

set of systems. It was commented that 

this enormous program would result 

due to the application of §§ 50.69(d)(1) 

and 50.69(e)(2) to maintain credited 

performance within the PRA and 

thereby enable the selected set of SSCs 

to be categorized as low safety 

significant. As the Commission has 

already noted, § 50.69(d)(1) obligates 

licensees to have a basis to support the 

performance of RISC–1 and RISC–2 

SSCs credited in the PRA used in the 

categorization process, including the 

performance credited for beyond design 

basis conditions. This is an important 

aspect of the rule. The categorization 

process will result in a number of 

safety-related SSCs being determined to 

be of low safety significance (i.e., RISC–

3) and subject to reduced treatment. 

This determination of low safety 

significance will implicitly take credit 

for the performance capability of other 

SSCs in the PRA, some or all of which 

may not be included in the scope of the 

licensee’s categorization process (due to 

the allowance for licensees to 

selectively implement the rule and to 

phase that implementation over time). 

To maintain the validity of the 

categorization process, and more 

importantly to maintain any potential 

risk increase as small, it is necessary to 

maintain the ‘‘credited’’ SSCs per 

§ 50.69, and this means the application 

of §§ 50.69(d)(1) and 50.69(e)(2) 

requirements as suggested by the 

comment. 

II.1.6.7 Adequate Protection 

Comments 

The NRC received several comments 

indicating that the proposed regulation 

would not maintain adequate protection 

of public health and safety. The 

Commission disagrees with these 

comments and concludes that both the 

proposed rule requirements and the 

final rule requirements maintain 

adequate protection for the reasons 

discussed in Section III.7.0 of this 

notice. 

II.1.6.8 License Amendment 

A commenter stated that the 

requirement to prepare, submit, and 

then receive approval of a license 

amendment to implement § 50.69 is a 

disincentive to its use. The commenter 

argued that, in light of the desire to 

move to a more performance-based 

regulatory regime, voluntary 

implementation of § 50.69 should be 

developed by licensees using the 

requirements in the rule and any 

attendant regulatory guidance, with 

routine NRC inspection serving to verify 

acceptable compliance. The 

Commission has decided not to revise 

§ 50.69 in response to this comment. 

The Commission continues to conclude 

that (as discussed in Section III.6.0 of 

this rule) the review of the license 

amendment submittal will involve 

substantial engineering judgment on the 

part of NRC reviewers, inasmuch as the 

rule does not contain objective, nondiscretionary criteria for assessing the 

adequacy of the PRA process, PRA 

68016 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

review results and sensitivity studies. 

Consistent with the Commission’s 

decision in Cleveland Electric 

Illuminating Co. (Perry Nuclear Power 

Plant, Unit 1), CLI–96–13, 44 NRC 315 

(1996), the final rule requires NRC 

approval to be provided by issuance of 

a license amendment.

III. Final Rule 

The Commission is establishing 

§ 50.69 as an alternative set of 

requirements whereby a licensee or 

applicant may undertake categorization 

of its SSCs consistent with the 

requirements in § 50.69(c) and adjust 

treatment requirements per § 50.69(d) 

based upon the resulting significance. 

Under this approach, a licensee or 

applicant is allowed to remove the 

special treatment requirements listed in 

§ 50.69(b) for SSCs that are determined 

to be of low safety significance while 

potentially enhancing requirements for 

treatment of other SSCs that are found 

to be safety significant. The 

requirements establish a process by 

which a licensee categorizes SSCs using 

a risk-informed process, adjusts 

treatment requirements consistent with 

the relative significance of the SSC, and 

manages the process over the lifetime of 

the plant. To implement these 

requirements, a risk-informed 

categorization process is employed to 

determine the safety significance of 

SSCs and place the SSCs into one of 

four RISC categories. It is important that 

this categorization process be robust to 

enable the Commission to remove 

requirements for SSCs determined to be 

of low safety significance. The 

determination of safety significance is 

performed by an integrated decisionmaking process which uses both risk 

insights and traditional engineering 

insights. The safety functions include 

both the design basis functions (derived 

from the ‘‘safety-related’’ definition, 

which includes external events), as well 

as functions credited for severe 

accidents (including external events). 

Treatment requirements for the SSCs are 

applied as necessary to maintain 

functionality and reliability and are a 

function of the category into which the 

SSC is categorized. Finally, assessment 

activities are conducted to make 

adjustments to the categorization and 

treatment processes as needed so that 

SSCs continue to meet applicable 

requirements. The rule also contains 

requirements for obtaining NRC 

approval of the categorization process 

and for maintaining plant records and 

reports. 

III.1.0 Categorization of SSCs 

Section 50.69 defines four RISC 

categories into which SSCs are 

categorized. Four categories were 

chosen because it is the simplest 

approach for transitioning between the 

previous SSC classification scheme and 

the new scheme used in § 50.69. The 

depiction in Figure 1 provides a 

conceptual understanding of the new 

RISC categories. The figure depicts the 

current safety-related versus nonsafetyrelated SSC categorization scheme with 

an overlay of the new risk-informed 

categorization. In the traditional 

deterministic approach, SSCs were 

generally categorized as either ‘‘safetyrelated’’ (as defined in § 50.2) or 

nonsafety-related. This division is 

shown by the vertical line in the figure. 

Risk insights, including consideration of 

severe accidents, can be used to identify 

SSCs as being safety significant or low 

safety significant (shown by the 

horizontal line). Hence, the application 

of a risk-informed categorization results 

in SSCs being grouped into one of four 

categories as represented by the four 

boxes in Figure 1. 

Box 1 of Figure 1 depicts safetyrelated SSCs that a risk-informed 

categorization process determines are 

significant contributors to plant safety. 

These SSCs are termed RISC–1 SSCs. 

RISC–2 SSCs, depicted by box 2 in 

Figure 1, are nonsafety-related SSCs that 

the risk-informed categorization 

determines to be significant contributors 

to plant safety. The third category are 

those SSCs that are safety-related SSCs 

and that a risk-informed categorization 

process determines are not significant 

individual contributors to plant safety. 

These SSCs are termed RISC–3 SSCs 

and are depicted by box 3 in Figure 1. 

Finally, there are SSCs that are 

nonsafety-related and that a riskinformed categorization process 

determines are not significant 

contributors to plant safety. These SSCs 

are termed RISC–4 SSCs and are 

depicted by box 4 in Figure 1.

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68017

Section 50.69 defines the terminology 

‘‘safety significant function’’ as 

functions whose loss or degradation 

could have a significant adverse effect 

on defense-in-depth, safety margins, or 

risk. This definition was chosen to be 

consistent with the concepts described 

in RG 1.174. The rule maintains more 

treatment requirements on SSCs that 

perform safety significant functions 

(RISC–1 and RISC–2 SSCs) than on 

SSCs that perform low safety significant 

functions to ensure that defense-indepth and safety margins are 

maintained. The rule also requires that 

the licensee or applicant provide 

reasonable confidence that the change 

in risk associated with implementation 

of § 50.69 will be small. 

III.2.0 Methodology for Categorization 

The cornerstone of § 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 § 50.69 are to achieve 

this objective. Essentially, the process is 

structured to ensure that all relevant 

information pertaining to SSC safety 

significance is considered by a panel 

(referred to as either an expert panel or 

an integrated decision-making panel 

(IDP)) that has the expertise and 

capabilities for making a sound decision 

regarding the SSC’s categorization, and 

that the assembled information is 

considered in a manner that ensures the 

Commission’s criteria for risk-informed 

applications are satisfied (i.e., defensein-depth is maintained, reasonable 

confidence that safety margins are 

maintained, reasonable confidence that 

any risk increase is small, and a 

monitoring and performance assessment 


ER22NO04.000</GPH>
68018 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

strategy is used). This process enables 

SSCs to be placed in the correct RISC 

category so that the appropriate 

treatment requirements will be applied 

commensurate with the SSC’s safety 

significance. A safety significant SSC is 

an SSC that performs a safety significant 

function as defined in § 50.69. The rule 

requires that SSC safety significance be 

determined using quantitative 

information from a PRA that reasonably 

represents the as-built, as-operated, 

current plant configuration, and which 

at a minimum covers internal events at 

full power. The categorization process 

must address both internal events and 

external events for all modes of 

operation and can use other available 

risk analyses and traditional engineering 

information to supplement the 

quantitative PRA results to address 

modes and events not within the scope 

of the PRA.

Section 50.69(c)(1)(i) ensures that the 

PRA is adequate for this application. 

Section 50.69(c)(1)(iii) requires that 

defense-in-depth is maintained as part 

of the categorization process. Section 

50.69(c)(1)(iv) requires that the revised 

treatment applied to RISC–3 SSCs be 

considered for its potential impact on 

risk. As an example, the Commission’s 

position is that the containment and its 

systems are important in the 

preservation of defense-in-depth (in 

terms of both large early and large late 

releases). As part of maintaining 

defense-in-depth, a licensee must 

demonstrate that the function of the 

containment as a barrier (including 

fission product retention and removal) 

is not significantly degraded when SSCs 

that support the functions are moved to 

RISC–3. 

Section 50.69(c)(2) requires the risk 

insights and other traditional 

information to be evaluated by the IDP 

and this panel must be comprised of 

expert, plant-knowledgeable members 

whose expertise includes PRA, safety 

analysis, plant operation, design 

engineering, and system engineering. 

Because the IDP makes the final 

determination about the safety 

significance of an SSC, the Commission 

concludes that the requirements in 

§ 50.69(c)(2) are necessary for the 

composition of the panel to be 

experienced personnel who possess 

diverse knowledge and insights in plant 

design and operation and who are 

capable in the use of deterministic 

knowledge and risk insights to 

categorize SSCs. 

As mentioned previously, the § 50.69 

categorization process requires that 

available deterministic and probabilistic 

information pertaining to SSC safety 

significance be considered in the 

decision process. The information 

considered must reasonably reflect the 

as-built and as-operated plant so that 

the decisions are based upon correct 

information, leading to proper 

categorization. Where applicable, the 

information is to come from a PRA that 

is adequate for this application (i.e., 

categorization of SSC safety 

significance). From this perspective, the 

IDP decision process can be viewed as 

an extension of the previous process for 

determining SSC safety classification 

(i.e., safety-related or nonsafety-related), 

in that it is making use of relevant risk 

information that was not considered or 

not available when the SSCs were 

initially classified. The IDP makes the 

final determination of the safety 

significance of SSCs using a process that 

takes all this information into 

consideration, in a structured, 

documented manner. The structure 

provides consistency to decisions that 

may be made over time and the 

documentation gives both the licensee 

and the NRC the ability to understand 

the basis for the categorization decision, 

should questions arise at a later date. 

Section 50.69(c)(1)(ii) contains 

general requirements for consideration 

of SSCs, modes of operation, and 

initiating events not modeled in the 

PRA. As a result, the implementing 

guidance plays a significant role in 

effective implementation and bolsters 

the need for NRC review and approval 

of the categorization process before 

implementation. 

The PRA used to provide the risk 

information to the categorization 

process is required to be subjected to a 

peer review. The peer review focuses on 

the PRA’s completeness and technical 

adequacy for determining the 

importance of particular SSCs, 

including consideration of the scope, 

level of detail, and technical quality of 

the PRA model, the assumptions made 

in the development of the results, and 

the uncertainties that impact the 

analysis. This provides confidence that 

for IDP decisions that use PRA 

information, the results of the 

categorization process provide a valid 

representation of the risk importance of 

SSCs. 

Before a licensee may implement 

§ 50.69, the NRC must approve the 

categorization process through a license 

amendment. This is necessary because 

of the importance of the PRA and 

categorization process to successful 

implementation of the rule. This review 

and approval of the categorization 

process is a one-time, process approval 

(i.e., the approval is not restricted to a 

set of systems or structures, and can be 

applied to any system or structure in the 

plant and the licensee is not required to 

come back to the NRC for review of the 

categorization process provided that 

licensee remains within the scope of the 

NRC’s safety evaluation). The NRC’s 

review of the § 50.69 submittal will 

determine whether § 50.69 requirements 

are satisfied and consider the adequacy 

of the PRA; focusing on the results of 

the peer review and the actions taken by 

the licensee to address any peer review 

findings. The Commission has 

determined that a focused NRC review 

of the PRA is necessary because there 

are key assumptions and modeling 

parameters that can have a significant 

impact on the results so that NRC 

review of their adequacy for this 

application is considered necessary to 

verify that the overall categorization 

process will yield acceptable decisions. 

Section 50.69(c)(1)(iv) requires 

reasonable confidence that the increase 

in the overall plant core damage 

frequency (CDF) and large early release 

frequency (LERF) resulting from 

potential decreases in the reliability of 

RISC–3 SSCs as a result of the changes 

in treatment be small. The rule further 

requires the licensee (or applicant) to 

describe the evaluations to be performed 

to meet this requirement. As presented 

in RG 1.174, the NRC considers small 

changes to be relative and to depend on 

the current plant CDF and LERF (hence 

we also refer to ‘‘acceptably small’’

changes in other portions of this notice 

since small can be different for different 

plants with different baseline levels of 

risk). For plants with total baseline CDF 

of 10–4 per year or less, small means 

CDF increases of up to 10–5 per year 

and for plants with total baseline CDF 

greater than 10–4 per year, small means 

CDF increases of up to 10–6 per year. 

However, if there is an indication that 

the CDF may be considerably higher 

than 10–4 per year, the focus of the 

licensee should be on finding ways to 

decrease rather than increase CDF and 

the licensee may be required to present 

arguments as to why steps should not be 

taken to reduce CDF for the reduction in 

special treatment requirements to be 

considered. For plants with total 

baseline LERF of 10¥5 per year or less, 

small LERF increases are considered to 

be up to 10¥6 per year, and for plants 

with total baseline LERF greater than 

10–5 per year, small LERF increases are 

considered to be up to 10–7 per year. 

However, if there is an indication that 

the baseline CDF or LERF may be 

considerably higher than 10¥4 or 10¥5, 

respectively, the licensee either must 

find ways to reduce risk and present the 

arguments to the NRC staff before 

implementation of § 50.69, otherwise it 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68019

is likely that the NRC will deny the 

§ 50.69 application. This is consistent 

with the guidance in Section 2.2.4 of RG 

1.174. It should be noted that this 

allowed increase shall be applied to the 

overall categorization process, even for 

those licensees that will implement 

§ 50.69 in a phased manner. This means 

that the allowable potential increase in 

risk must be determined in a cumulative 

way for all SSCs being categorized 

under § 50.69. 

Section 50.69 is structured to 

maintain the design basis functional 

requirements of the plant. These 

requirements (that maintain design basis 

functional requirements) when 

considered in conjunction with the 

requirements to provide reasonable 

confidence that the potential change in 

risk is small (as previously discussed), 

also provide reasonable confidence that 

safety margins are maintained. 

Specifically, licensees are required to 

ensure with reasonable confidence that 

RISC–3 SSCs remain capable of 

performing their design basis functions 

and these SSCs must remain capable of 

performing their design basis function, 

e.g., by providing a reliability that is not 

significantly degraded, to provide 

reasonable confidence that any 

increases in CDF or LERF will be 

acceptably small. 

Section 50.69(c)(1)(iv) requires 

applicants and licensees to perform 

evaluations to assess the potential 

impact on risk from changes to 

treatment. Further, § 50.69(d)(2) requires 

that the treatment applied to RISC–3 

SSCs be consistent with the 

categorization process. For SSCs 

modeled in the PRA, the licensee or 

applicant might conduct a risk 

sensitivity study that assesses the 

impact of changes in SSC failure 

probabilities or reliabilities that might 

occur due to the revised treatment. For 

example, a licensee could increase the 

failure rates of RISC–3 SSCs by 

appropriate factors to provide insights 

into the potential changes in risk that 

might result from reduced treatment 

(e.g., reduced maintenance, testing, 

inspection, and quality assurance). For 

other SSCs, other types of evaluations 

would be used to provide the basis for 

concluding that the potential increase in 

risk would be small. Under 

§ 50.69(b)(2)(iv), a licensee will need to 

submit its basis supporting the 

evaluations that estimate the potential 

change in risk. A licensee is required by 

§ 50.69(b)(2)(iv) to consider potential 

effects of common-cause interaction 

susceptibility and potential impacts 

from known degradation mechanisms. 

The rule focuses on common-cause 

effects because significant increases in 

common-cause failures could invalidate 

the evaluations performed to show that 

any potential change in risk due to 

implementation of § 50.69 would be 

small. With respect to known 

degradation mechanisms, this is an 

acknowledgment that certain treatment 

requirements have evolved over time to 

deal with these mechanisms (e.g., use of 

particular inspection techniques or 

frequencies), and that when 

contemplating changes to treatment, the 

lessons from this experience are to be 

taken into account. 

For SSCs categorized by means other 

than PRA models, the licensee needs to 

provide a basis to conclude that any 

potential increase in risk that might 

result from reduced treatment would be 

small. These requirements are included 

in § 50.69 so that a licensee has a basis 

for concluding that the evaluations 

performed to provide reasonable 

confidence that only a acceptably small 

change in risk will result remain valid. 

In addition, the rule requires that 

implementation be performed for an 

entire system or structure and not for 

selected components within a system or 

structure. This required scope ensures 

that all safety functions associated with 

a system or structure are properly 

identified and evaluated when 

determining the safety significance of 

individual components within a system 

or structure and that the entire set of 

components that comprise a system or 

structure are considered and addressed. 

III.3.0 Treatment Requirements 

The final rule applies treatment 

requirements to SSCs commensurate 

with their safety significance.

III.3.1 RISC–1 and RISC–2 Treatment 

For SSCs determined by the IDP to be 

safety significant (i.e., RISC–1 and 

RISC–2 SSCs), § 50.69 maintains the 

current regulatory requirements (i.e., it 

does not remove any requirements from 

these SSCs) for special treatment. These 

current requirements are adequate for 

addressing design basis performance of 

these SSCs. Additionally, § 50.69(d)(1) 

requires that sufficient treatment be 

applied to support the credit taken for 

these SSCs for beyond design basis 

events. For example, in developing the 

PRA model, a licensee must determine 

the availability, capability, and 

reliability of RISC–1 and RISC–2 SSCs 

in performing specific functions under 

various plant conditions. These 

functions may be beyond the design 

basis for individual SSCs. Further, the 

conditions under which those functions 

are to be performed may exceed the 

design basis conditions for the 

applicable SSCs. Section 50.69(d)(1) 

requires the treatment applied to RISC–

1 and RISC–2 SSCs to be consistent with 

the performance credited in the 

categorization process. This includes 

credit with respect to prevention and 

mitigation of severe accidents. In some 

cases, licensees might need to enhance 

the treatment applied to RISC–1 or 

RISC–2 SSCs to support the credit taken 

in the categorization process, or 

conversely adjust the credit for 

performance of the SSC in the 

categorization process to reflect actual 

treatment practices and/or documented 

performance capability. In addition, 

§ 50.69(e) requires monitoring and 

adjustment of treatment processes or 

categorization decisions as needed 

based upon operational experience. 

III.3.2 RISC–3 Treatment 

Section 50.69(d)(2) imposes 

requirements that are intended to 

maintain RISC–3 SSC design basis 

capability. Although individually RISC–

3 SSCs are not significant contributors 

to plant safety, they do perform 

functions necessary to respond to 

certain design basis events of the 

facility. Thus, collectively, RISC–3 SSCs 

can be safety significant and as such, it 

is important to maintain their design 

basis functional capability. Maintenance 

of RISC–3 design basis functionality is 

important to ensure that defense-indepth and safety margins are 

maintained. As a result, § 50.69(d)(2) 

requires that licensees or applicants 

ensure with reasonable confidence that 

RISC–3 SSCs remain capable of 

performing their safety-related functions 

under design basis conditions, 

including seismic conditions and 

environmental conditions and effects 

throughout their service life. To support 

this requirement, § 50.69(d)(2) contains 

inspection, testing, and corrective action 

requirements, and in addition requires 

that the treatment of RISC–3 SSCs be 

consistent with the categorization 

process. The requirements are 

performance-based and give licensees 

the flexibility to implement treatment 

that they have determined is needed, 

commensurate with the low safety 

significance of the SSCs in order to 

provide reasonable confidence that their 

safety-related functional capability is 

maintained. In this context, ‘‘reasonable 

confidence’’ is a somewhat reduced 

level of confidence as compared with 

the relatively high level of confidence 

provided by the current special 

treatment requirements. These 

alternative treatment requirements for 

RISC–3 SSCs represent a relaxation of 

those special treatment requirements 

that are removed for RISC–3 SSCs by the 

rule. For example, the alternative 

68020 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

treatment requirements for RISC–3 SSCs 

in § 50.69 are less detailed than 

provided in the special treatment 

requirements and allow significantly 

more flexibility by licensees in treating 

RISC–3 SSCs. The Commission is 

allowing greater flexibility and a lower 

level of assurance to be provided for 

RISC–3 SSCs in recognition of their low 

individual safety significance and this 

recognition includes a consideration for 

the potential change in reliability that 

might occur when treatment is reduced 

from what had previously been required 

by the special treatment requirements. 

In implementing the rule 

requirements, licensees will need to 

obtain data or information sufficient to 

make a technical judgement that RISC–

3 SSCs will remain capable of 

performing their safety-related functions 

under design basis conditions, and to 

enable the licensee to take actions to 

restore equipment performance 

consistent with corrective action 

requirements included in the rule. 

Effective implementation of the 

treatment requirements should result in 

reasonable confidence that RISC–3 SSCs 

will perform their safety-related 

function under normal and design basis 

conditions. This level of confidence is 

both less than that associated with 

RISC–1 SSCs, which are subject to all 

special treatment requirements, and 

consistent with the low individual 

safety significance of RISC–3 SSCs. 

It is noted that changes that affect any 

non-treatment aspects of an SSC (e.g., 

changes to the SSC design basis 

functional requirements) are still 

required to be evaluated in accordance 

with other regulatory requirements, 

such as § 50.59. The Commission, in 

developing § 50.69, is drawing a 

distinction between treatment (managed 

through § 50.69) and design changes 

(managed through other processes, such 

as § 50.59). As previously noted, this 

rulemaking is only risk-informing the 

scope of special treatment requirements. 

The process and requirements 

established in § 50.69 do not extend to 

making changes to the design basis 

functional requirements of SSCs. 

III.3.3 RISC–4 Treatment 

Section § 50.69 does not impose any 

new treatment requirements on RISC–4 

SSCs. Instead, RISC–4 SSCs are simply 

removed from the scope of any 

applicable special treatment 

requirements identified in § 50.69(b)(1). 

This is justified in view of their low 

significance considering both safetyrelated and risk information. 

Requirements applicable to RISC–4 

SSCs not removed by § 50.69(b)(1) 

continue to apply. Any changes (beyond 

changes to special treatment 

requirements) must be made per 

existing design change control 

requirements including § 50.59, as 

applicable. 

III.4.0 Removal of RISC–3 and RISC–4 

SSCs From the Scope of Special 

Treatment Requirements 

Through the application of § 50.69, 

RISC–3 and RISC–4 SSCs are removed 

from the scope of the specific special 

treatment requirements listed in 

§ 50.69(b)(1). The special treatment 

requirements were originally imposed to 

provide a high level of assurance that 

safety-related SSCs would perform 

when called upon with high reliability. 

As previously noted, the requirements 

include extensive quality assurance 

requirements and qualification testing 

requirements, as well as inservice 

inspection and testing requirements. 

These requirements can be quite 

demanding and expensive, as indicated 

in the data provided in the regulatory 

analysis on procurement costs. The 

Commission concluded that, in light of 

the low individual safety significance of 

RISC–3 SSCs, it is unnecessary to have 

the same high level of assurance that 

they would perform as designed. This is 

because some increased likelihood of 

their individual failure can be tolerated 

without significant impact to safety. 

Thus, the Commission decided to 

remove the RISC–3 and RISC–4 SSCs 

from those detailed, specific 

requirements that provided the high 

level of assurance. However, the 

functional requirements for these SSCs 

remain. As an example, a RISC–3 

component must still be designed to 

withstand any harsh environment it 

would experience under a design basis 

event, but the NRC will not require that 

this capability be demonstrated by a 

qualification test. Further, the 

performance (and treatment) of these 

RISC–3 SSCs remain under regulatory 

control, but in a different way. Instead 

of the special treatment requirements, 

the Commission has set forth more 

general requirements by which a 

licensee is to maintain functionality. 

These requirements give the licensee 

more latitude in applying treatment to 

maintain the design basis functional 

capability of the RISC–3 SSCs. The more 

general requirements that the 

Commission is specifying for the RISC–

3 SSCs include inspection, testing, and 

corrective action, as a means of 

maintaining functionality. As discussed 

elsewhere in the SOC of this rule, the 

Commission concludes that the 

requirements in § 50.69 will maintain 

adequate protection of public health and 

safety if effectively implemented by 

licensees. Hence, implementation of 

§ 50.69 should result in a better focus 

for both the licensee and the regulator 

on issues that pertain to plant safety and 

is consistent with the Commission’s 

policy statement for the use of PRA. 

In some cases, the Commission 

concluded that the RISC–3 and RISC–4 

SSCs could be removed from the scope 

of specific special treatment 

requirements, while in other cases the 

Commission concluded that only partial 

removal was appropriate. Finally, there 

was a set of requirements initially 

identified as special treatment for which 

the Commission is not removing RISC–

3 and RISC–4 SSCs from their scopes. 

These requirements are discussed in 

Section III.4.10.

III.4.1 Reporting Requirements Under 

10 CFR Part 21 and § 50.55(e) 

Section 206 of the Energy 

Reorganization Act of 1974 (ERA) 

requires the directors and responsible 

officers of nuclear power plant licensees 

and firms supplying ‘‘components of 

any facility or activity * * * licensed or 

otherwise regulated by the 

Commission’’ to ‘‘immediately report’’

to the Commission if they have 

information that such facility, activity, 

or basic components supplied to such 

facility or activity either fails to comply 

with the AEA, or Commission rule, 

regulation, order or license ‘‘relating to 

substantial safety hazards,’’ or contains 

a ‘‘defect which could create a 

substantial safety hazard * * *’’ Id.,

paragraph (a). Congress adopted Section 

206 to ensure that individuals, and 

responsible directors and officers of 

licensees and firms supplying important 

components to nuclear power plants 

notify the NRC in a timely fashion of 

potentially significant safety problems 

or noncompliance with NRC 

requirements. The NRC then may assess 

the reported information and take any 

necessary regulatory action in a timely 

fashion to protect public health and 

safety or common defense and security. 

Congress did not include definitions for 

the terms, ‘‘components,’’ ‘‘basic 

components,’’ or ‘‘substantial safety 

hazard,’’ in Section 206, but instead 

directed the Commission to issue 

regulations defining these terms. 

The Commission’s regulations 

implementing Section 206 appear in 10 

CFR part 21 and § 50.55(e) for license 

holders and construction permit 

holders, respectively. The Commission 

established definitions of ‘‘basic 

component,’’ ‘‘defect,’’ and ‘‘substantial 

safety hazard’’ in part 21 on the premise 

that the deterministic regulatory 

paradigm embedded in the 

Commission’s regulations would 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68021

continue to be the appropriate basis for 

determining the safety significance of an 

SSC, and therefore, the extent of the 

reporting obligation under Section 206. 

This is most evident in the § 21.3 

definition of ‘‘basic component,’’ which 

is similar to the definition of ‘‘safetyrelated’’ SSCs in § 50.2 (originally 

embodied in § 50.49). Part 21 also 

recognizes that Congress did not intend 

that every potential noncompliance or 

‘‘defect’’ in a component raises such 

significant safety issues that the NRC 

must be informed of every identified or 

potential noncompliance or defect. 

Instead, Congress limited the Section 

206 reporting requirement to those 

instances of noncompliance and defects 

that represent a ‘‘substantial safety 

hazard.’’ Thus, part 21 limits the 

reporting requirement to instances of 

noncompliance and defects representing 

‘‘substantial safety hazard,’’ which part 

21 defines as:

A loss of safety function to the extent 

there is a major reduction in the degree 

of protection afforded to public health 

and safety for any facility or activity 

licensed, other than for export, pursuant 

to parts 30, 40, 50, 60, 61, 63, 70, 71, 

or 72 of this chapter.

Finally, Part 21 establishes that a 

licensee or vendor should ‘‘immediately 

report’’ potential noncompliance or 

defects to the NRC in a telephonic 

‘‘notification’’ (see § 21.3) within two (2) 

days of receipt of information 

identifying a noncompliance or defect 

in a basic component (see § 21.21(d)). In 

addition, part 21 requires that vendors/

suppliers of basic components must 

make notifications to purchasers or 

licensees of a reportable noncompliance 

or deviation within five (5) working 

days of completion of evaluations for 

determining whether noncompliance or 

deviation constitutes a substantial safety 

hazard (see § 21.21(b)). Thus, Part 21 

establishes a reporting scheme for 

immediate reporting of the most safety 

significant noncompliances and defects, 

as contemplated by Section 206 of the 

ERA. 

Section 50.69 substitutes a riskinformed 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 

§ 50.55(e) reporting under § 50.69 and 

whether changes to part 21 and/or 

§ 50.55(e) are necessary to ensure proper 

reporting of substantial safety hazards 

caused by these SSCs; and (2) the 

appropriate reporting obligations of 

licensees and vendors under § 50.69, 

and whether changes to part 21 and/or 

§ 50.55(e) are necessary to impose the 

intended reporting obligations on these 

entities under § 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 riskinformed approach embodied in § 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 § 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 properlycategorized RISC–1 SSCs in all 

likelihood fall within the Commission’s 

definition of ‘‘basic components’’ and 

are currently subject to part 21 and 

§ 50.55(e) provides confirmation that the 

Commission’s determination is prudent. 

Similarly, the Commission believes 

that SSCs categorized as RISC–4 should 

continue to be beyond the scope of, and 

not be subject to, part 21 and § 50.55(e). 

SSCs properly categorized as RISC–4 

have little or no risk significance. It is 

highly unlikely that any significant 

regulatory action would be taken by the 

NRC based upon information on defects 

or instances of noncompliance in RISC–

4 SSCs so reporting them serves no 

regulatory purpose. Again, the fact that 

SSCs properly categorized as RISC–4 do 

not otherwise fall within the definition 

of ‘‘basic component’’ and, therefore, are 

not subject to part 21 and § 50.55(e) 

provides some confirmation of the 

prudence of the Commission’s 

determination. 

Thus, the most problematic issue from 

the standpoint of regulatory coherence 

is determining the appropriate scope of 

reporting for RISC–2 and RISC–3 SSCs. 

For the following reasons, the 

Commission proposes that neither 

RISC–2 nor RISC–3 SSCs be subject to 

part 21 and § 50.55(e) reporting 

requirements. 

The Commission begins by 

considering the regulatory objective of 

Part 21 and § 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 § 21.3) so that the NRC 

may make a determination if such a 

safety hazard requires that immediate 

NRC regulatory action be taken at one or 

more nuclear power plants to ensure 

adequate protection to public health and 

safety or common defense and security. 

The second is to ensure that nuclear 

power plant licensees are immediately 

informed of a potentially significant 

noncompliance or defect in supplied 

components. This reporting allows a 

licensee using these components to 

immediately evaluate the 

noncompliance or defect to determine if 

a safety hazard exists at the plant and 

take timely corrective action as 

necessary. In both cases, the regulatory 

objective is limited to components that 

have the highest significance with 

respect to ensuring adequate protection 

to public health and safety and common 

defense and security and whose failure 

or lack of proper functioning could 

create an imminent safety hazard so that 

immediate evaluation of the situation 

and implementation of necessary 

corrective action is necessary to ensure 

adequate protection. In the context of a 

construction permit, the safety hazard is 

two-fold: 

(1) That a noncompliance or defect 

could be incorporated into construction 

where it could never be detected; and, 

(2) That a noncompliance or defect 

would, upon initial operation and 

without prior indications of failure, 

create a substantial safety hazard.

The Commission believes that the 

regulatory objectives embodied in part 

21 and § 50.55(e) reporting remain the 

same regardless of whether the nuclear 

power plant is operating under the 

existing, deterministic regulatory system 

or the alternative, risk-informed system 

embodied in § 50.69. In both cases, the 

reporting scheme should focus on 

immediate reporting to the NRC and 

licensee of potentially significant 

noncompliances and defects that could 

create a substantial safety hazard 

requiring immediate evaluation and 

corrective action to ensure continuing 

adequate protection. Accordingly, in 

determining whether RISC–2 and RISC–

3 SSCs should be subject to part 21 

reporting, the Commission assessed 

whether failure or malfunction of these 

SSCs could reasonably lead to a safety 

hazard so that immediate evaluation of 

the situation and implementation of 

necessary corrective action is necessary 

to ensure adequate protection. 

For RISC–2 SSCs, the Commission 

does not believe their failure or 

malfunction could reasonably lead to a 

safety hazard so that immediate licensee 

and NRC evaluation of the situation and 

implementation of necessary corrective 

action is necessary to ensure adequate 

68022 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

2 In Generic letter 88–20, dated November 23, 

1988, licensees were requested to perform 

individual plant examinations to identify plantspecific 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 costbeneficial improvements to reduce risk. In 

supplement 4 to the generic letter dated June 28, 

1991, this request was extended to include external 

events (e.g., earthquakes, fires, floods). The NRC 

staff reviewed the plant-specific responses and 

prepared a staff evaluation report on each submittal. 

Further, the set of results were presented in 

NUREG–1560, IPE Program: Perspectives on Reactor 

Safety and Plant Performance. A similar report on 

IPEEE results was issued as NUREG–1742. In 

addition, as discussed in SECY–00–0062, the staff 

has conducted IPE follow-up activities with owners 

groups and licensees to confirm that identified 

improvements have been implemented and if any 

other actions were warranted.

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 § 50.69 determines the relative 

significance of SSCs, with those in 

RISC–1 and RISC–2 being more 

significant than those in RISC–3 or 

RISC–4. This does not mean that any 

RISC–2 SSC would rise to the level of 

necessitating immediate action if defects 

were identified. 

RISC–1 SSCs are viewed as being of 

sufficient safety significance to require 

part 21 reporting. It is the capability 

provided by these RISC–1 SSCs for 

purposes of satisfying safety-related 

functional requirements that also leads 

to RISC–1 SSCs being safety-significant, 

as these are key functions in prevention 

and mitigation of severe accidents. 

Thus, RISC–1 SSCs are generally 

significant for a range of events and 

conditions and, as the primary means of 

accident prevention and mitigation, the 

Commission wants to continue to 

achieve the high level of quality, 

reliability, preservation of margins, and 

assurance of performance of current 

regulatory requirements. 

By contrast, RISC–2 SSCs are less 

important than RISC–1 SSCs because 

they do not play a role in prevention 

and mitigation of design basis events 

(i.e., the SSCs that assure the integrity 

of the reactor coolant boundary, the 

capability to shut down the reactor and 

maintain it in a safe shutdown 

condition, or the capability to prevent or 

mitigate the consequences of accidents 

that could result in potential offsite 

exposures comparable to the applicable 

exposure guidelines set forth in 

§ 50.34(a)(1) or § 100.11). For example, 

they are not part of the reactor 

protection system or engineered safety 

features that perform critical safety 

functions such as reactivity control, 

inventory control and heat removal. 

When viewed from a deterministic 

standpoint, RISC–2 SSCs are not 

considered to rise to the level of a 

potential substantial safety hazard. 

From the risk-informed perspective, 

SSCs may end up classified as RISC–2 

for a number of reasons. The 

classification might occur because: (1) 

they contribute to plant risk by 

initiating transients that could lead to 

severe accidents (if multiple failures of 

other mitigating SSCs were to occur); or 

(2) they can reduce risk by providing 

backup mitigation to RISC–1 SSCs in 

response to an event. 

The Commission recognizes that 

noncompliance by, or defects in, RISC–

2 SSCs, which could increase risk, such 

as by more frequent initiation of a 

transient, may appear to constitute a 

‘‘substantial safety hazard.’’ However, 

upon closer examination, the 

Commission believes otherwise. The 

risk significance of such ‘‘transientinitiating’’ RISC–2 SSCs depends upon 

their frequency of initiation, with 

resultant consequences depending upon 

the failure of multiple other components 

of varying types in different systems. 

Further, their risk significance, as 

identified by the categorization process, 

is a result of the reliability (failure rates) 

currently being achieved for these SSCs 

treated as commercial-grade 

components, which includes the 

possibility of noncompliances and 

defects. Because requirements on RISC–

2 SSCs are not being reduced, there is 

no reason to believe that their 

performance would degrade as a result 

of implementation of § 50.69. In fact, by 

better understanding of their safety 

significance, and through the added 

requirements in this rule for RISC–2 

SSCs to achieve consistency between 

their categorization and treatment, 

performance should, at a minimum, be 

maintained and in some cases, 

enhanced. As discussed in Sections III.3 

and III.5 of this rule, the Commission is 

imposing additional regulatory controls 

on RISC–2 SSCs to prevent their 

performance from degrading. In 

addition, the Commission is requiring: 

(1) that licensees evaluate treatment 

being applied for consistency with the 

performance credited in the 

categorization; (2) monitoring of the 

performance of these SSCs; (3) 

corrective actions; and (4) reporting 

when a loss of a safety significant 

function occurs. Thus, there are 

requirements for corrective action by the 

licensee if noncompliances involving 

these SSCs are identified. The 

Commission concludes that these 

requirements are sufficient to preclude 

the need for Part 21 reporting, because 

no RISC–2 SSC is so significant as to 

necessitate immediate Commission (or 

licensee) action. 

For RISC–2 SSCs that provide backup 

mitigation to RISC–1 SSCs, the 

Commission also finds it prudent and 

desirable from a risk-informed 

standpoint to provide an enhanced level 

of assurance that RISC–2 SSCs can 

perform their safety significant 

functions, but the failure or malfunction 

of these RISC–2 SSCs does not raise a 

concern about imminent safety hazards. 

Moreover, over the last several years, 

the current fleet of power reactors have 

been subjected to a number of risk 

studies, including NUREG–1150, and 

other generic and plant-specific reviews. 

While some safety improvements have 

been identified as a result of these 

reviews, none has been of such 

significance as to require immediate 

action. This essentially means that no 

SSCs categorized as RISC–2 would rise 

to the level of significance that their 

failure or lack of functionality would 

constitute a substantial safety hazard 

requiring immediate NRC regulatory 

action. For example, in the case of two 

key risk scenarios, Station Blackout and 

Anticipated Transient Without Scram, 

the Commission imposed regulatory 

requirements to reduce risk from these 

events. However, the rules were issued 

as cost-beneficial safety improvements. 

The Commission believes its conclusion 

about the relative significance of RISC–

2 SSCs is also supported by plantspecific risk studies, such as the 

Individual Plant Examination (IPE) and 

Individual Plant Examination of 

External Events (IPEEE),2 conducted to 

identify (and correct) any plant-specific 

vulnerabilities to severe accident risk. 

NRC’s review of the licensee submittals 

has not identified any situations 

requiring immediate action for 

protection of public health and safety. 

In addition, as part of license renewal 

reviews, the NRC reviews severe 

accident mitigation alternatives 

(SAMAs), to identify and evaluate plant 

design changes with the potential for 

improving severe accident safety 

performance. In the license renewals 

completed to date, only a few candidate 

SAMAs have been found to be costbeneficial (and none were considered 

necessary to provide adequate 

protection of public health and safety).

In light of risk assessments and 

actions that have already been 

implemented, the Commission believes 

there would be no SSCs categorized 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68023

3NUREG–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.

under § 50.69 as RISC–2 whose failure 

would represent a significant and 

substantial safety hazard so that 

immediate notification under Part 21 

and NRC regulatory action is required. 

Accordingly, the results of these risk 

assessments provide additional 

confidence to the Commission that Part 

21 requirements need not be imposed 

on RISC–2 SSCs. 

The Commission also considered if 

notification of component defects 

should be required from the perspective 

of other potentially-affected licensees. 

The set of SSCs that are RISC–2 would 

vary from site to site because it depends 

upon the specifics of plant design and 

operation, particularly for the balanceof-plant which typically differs more 

from plant to plant than does the 

nuclear steam supply portion. Further, 

the suppliers of these components 

would vary. Therefore, the specific type 

of notifications under Part 21, for the 

purposes of NRC assessment of generic 

implications of component defects and 

to assure notification of licensees with 

the same components in service, would 

not fulfill a useful regulatory function. 

The Commission notes that although 

Part 21 and § 50.55(e) (component 

defect) reporting will not be required for 

RISC–2 SSCs, § 50.69(g) contains 

enhanced reporting requirements 

applicable to loss of system function 

attributable to, inter alia, failure or lack 

of function of RISC–2 SSCs. This is 

discussed in greater detail in Section 

III.5. 

Therefore, because of the more 

supporting role that the RISC–2 SSCs 

play with respect to ensuring critical 

safety functions, a noncompliance or 

defect in a RISC–2 SSC would not result 

in a substantial safety hazard such that 

immediate licensee and NRC evaluation 

of the situation and implementation of 

corrective action is necessary to ensure 

adequate protection. Thus, the 

Commission believes that a 

noncompliance or defect in a RISC–2 

SSC does not constitute a substantial 

safety hazard for which reporting is 

necessary under Part 21. Accordingly, 

the Commission concludes that 

reporting requirements to comply with 

Section 206 of the ERA are not 

necessary for RISC–2 SSCs and that the 

scope of Part 21 and § 50.55(e) reporting 

requirements exclude RISC–2 SSCs.

The Commission also concludes that 

RISC–3 SSCs should not be subject to 

Part 21 and § 50.55(e) reporting. A 

failure of a properly-categorized RISC–

3 SSC should result in only a small 

change in risk and should not result in 

a major degradation of essential safetyrelated equipment (see NUREG–0302, 

Rev. 1) 3. As previously discussed, the 

body of regulatory requirements (i.e., 

the retained requirements and the 

requirements contained in this rule) are 

sufficient, if effectively implemented, so 

that simultaneous failures in multiple 

systems (as would be necessary to lead 

to a substantial safety hazard involving 

RISC–3 SSCs) would not occur. Further, 

the broad applicability of information 

from a single RISC–3 SSC that would be 

provided under Part 21 and § 50.55(e) 

reporting would be questionable 

because of the significant changes in 

treatment for RISC–3 SSCs allowed 

under § 50.69. Accordingly, the 

Commission concludes that RISC–3 

SSCs should not be subject to reporting 

requirements of Part 21 and § 50.55(e).

The Commission concludes that Part 

21 reporting requirements extend only 

to RISC–1 SSCs because they are 

important in ensuring public health and 

safety. RISC–2 SSCs are not subject to 

reporting because they play a lesser role 

than RISC–1 SSCs in protection of 

public health and safety and with the 

significant changes in treatment allowed 

under § 50.69, no regulatory purpose 

would be served by Part 21 reporting (as 

previously discussed). Individually, 

RISC–3 and RISC–4 SSCs have little or 

no risk significance and no regulatory 

purpose would be served by subjecting 

RISC–3 and RISC–4 SSCs to Part 21 and 

§ 50.55(e). 

The Commission does not believe that 

any changes to Part 21 or § 50.55(e) are 

necessary to accomplish its conclusions 

with respect to RISC–2 and RISC–3 

SSCs. The Commission believes this is 

consistent with the statutory 

requirements in Section 206 of the ERA. 

Section 206 does not contain any 

definition of ‘‘substantial safety 

hazard,’’ but contains a direction to the 

Commission to define this term by 

regulation. Nothing in the legislative 

history suggests that Congress had in 

mind a fixed and unchanging concept of 

‘‘substantial safety hazard’’ or that the 

term was limited to deterministic 

regulatory principles. Hence, the 

Commission has broad discretion and 

authority to determine the appropriate 

scope of reporting under Section 206. 

The Commission believes that the 

current definition of ‘‘substantial safety 

hazard’’ in § 21.3 is broadly written to 

permit the Commission to interpret it as 

applying, in the context of a riskinformed regulatory approach, only to 

RISC–1 SSCs. Section 50.69 embodies a 

risk-informed regulatory paradigm that 

is different in key respects from the 

Commission’s historical deterministic 

approach and applies the risk-informed 

approach to classifying a nuclear power 

plant’s SSCs according to the SSC’s risk 

significance. SSCs that are classified as 

RISC–1 are those that represent the most 

important SSCs from both a risk and 

deterministic standpoint: they perform 

the key functions of preventing, 

controlling, and mitigating accidents 

and controlling risk. Failure of RISC–1 

SSCs represent, from a risk-informed 

regulatory perspective, the most 

important and significant safety 

concerns (i.e., a ‘‘substantial safety 

hazard’’). Therefore, the Commission 

believes that, in the context of the riskinformed regulatory approach embodied 

in § 50.69, it is reasonable for the 

Commission to interpret ‘‘substantial 

safety hazard’’ as applying only to 

RISC–1 SSCs and that reporting under 

Section 206 may be limited to RISC–1 

SSCs. 

The Commission considered two 

alternative approaches for limiting the 

reporting requirements in Part 21 and 

§ 50.55(e) to RISC–1 SSCs: 

(1) Interpreting ‘‘basic component’’ to 

encompass a risk-informed view of what 

SSCs the term encompasses; and, 

(2) Including a second definition of 

‘‘basic component’’ in § 21.3, which 

would apply only to those portions of a 

plant that have been categorized in 

accordance with § 50.69 and would be 

defined as an SSC categorized as RISC–

1 under § 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 ‘‘safetyrelated SSC’’ in § 50.2. The § 50.2 

definition of ‘‘afety-related SSC’’ refers 

to the ability of the SSC to remain 

functional during ‘‘design basis events.’’

The term, ‘‘design basis events’’ in 

Commission practice has referred to the 

deterministic approach of defining the 

events and conditions (e.g., shutdown, 

normal operation, and accident) for 

which an SSC is expected to function 

(or not fail). Identification of design 

basis events is inherently different 

conceptually when compared to a riskVerDate jul<14>2003 20:18 Nov 19, 2004 Jkt 205001 PO 00000 Frm 00017 Fmt 4701 Sfmt 4700 E:\FR\FM\22NOR2.SGM 22NOR2
68024 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

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 

§ 21.3 could be misunderstood as a 

change to the reporting requirements for 

licensees who choose not to comply 

with § 50.69. 

III.4.1.2 Reporting Obligations of 

Vendors for RISC–3 SSCs 

The reporting requirements of Section 

206 apply to individuals, directors, and 

responsible officers of a firm 

constructing, owning, operating or 

supplying the basic components of any 

NRC-licensed facility or activity. 

Nuclear power plant licensees and 

nuclear power plant construction permit 

holders who are subject to reporting 

under Section 206, Part 21, and 

§ 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 § 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 § 50.55(e) for RISC–2, RISC–

3 or RISC–4 SSCs. 

Thus, a vendor who supplied a safetyrelated 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 

§ 50.55(e) reporting requirements. 

However, as a practical matter that 

vendor would likely continue to comply 

with Part 21 or § 50.55(e). Vendors are 

informed of their Part 21 or § 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 

§ 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 § 50.69, whether the SSCs 

that they provided under contract as 

safety-related are now categorized as 

RISC–3, thereby removing the vendor’s 

reporting obligation under Part 21 or 

§ 50.55(e). Failing to inform a vendor 

that a safety-related SSC that it provided 

is no longer subject to Part 21 or 

§ 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 § 50.55(e). 

To address the potential for 

unnecessary reporting under § 50.69, the 

Commission considered including a 

new requirement in either § 50.69 or 

Part 21 and § 50.55(e). The new 

provision would require the licensee or 

construction permit holder to inform a 

vendor that a safety-related SSC that it 

provided has been categorized as RISC–

3. After consideration, the Commission 

believes that it is unlikely that this 

provision would result in any great 

reduction in the potential scope of 

reporting by vendors. The NRC does not 

receive many Part 21 reports, so the 

overall reporting burden to be reduced 

may be insubstantial. Furthermore, the 

Commission believes that the proposal 

could cause confusion, inasmuch as a 

vendor may supply many identical 

components to a licensee/holder, with 

some of the items intended for use in 

SSCs categorized as RISC–3 and other 

items intended in nonsafety-related 

applications. A vendor would have 

some difficulty in determining whether 

the problem with the supplied SSC 

potentially affects the SSC categorized 

as RISC–3 (as opposed to the supplied 

SSC used in nonsafety-related 

applications). The Commission also 

believes there may be some value in 

notification of the NRC when defects are 

identified, as they may reveal issues 

about the quality processes or 

implications for basic components at 

other facilities. Finally, the NRC notes 

that the vendor has already been 

compensated by the licensee for the 

burden associated with Part 21 and 

§ 50.55(e) as part of the initial 

procurement process. For these reasons, 

the Commission is not adopting a 

provision in § 50.69, Part 21, or 

§ 50.55(e) requiring a licensee or 

construction permit holder to inform a 

vendor of safety-related SSCs that its 

SSCs have been categorized as RISC–3.

III.4.1.3 Criminal Liability Under 

Section 223.b. of the AEA 

As discussed earlier, Section 206 of 

the ERA authorizes the imposition of 

civil penalties for a licensee’s and 

vendor’s failure to report instances of 

noncompliance or defects in ‘‘basic 

components’’ that create a ‘‘substantial 

safety hazard.’’ However, in addition to 

the civil penalties authorized by Section 

206, criminal penalties may be imposed 

under Section 223.b. of the AEA on an 

individual director, officer, or employee 

of a firm that supplies components to a 

nuclear power plant, that knowingly 

and willfully violate regulations that 

results (or could have resulted) in a 

‘‘significant impairment of a basic 

component * * *.’’ Licensees, 

applicants, and vendors should note the 

difference in the definition of ‘‘basic 

component’’ in Part 21 versus the 

definition set forth in Section 223.b: 

For the purposes of this subsection, 

the term ‘‘basic component’’ means a 

facility structure, system, component or 

part thereof necessary to assure—

(1) The integrity of the reactor coolant 

pressure boundary; 

(2) The capability to shutdown the 

facility and maintain it in a safe 

shutdown condition; or 

(3) The capability to prevent or 

mitigate the consequences of accidents 

that could result in an unplanned offsite 

release of quantities of fission products 

in excess of the limits established by the 

Commission. 

The U.S. Department of Justice is 

responsible for prosecutorial decisions 

involving violations of Section 223.b. 

III.4.1.4 Posting Requirements 

Both AEA Section 223.b and ERA 

Section 206 require posting of their 

statutory requirements at the premises 

of all licensed facilities. This is 

implemented through 10 CFR parts 19 

and 21. 

As a result of implementation of 

§ 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 § 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 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68025

with postulated accidents is contained 

in GDC–4. Section 50.49 was written to 

provide specific programmatic 

requirements for a qualification program 

and documentation for electrical 

equipment, and thus, is a special 

treatment requirement. 

Section 50.49(b) imposes 

requirements on licensees to have an 

environmental qualification program 

that meets the requirements contained 

therein. It defines the scope of electrical 

equipment important to safety that must 

be included under the environmental 

qualification program. Further, this 

regulation specifies methods to be used 

for qualification of the equipment for 

identified environmental conditions and 

documentation requirements. 

RISC–3 and RISC–4 SSCs are removed 

from the scope of the requirements of 

§ 50.49 by § 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 § 50.49 for 

testing, documentation files and 

application of margins, are not 

necessary (for the reasons stated in 

Section III.4.0). The requirements in 

GDC–4 as they relate to RISC–3 and 

RISC–4 SSCs, and the design basis 

requirements for these SSCs, including 

the environmental conditions such as 

temperature and pressure, remain in 

effect. Thus, these SSCs must continue 

to remain capable of performing their 

safety-related functions under design 

basis environmental conditions. 

III.4.3 Section 50.55a(f), (g), and (h)

Codes and Standards 

Section 50.69(b)(2)(iv) removes RISC–

3 SSCs from the scope of certain 

provisions of § 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 

§ 50.55a. Each of the subsections being 

removed is discussed in the paragraphs 

below. 

Section 50.55a(f) incorporates by 

reference provisions of the ASME Code, 

as endorsed by NRC, that contains 

inservice testing requirements. These 

are special treatment requirements. 

Through this rulemaking, RISC–3 SSCs 

are removed from the scope of these 

requirements and instead are subject to 

the requirements in § 50.69(d)(2). For 

the reasons discussed in Section III.4.0, 

the Commission has determined that for 

low safety significant SSCs, it is not 

necessary to impose the specific 

detailed provisions of the Code, as 

endorsed by NRC, and these 

requirements can be replaced by the 

more ‘‘high-level’’ alternative treatment 

requirements, which allow greater 

flexibility to licensees in 

implementation. 

Section 50.55a(g) incorporates by 

reference provisions of the ASME Code, 

as endorsed by NRC, that contain the 

inservice inspection, and repair and 

replacement requirements for ASME 

Class 2 and Class 3 SSCs. The 

Commission will not remove the repair 

and replacement provisions of the 

ASME Code required by § 50.55a(g) for 

ASME Class 1 SSCs, even if they are 

categorized as RISC–3, because those 

SSCs constitute principal fission 

product barriers as part of the reactor 

coolant system or containment. For 

Class 2 and Class 3 SSCs that are shown 

to be of low safety significance and 

categorized as RISC–3, the additional 

assurance obtained from the specific 

provisions of the ASME Code is not 

considered necessary. However, the 

Commission has not removed the 

requirements for fracture toughness 

specified for ASME Class 2 and Class 3 

SSCs because fracture toughness is a 

significant design parameter for the 

material used to construct the SSC. 

Fracture toughness is a property of the 

material that prevents premature failure 

of an SSC at abrupt geometry changes, 

or at small undetected flaws. Adequate 

fracture toughness of SSCs is necessary 

to prevent common cause failures due to 

design basis events, such as 

earthquakes. 

Section 50.55a(h) incorporates by 

reference the requirements in either 

IEEE 279, ‘‘Criteria for Protection 

Systems for Nuclear Power Generating 

Stations,’’ or IEEE 603–1991, ‘‘IEEE 

Standard Criteria for Safety Systems for 

Nuclear Power Generating Stations.’’

Within these IEEE standards are special 

treatment requirements. Specifically, 

Sections 4.3 and 4.4 of IEEE 279 and 

Sections 5.3 and 5.4 of IEEE 603–1991 

contain quality and environmental 

qualification requirements. RISC–3 

SSCs are being removed from the scope 

of this special treatment requirement.

III.4.4 Section 50.65 Monitoring the 

Effectiveness of Maintenance 

The Commission is removing RISC–3 

and RISC–4 SSCs from the scope of the 

requirements of § 50.65 (except for 

paragraph (a)(4)). The basis for this 

removal is provided in Section III.4.0 

and the following discussion. 

Section 50.65, the Maintenance Rule, 

imposes requirements for licensees to 

monitor the effectiveness of 

maintenance activities for safety 

significant plant equipment to minimize 

the likelihood of failures and events 

caused by the lack of effective 

maintenance. Specifically, § 50.65 

requires the performance of SSCs 

defined in § 50.65(b) to either be 

monitored against licensee established 

goals in a manner sufficient to provide 

confidence that the SSCs are capable of 

fulfilling their intended functions, or 

demonstrated to be effectively 

controlled through the performance of 

appropriate preventative maintenance. 

The rule further requires that where 

performance does not match the goals, 

appropriate corrective action shall be 

taken. Included within the scope of 

§ 50.65(b) are SSCs that are relied upon 

to remain functional during design basis 

events or in emergency operating 

procedures and nonsafety-related SSCs 

whose failure could result in the failure 

of a safety function or cause a reactor 

scram or activation of a safety-related 

system. 

Sections 50.65(a)(1), (a)(2), and (a)(3) 

impose action requirements; thus, they 

are special treatment requirements. 

Upon implementation of § 50.69, a 

licensee is not required to apply 

maintenance rule monitoring, goal 

setting, corrective action, alternate 

demonstration, or periodic evaluation 

treatments required by § 50.65(a)(1), 

(a)(2), and (a)(3) to RISC–3 and RISC–4 

SSCs. The rule includes provisions for 

a licensee to use performance 

information to feedback into its 

processes to adjust treatment (or 

categorization) when results so indicate 

in § 50.69(e)(3). However, this 

requirement does not require the 

specific monitoring and goal setting as 

required in § 50.65, in consideration of 

the lower safety significance of these 

SSCs. 

RISC–1 and RISC–2 SSCs that are 

currently within the scope of § 50.65(b) 

remain subject to existing maintenance 

rule requirements. Furthermore, 

§ 50.69(e)(2) requires additional 

monitoring, evaluation and appropriate 

action for these SSCs. 

The removal of RISC–3 and RISC–4 

SSCs from the scope of requirements 

does not include § 50.65(a)(4), which 

contains requirements to assess and 

manage the increase in risk that may 

result from maintenance activities. The 

requirements in § 50.65(a)(4) remain in 

effect. Section 50.65(a)(4) already 

includes provisions by which a licensee 

can limit the scope of the assessment 

required to SSCs that a risk-informed 

evaluation process has shown to be 

significant to public health and safety. 

Thus, there is no need to revise the 

requirements to permit a licensee to 

apply requirements commensurate with 

SSC safety-significance. 

68026 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

III.4.5 Sections 50.72 and 50.73

Reporting Requirements 

This rule removes the requirements in 

§ 50.72 and § 50.73 for RISC–3 and 

RISC–4 SSCs. Sections 50.72 and 50.73 

contain requirements for licensees to 

report events involving certain SSCs. 

These reporting requirements are 

special treatment requirements. The 

NRC requires event reports in part so 

that it can follow-up on corrective 

action for these circumstances. Through 

this rulemaking, the Commission is 

removing RISC–3 and RISC–4 SSCs 

from the scope of these requirements. 

The broad applicability of information 

obtained under § 50.72 and § 50.73 for 

RISC–3 SSCs would be questionable 

because of the significant changes in 

treatment allowed under § 50.69 (see the 

similar discussion for Part 21 in Section 

III.4.1.1). Therefore, the Commission 

does not consider the burden associated 

with reporting events or conditions only 

affecting these SSCs to be warranted. 

III.4.6 10 CFR Part 50, Appendix B 

Quality Assurance Requirements 

This rule removes RISC–3 and RISC–

4 SSCs from the scope of requirements 

in Appendix B to 10 CFR part 50. 

Appendix B contains requirements for a 

quality assurance program meeting 

specified attributes. The intent of 

Appendix B to 10 CFR part 50, and the 

complementary regulations, is to 

provide quality assurance requirements 

for the design, construction, and 

operation of nuclear power plants. The 

quality assurance requirements of 

Appendix B are to provide adequate 

confidence that an SSC will perform 

satisfactorily in service. These 

requirements were developed to be 

applied to safety-related SSCs. In the 

implementation of Appendix B, a 

licensee is bound to detailed and 

prescriptive quality requirements to 

apply to activities affecting those SSCs. 

As such, these requirements meet the 

Commission’s definition of special 

treatment requirements. These 

requirements are removed from 

application to RISC–3 and RISC–4 SSCs 

because their low individual safety 

significance does not warrant the level 

of quality requirements that currently 

exist with Appendix B. 

III.4.7 10 CFR Part 50, Appendix J 

Containment Leakage Testing 

Section 50.69(b)(1)(x) removes a 

subset of RISC–3 and RISC–4 SSCs from 

the scope of the requirements in 

Appendix J to 10 CFR part 50 that 

pertain to containment leakage testing. 

Specifically, RISC–3 and RISC–4 SSCs 

that meet specified criteria in 

§ 50.69(b)(1)(x) are removed from the 

scope of the requirements for Type B 

and Type C testing. It is important to 

note that this removes only the 

Appendix J leakage testing requirements 

from these SSCs. These SSCs must still 

be capable of performing their design 

basis functions (i.e., to close or isolate 

containment). The basis for the removal 

of the Appendix J leakage testing 

requirements follows. 

One of the conditions of all operating 

licenses for water-cooled power reactors 

as specified in § 50.54(o), is that primary 

reactor containments shall meet the 

containment leakage test requirements 

set forth in Appendix J to 10 CFR part 

50. These test requirements provide for 

preoperational and periodic verification 

by tests of the leak-tight integrity of the 

primary reactor containment, and 

systems and components that penetrate 

containment of water-cooled power 

reactors and establish the acceptance 

criteria for these tests. As such, these 

tests are special treatment requirements. 

The purposes of the tests are to assure 

that: 

(1) Leakage through the primary 

reactor containment, or through systems 

and components penetrating primary 

containment, shall not exceed allowable 

leakage rate values as specified in the 

technical specifications; and 

(2) Periodic surveillance of reactor 

containment penetrations and isolation 

valves is performed so that proper 

maintenance and repairs are made 

during the service life of the 

containment, and systems and 

components penetrating primary 

containment. 

Appendix J includes two options; 

Option A and Option B. Option A 

includes prescriptive requirements 

while Option B identifies performancebased requirements and criteria for 

preoperational and subsequent periodic 

leakage rate testing. A licensee may 

choose either option for meeting the 

requirements of Appendix J. 

The discussion contained in 

Appendix J to 10 CFR part 50 can be 

divided into two categories. Parts of 

Appendix J contain testing 

requirements. Other parts contain 

information, such as definitions or 

clarifications, necessary to explain the 

testing requirements. A review of 

Appendix J did not identify any 

technical requirements other than those 

describing the methods of the required 

testing. Therefore, Appendix J was 

considered to be, in its entirety, a 

special treatment requirement.

Although the 1995 revision to 

Appendix J was characterized as riskinformed, the changes were not as 

extensive as those expected by inclusion 

of Appendix J within the scope of 

§ 50.69. The 1995 revision to Appendix 

J primarily decreased testing 

frequencies, whereas risk-informing the 

scope of SSCs that are subject to 

Appendix J testing removes some 

components from testing (i.e., to the 

extent that defense-in-depth is 

maintained in accordance with the riskinformed categorization process). 

III.4.7.1 Types of Tests Required by 

Appendix J 

Appendix J testing is divided into 

three types: Type A, Type B, and Type 

C. Type A tests are intended to measure 

the primary reactor containment overall 

integrated leakage rate after the 

containment has been completed and is 

ready for operation and at periodic 

intervals thereafter. Type B tests are 

intended to detect local leaks and to 

measure leakage across each pressurecontaining or leakage-limiting 

boundary. Primary reactor containment 

penetrations required to be Type B 

tested are identified in Appendix J. 

Type C tests are intended to measure 

containment isolation valve (CIV) 

leakage rates. The containment isolation 

valves required to be Type C tested are 

identified in Appendix J. 

III.4.7.2 Reduction in Scope for 

Appendix J Testing 

Type A Testing: The Commission is 

not changing the Type A testing 

requirements of Appendix J. 

Type B Testing: The Commission is 

not changing the Type B testing 

requirements for air lock door seals, 

including door operating mechanism 

penetrations that are part of the 

containment pressure boundary and 

doors with resilient seals or gaskets, 

except for seal-welded doors. However, 

the Commission concludes that Type B 

testing is not necessary for other 

penetrations that are determined to be of 

low safety significance and that meet 

one or both of the following criteria: 

1. Penetrations pressurized with the 

pressure being continuously monitored. 

2. Penetrations are 1 inch nominal 

size or less. 

Type C Testing: The Commission 

concludes that Type C testing is not 

necessary for valves that are determined 

to be of low safety significance and that 

meet one or more of the following 

criteria: 

1. The valve is required to be open 

under accident conditions to prevent or 

mitigate core damage events. 

2. The valve is normally closed and in 

a physically closed, water-filled system. 

3. The valve is in a physically closed 

system whose piping pressure rating 

exceeds the containment design 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68027

pressure rating and is not connected to 

the reactor coolant pressure boundary. 

4. The valve size is 1-inch nominal 

pipe size or less. 

The Commission has made a 

determination that the size specified in 

§ 50.69(b)(x) and identified above is 

acceptable. At this time, the NRC has 

not determined that a larger size is 

acceptable for application to § 50.69, nor 

has the NRC received such a proposal. 

At this time, for the Commission to 

entertain a larger penetration/CIV size, 

and subsequently revise the rule 

language to reflect any such review 

(assuming that such a size is acceptable) 

would likely cause the NRC to re-notice 

§ 50.69 for stakeholder comment. 

Licensees and applicants are free to 

pursue exemptions (to § 50.69(b)(x)) to 

this criteria if they conclude a larger 

penetration opening can be justified for 

their containment design. If such a 

proposal is ultimately reviewed and 

accepted, and can be applied 

generically, the NRC will consider a 

revision to § 50.69 to reflect the new 

criteria. 

III.4.7.3 Basis for Reduction of Scope 

The first category of penetrations 

which are excluded from Type B testing 

are penetrations that are pressurized 

with the pressures in the penetrations 

being continuously monitored by 

licensees. This monitoring would detect 

significant leakage from the 

penetrations. The monitoring of the 

pressures in the penetrations, in 

conjunction with the requirements for 

RISC–3 SSCs (including taking 

corrective action when an SSC fails), 

ensures with reasonable confidence, 

without the need for Type B testing, that 

these penetrations are functional. 

The second category of penetrations 

excluded from Type B testing are 

penetrations that are 1 inch nominal 

size or less. These penetrations do not 

contribute to large early releases. 

Accordingly, the failure of such 

penetrations does not contribute in a 

significant way to safety or increased 

risk. The Commission concludes that 

such penetrations will not be subject to 

Type B testing. 

Regarding Type C containment 

leakage testing, the Commission finds 

that for the four categories of 

containment isolation valves identified 

in § 50.69(b)(1)(x), the removal of Type 

C testing requirements is reasonable 

because even without Type C testing, 

the probability of significant leakage 

during an accident (i.e., leakage to the 

extent that public health and safety is 

affected) is small. 

Appendix J to 10 CFR part 50 deals 

only with leakage rate testing of the 

primary reactor containment and its 

penetrations. It assumes that CIVs are in 

their safe position. No failure is 

assumed that causes the CIVs to be open 

when they are supposed to be closed. 

The valve would be open if needed to 

transmit fluid into or out of containment 

to mitigate an accident or closed if not 

needed for this purpose. For purposes of 

this evaluation, it is assumed that an 

open valve is capable of being closed. 

The licensee or applicant implementing 

§ 50.69 must apply treatment to RISC–

3 CIVs that ensures with reasonable 

confidence that those valves are capable 

of performing their safety-related 

function to close under design basis 

conditions. Testing to ensure the 

capability of CIVs to reach their safe 

position is not within the scope of 

Appendix J and as such is not within 

the scope of this evaluation. Therefore, 

the valves addressed by this evaluation 

are considered to be closed, but may be 

leaking. The increase in risk due to 

these SSCs being removed from the 

scope of Appendix J requirements is 

negligible. 

The acceptability of the removal of 

Appendix J leakage testing for the RISC–

3 CIVs is based on the assumption that 

those valves are capable of achieving the 

full seated position by means of the 

actuator. Therefore, even though a 

RISC–3 CIV might be exempt from 

Appendix J leakage testing, the RISC–3 

CIV must meet the treatment 

requirements in § 50.69(d) to provide 

reasonable confidence that the CIV can 

perform its safety function (e.g., to 

close) under design basis conditions. 

Because it is likely that most CIVs will 

be categorized as RISC–3, the licensee or 

applicant must evaluate the proposed 

change in the treatment of RISC–3 CIVs 

to ensure that defense-in-depth is 

maintained by ensuring with reasonable 

confidence that the RISC–3 CIVs are 

capable of performing their safetyrelated functions under design basis 

conditions. Although the licensee or 

applicant is allowed flexibility in 

addressing this issue, the rule requires 

that the licensee or applicant ensure 

with reasonable confidence the 

capability of RISC–3 CIVs to perform 

their safety functions to maintain 

defense-in-depth as discussed in RG 

1.174. 

Past studies (e.g., NUREG–1150, 

‘‘Severe Accident Risks: An Assessment 

for Five U.S. Nuclear Power Plants; 

Final Summary Report,’’ dated 

December 1990) show that the overall 

reactor accident risks are not sensitive 

to variations in containment leakage 

rate. This is because reactor accident 

risk is dominated by accident scenarios 

in which the containment either fails or 

is bypassed. These very low probability 

scenarios dominate predicted accident 

risks due to their high consequences. 

The Commission examined the effect 

of containment leakage on risk in more 

detail as part of the Appendix J to 10 

CFR part 50, Option B, rulemaking. The 

results of these studies are applicable to 

this evaluation. NUREG–1493, 

‘‘Performance-Based Containment LeakTest Program,’’ dated September 1995, 

calculated the containment leakage 

necessary to cause a significant increase 

in risk and found that the leakage rate 

must typically be approximately 100 

times the Technical Specification leak 

rate, La. It is improbable that even the 

leakage of multiple valves in the 

categories under consideration would 

exceed this amount. Operating 

experience shows that most measured 

leaks are much less than 100 times La. 

A more direct estimate of the increase 

in risk for the revision to Appendix J 

can be obtained from the Electric Power 

Research Institute (EPRI) report TR–

104285, ‘‘Risk Impact Assessment of 

Revised Containment Leak Rate Testing 

Intervals,’’ dated August 1994. This 

report examined the change in the 

baseline risk (as determined by a plant’s 

IPE risk assessment) due to extending 

the leakage rate test intervals. For the 

pressurized water reactor (PWR) large 

dry containment examined in the EPRI 

report, for example, the percent increase 

in baseline risk from extending the Type 

C test interval from 2 years to 10 years 

was less than 0.1 percent. While this 

result was for a test interval of 10 years 

vs. the current proposal to do no more 

Type C testing of the subject valves for 

the life of a plant, the analysis may 

reasonably apply to this situation 

because it contains several conservative 

assumptions that offset the 10-year time 

interval. These assumptions include the 

following: 

1. The study used leakage rate data 

from operating plants. Any leakage over 

the plant’s administrative leakage limit 

was considered a leakage failure. An 

administrative limit is a utility’s 

internal limit and does not imply 

violation of any Appendix J limits. 

Therefore, the probability of a leakage 

failure is overestimated.

2. Failure of one valve to meet the 

administrative limit does not imply that 

the penetration would leak because 

containment penetrations typically have 

redundant isolation valves. While one 

valve may leak, the other valve may 

remain leak-tight. The study assumed 

that failure of one valve in a series failed 

the penetration. Therefore, the 

probability of a penetration leak is 

overestimated. 

68028 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

3. The analysis assumed possible 

leakage of all valves subject to Type C 

testing, not just those subject to the 

relief per § 50.69. 

According to this analysis, the 

removal of SSCs from the scope of 

Appendix J requirements does not have 

a significant effect on risk. The NUREG–

1493 analysis shows that the amount of 

leakage necessary to significantly 

increase risk is two orders of magnitude 

greater than a typical Technical 

Specification leakage rate limit. 

Therefore, the risk to the public will not 

significantly increase due to the relief 

from the requirements of Appendix J to 

10 CFR part 50. 

III.4.8 Appendix A to 10 CFR Part 100 

(and Appendix S to 10 CFR Part 50 

(Seismic Requirements)) 

Section 50.69(b)(1)(xi) removes RISC–

3 and RISC–4 SSCs from the 

requirement in Appendix A to 10 CFR 

part 100 to demonstrate that SSCs are 

designed to withstand the safe 

shutdown earthquake (SSE) by 

qualification testing or specific 

engineering methods. GDC–2 requires 

that SSCs ‘‘important to safety’’ be 

capable of withstanding the effects of 

natural phenomena, such as 

earthquakes. The requirements of 10 

CFR part 100 pertain to reactor site 

criteria and Appendix A addresses 

seismic and geologic siting criteria used 

by the Commission to evaluate the 

suitability of plant design bases 

considering these characteristics. 

Sections VI(a)(1) and (2) of Appendix A 

to 10 CFR part 100 address the 

engineering design for the SSE and 

operating basis earthquake (OBE), 

respectively. Section 50.69 excludes 

RISC–3 and RISC–4 SSCs from the 

scope of the requirements of Sections 

VI(a)(1) and (2) of Appendix A to 10 

CFR part 100, only to the extent that the 

rule requires testing and specific types 

of analyses to demonstrate that safetyrelated SSCs are designed to withstand 

the SSE and OBE. It is only these 

aspects of Appendix A to 10 CFR part 

100 that are considered special 

treatment requirements. As discussed in 

Section III.4.0 of this rulemaking, 

because of the low individual safety 

significance of the RISC–3 and RISC–4 

SSCs, the additional assurance provided 

by qualification testing (or specific 

methods of analysis) is not considered 

necessary. 

Appendix A to part 100 is applicable 

for current operating reactors. The 

seismic design requirements are set 

forth in Appendix S to part 50 for new 

plant applications. The NRC has 

determined that Appendix S does not 

need to be included within the scope of 

§ 50.69 because the wording of the 

requirements with respect to 

‘‘qualification’’ by testing or specific 

types of analysis is not present in 

Appendix S. Therefore, a revision to the 

regulations is not necessary to permit a 

licensee to implement means other than 

qualification testing or the specified 

methods to demonstrate SSC capability. 

III.4.9 Section 50.46a(b) Appendix B 

Requirements for Reactor Coolant 

System Vents 

The Commission established new 

requirements for combustible gas 

control in § 50.44 using risk insights and 

issued the revised rule on September 16, 

2003 (68 FR 54123). As part of the 

§ 50.44 rulemaking, portions of the old 

§ 50.44 were relocated to more 

appropriate regulations. In particular, 

requirements formerly located in § 50.44 

were relocated to § 50.46a(b) concerning 

the design of vents and associated 

controls, instruments, and power 

sources and the need for these 

components to conform to 10 CFR part 

50 Appendix B. This rule removes 

RISC–3 SSCs from the scope of 

Appendix B quality assurance 

requirements, as discussed in Section 

III.4.6. These same arguments apply to 

the requirements in § 50.46a(b) where 

Appendix B is being imposed on a 

specific set of components. As such, this 

rule removes the RISC–3 and RISC–4 

SSCs from the scope of Appendix B 

requirements contained in § 50.46a(b). 

This applies only to the requirements 

relating to Appendix B in § 50.46a(b); 

the remaining requirements of § 50.46a 

remain unchanged. 

III.4.10 Requirements Not Removed by 

§ 50.69(b)(1) 

In the following paragraphs, the 

Commission discusses certain rules that 

were considered as candidates for 

removal as requirements for RISC–3 and 

RISC–4 SSCs during development of 

this rulemaking. These rules were 

identified as candidate rules in SECY–

99–256. They are not part of this 

rulemaking for the reasons stated. 

III.4.10.1 Section 50.34 Contents of 

Applications 

Section 50.34 identifies the required 

information that applicants must 

provide in preliminary and final safety 

analysis reports. Because § 50.69 

contains the documentation 

requirements for licensees and 

applicants who choose to implement 

§ 50.69, and these requirements do not 

conflict with § 50.34, it is not necessary 

to revise § 50.34 to implement § 50.69. 

III.4.10.2 Section 50.36 Technical 

Specifications 

Section 50.36 establishes operability, 

surveillance, limiting conditions for 

operation and other requirements on 

certain SSCs. Because this rule specifies 

testing and related requirements, it was 

considered as a candidate special 

treatment rule. However, the 

Commission concluded that it was not 

appropriate to revise § 50.36 for several 

reasons. 

Currently, the NRC staff and the 

industry are developing risk-informed 

improvements to technical 

specifications. These improvements, or 

initiatives, are intended to maintain or 

improve safety while reducing 

unnecessary burden, and to bring 

technical specifications into congruence 

with the Commission’s other riskinformed regulatory requirements, in 

particular risk management 

requirements of the Maintenance Rule 

in 10 CFR 50.65(a)(4). Eight initiatives 

for fundamental improvements to the 

Standard Technical Specifications (TS) 

have been proposed. Two of the 

initiatives have been approved and 

offered to licensees for adoption, and six 

are being developed by the industry and 

NRC staff. All of the initiatives involve, 

to some prescribed degree, assessing 

and managing plant risk using a 

configuration risk management program 

consistent with and in some cases 

exceeding the requirements of the 

Maintenance Rule in 10 CFR 50.65. The 

two approved initiatives involve: 

permitting the extension of up to one 

surveillance interval of an inadvertently 

missed surveillance; and, permitting 

plant mode transitions with inoperable 

equipment, anticipating the imminent 

return of the equipment to operability. 

The six initiatives under development 

involve: shutting down to hot shutdown 

rather than cold shutdown to repair 

equipment; permitting the temporary 

extension of allowed outage times; 

permitting the determination of 

surveillance frequencies through the use 

of an approved methodology; permitting 

time to restore equipment operability 

rather than immediately shutting down; 

providing extended time to restore 

support systems to operability; and, 

revising the scope of technical 

specifications to include only on risk 

significant systems, which would 

require rulemaking. 

Improved standard TSs have already 

resulted in the relocation of 

requirements for less important SSCs to 

other documents. Given the ongoing 

regulatory efforts to risk-inform the TSs, 

it was not considered necessary to scope 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68029

§ 50.36 into § 50.69 as a special 

treatment requirement.

III.4.10.3 Section 50.44 Combustible 

Gas Control 

During the effort to identify candidate 

special treatment rules (refer to SECY–

99–256), certain provisions within 

§ 50.44 were identified as containing 

special treatment requirements in that 

they specified conformance with 

Appendix B for particular design 

features, specified requirements for 

qualification, and related statements. 

For proposed § 50.69, the Commission 

elected not to identify § 50.44 as a 

special treatment rule, and instead 

decided to wait on the outcome of the 

effort to risk inform § 50.44. The 

Commission subsequently rebaselined 

the requirements in § 50.44 using risk 

insights and issued the revised rule on 

September 16, 2003 (68 FR 54123). As 

a result, the NRC concludes that there 

is no need to include § 50.44 within the 

scope of § 50.69. However, as part of the 

September 16, 2003, rulemaking, 

portions of the old § 50.44 were 

relocated to more appropriate 

regulations. In particular, requirements 

were relocated to § 50.46a(b) concerning 

the design of vents and associated 

controls, instruments, and power 

sources and the need for these 

components to conform to 10 CFR Part 

50 Appendix B. Because this aspect of 

the relocated requirements is a special 

treatment requirement (and this same 

requirement was also identified in the 

old § 50.44 as being a special treatment 

requirement) it is now captured within 

the scope of § 50.69(b)(1) as discussed in 

Section III.4.9. 

III.4.10.4 Section 50.48 (Appendix R 

and GDC 3) Fire Protection 

Initially, fire protection requirements 

were considered to be within the scope 

of this rulemaking effort. There are 

augmented quality provisions applied to 

fire protection systems and these 

augmented quality provisions are 

considered special treatment 

requirements. However, these 

provisions are not contained in the 

Commission’s regulations and therefore 

a revision to the rules (i.e., to scope 

them into § 50.69) is not required to 

support a change (i.e., changes to these 

requirements can be made without a 

revision to the rules). Additionally, the 

Commission has issued a final rule that 

would allow licensees to voluntarily 

adopt National Fire Protection 

Association (NFPA)–805 requirements 

in lieu of other fire protection 

requirements. NFPA–805 sets forth 

requirements for establishing and 

implementing a risk-informed fire 

protection program. Inasmuch as the 

NRC has addressed fire protection in 

another rulemaking, fire protection 

requirements were not included in the 

scope of the § 50.69 rulemaking. 

III.4.10.5 Section 50.59 Changes, 

Tests, and Experiments 

There is no change is being made to 

§ 50.59 as a result of § 50.69, however, 

the Commission does not believe that a 

§ 50.59 evaluation need be performed 

when a licensee implements § 50.69 and 

thereby changes the special treatment 

requirements applied to RISC–3 and 

RISC–4 SSCs. Accordingly, § 50.69(f) 

contains language that removes the 

requirement for licensees to perform 

§ 50.59 evaluations for the changes in 

special treatment that stem from § 50.69 

implementation. The process of 

adjusting treatment for RISC–3 and 

RISC–4 SSCs does not need to be subject 

to § 50.59 because the rulemaking 

already provides the decision process 

for categorization and determination of 

revision to requirements resulting from 

the categorization. Because it is only in 

the area of treatment for RISC–3 and 

RISC–4 SSCs that might be viewed as 

involving a reduction in requirements, 

these are the only aspects for which this 

rule provision applies. As required by 

§ 50.69(f), the licensee or applicant will 

be required to update the FSAR 

appropriately to reflect incorporation of 

its treatment processes into the FSAR. 

However, it is important to recognize 

that changes that may affect any nontreatment 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 § 50.59. The 

Commission, in developing § 50.69, is 

drawing a distinction between treatment 

(managed through § 50.69) and design 

changes (managed through other 

processes such as § 50.59). As 

previously noted, this rulemaking is 

only risk-informing the scope of special 

treatment requirements. The process 

and requirements established in § 50.69 

do not extend to making changes to the 

non-treatment portion of the design 

basis of SSCs. 

III.4.10.6 Appendix A to 10 CFR Part 

50 General Design Criteria (GDC) 

The NRC has concluded that the GDC 

of Appendix A to 10 CFR Part 50 do not 

need to be revised because they specify 

design requirements and do not specify 

special treatment requirements. Because 

this rulemaking is not revising the nontreatment portion of the design basis of 

the facility, the GDC should remain 

intact and are not within the scope of 

§ 50.69. This subject is discussed in 

more detail in the NRC’s action on the 

South Texas exemption request, in 

which their request for exemption from 

certain GDCs was denied as being 

unnecessary to accomplish what was 

proposed (see Section IV.2.0). 

III.4.10.7 10 CFR Part 52 Early Site 

Permits, Standard Design Certifications 

and Combined Operating Licenses 

Part 52 cross-references regulations 

from other parts of Chapter 10 of the 

Code of Federal Regulations, most 

notably Part 50. Therefore, it was 

initially considered for inclusion in this 

rulemaking effort. However, the 

‘‘applicability’’ paragraph (§ 50.69(b)) 

makes clear that § 50.69 is available to 

applicants for, and holders of a facility 

license. Accordingly, there is no need to 

revise Part 52 to assure the availability 

of § 50.69. There are issues associated 

with Part 52 design certifications and 

these are currently excluded from the 

group of entities who may adopt the 

provisions of 50.69 as discussed in 

Section V.3.0. 

II.4.10.8 10 CFR Part 54 License 

Renewal 

10 CFR part 54, which sets forth the 

license renewal requirements for 

nuclear power reactors, was identified 

as a candidate special treatment 

requirement in SECY–99–256. The Part 

54 aging management requirements are 

special treatment requirements in that 

they provide assurance that SSCs will 

continue to meet their licensing basis 

requirements during the renewed 

license period. Section 54.4 explicitly 

defines the scope of the license renewal 

rule using the traditional deterministic 

approach. Part 54 imposes aging 

management requirements in § 54.21 on 

the scope of SSCs meeting § 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 § 50.69 treatment 

provides adequate aging management in 

accordance with § 54.21. The NRC staff 

suggested that no changes are necessary 

to Part 54 to implement § 50.69 either 

before renewing a licensing or after 

license renewal. 

The goal of the license renewal 

program is to establish a stable, 

predictable, and efficient license 

renewal process. The Commission 

believes that a revision of Part 54 at this 

time could have a significant effect on 

the stability and consistency of the 

processes established for preparation of 

license renewal applications and for 

NRC staff review. Further, as discussed 

68030 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

below, the Commission believes that the 

requirements in Part 54 are compatible 

with the § 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 § 50.69 for a facility 

that has already received a renewed 

license. Thus, Part 54 requires no 

changes at this time. However, in the 

future, the Commission will consider 

whether revisions to the scope of Part 54 

are appropriate.

The 1995 amendment to Part 54 

excluded active components to ‘‘reflect 

a greater reliance on existing licensee 

programs that manage the detrimental 

effects of aging on functionality, 

including those activities implemented 

to meet the requirements of the 

maintenance rule’’ (May 8, 1995; 60 FR 

22471). Although § 50.69 removes 

RISC–3 components from the scope of 

the maintenance rule requirements in 

§ 50.65(a)(1), (a)(2), and (a)(3), a licensee 

is required under § 50.69(d)(2) to 

provide confidence in the capability of 

RISC–3 SSCs to perform their safetyrelated functions under design-basis 

conditions when challenged. The SOC 

for Part 54 also indicated the 

Commission’s recognition that risk 

insights could be used in evaluating the 

robustness of an aging management 

program (May 8, 1995; 60 FR 22468). 

III.4.10.9 Other Requirements 

In the ANPR and related documents, 

the NRC staff and stakeholders 

suggested a number of other regulatory 

requirements that might be candidates 

for inclusion in § 50.69. These included 

§ 50.12 (exemptions), § 50.54(a), (p), and 

(q) (plan change control), and § 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 

§ 50.69. 

III.5.0 Feedback, Documentation, and 

Reporting Requirements 

The validity of the categorization 

process relies on ensuring that the 

performance and condition of SSCs 

continue to be maintained consistent 

with applicable assumptions. Changes 

in the level of treatment applied to an 

SSC might result in changes in the 

reliability of the SSCs credited in the 

categorization process. Additionally, 

plant changes, changes to operational 

practices, and plant and industry 

operational experience may impact 

categorization process results. 

Consequently, the rule contains 

requirements for updating the 

categorization and treatment processes 

when conditions warrant to assure that 

continued SSC performance is 

consistent with the categorization 

process and results. 

Specifically, the rule requires 

licensees to review the changes to the 

plant, operational practices, applicable 

plant and industry operational 

experience, and, as appropriate, update 

the PRA and SSC categorization. The 

review must be performed in a timely 

manner but no longer than once every 

two refueling outages. In addition, 

licensees are required to obtain 

sufficient information on SSC 

performance to verify that the 

categorization process and its results 

remain valid. For RISC–1 SSCs, much of 

this information may be obtained from 

present programs for inspection, testing, 

surveillance, and maintenance. 

However, for RISC–2 SSCs and for 

RISC–1 SSCs credited for beyond design 

basis accidents, licensees need to ensure 

that sufficient information is obtained. 

For RISC–3 SSCs, there is a relaxation 

of the requirements for obtaining 

information when compared to the 

applicable special treatment 

requirements. However, sufficient 

information still needs to be obtained. 

The rule requires considering 

performance data, determining if 

adverse changes in performance have 

occurred, and making the necessary 

adjustments so that desired performance 

is achieved so that the evaluations 

conducted to meet § 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 

§ 50.69. For safety significant SSCs, all 

current requirements continue to apply 

and, as a consequence, Appendix B 

corrective action requirements are 

applied to the design basis aspects of 

RISC–1 SSCs to ensure that conditions 

adverse to quality are corrected. For 

both RISC–1 and RISC–2 SSCs, 

requirements are included in 

§ 50.69(e)(2) for monitoring and for 

taking action when SSC performance 

degrades. 

When a licensee or applicant 

determines that a RISC–3 SSC does not 

meet its established acceptance criteria 

for performance of design basis 

functions, the rule requires that a 

licensee perform timely corrective 

action (§ 50.69(d)(2)(ii)). Further, as part 

of the feedback process, the review of 

operational data may reveal 

inappropriate credit for reliability or 

performance and a licensee would need 

to re-visit the findings made in the 

categorization process or modify the 

treatment for the applicable SSCs 

(§ 50.69(e)(3)). These provisions would 

then restore the facility to the 

conditions that were considered in the 

categorization process and would also 

restore the capability of the SSCs to 

perform their functions. 

Section 50.69(f) requires the licensee 

or applicant to document the basis for 

its categorization of SSCs before 

removing special treatment 

requirements. Section 50.69(f) also 

requires the licensee or applicant to 

update the final safety analysis report to 

reflect which systems have been 

categorized. 

Finally, § 50.69(g) requires reporting 

of events or conditions that prevented, 

or would have prevented, a RISC–1 or 

RISC–2 SSC from performing a safety 

significant function. Because the 

categorization process has determined 

that RISC–2 SSCs are of safety 

significance, NRC is interested in 

reports about circumstances where a 

safety significant function was, or 

would have been, prevented because of 

events or conditions. This reporting will 

enable NRC to be aware of situations 

impacting those functions found to be 

significant under § 50.69, so that NRC 

can take any actions deemed 

appropriate. 

Properly implemented, these 

requirements ensure that the validity of 

the categorization process and results 

are maintained throughout the 

operational life of the plant. 

III.6.0 Implementation Process 

Requirements 

The Commission is making the 

provisions of § 50.69 available to both 

applicants for licenses and to holders of 

facility licenses for light-water reactors. 

The rule is limited to light-water 

reactors because the Commission does 

not yet have substantial experience or 

information sufficient to develop riskinformed requirements applicable to 

non-light water reactors. Consequently, 

the technical aspects of the rule (e.g.,

providing reasonable confidence that 

risk increases are small), including the 

implementation guidance, are specific 

to light-water reactor designs. 

Section 50.69 relies on a robust 

categorization process to provide 

reasonable confidence that the safety 

significance of SSCs is correctly 

determined. To ensure a robust 

categorization is employed, § 50.69 

requires the categorization process to be 

reviewed and approved by the NRC 

before implementation of § 50.69 by 

following the license amendment 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68031

process of § 50.90 or as part of the 

license application review. While 

detailed regulatory guidance has been 

developed to provide guidance for 

implementing categorization consistent 

with the rule requirements, the 

Commission concluded that a prior 

review and approval was still necessary 

to enable the NRC staff to review the 

scope and quality of the plant-specific 

PRA; taking into account industry peer 

review results. The NRC staff will also 

review other evaluations and 

approaches that may be used, such as 

margins-type analyses, as well as 

examine any aspects of the proposed 

categorization process that are not 

consistent with the NRC’s regulatory 

guidance for implementing § 50.69. 

Thus, the rule requires that a licensee 

who wishes to implement § 50.69 

submit an application for license 

amendment to the NRC containing 

information about the categorization 

process and about the industry peer 

review process employed. An applicant 

would submit this information as part of 

its license application. The NRC will 

approve, by license amendment, a 

request to allow a licensee to implement 

§ 50.69 if it is satisfied that the 

categorization process to be used meets 

the requirements in § 50.69. 

NEI submitted a paper, ‘‘License 

Amendments: Analysis of Statutory and 

Legal Requirements’’ (NEI Analysis) in a 

July 10, 2002, letter to the Director of 

the Office of Nuclear Reactor Regulation 

(NRR). In this analysis, NEI contends 

that approval of a licensee’s/applicant’s 

request to implement § 50.69 need not 

be accomplished by a license 

amendment. NEI essentially argues that 

the rule does not increase the licensee’s 

operating authority, but merely provides 

a ‘‘different means of complying with 

the existing regulations * * *’’ Id., p.8. 

The Commission disagrees with this 

position, inasmuch as § 50.69 permits 

the licensee/applicant, once having 

obtained approval from the NRC, to 

depart from compliance with the 

‘‘special treatment’’ requirements set 

forth in those regulations delineated in 

§ 50.69. NEI also argues that the NRC’s 

review and approval of the SSC 

categorization process under § 50.69 is 

analogous to the review and approval 

process in Cleveland Electric 

Illuminating Co. (Perry Nuclear Power 

Plant, Unit 1), CLI–96–13, 44 NRC 315 

(1996), which the Commission 

determined did not require a license 

amendment. Unlike the Perry case, 

where the license already provided for 

the possibility of material withdrawal 

schedule changes and the governing 

ASTM standard set forth objective, nondiscretionary criteria for changes to the 

withdrawal schedule, § 50.69 does not 

contain these criteria for assessing the 

adequacy of the categorization process, 

PRA peer review results, and the basis 

for sensitivity studies. Hence, the NRC’s 

approval of a request to implement 

§ 50.69 will involve substantial 

professional judgment and discretion. 

The Commission does not agree with 

NEI’s assertion that the NRC’s approval 

of a request to implement § 50.69 may 

be made without a license amendment 

in accordance with the Perry decision.

The Commission does not believe it is 

necessary to perform a prior review of 

the treatment processes to be 

implemented for RISC–3 SSCs in lieu of 

the special treatment requirements. 

Instead, the NRC has developed § 50.69 

to contain requirements that ensure the 

categorization process is sufficiently 

robust to provide reasonable confidence 

that SSC safety significance is correctly 

determined; sufficient requirements on 

RISC–3 SSCs to provide a level of 

assurance that these SSCs remain 

capable of performing their design basis 

functions commensurate with their 

individual low safety significance; and 

requirements for obtaining information 

concerning the performance of these 

SSCs to help enable corrective actions to 

be taken before RISC–3 SSC reliability 

degrades beyond the values used in the 

evaluations conducted to satisfy 

§ 50.69(c)(1)(iv). The NRC concludes 

that compliance with these 

requirements, in conjunction with 

inspection of § 50.69 licensees, is a 

sufficient level of regulatory oversight 

for these SSCs. 

The Commission included 

requirements in the rule for 

documenting categorization decisions to 

facilitate NRC oversight of a licensee’s 

or applicant’s implementation of the 

alternative requirements. The rule also 

includes provisions to have the FSAR 

and other documents updated to reflect 

the revised requirements and progress 

in implementation. These requirements 

will allow the NRC and other 

stakeholders to remain knowledgeable 

about how a licensee is implementing 

its regulatory obligations as it 

transitions from past requirements to 

the revised requirements in § 50.69. As 

part of these provisions, the 

Commission has concluded that 

requiring evaluations under § 50.59 (for 

changes to the facility or procedures as 

described in the FSAR) or under 

§ 50.54(a) (for changes to the quality 

assurance plan) is not necessary for 

those changes directly related to 

implementation of § 50.69. For 

implementation of treatment processes 

for low safety significant SSCs, in 

accordance with the rule requirements 

contained in § 50.69, the Commission 

concludes that requiring further review 

if NRC approval might be required for 

these changes is an unnecessary burden. 

Thus, a licensee is permitted to make 

changes concerning treatment 

requirements that might be contained in 

these documents. The Commission is 

limiting this relief to changes directly 

related to implementation (with respect 

to treatment processes). Changes that 

affect any non-treatment aspects of an 

SSC (e.g., changes to the SSC design 

basis functional requirements) are still 

required to be evaluated in accordance 

with other regulatory requirements such 

as § 50.59. This rulemaking is only riskinforming the scope of special treatment 

requirements. The process and 

requirements established in § 50.69 do 

not extend to making changes to the 

non-treatment portion of the design 

basis. 

III.7.0 Adequate Protection 

The Commission concludes that 

§ 50.69 provides reasonable assurance of 

adequate protection of public health and 

safety because the principles listed 

below were used in the development of 

§ 50.69 and because these principles 

will continue to be employed in the 

NRC’s continuing regulatory oversight 

of § 50.69 implementation. Those 

principles are: 

(a) Reasonable confidence that the 

net increase in plant risk is small; 

(b) Defense-in-depth is maintained; 

(c) Reasonable confidence that safety 

margins are maintained; and 

(d) Monitoring and performance 

assessment strategies are used. 

These principles were established in 

RG 1.174, which provided guidance on 

an acceptable approach to risk-informed 

decision-making consistent with the 

1995 Commission policy on the use of 

PRA. Section 50.69 was developed to 

incorporate these principles, both to 

ensure consistency with Commission 

policy, and to ensure that the rule 

maintains adequate protection of public 

health and safety. 

The following discusses how § 50.69 

meets the four criteria, and as a result, 

maintains adequate protection of public 

health and safety. 

III.7.1 Net Increase in Risk Is Small 

Section 50.69(c) requires the use of a 

robust, risk-informed categorization 

process that ensures that all relevant 

information concerning the safety 

significance of an SSC is considered by 

a competent and knowledgeable panel 

who makes the final determination of 

the safety significance of SSCs. The NRC 

review and approval of the licensee’s 

68032 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

categorization process ensures that it 

meets the requirements of § 50.69(c) and 

that, as a result, the correct SSC safety 

significance is determined with high 

confidence. Correctly determining safety 

significance of an SSC provides 

confidence that special treatment 

requirements are only removed from 

SSCs with low individual safety 

significance and that these requirements 

continue to be satisfied for SSCs of 

safety significance. The rule requires 

that the potential net increase in risk 

from implementation of § 50.69 be 

assessed and that reasonable confidence 

be provided that this risk change is 

small. These requirements to provide 

reasonable confidence that the net 

change in risk is acceptably small as 

part of the categorization decision, in 

conjunction with the rule requirements 

for maintaining design basis functions 

and the processes noted below for 

feedback and adjustment over time, all 

contribute to preventing risk from 

increasing beyond the ranges that the 

NRC has considered to be appropriate as 

discussed in the RG 1.174 acceptance 

guidelines. As a result, these 

requirements are a contributing element 

for maintaining adequate protection of 

public health and safety. 

III.7.2 Defense-in-Depth Is Maintained 

Section 50.69 (c)(1)(iii) requires that 

defense-in-depth be maintained as part 

of the categorization requirements of 

§ 50.69(c)(1) and as a result, defense-indepth is considered explicitly in the 

categorization process. Thus, SSCs that 

otherwise might be considered low 

safety significant, but are important to 

defense-in-depth as discussed in the 

implementation guidance, will be 

categorized as safety significant (and 

will remain subject to special treatment 

requirements). For safety significant 

SSCs (i.e., RISC–1 and RISC–2 SSCs), all 

current special treatment requirements 

remain (i.e., the rule does not remove 

any of these requirements) to provide 

high confidence that they can perform 

design basis functions. Additionally, 

§ 50.69(d)(1) requires sufficient 

treatment be applied to support the 

credit taken for these SSCs for beyond 

design basis events. For RISC–3 SSCs, 

§ 50.69 imposes high-level treatment 

requirements that when effectively 

implemented, maintain the capability of 

RISC–3 SSCs to perform their design 

basis functions. Thus, the complement 

of SSCs installed at the facility that 

provide defense-in-depth will continue 

to be available and capable of 

performing the functions necessary to 

support defense-in-depth. The rule does 

not change the design basis functional 

requirements of the facility, which were 

established based upon defense-indepth considerations. Accordingly, the 

Commission concludes that § 50.69 

maintains defense-in-depth. 

III.7.3 Safety Margins Are Maintained 

Section 50.69(c)(1)(iv) requires that 

evaluations be performed that provide 

reasonable confidence that sufficient 

safety margins are maintained. This is 

provided by a combination of: 

(1) Maintaining all existing functional 

and treatment requirements on RISC–1 

and RISC–2 SSCs and additionally 

ensuring, through the application of 

sufficient treatment and feedback 

requirements, that any credit for these 

SSCs for beyond design basis conditions 

is valid and maintained; 

(2) Maintaining the design basis 

functional requirements of the facility 

for all SSCs, including RISC–3 SSCs as 

described in Section III.7.2; and 

(3) Requiring a licensee to have 

reasonable confidence that the overall 

increase in risk that may result due to 

implementation of § 50.69 is small. 

Maintaining all current requirements 

on RISC–1 and RISC–2 SSCs and 

requiring sufficient treatment be applied 

to support the credit taken for these 

SSCs for beyond design basis events 

provides assurance that the safety 

significant SSCs continue to perform as 

credited in the categorization process. 

Maintaining design basis functional 

requirements for RISC–3 SSCs ensures 

that these SSCs continue to be designed 

to criteria that enable them to perform 

their design basis functions. The 

reduction in treatment applied to RISC–

3 SSCs results in an increased level of 

uncertainty concerning the functionality 

of RISC–3 SSCs. This reduction in 

treatment may result in an increase in 

RISC–3 SSC failure rates (i.e., a 

reduction in RISC–3 SSC reliability). To 

address this possibility and its 

relationship to safety margin, § 50.69 

requires that there be reasonable 

confidence that any potential increases 

in CDF and LERF that might stem from 

changes in RISC–3 SSC reliability due to 

reduced treatment permitted by § 50.69, 

be small. As discussed in Section III.7.4, 

the rule requires (through monitoring 

requirements) that the SSCs must be 

maintained so that they continue to be 

capable of performing their design basis 

functions. For these reasons, the 

Commission concludes that § 50.69 

maintains sufficient safety margins.

III.7.4 Monitoring and Performance 

Assessment Strategies Are Used 

Section 50.69(e) contains 

requirements that ensure that the riskinformed categorization and treatment 

processes are updated and maintained 

over time. Data that reflect operational 

practices, the facility configuration, 

plant and industry experience, and SSC 

performance are required to be fed back 

into the PRA and the categorization 

process on a periodic basis and when 

appropriate, adjustments to the 

categorization and/or treatment 

processes are required to maintain the 

validity of these processes. In addition, 

§ 50.69(g) contains requirements that 

reports are made to NRC of conditions 

preventing RISC–1 and RISC–2 SSCs 

from performing their safety significant 

functions. Together, these requirements 

maintain the validity of the riskinformed categorization and treatment 

processes so that the above criteria will 

continue to be satisfied over the life of 

the facility. 

III.7.5 Summary and Conclusions 

Section § 50.69 contains requirements 

that: 

1. Provide reasonable confidence that 

any net risk increase from 

implementation of its requirements is 

small; 

2. Maintain defense-in-depth; 

3. Provide reasonable confidence that 

safety margins are maintained; and 

4. Require the use of monitoring and 

performance assessment strategies. 

Together, these requirements result in 

a rule that is consistent with the 

Commission’s policy on the use of PRA 

and, more importantly, maintains 

adequate protection of public health and 

safety. 

IV. Pilot Activities 

IV.1.0 Pilot plants 

To aid in the development of the rule 

and associated implementation 

guidance, several plants volunteered to 

conduct pilot activities with the 

objective of exercising the proposed NEI 

implementation guidance and using the 

feedback and lessons-learned to 

improve both the implementation 

guidance and the governing regulatory 

framework. There were two separate 

pilot efforts. The first pilot effort 

focused on the categorization guidance 

and IDP performance. This effort is 

discussed in Section IV.1.1 

Categorization Pilot. The second pilot 

effort is ongoing and is focused on the 

§ 50.69 submittal and its review. This 

pilot effort is discussed in Section IV.1.2 

Submittal Pilot. 

IV.1.1 Categorization Pilot 

The categorization pilot effort was 

supported by three of the industry 

owners groups who identified pilots for 

their reactor types and participated by 

piloting sample systems using the draft 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68033

NEI implementation guidance. 

Supporting the pilot effort were the 

Westinghouse Owners Group with lead 

plants Wolf Creek and Surry, the BWR 

Owners Group with lead plant Quad 

Cities, and the CE Owners Group with 

lead plant Palo Verde. The B&W Owners 

Group did not participate, but did 

follow the pilot activities. 

The NRC staff’s participation and 

principal point of interaction in the 

pilot effort was primarily in observation 

of the deliberations of the IDP. By 

observing the IDP, the NRC staff was 

able to view the culmination of the 

categorization effort and gain good 

insights regarding both the robustness of 

the categorization process in general 

and the IDP decision-making process 

specifically. Following each of the pilot 

IDPs, the NRC staff developed and 

issued a trip report containing its 

observations. 

The following points set forth the 

principal lessons learned and key 

feedback from the NRC staff’s 

observations of the pilot activities: 

• Potential treatment changes and 

their potential effects need to be 

understood by the IDP as part of the 

deliberations on categorization. 

• The pilots showed the importance 

of documenting IDP decisions and the 

basis for them. The rule contains a 

requirement for the categorization basis 

to be documented (and records retained) 

in § 50.69(f). 

• The pilots experienced difficulty in 

explicit consideration about safety 

margins, especially in view of the fact 

that functionality must be retained. In 

the first draft rule language posted, 

requirements were included for the IDP 

to consider safety margins in its 

deliberations. On the basis of the pilot 

experience, NRC adjusted its approach 

to safety margins to include this in the 

section of the rule that requires 

consideration of effects of changes in 

treatment and the use of evaluations as 

the means of providing reasonable 

confidence safety margins are 

maintained. 

• The need for a number of 

improvements to the industry 

implementation guidance provided in 

NEI 00–04 were noted. For example, 

two areas for improvement were the 

defense-in-depth matrix presented 

therein and the need for more specific 

guidance on making decisions where 

quantitative information is not 

available. These lessons learned were 

factored into the revised version of NEI 

00–04. 

• During the pilot activity, pressure 

boundary (‘‘passive’’) functions were 

also categorized using the draft version 

of an ASME Code Case on categorization 

available at that time. A separate 

categorization process was used for 

these passive functions because it was 

recognized by pilot participants that the 

approach for these SSCs must be 

somewhat different than for ‘‘active’’

functions due to considerations such as 

spatial interaction. Specifically, if a 

pressure boundary SSC failed, the 

resulting high-energy release or flooding 

might impact other equipment in 

physical proximity, so the process 

needed to account for those effects in 

addition to the significance of the SSC 

that initially failed. Improvements to the 

ASME Code Case for categorization of 

piping (and related components) were 

identified and fed back into the code 

development process. 

• The pilot experiences also revealed 

the intricacies of the relationship 

between ‘‘functions’’ (which play a role 

in decisions on safety significance) and 

‘‘components’’ (importance measures 

are associated with components and 

treatment is also generally applied on a 

component basis). Because a particular 

component may support more than one 

function, the categorization of the 

component needs to correspond with 

the most significant function and means 

must be provided for a licensee to 

‘‘map’’ the components to the functions 

they support. 

• At each pilot, the NRC noted that 

the IDP needed to include consideration 

of long term containment heat removal 

in characterizing SSCs. The NRC 

considers retention of long term 

containment heat removal capability 

important to defense-in-depth for light 

water reactors. 

• Finally, a number of lessons were 

learned about how to conduct the IDP 

process, such as training needs, 

materials to be provided to the panel, 

etc. As a result of this feedback, NEI 

revised NEI 00–04 (discussed in Section 

VI).

IV.1.2 Submittal Pilot 

The submittal pilot effort is a 

currently ongoing effort that focuses on 

the § 50.69 submittal and the NRC staff’s 

review and approval of that submittal. 

This pilot effort is supported by the 

Westinghouse Owners Group with lead 

plants Wolf Creek and Surry. The 

objectives of this pilot effort are to: 

• Enable the staff to develop reviewer 

guidance for review and approval of the 

§ 50.69 submittal. 

• To acquire experience with the use 

of RG 1.201 and use this experience to 

improve the guidance and address the 

technical interpretation/implementation 

issues identified in RG 1.201. 

• Enable industry to develop (beyond 

RG 1.201/NEI 00–04) the specific 

information that will be required for a 

license amendment submittal that will 

be submitted for prior staff review and 

approval for implementing § 50.69. 

The NRC staff will use the results of 

this pilot effort to improve RG 1.201 and 

to develop the reviewer guidance for 

§ 50.69 submittals. Industry expects to 

use the results of the pilot to develop a 

template for a § 50.69 license 

amendment submittal. 

IV.2.0 South Texas Exemption as 

Proof of Concept 

A major element of the rulemaking 

plan described in SECY–99–256 was the 

review of the STPNOC exemption 

request. The review of the STPNOC 

exemption request was viewed as a 

proof-of-concept prototype for this 

rulemaking rather than a pilot because 

it preceded development of draft rule 

language or related implementation 

guidance. 

By letter dated July 13, 1999, STPNOC 

requested approval of exemption 

requests to enable implementation of 

processes for categorizing the safety 

significance of SSCs and treatment of 

those SSCs consistent with its 

categorization process. The STPNOC 

process included many similar elements 

to that described in this rulemaking, but 

with some differences. Their process 

identified SSCs as being either high, 

medium, low or non-risk significant. 

The scope of the exemptions requested 

included only those safety-related SSCs 

that have been categorized as low safety 

significant or as non-risk significant 

using STPNOC’s categorization process. 

The licensee indicated that the 

categorization and treatment processes 

would be implemented over the 

remaining licensed period of the 

facility. Thus, the basis for the 

exemptions granted was the NRC staff’s 

approval of the licensee’s categorization 

process and alternative treatment 

elements, rather than a comprehensive 

review of the final categorization and 

treatment of each SSC (review of the 

process rather than the results is also 

the approach planned under the 

rulemaking). As a result of discussions 

with the staff on a number of topics, 

STPNOC submitted a revised exemption 

request on August 31, 2000. 

On November 15, 2000, the NRC staff 

issued a draft safety evaluation 

(SE)(ADAMS accession number 

ML003761558), based on the revised 

exemption requests. Following the 

licensee’s response to the draft SE, the 

staff prepared SECY–01–0103 dated 

June 12, 2001 (ADAMS accession 

number ML011560317), to inform the 

Commission of the staff’s finding 

regarding the STPNOC exemption 

68034 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

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 § 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 

§ 50.69, the NRC continues to require a 

robust categorization with a detailed 

staff review. 

The rule specifies the requirement 

that the licensee shall ensure with 

reasonable confidence functionality and 

further specifies some high-level 

requirements for RISC–3 SSC treatment. 

Under § 50.69, the NRC will not review 

and approve licensee’s RISC–3 

treatment programs. Licensees will have 

to establish appropriate performancebased SSC treatment to maintain the 

validity of the categorization process 

and its results. The rule requires that 

licensees adjust the categorization or 

treatment processes, as appropriate, in 

response to the SSC performance 

information obtained as part of the 

treatment process. 

V. Section-by-Section Analysis 

V.1.0 Section 50.8 Information 

Collection 

This rule includes a revision to 

§ 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 § 50.69 includes information 

collection requirements, a conforming 

change to § 50.8(b) is necessary to list 

§ 50.69 as one of these rules. See also 

Section XII of the SOC for discussion 

about information collection 

requirements of § 50.69. 

V.2.0 Section 50.69(a) Definitions 

Section 50.69(a) provides the 

definition for the four RISC categories 

and the definition of the term ‘‘safety 

significant function.’’ RISC–1 SSCs are 

safety-related SSCs (as defined in § 50.2) 

and that are found to be safety 

significant (using the risk-informed 

categorization process being established 

by this rule). RISC–2 SSCs are SSCs that 

do not meet the safety-related 

definition, but determined to be safety 

significant. RISC–3 SSCs are safetyrelated SSCs that are determined to be 

low safety significant on an individual 

basis. Finally, RISC–4 SSCs are SSCs 

that are not safety-related and that are 

determined to be low safety significant. 

The NRC selected the terms ‘‘safety 

significant’’ and ‘‘low safety significant’’

as the best representations of their 

meaning. Every component (if 

categorized) is either safety significant 

or low safety significant. The ‘‘low’’

category could include those SSCs that 

have no safety significance, as well as 

some SSCs that individually are not 

safety significant, but collectively can 

have a significant impact on plant safety 

(and hence the need for maintaining the 

design basis capability of these SSCs). 

Similarly, within the category of ‘‘safety 

significant,’’ some SSCs have more 

safety significance than others; so it did 

not appear appropriate to call them all 

‘‘high safety significant.’’ The RISC 

definitions of paragraph (a) are used in 

subsequent paragraphs of § 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 

§ 50.69 considers both probabilistic and 

deterministic information in the 

decision process. Thus, it is more 

accurate to represent the outcome as a 

determination of overall safety 

significance, that includes the 

consideration of risk, as opposed to 

characterizing the outcome as purely 

‘‘risk-significance.’’

Those functions that are not 

determined to be safety significant are 

considered to be low safety-significant. 

The determination as to which 

functions are safety significant is done 

by following the categorization process 

outlined in paragraph (c), as 

implemented following the guidance in 

RG 1.201, ‘‘Guidelines for Categorizing 

Structures, Systems, and Components in 

Nuclear Power Plants According to their 

Safety Significance.’’

V.3.0 Section 50.69(b) Applicability 

Section 50.69(b) may be voluntarily 

implemented by: 

(1) A holder of a license to operate a 

light water reactor (LWR) nuclear power 

plant under this part; 

(2) Holders of Part 54 renewed LWR 

licenses; 

(3) An applicant for a construction 

permit or operating license under this 

part; and 

(4) An applicant for a design 

approval, a combined license, or 

manufacturing license under Part 52 of 

this chapter. 

For current licensees, implementation 

will be through a license amendment as 

set forth in § 50.90. This review and 

approval of the categorization process is 

a one-time process approval (i.e., the 

approval is not restricted to a set of 

systems or structures, and instead can 

be applied to any system or structure in 

the plant). The licensee is not required 

to come back to the NRC for review of 

the categorization process provided they 

remain within the scope of the NRC’s 

safety evaluation. Until the request is 

approved, a licensee is free to develop 

(at their own risk) the § 50.69 processes 

and perform the § 50.69 categorization. 

However, they must continue to follow 

existing requirements until approval. 

Upon approval of the categorization 

process, the licensee can implement the 

results of the categorization process 

including the revised § 50.69 treatment 

requirements.

For part 54 license holders, 

implementation is the same as that for 

a holder of an operating license under 

part 50, that is, to apply for an 

amendment to the (renewed) license. 

For the case where a licensee renewed 

its license first and then implemented 

§ 50.69, a licensee might revise some 

aging management programs for RISC–3 

SSCs, consistent with the requirements 

of § 50.69. The Commission believes 

that there should be little or no 

impediment for doing so because the 

categorization process that allows for 

the reduction in the special treatment 

requirements for RISC–3 components is 

expected to provide an appropriate level 

of safety for the respective structures, 

systems and components. 

In the development of § 50.69, 

questions were considered regarding the 

impact to licensees that implement 

§ 50.69 and subsequently apply to 

renew their license. Because part 54 

includes scoping criteria that bring 

safety-related components within its 

scope, these components could not be 

exempted without amending part 54 to 

allow for their exclusion. However, 

there are still options available to 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68035

applicants for renewal that have 

implemented § 50.69 first. Because 

§ 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 

(§ 50.69(d)(2)) provide a sufficient 

demonstration that aging management 

of the components will be achieved 

during the renewal period to ensure the 

functionality of the structure, system, or 

component. In addition, in the 1995 

amendment to part 54, the Commission 

recognized that risk insights could be 

used in evaluating the robustness of an 

aging management program. The NRC 

staff has already received and accepted 

one proposal (Arkansas Unit 1) for a 

risk-informed program for small-bore 

piping which demonstrates that risk 

arguments can be used to a degree. 

Adopting § 50.69 requirements for an 

applicant for a construction permit or 

operating license under this part 

requires that the applicant first design 

the facility to meet the current part 50 

requirements. Specifically, to use the 

§ 50.69 requirements requires that SSCs 

first be classified into the traditional 

safety-related and nonsafety-related 

classifications. This establishes the 

design basis functional requirements for 

the facility, which as previously stated, 

§ 50.69 is not changing. Once the SSC 

categorization has been done consistent 

with the safety-related definition in 

§ 50.2, then § 50.69 can be used to 

categorize SSCs into RISC–1, RISC–2, 

RISC–3, and RISC–4 and the alternative 

treatment requirements of § 50.69 

implemented. A new applicant who 

chooses to adopt the § 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 § 50.69 requirements 

before initial plant operation. 

An applicant for a design approval, a 

combined license, or manufacturing 

license under part 52 of this chapter 

may adopt § 50.69 requirements. An 

applicant for a design approval, or 

manufacturing license would follow a 

process very similar (from the 

standpoint of § 50.69) to that described 

above for an applicant for a construction 

permit or operating license under part 

50 (i.e., SSCs must first be classified into 

the traditional safety-related and 

nonsafety-related classifications which 

establishes the design basis functional 

requirements for the facility and then 

§ 50.69 can be used to categorize SSCs 

into RISC–1, RISC–2, RISC–3, and 

RISC–4). Because § 50.69 includes 

elements of procurement and 

installation, as well as inservice 

activities, implementation of the rule by 

a holder of a manufacturing license or 

by a part 52 applicant that references 

such a design would place restrictions 

on the eventual operator of the facility. 

The entity that actually constructs and 

operates the facility would also have to 

implement § 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 

§ 50.55a, which may not be compatible 

with the facility as manufactured by the 

manufacturing licensee. 

An applicant for a part 52 combined 

license can apply § 50.69 to a referenced 

design certification that did not comply 

with § 50.69 provided the design is a 

LWR design that used the safety-related 

definition in § 50.2. An applicant who 

references a certified design and wishes 

to implement § 50.69 would include the 

specified information in § 50.69(b)(2) as 

part of its application for a license. This 

does not mean that an applicant would 

actually construct the facility per all 

parts 50 and 100 requirements first, 

before applying § 50.69. Instead, the 

facility needs to be designed per these 

requirements, but following approval of 

the application request under 

§ 50.69(b)(4), RISC–3 SSCs could be 

procured per the requirements of 

§ 50.69(d). 

The final rule excludes applicants for 

standard design certifications from the 

group of entities who may take 

advantage of the provisions of § 50.69. 

In considering whether to extend the 

applicability of § 50.69 to design 

certifications, the Commission 

identified a number of difficult issues 

which would have to be resolved to 

support such an extension. For example, 

it is unclear whether the dynamic 

process of recategorizing SSCs under 

§ 50.69 would be inconsistent with the 

special change restrictions in § 52.63(a), 

thereby requiring the inclusion of a 

special change provision in the 

individual design certification rule. 

Inasmuch as the proposed rule did not 

include a provision that would have 

allowed design certification applicants 

to use § 50.69, the NRC has not had the 

benefit of the views of the industry and 

the public on these issues. Moreover, 

the industry has not expressed any 

interest in submitting a design 

certification using the principles of 

§ 50.69. Accordingly, the final rule does 

not address the issue of applying § 50.69 

to new design certifications; issues 

associated with the application of 

§ 50.69 to design certification 

rulemaking can be addressed on a caseby-case basis as necessary. In the future, 

the Commission could initiate 

rulemaking to extend § 50.69 to new 

design certifications after the NRC has 

had some experience in this area. For 

much the same reasons, the rule does 

not provide a process for changing an 

existing design certification rule to 

voluntarily comply with § 50.69. In 

addition, a rulemaking would be 

necessary to change an existing certified 

design (see Section VIII of Appendix A 

to 10 CFR part 52), and it is unlikely 

that such a change would satisfy the 

requirements of § 52.63(a)(1). A request 

for a generic change to adopt § 50.69 

would not meet the special backfit 

requirements of Section VIII. Therefore, 

the NRC would not review the request. 

Additionally, the NRC would not want 

to expend resources reviewing changes 

to designs that may not be referenced. 

However, applicants for COLs that 

reference a certified design could adopt 

§ 50.69 and the rule provides for that 

approach. 

The rule provisions were devised to 

provide means for licensees and 

applicants for light water reactors to 

implement § 50.69. In view of some of 

the specific provisions of the rule, for 

example, ‘‘safety-related’’ definition and 

use of CDF/LERF metrics, the 

Commission is making this rule 

applicable only to light-water reactor 

designs. 

V.3.1 Section 50.69(b)(1) Removal of 

RISC–3 and RISC–4 SSCs From the 

Scope of Treatment Requirements 

Section 50.69 (b)(1) lists the specific 

special treatment requirements from 

whose scope the RISC–3 and RISC–4 

SSCs are being removed through the 

application of § 50.69. In this paragraph, 

each regulatory requirement (or portions 

thereof) removed by this rulemaking is 

listed in a separate item, numbered from 

§ 50.69(b)(1)(i) through (b)(1)(xi). The 

basis for removal of these requirements 

was discussed in Section III.4. These 

requirements are being removed due to 

the low safety significance of RISC–3 

and RISC–4 SSCs as determined by an 

approved risk-informed categorization 

process meeting the requirements of 

§ 50.69(c). The special treatment 

requirements for RISC–3 SSCs are 

replaced with the high-level, 

performance-based requirements in 

§ 50.69(d)(2) that require the licensee to 

provide reasonable confidence that 

RISC–3 SSCs will continue to be 

capable of performing their safetyrelated functions under design basis 

conditions. These performance-based 

RISC–3 requirements in paragraph (d)(2) 

are discussed below in greater detail. 

Note that special treatment 

requirements are not removed from any 

68036 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

SSCs until the NRC approves the 

categorization process and a licensee (or 

applicant) has categorized those SSCs 

using the requirements of § 50.69(c) to 

provide the documented basis for the 

decision that they are of low safety 

significance.

V.3.2 Section 50.69 (b)(2)

Application Process 

Section 50.69(b)(2) requires a licensee 

who voluntarily seeks to implement 

§ 50.69 to submit an application for a 

license amendment under § 50.90 that 

contains the following information: 

(i) A description of the categorization 

process that meets the requirements of 

§ 50.69(c). 

(ii) A description of the measures 

taken to assure that the quality and level 

of detail of the systematic processes that 

evaluate the plant for internal and 

external events during normal 

operation, low power, and shutdown 

(including the plant-specific PRA, 

margins-type approaches, or other 

systematic evaluation techniques used 

to evaluate severe accident 

vulnerabilities) are adequate for the 

categorization of SSCs. 

(iii) Results of the PRA review process 

to be conducted to meet § 50.69(c)(1)(i). 

(iv) A description of, and basis for 

acceptability of, the evaluations to be 

conducted to satisfy § 50.69(c)(1)(iv). 

The evaluations must include the effects 

of common cause interaction 

susceptibility, and the potential impacts 

from known degradation mechanisms 

for both active and passive functions, 

and address internally and externally 

initiated events and plant operating 

modes (e.g., full power and shutdown 

conditions). 

Regarding the categorization process 

description, the NRC expects that most 

licensees and applicants will commit to 

RG 1.201 which endorses NEI 00–04, 

with some conditions and exceptions. If 

a licensee or applicant wishes to use a 

different approach, the submittal must 

provide a sufficient description of how 

the categorization would be conducted. 

As part of the submittal, a licensee or 

applicant is to describe the measures 

they have taken to assure that the plantspecific PRA, as well as other methods 

used, are adequate for application to 

§ 50.69. The measures described include 

such items as any peer reviews 

performed, any actions taken to address 

peer review findings that are important 

to categorization, and any efforts to 

compare the plant-specific PRA to the 

ASME PRA standard. The NRC has 

developed reviewer guidance applicable 

to these submittals that is described in 

Section VI. The licensee or applicant 

must also describe what measures they 

have used for the methods other than a 

PRA to determine their adequacy for 

this application. 

Further, the licensee or applicant is 

required to include information about 

the evaluations they intend to conduct 

to provide reasonable confidence that 

the potential increase in risk would be 

small. This includes any risk sensitivity 

study for RISC–3 SSCs, including the 

basis for whatever change in reliability 

is being assumed for these analyses. A 

licensee must provide sufficient 

information to the NRC, describing the 

risk sensitivity study and other 

evaluations and the basis for their 

acceptability as appropriately 

representing the potential increase in 

risk from implementation of the 

requirements in this rule. 

RISC–3 SSCs are defined as having 

low individual safety significance under 

§ 50.69. Licensees and applicants must 

implement effective treatment, 

consisting of, at a minimum, inspection, 

testing, and corrective action, to 

maintain RISC–3 SSC functionality as 

required by § 50.69(d)(2). This treatment 

need not be described to the NRC as part 

of the § 50.69 submittal as provided in 

§ 50.69(b)(2). 

V.3.3 Section 50.69 (b)(3) Approval 

for Licensees 

Section 50.69(b)(3) provides that the 

Commission will approve a licensee’s 

implementation of this section by 

license amendment if it determines that 

the proposed process for categorization 

of RISC–1, RISC–2, RISC–3, and RISC–

4 SSCs satisfies the requirements of 

§ 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 plantspecific 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 § 50.69 by issuing 

a license amendment. 

V.3.4 Section 50.69(b)(4) Process for 

Applicants 

Section 50.69(b)(4) requires that an 

applicant for a license, standard design 

approval, or manufacturing license that 

chooses to implement § 50.69 must 

submit the information listed in 

§ 50.69(b)(2) as part of its application. 

The rule is structured to transition from 

the ‘‘safety-related’’ classification (and 

related treatment requirements) to a 

‘‘safety significant’’ classification. Thus, 

an applicant would first need to design 

the facility to meet applicable Part 50 

design requirements and then apply the 

requirements of § 50.69. This 

information must be submitted in 

addition to other technical information 

necessary to meet § 50.34. The NRC will 

provide its approval of implementation 

of § 50.69, if it concludes that the rule 

requirements are met, as part of its 

action on the application. 

V.4.0 Section 50.69(c) Categorization 

Process Requirements 

Section 50.69(c) establishes the 

requirements for the risk-informed 

categorization process including 

requirements for the supporting PRA. 

Licensees or applicants who wish to 

adopt the requirements of § 50.69 will 

need to make a submittal (per 

§ 50.69(b)(2) or § 50.69(b)(4) 

respectively) that discusses how their 

proposed categorization process, 

supporting PRA, and evaluations meet 

the § 50.69(c) requirements. As 

described in Section III.2.0, these 

requirements are intended to ensure that 

the risk-informed § 50.69 categorization 

process determines the appropriate 

safety significance of SSCs with high 

confidence. The introductory paragraph 

of § 50.69(c) states that SSCs must be 

categorized as RISC–1, RISC–2, RISC–3, 

or RISC–4 by a process that determines 

whether the SSC performs one or more 

safety significant functions and 

identifies those functions. 

V.4.1 Section 50.69(c)(1)(i) Results 

and Insights From a Plant-Specific 

Probabilistic Risk Assessment 

Section 50.69(c)(1)(i) contains the 

requirements for the PRA itself, and 

how it is to be used in the categorization 

process. The PRA must have sufficient 

capability and quality to support the 

categorization of the SSCs. Section 

V.4.1.1 discusses these requirements in 

more detail. The PRA and associated 

sensitivity studies are used primarily in 

the categorization of the SSCs as to their 

safety significance as discussed in 

Section V.4.1.2, and the PRA is also 

used to perform evaluations to assess 

the potential risk impact of the 

proposed change in treatment of the 

RISC–3 SSCs, as discussed in Section 

V.4.4.

V.4.1.1 Scope, Capability, and Quality 

of the PRA To Support the 

Categorization Process 

As required in § 50.69(c)(1)(ii), 

initiating events from sources both 

internal and external to the plant and 

for all modes of operation, including 

low power and shutdown modes, must 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68037

be considered when performing the 

categorization of SSCs. It is recognized 

that few licensees have fully developed 

PRA models that cover such a scope. 

However, as a minimum, the PRA to be 

used to support categorization under 

§ 50.69(c)(1) must model internal 

initiating events occurring at full power 

operations. The PRA will have to be 

able to calculate both core damage 

frequency and large early release 

frequency to meet the requirement in 

§ 50.69(c)(iv). The PRA must reasonably 

represent the current configuration and 

operating practices at the plant to meet 

§ 50.69(c)(1)(ii). The PRA model should 

be of sufficient technical quality and 

level of detail to support the 

categorization process. This means that 

it represents a coherent, integrated 

model, and has sufficient detail to 

support the categorization of SSCs into 

the safety significant and low safety 

significant categories. 

The quality and scope of the plantspecific PRA will be assessed by the 

NRC taking into account appropriate 

standards and peer review results. The 

NRC has prepared a regulatory guide 

(RG 1.200) on determining the technical 

adequacy of PRA results for riskinformed activities. As one step in the 

assurance of technical quality, the PRA 

must have been subjected to a peer 

review process assessed against a 

standard or set of acceptance criteria 

that is endorsed by the NRC. Thus, the 

NRC will rely on the NEI Peer Review 

Process, as modified in the NRC’s 

approval, or the ASME/ANS Peer 

Review Process, as modified in the 

NRC’s approval both of which are (or 

will be) documented in RG 1.200. As 

discussed in Section VI, NRC has also 

developed review guidelines for 

considering the sufficiency of a PRA 

that was subjected to the NEI peer 

review process for this application in 

§ 50.69. This guidance was developed 

based on an earlier draft version of NEI 

00–04 and could be useful in ensuring 

the adequacy of the PRA for this 

application. The submittal requirements 

listed in § 50.69(b)(2) include a 

requirement to provide information 

about the quality of the PRA analysis 

and other supporting analyses and about 

the peer review results. 

V.4.1.2 Risk Categorization Process 

Based on PRA Information 

For SSCs modeled in the PRA, the 

typical categorization process relies on 

the use of importance measures as a 

screening method to assign the 

preliminary safety significance of SSCs. 

(Other methodologies such as success 

path identification methodologies can 

also be used, however, this discussion 

will focus on the use of importance 

measures because these are the most 

commonly used methods to identify 

safety significance of SSCs using a PRA, 

for example, in the implementation of 

§ 50.65). The determination of the safety 

significance of SSCs by importance 

measures is also important because it 

can identify potential risk outliers and 

therefore, changes that exacerbate these 

outliers can be avoided; and it can 

facilitate IDP deliberations of SSCs that 

are not modeled in the PRA, for 

example, events from the ranked list can 

be used as surrogates for those SSCs that 

are not modeled or are only implicitly 

modeled in the PRA. 

For SSCs modeled in the PRA, SSC 

importance is effectively determined 

(see § 50.69(c)(1)(iv)) based on both CDF 

and LERF. Importance measures should 

be chosen so that the IDP can be 

provided with information on the 

relative contribution of an SSC to total 

risk. Examples of importance measures 

that can accomplish this are: the 

Fussell-Vesely (F–V) importance and 

the Risk Reduction Worth (RRW) 

importance. Importance measures 

should also be used to provide the IDP 

with information on the margin 

available should an SSC fail to function. 

The Risk Achievement Worth (RAW) 

importance and the Birnbaum 

importance are example measures that 

are suitable for this purpose. 

In choosing screening criteria to be 

used with the PRA importance 

measures, it should be noted that 

importance measures do not directly 

relate to changes in the absolute value 

of risk. Therefore, the final criteria for 

categorizing SSCs into the safety 

significant and the low safety significant 

categories must be based on an 

assessment of the potential overall 

impact of SSC categorization and a 

comparison of this potential impact to 

the acceptance guidelines for changes in 

CDF and LERF. However, typically in 

the initial screening stages, an SSC with 

F–V < 0.005 based on CDF and LERF, 

and RAW < 2 based on CDF and LERF 

can be considered as potentially low 

safety-significant. In addition, the 

appropriateness of the importance 

measures in specifically addressing SSC 

CCF contributions and associated 

screening criteria should be considered. 

IDP consideration of § 50.69(c)(1)(ii), 

(c)(1)(iii), and (c)(1)(iv) should be 

carried out to confirm the low safety 

significance of these SSCs. 

In determining the safety significance 

of SSCs, consideration should be given 

to the potential for the multiple failure 

modes for the SSC. PRA basic events 

represent specific failure events and 

failure modes of SSCs. The 

determination of SSC safety significance 

should take into account the effects of 

all associated basic PRA events (such as 

failure to start and failure to run), 

including indirect contributions through 

associated common cause failure (CCF) 

events. 

Because importance measures are 

typically evaluated on the basis of 

individual events, single-event 

importance measures have the potential 

to dismiss all elements of a system or 

group despite the system or group 

having a high importance when taken as 

a whole. Conversely, there may be 

grounds for screening groups of SSCs, 

owing to the unimportance of the 

systems of which they are elements. 

One approach around this problem is to 

first determine the importance of system 

functions performed by the selected 

plant systems. If necessary, each 

component in a system is then 

evaluated to identify the system 

function(s) supported by that 

component. SSCs may be initially 

assigned the same category as the most 

limiting system function they support. 

System operating configuration, 

reliability history, recovery time 

available, and other factors can then be 

considered when evaluating the effect 

on categorization from an SSC’s 

redundancy or diversity. The primary 

consideration in the process is whether 

the failure of an SSC will fail or severely 

degrade the safety function. If the 

answer is no, then a licensee may factor 

into the categorization the SSC’s 

redundancy, as long as the SSC’s 

reliability credited in the categorization 

process and that of its redundant 

counterpart(s) have been taken into 

account. 

When the PRA used in the importance 

analyses includes models for external 

initiating events and/or plant operating 

modes other than full power, caution 

should be used when considering the 

results of the importance calculations. 

The PRA models for external initiating 

events (e.g., events initiated by fires or 

earthquakes) and for low power and 

shutdown plant operating modes may 

be more conservative and have a greater 

degree of uncertainty than for internal 

initiating events. Use of conservative 

models can influence the calculation of 

importance measures by moving more 

SSCs into the low safety significance 

category. Therefore, when PRA models 

for external event initiators and for the 

low power and shutdown modes of 

operation are available and used, the 

importance measures should be 

evaluated for each analysis separately 

and collectively, and the results of these 

evaluations should be provided to the 

IDP. 

68038 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

As part of the demonstration of PRA 

adequacy, the sensitivity of SSC 

importance to uncertainties in the 

parameter values for component 

availability/reliability, human error 

probabilities, and CCF probabilities 

should be evaluated. Results of these 

sensitivity analyses should be provided 

to the IDP. The following should be 

considered in IDP deliberations on the 

sensitivity study results: 

(1) The change in event importance 

when the parameter value is varied over 

its uncertainty range for the event 

probability can in some cases provide 

SSC categorization results that are 

different. Therefore, in considering the 

sensitivity of component categorization 

to uncertainties in the parameter values, 

the IDP should ensure that SSC 

categorization is not affected by data 

uncertainties. 

(2) PRAs typically model recovery 

actions, especially for dominant 

accident sequences. Estimating the 

success probability for the recovery 

actions involves a certain degree of 

subjectivity. The concerns in this case 

stem from situations where very high 

success probabilities are assigned to a 

sequence, resulting in related 

components being ranked as low risk 

contributors. Furthermore, it is not 

desirable for the categorization of SSCs 

to be impacted by recovery actions that 

sometimes are only modeled for the 

dominant scenarios. Sensitivity analyses 

should be used to show how the SSC 

categorization would change if recovery 

actions were removed. The IDP should 

ensure that the categorization is not 

unduly impacted by the modeling of 

recovery actions. 

(3) CCFs are modeled in PRAs to 

account for dependent failures of 

redundant components within a system. 

CCF probabilities can impact PRA 

results by enhancing or obscuring the 

importance of components. A 

component may be ranked as a high risk 

contributor mainly because of its 

contribution to CCFs or a component 

may be ranked as a low risk contributor 

mainly because it has negligible or no 

contribution to CCFs. The IDP should 

ensure that the categorization is not 

unduly impacted by the modeling of 

CCFs. The IDP should also be aware that 

removing or relaxing requirements may 

increase the CCF contribution, thereby 

changing the risk impact of an SSC.

V.4.2 Section 50.69(c)(1)(ii)

Integrated Assessment of SSC Function 

Importance 

Section 50.69(c)(1)(ii) contains 

requirements for an integrated, 

systematic process to address events 

including those not modeled in the 

PRA, including both design basis and 

severe accident functions. For various 

reasons, many SSCs in the plant will not 

be modeled explicitly in the PRA. 

Therefore, the categorization process 

must determine the safety significance 

of these SSCs by other means. Because 

importance measures are not available 

for use as screening, other criteria or 

considerations must be used by the IDP 

to determine the significance. Guidance 

on how these deliberations should be 

conducted is included in the NRC 

regulatory guidance associated with this 

rule, and in the industry guidance. 

Section 50.69(c)(1)(ii) requires that all 

aspects of the processes used to 

categorize SSC must ‘‘reasonably 

reflect’’ the current plant configuration, 

operating practices, and applicable 

operating experience. The terminology, 

‘‘reasonably reflect,’’ was selected to 

allow for appropriate PRA modeling and 

also to make clear that the PRA and 

categorization processes do not need to 

be instantaneously revised when a plant 

change occurs (see also requirements in 

§ 50.69(e)(1) on PRA updating). 

V.4.3 Section 50.69(c)(1)(iii)

Maintaining Defense-in-Depth 

Section 50.69(c)(1)(iii) requires that 

the categorization process maintain 

defense-in-depth. To satisfy this 

requirement, when categorizing SSCs as 

low safety significant, the IDP must 

demonstrate that defense-in-depth is 

maintained. Defense-in-depth is 

adequate if the overall redundancy and 

diversity among the plant’s systems and 

barriers is sufficient to ensure the risk 

acceptance guidelines discussed in 

Section V.4.4 are met, and that: 

• Reasonable balance is preserved 

among prevention of core damage, 

prevention of containment failure or 

bypass, and mitigation of consequences 

of an offsite release. 

• System redundancy, independence, 

and diversity is preserved 

commensurate with the expected 

frequency of challenges, consequences 

of failure of the system, and associated 

uncertainties in determining these 

parameters. 

• There is no over-reliance on 

programmatic activities and operator 

actions to compensate for weaknesses in 

the plant design. 

• Potential for common cause failures 

is taken into account. 

The Commission’s position is that the 

containment and its systems are 

important in the preservation of 

defense-in-depth (in terms of both large 

early and large late releases). Therefore, 

as part of meeting the defense-in-depth 

principle, a licensee should demonstrate 

that the function of the containment as 

a barrier (including fission product 

retention and removal) is not 

significantly degraded when SSCs that 

support the functions are moved to 

RISC–3 (e.g., containment isolation or 

containment heat removal systems). The 

concepts used to address defense-indepth for functions required to prevent 

core damage may also be useful in 

addressing issues related to those SSCs 

that are required to preserve long-term 

containment integrity. Where a licensee 

categorizes containment isolation valves 

or penetrations as RISC–3, the licensee 

should address the impact of the change 

in treatment to ensure that defense-indepth continues to be satisfied. Where 

the impact of changes in treatment does 

not support the reliability assumptions 

in the categorization process, the 

licensee should resolve this situation by 

adjustments to the categorization 

process assumptions or treatment of the 

component. 

V.4.4 Section 50.69(c)(1)(iv) Include 

Evaluations To Provide Reasonable 

Confidence That Sufficient Safety 

Margins Are Maintained and That Any 

Potential Increases in CDF and LERF 

Resulting From Changes in Treatment 

Permitted by Implementation of 

§ 50.69(b)(1) and § 50.69(d)(2) Are Small 

Section 50.69(c)(1)(iv) specifies that 

the categorization process include 

evaluations to provide reasonable 

confidence that as a result of 

implementation of revised treatment 

permitted for RISC–3 SSCs, sufficient 

safety margins are maintained and any 

potential increases in CDF and LERF are 

small. Safety margins can be maintained 

if the licensee maintains the 

functionality of the SSCs following 

implementation of the revised 

requirements and if periodic inspection, 

testing, and corrective action activities 

are adequate to prevent, detect and 

correct significant SSC performance and 

reliability degradation. Later sections of 

this SOC provide discussion on the 

treatment the licensee will implement to 

ensure with reasonable confidence that 

RISC–3 SSCs remain capable of 

performing their safety functions under 

design basis conditions. The 

requirements of the rule to show that 

sufficient safety margins are maintained 

and that potential increases in risk are 

acceptably small are discussed below. 

As part of their submittal, a licensee 

or applicant is to describe the 

evaluations to be conducted for 

purposes of providing reasonable 

confidence that there would be no more 

than an acceptably small (potential) 

increase in risk. For SSCs included in 

the PRA, the Commission expects a risk 

sensitivity study (evaluation) to be 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68039

performed to provide a basis for 

concluding that if the reliability of these 

RISC–3 SSCs should collectively 

degrade because of the changes in 

treatment, the potential risk increase 

would be small. Satisfying the rule 

requirement that the risk increase is 

acceptably small presumes that the 

increase in failure rates credited in the 

PRA risk sensitivity study bounds any 

reasonable estimate of the increase that 

may be expected as a result of the 

changes in treatment; also considering 

the feedback and corrective action 

aspects of the rule. 

The categorization process 

encompasses both active and passive 

functions of SSCs. Section 

50.69(b)(2)(iv) includes the requirement 

that the change-in-risk evaluations 

performed to satisfy § 50.69(c)(1)(iv) 

must address potential impacts from 

known degradation mechanisms on both 

active and passive functions. The 

manner of addressing these potential 

impacts may be either qualitative or 

quantitative and may rely on the 

maintenance of current programs that 

address these degradation mechanisms 

(e.g., microbiologically-induced 

corrosion, flow-assisted corrosion) and/

or may incorporate existing riskinformed approaches (e.g., riskinformed inservice inspection). 

One mechanism that could lead to 

large increases in CDF/LERF is 

extensive, across system common cause 

failures. These CCFs could occur where 

the mechanisms that lead to failure, in 

the absence of special treatment, are 

sufficiently rapidly developing or are 

not self-revealing that there would be 

few opportunities for early detection 

and corrective action. Thus, when 

deciding how much to assume that SSC 

reliability might change, the applicant 

or licensee is expected to consider 

potential effects of common-cause 

interaction susceptibility, including 

cross-system interactions and potential 

impacts from known degradation 

mechanisms; while also considering the 

feedback and corrective actions aspects 

of the rule. 

Those aspects of treatment that are 

necessary to prevent SSC degradation or 

failure from known degradation 

mechanisms, to the extent that the 

results of the evaluations are 

invalidated, must be retained. 

Identifying those aspects will involve an 

understanding of what the degradation 

mechanisms are and what elements of 

treatment are sufficient to prevent the 

degradation.

The treatment for all RISC–3 SSCs 

may not be the same. As an example, 

motor operated valves (MOVs) operating 

in a severe environment (e.g., in the 

steam tunnel) would be more 

susceptible to failure because of grease 

degradation if they were not regularly 

maintained and tested. However, not all 

MOVs, even if they have the same 

design and are identical in other 

respects, will be exposed to the same 

environment. Therefore, the other 

MOVs may not be as susceptible to 

failure as those in the steam tunnel and 

less frequent maintenance and testing 

would be acceptable. While it may be 

simpler to increase the unreliability or 

unavailability of all the RISC–3 SSCs by 

a certain bounding factor to demonstrate 

that the change in risk is acceptably 

small, this example suggests that it may 

also be appropriate to use different 

factors for different groups of SSCs 

depending on the impact of reducing 

treatment on those SSCs. 

Section 50.69(c)(1)(iv) requires 

reasonable confidence that the increase 

in the overall plant CDF and LERF 

resulting from potential decreases in the 

reliability of RISC–3 SSCs as a result of 

the changes in treatment, be small. The 

rule further requires the licensee or 

applicant to describe the evaluations to 

be performed to meet this requirement. 

As presented in RG 1.174, the NRC 

considers small changes to be relative 

and to depend on the current plant CDF 

and LERF (hence we also refer to 

‘‘acceptably small’’ changes in other 

portions of this notice since small can 

be different for different plants with 

different baseline levels of risk). For 

plants with total baseline CDF of 10¥4

per year or less, small means CDF 

increases of up to 10¥5 per year and for 

plants with total baseline CDF greater 

than 10¥4 per year, small means CDF 

increases of up to 10¥6 per year. 

However, if there is an indication that 

the CDF may be considerably higher 

than 10¥4 per year, the focus of the 

licensee should be on finding ways to 

decrease rather than increase CDF and 

the licensee may be required to present 

arguments as to why steps should not be 

taken to reduce CDF for the reduction in 

special treatment requirements to be 

considered. For plants with total 

baseline LERF of 10¥5 per year or less, 

small LERF increases are considered to 

be up to 10¥6 per year, and for plants 

with total baseline LERF greater than 

10¥5 per year, small LERF increases are 

considered to be up to 10¥7 per year. 

However, if there is an indication that 

the baseline CDF or LERF may be 

considerably higher than 10¥4 or 10¥5, 

respectively, the licensee either must 

find ways to reduce risk and present the 

arguments to the staff before 

implementation of § 50.69, otherwise it 

is likely that the staff will reject the 

§ 50.69 application. This is consistent 

with the guidance in Section 2.2.4 of RG 

1.174. It should be noted that this 

allowed increase shall be applied to the 

overall categorization process, even for 

those licensees that will implement 

§ 50.69 in a phased manner. 

If a PRA model does not exist for the 

external initiating events or the low 

power and shutdown operating modes, 

justification should be provided, on the 

basis of bounding analyses or qualitative 

considerations, that the effect on risk 

(from the unmodeled events or modes of 

operation) is not significant and that the 

total effect on risk from modeled and 

unmodeled events and modes of 

operation is small, consistent with 

Section 2.2.4 of RG 1.174. 

V.4.5 Section 50.69(c)(1)(v) System 

or Structure Level Review 

Section 50.69(c)(1)(v) specifies that 

the categorization be done at the system 

or structure level; not for selected 

components within a system. A licensee 

or applicant is allowed to implement 

§ 50.69 for a subset of the plant systems 

and structures (i.e., partial 

implementation) and to phase in 

implementation over time. However, the 

implementation, including the 

categorization process, must address 

entire systems or structures; not selected 

components within a system or 

structure. Note that this requirement 

should be understood to exclude entire 

support systems (e.g., if system A is 

categorized as RISC–3, but is dependent 

on system B components which in turn 

have been categorized as RISC–1, then 

system A is understood not to include 

the system B components and is not to 

be categorized as RISC–1). This required 

scope ensures that all safety functions 

associated with a system or structure are 

properly identified and evaluated when 

determining the safety significance of 

individual components within a system 

or structure and that the entire set of 

components that comprise a system or 

structure are considered and addressed. 

V.4.6 Section 50.69(c)(2) Use of 

Integrated Decision-Making Panel 

Section 50.69(c)(2) sets forth the 

requirements for using an IDP to make 

the determination of safety significance, 

and for the composition of the IDP. The 

fundamental requirement for the 

categorization process (as stated in 

§ 50.69(c)(1)(ii)) is that it include use of 

an integrated systematic process. The 

determination of safety significance of 

SSCs is to be performed as part of an 

integrated decision-making process. By 

‘‘integrated decision-making process,’’

the Commission means a process that 

integrates both risk insights and 

68040 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

traditional engineering insights. In 

categorizing SSCs as low safetysignificant, defense-in-depth must be 

maintained (per § 50.69(c)(1)(iii)) and 

there must be reasonable confidence 

that sufficient safety margin is 

maintained by showing that any 

increases in risk are small per 

§ 50.69(c)(1)(iv). To account for each of 

these factors and to account for risk 

insights not found in the plant-specific 

PRA, § 50.69(c)(2) requires that the final 

categorization of each SSC be performed 

using an integrated decision-making 

panel (IDP). A structured and systematic 

process using documented criteria must 

be used to guide the decision-making 

process. Categorization is an iterative 

process based on expert judgment to 

integrate the qualitative and quantitative 

elements that impact SSC safety 

significance. The insights and varied 

experience of IDP members are relied on 

to ensure that the final result reflects a 

comprehensive and justifiable 

judgment. 

The panel must be composed of 

experienced personnel who possess 

diverse knowledge and insights in plant 

design and operation and who are 

capable in the use of deterministic 

knowledge and risk insights in making 

SSC classifications. The NRC places 

significant reliance on the capability of 

a licensee to implement a robust 

categorization process that relies heavily 

on the skills, knowledge, and 

experience of the people that implement 

the process, in particular on the 

qualifications of the members of the 

IDP. The IDP must be composed of a 

group of individuals who collectively 

have expertise in plant operation, 

design (mechanical and electrical) 

engineering, system engineering, safety 

analysis, and probabilistic risk 

assessment. At least three members of 

the IDP should have a minimum of five 

years experience at the plant, and there 

should be at least one member of the 

IDP who has worked on the modeling 

and updating of the plant-specific PRA 

for a minimum of three years. 

The IDP should be trained in the 

specific technical aspects and 

requirements related to the 

categorization process. Training should 

address at a minimum the purpose of 

the categorization; present treatment 

requirements for SSCs including 

requirements for design basis events; 

PRA fundamentals; details of the plantspecific PRA including the modeling, 

scope, and key assumptions, the 

interpretation of risk importance 

measures, and the role of sensitivity 

studies and the change-in-risk 

evaluations; and the defense-in-depth 

philosophy and requirements to 

maintain defense-in-depth. 

The licensee or applicant (through the 

IDP) shall document its decision criteria 

for categorizing SSCs as safety 

significant or low safety significant 

pursuant to § 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 § 50.69(e)(1). 

V.5.0 Section 50.69(d) Treatment 

Requirements for Structures, Systems, 

and Components 

Treatment requirements applicable to 

RISC–1, RISC–2, and RISC–3 SSCs are 

specified in § 50.69(d). Any regulatory 

requirements applicable to RISC–1, 

RISC–2, RISC–3, and RISC–4 SSCs not 

removed by § 50.69(b)(1) continue to 

apply. 

V.5.1 Section 50.69(d)(1) RISC–1 and 

RISC–2 Treatment 

Section 50.69(d)(1) requires that a 

licensee or applicant ensure that RISC–

1 and RISC–2 SSCs perform their 

functions consistent with the 

categorization process assumptions by 

evaluating treatment being applied to 

these SSCs to ensure that it supports the 

key assumptions in the categorization 

process that relate to their assumed 

performance. This rule language means 

that the licensee or applicant must 

evaluate the treatment associated with 

those key assumptions in the PRA that 

relate to performance of particular SSCs. 

For example, if a relief valve was being 

credited with capability to relieve water 

(as opposed to its design condition of 

steam), such an evaluation would look 

at whether the component has been 

determined to be able to perform as 

assumed.

Because RISC–1 and RISC–2 SSCs are 

the safety significant SSCs and their 

performance as credited in the PRA is 

important to maintaining an acceptable 

level of plant risk, given that special 

treatment requirements are being 

removed from RISC–3 SSCs, it is a key 

and necessary part of § 50.69 to ensure 

these SSCs can perform as credited in 

the PRA. However, the requirements in 

§ 50.69(d)(1) do not extend special 

treatment requirements to RISC–1 

beyond design basis functions and to 

RISC–2 SSCs. 

The performance conditions for 

beyond design basis capabilities of 

RISC–1 SSCs credited in the PRA are 

not subject to the requirements of 10 

CFR Part 50, Appendix B. However, 

plant SSCs credited for beyond design 

basis capabilities must have a valid 

technical basis for the credit (i.e., the 

failure rate/probability of the SSC 

performing the beyond design basis 

function) given in the PRA. Further, the 

basis for this credit should already be 

established and documented in the PRA 

supporting documentation so this 

should not be an additional burden for 

licensees to capture and implement. If 

an existing technical basis does not exist 

or is insufficient to support the credit 

taken for beyond design basis capability 

(e.g., the supporting test program does 

not test the SSC at the beyond design 

basis conditions), the licensee or 

applicant is required by § 50.69(d)(1) to 

develop a technical basis for the credit 

taken in the PRA potentially including 

a treatment program for the SSC that 

validates the capability credited. 

For SSCs categorized as RISC–1 or 

RISC–2, all existing applicable 

requirements continue to apply (i.e., no 

special treatment requirements are 

removed by § 50.69). This rule does not 

require licensees to evaluate the 

effectiveness of special treatment 

requirements for RISC–1 SSCs to ensure 

that they are capable of performing their 

design basis functions. The special 

treatment requirements in other NRC 

regulations address the design basis 

capability of RISC–1 SSCs. 

The categorization process will result 

in a number of safety-related SSCs being 

determined to be of low safety 

significance (i.e., RISC–3) and subject to 

reduced treatment. This determination 

of low safety significance will implicitly 

take credit for the performance 

capability of other SSCs in the PRA, 

some, or all of which, may not be 

included in the scope of the licensee’s 

categorization process (due to the 

allowance for licensees to selectively 

implement the rule and to phase that 

implementation over time). To maintain 

the validity of the categorization 

process, and more importantly to 

maintain any potential risk increase as 

small, it is necessary to maintain the 

‘‘credited’’ SSCs per § 50.69, and this 

means the application of § 50.69(d)(1) 

and § 50.69(e)(2) requirements. 

V.5.2 Section 50.69(d)(2) RISC–3 

Treatment 

Section 50.69(d)(2) requires that the 

licensee or applicant must ensure with 

reasonable confidence that RISC–3 SSCs 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68041

remain capable of performing their 

safety-related functions under design 

basis conditions, including seismic 

conditions and environmental 

conditions and effects throughout their 

service life. By ‘‘reasonable confidence’’, 

the Commission means that the licensee 

or applicant is required to provide a 

‘‘reasonable confidence’’ level with 

regard to maintaining the capability of 

RISC–3 safety-related functions. As 

indicated previously in this notice, 

‘‘reasonable confidence’’ is a level of 

confidence that is both less than that 

associated with RISC–1 SSCs which are 

subject to all the special treatment 

requirements, and consistent with their 

individual low safety significance. The 

term ‘‘ensure’’ is intended to convey the 

Commission’s determination that the 

licensee is under a legally-binding 

regulatory requirement to provide the 

requisite ‘‘reasonable confidence.’’

Although § 50.69(b)(1) removes for 

RISC–3 SSCs the environmental 

qualification requirements of § 50.49, it 

does not eliminate the requirements in 

10 CFR part 50, Appendix A, ‘‘General 

Design Criteria for Nuclear Power 

Plants,’’ that electric equipment 

important to safety be capable of 

performing their intended functions 

under the applicable environmental 

conditions. For example, GDC–4 of 10 

CFR part 50, Appendix A, requires that 

SSCs important to safety be designed to 

accommodate the effects of, and to be 

compatible with, the environmental 

conditions and effects associated with 

normal operation, maintenance, testing, 

and postulated accidents. To satisfy the 

provisions of GDC–4 of 10 CFR part 50, 

Appendix A, the licensee or applicant 

must address environmental conditions 

such as temperature, pressure, 

humidity, chemical effects, radiation, 

and submergence; and environmental 

effects such as aging and synergisms. In 

accordance with § 50.69(d)(2), the 

licensee or applicant must design 

electric equipment important to safety 

so they are capable of performing their 

intended functions under applicable 

environmental conditions and effects 

throughout their service life. If RISC–3 

electrical equipment is relied on to 

perform its safety-related function 

beyond its design life, § 50.69(d)(2) 

requires the licensee or applicant to 

have a basis for the continued capability 

of the equipment under adverse 

environmental conditions and effects. 

Under § 50.69, RISC–3 SSCs would 

continue to be required to function 

under design basis seismic conditions 

(such as design load combinations of 

normal and accident conditions with 

earthquake motions), but would not be 

required to be qualified by testing or 

specific engineering methods in 

accordance with the requirements stated 

in 10 CFR part 100, Appendix A. A 

licensee or applicant who adopts the 

rule would no longer be required to 

meet certain requirements in Appendix 

A to part 100, Sections VI(a)(1) and 

VI(a)(2), to the extent that those 

requirements have been interpreted as 

mandating qualification testing and 

specific engineering methods to 

demonstrate that RISC–3 SSCs are 

designed to withstand the Safe 

Shutdown Earthquake and Operating 

Basis Earthquakes. The rule does not 

remove the design requirements related 

to the capability of RISC–3 SSCs to 

remain functional considering Safe 

Shutdown Earthquake and Operating 

Basis Earthquake seismic loads, 

including applicable concurrent loads. 

The rule does not change the design 

input earthquake loads (magnitude of 

the loads and number of events) or the 

required load combinations used in the 

design of RISC–3 SSCs. For example, for 

the replacement of an existing safetyrelated SSC that is subsequently 

categorized as RISC–3, the same seismic 

design loads and load combinations 

would still apply. The rule would 

permit the licensee or applicant to select 

a technically defensible method to show 

that RISC–3 SSCs will remain functional 

when subject to design earthquake 

loads. Several public comments on the 

proposed rule supported the use of 

earthquake experience data as a method 

to demonstrate SSCs will remain 

functional during earthquakes. If the 

licensee or applicant chooses to use 

only earthquake experience data to 

demonstrate that the SSC will perform 

its safety-related function, with no 

further engineering evaluation, then the 

earthquake experience data must 

envelope the SSC design basis, 

including the number of earthquake 

events and the design load 

combinations. Additionally, if the SSC 

is required to function during or after 

the earthquake, the experience data 

would need to contain explicit 

information that the SSC actually 

functioned during or after the design 

basis earthquake events as required by 

the SSC design basis. The successful 

performance of an SSC after the 

earthquake event does not demonstrate 

it would have functioned during the 

event. Implementation of § 50.69 does 

not change the seismic design basis for 

USI A–46 facilities and, therefore, does 

not impose additional requirements on 

these facilities. 

Section 50.69(d)(2) should not be 

interpreted to extend or expand design 

basis conditions to SSCs where such 

conditions were not previously part of 

its design basis. 

Section 50.69(d)(2) requires that the 

treatment of RISC–3 SSCs be consistent 

with the categorization process. This 

rule language means that, when 

establishing the treatment for RISC–3 

SSCs, the licensee or applicant must 

take into account the assumptions in the 

categorization process regarding the 

design basis capability and reliability of 

RISC–3 SSCs to perform their safetyrelated functions throughout their 

service life. The evaluation by the 

licensee or applicant of the consistency 

of the treatment of RISC–3 SSCs with 

the categorization process may be 

qualitative so long as it provides 

reasonable confidence of the design 

basis capability of RISC–3 SSCs, based 

on plant-specific and industry-wide 

operational experience and vendor 

information. In establishing treatment 

for RISC–3 SSCs, the licensee or 

applicant is responsible for addressing 

applicable vendor recommendations 

and operational experience such that 

the treatment established for RISC–3 

SSCs provides reasonable confidence for 

design basis capability. For example, 

operational experience might be 

described in NRC information notices or 

identified in responses to NRC bulletins, 

generic letters, or other licensee 

commitment documents. The treatment 

applied to RISC–3 SSCs must also 

support the assumptions used in 

justifying the removal of requirements 

applicable to those SSCs. For example, 

where a licensee or applicant intends as 

part of implementing § 50.69 to 

eliminate leakage testing required in 10 

CFR Part 50, Appendix J, for 

containment isolation valves, the 

treatment applied to those valves must 

support the assumption that they are 

capable of closing under design basis 

conditions. 

Some public comments on the 

proposed rule suggested that a reference 

to general industrial practices would be 

sufficient to satisfy the requirements for 

the treatment for RISC–3 SSCs. 

However, as described in NUREG/CR–

6752, ‘‘A Comparative Analysis of 

Special Treatment Requirements for 

Systems, Structures, and Components 

(SSCs) of Nuclear Power Plants with 

Commercial Requirements of NonNuclear Power Plants,’’ significant 

variation exists in the application of 

industrial practices at nuclear power 

plants. Hence, a simple reference to 

these practices does not provide a basis 

to satisfy the rule’s requirements. To 

satisfy the requirement that the 

treatment of RISC–3 SSCs be consistent 

with the categorization process, the 

licensee or applicant must establish 

68042 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

treatment that provides reasonable 

confidence SSCs perform their safetyrelated functions under design basis 

conditions and is consistent with the 

assumptions in the categorization 

process (e.g., reliability levels). The 

licensee or applicant must either 

establish treatment that provides this 

level of reliability or use consensus 

standards that provide a proven level of 

reliability based on experience. In using 

consensus standards, the licensee or 

applicant must note that combining or 

omitting provisions of standards might 

result in ineffective implementation of 

§ 50.69 by causing RISC–3 SSCs to be 

incapable of performing their design 

basis safety functions. The NRC 

considers the ASME code cases 

endorsed in § 50.55a and listed in RG 

1.84, 1.147, and 1.192 to be one 

acceptable method of establishing 

treatment of RISC–3 SSCs, where 

applicable, in that those applicable 

endorsed code cases adjust treatment 

based on the safety significance of the 

components.

Under § 50.69, most special treatment 

requirements will be removed from 

RISC–3 SSCs, which will typically 

comprise a large percentage of safetyrelated SSCs in a nuclear power plant. 

These special treatment requirements 

will be replaced with the high-level 

treatment requirements in § 50.69(d)(2) 

that will allow significant reduction in 

the treatment applied to RISC–3 SSCs. 

This reduction in treatment can 

introduce common-cause concerns and 

weaken defenses against them. 

Therefore, § 50.69(d)(2) requires that 

inspection, testing and corrective action 

be provided for RISC–3 SSCs. The 

inspection and testing requirement in 

§ 50.69(d)(2)(i) is to provide sufficient 

performance data for RISC–3 SSCs to 

determine if the reduction in treatment 

has adversely affected their design basis 

capability and to provide reasonable 

confidence that the SSC can perform its 

safety function throughout their service 

life. The corrective action requirement 

in § 50.69(d)(2)(ii) is to address SSC 

failures and provide reasonable 

confidence in avoiding future problems. 

These requirements are necessary to 

provide reasonable confidence that 

RISC–3 safety related functional 

capability is maintained and thereby 

avoid adverse impacts on the reliability 

and availability of multiple RISC–3 

SSCs, which could reduce plant safety 

beyond the categorization process 

assumptions or results and invalidate 

the risk sensitivity results. 

A licensee or applicant may not 

simply assume that a sensitivity study 

that increases the failure probability for 

all RISC–3 SSCs simultaneously, with 

no additional basis to support it, would 

necessarily bound the potential change 

in risk that could result due to 

implementation of § 50.69. There is a 

potential that risk due to 

implementation of § 50.69 could 

increase as a result of the reduction in 

treatment due to common-cause 

interactions or degradation, and this 

impact might not be uniform across the 

population of RISC–3 SSCs. For 

example, if a licensee were to simply 

eliminate maintenance, testing, or 

lubrication of pumps or valves, it could 

significantly impact performance of 

those specific components and the 

impact might exceed the cumulative 

impact of individually reducing the 

reliability of all RISC–3 SSCs by a few 

percent or less. Public comments on the 

proposed rule indicated that crosssystem common-cause interactions and 

degradation mechanisms are typically 

addressed through the treatment 

processes applied to plant equipment, 

rather than being addressed in the 

categorization process. In satisfying the 

rule, the licensee or applicant must 

consider potential common-cause 

interactions and degradation 

mechanisms in establishing treatment 

for RISC–3 SSCs so there is a reasonable 

basis to support the assumptions made 

for the risk sensitivity study. 

V.5.2.1 Section 50.69(d)(2)(i)

Inspection and Testing 

Section 50.69(d)(2)(i) requires the 

licensee to conduct periodic inspection 

and testing activities to determine 

whether RISC–3 SSCs will remain 

capable of performing their safetyrelated functions under design basis 

conditions. 

The prescriptive special treatment 

requirements in §§ 50.55a and 50.65 for 

inspection, testing, and surveillance 

have been removed for RISC–3 SSCs. In 

lieu of those prescriptive requirements, 

the final rule requires the licensee or 

applicant to implement inspection and 

testing of RISC–3 SSCs sufficient to 

provide reasonable confidence that 

RISC–3 SSCs remain capable of 

performing their safety-related functions 

under design basis conditions 

throughout their service life. The 

licensee or applicant may apply 

industrial practices for the treatment of 

RISC–3 SSCs if those practices maintain 

the capability of the RISC–3 SSCs to 

perform their design-basis safety 

functions. 

With respect to RISC–3 pumps and 

valves, the rule language in 

§ 50.69(d)(2)(i) means that the licensee 

or applicant must implement periodic 

testing or inspection sufficient to 

provide reasonable confidence that 

these pumps and valves will be capable 

of performing their safety-related 

functions under design basis conditions. 

To determine that the pump or valve 

will remain capable of performing its 

safety-related function, the licensee or 

applicant will need to obtain sufficient 

operational information or performance 

data to provide with reasonable 

confidence that the RISC–3 pumps and 

valves will be capable of performing 

their safety-related functions if called 

upon to function under operational or 

design basis conditions over the interval 

between periodic testing or inspections. 

In addition, the operational information 

and performance data must be sufficient 

to satisfy the requirements of 

§ 50.69(d)(2)(i) for use in identifying the 

need for corrective action under 

§ 50.69(d)(2)(ii) and in providing 

information for feedback to the 

categorization and treatment processes 

under § 50.69(e)(3). 

In some cases, a licensee or applicant 

implementing § 50.69 might apply more 

rigorous test methods than previously 

applied to satisfy the ASME Code 

inservice testing provisions because 

§ 50.69 does not specify restrictive time 

limits on test intervals that were 

provided in the ASME Code. As a result, 

§ 50.69 allows significant flexibility by 

the licensee or applicant in verifying the 

design basis capability of their safetyrelated SSCs categorized as RISC–3. 

However, the licensee or applicant 

needs to consider the lessons learned 

over the last 20 years regarding SSC 

performance in establishing the 

treatment for RISC–3 SSCs. Contrary to 

suggestions in some public comments 

on the proposed rule, operating 

experience and research does not 

support an assumption that exercising a 

valve or pump will provide reasonable 

confidence of design-basis capability in 

that such exercising will not detect 

service-induced aging or degradation 

that could prevent the component from 

performing its design basis functions in 

the future, and therefore is insufficient 

by itself to satisfy § 50.69(d)(2)(i). The 

licensee or applicant may develop the 

type and frequency of tests or 

inspections for RISC–3 pumps and 

valves provided they are sufficient to 

conclude that the pump or valve will 

perform its safety-related function 

throughout the service life. The 

provisions for risk-informed inspection 

and testing in applicable ASME code 

cases (as incorporated in § 50.55a) 

would constitute one effective approach 

for satisfying the § 50.69 requirements. 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68043

V.5.2.2 Section 50.69(d)(2)(ii)

Corrective Action Process 

Section 50.69(d)(2)(ii) requires that 

conditions that would prevent a RISC–

3 SSC from performing its safety-related 

functions under design basis conditions 

must be corrected in a timely manner. 

In the case of significant conditions 

adverse to quality, the rule requires that 

measures be taken to provide reasonable 

confidence that the cause of the 

condition is determined and corrective 

action taken to preclude repetition. 

Significant conditions adverse to quality 

include common-cause concerns for 

multiple RISC–3 SSCs or concerns 

related to the validity of the 

categorization process or its results. For 

example, if measuring and test 

equipment is found to be in error or 

defective, the licensee or applicant will 

be responsible for determining the 

functionality of safety-related SSCs 

checked using that equipment to 

prevent the occurrence of commoncause problems that might invalidate 

the categorization process assumptions 

and results. Effective implementation of 

the corrective action process would 

include timely response to information 

from plant SSCs, overall plant 

operations, and industry generic 

activities that might reveal performance 

concerns for RISC–3 SSCs on both an 

individual and common-cause basis. 

Contrary to some public comments on 

the proposed rule, the corrective action 

process alone is insufficient to monitor 

the effects of reduced treatment on 

RISC–3 SSCs, and therefore the 

Commission has incorporated feedback 

requirements into § 50.69. 

V.6.0 Section 50.69(e) Feedback and 

Process Adjustment 

Section 50.69(e)(1) requires the 

licensee or applicant to review changes 

to the plant, operational practices, 

applicable plant and industry 

operational experience and, as 

appropriate, update the PRA and SSC 

categorization and treatment processes, 

in a timely manner, but no longer than 

every two refueling outages for RISC–1, 

RISC–2, RISC–3, and RISC–4 SSCs. The 

date the NRC grants the license 

amendment to implement 10 CFR 50.69 

begins the updating interval and 

provides a recognizable date for the 

periodic updating of the categorization 

and treatment processes. Depending on 

the timing of license amendment 

issuance (for example, just before a 

refueling outage), the licensee or 

applicant might have minimal plant 

changes, operational practices, or 

operational experience to review in 

updating the categorization and 

treatment processes in the early phases 

of implementing the rule. If plant 

changes, operational practices, or 

operational experience would result in 

a significant adverse impact on plant 

safety or public health and safety, the 

licensee or applicant must update the 

categorization or treatment processes in 

a timely manner without waiting for the 

two refueling outage schedule. The 

information collected under 

§ 50.69(e)(2) and (e)(3) would be among 

the information used to determine the 

need for updating the categorization or 

treatment processes in a timely manner 

required under § 50.69(e)(1). The plant 

and industry operational experience 

referred to in § 50.69(e)(1) includes the 

data collected under § 50.69(e)(3) for 

RISC–3 SSCs. In addition to the periodic 

updating of the quantitative reliability 

information, the feedback of plant 

operational experience is intended to 

include qualitative information on the 

performance of plant SSCs obtained 

through the corrective action program 

and processes as well as from applicable 

vendor recommendations and 

operational experience. For example, 

lessons learned from operational 

experience might be described in NRC 

information notices or implemented in 

response to NRC bulletins or generic 

letters. The evaluation of the 

categorization process includes 

verifying the continued validity of the 

risk sensitivity study and the associated 

SSC performance assumptions.

Section 50.69(e)(2) requires the 

licensee or applicant to monitor the 

performance of RISC–1 and RISC–2 

SSCs and make adjustments as 

necessary to either the categorization 

(i.e., by moving other RISC–3 or RISC–

4 SSCs back into RISC–1 or RISC–2 

until the change in risk is acceptably 

small) or treatment processes so the 

categorization process and results are 

maintained valid. To meet this 

requirement, the licensee or applicant 

must monitor all unavailabilities and 

functional failures so they can 

determine when adjustments to the 

categorization or treatment processes are 

needed. The licensee or applicant will 

also need to monitor SSCs that are 

credited in the PRA for performing 

beyond design basis functions (if 

applicable) that are not necessarily 

included in the scope of an existing 

maintenance rule program. 

The categorization process will result 

in a number of safety-related SSCs being 

determined to be of low safety 

significance (i.e., RISC–3) and subject to 

reduced treatment. This determination 

of low safety significance will implicitly 

take credit for the performance 

capability of other SSCs in the PRA, 

some, or all of which, may not be 

included in the scope of the licensee’s 

categorization process (due to the 

allowance for licensees to selectively 

implement the rule and to phase that 

implementation over time). To maintain 

the validity of the categorization 

process, and more importantly to 

maintain any potential risk increase as 

small, it is necessary to maintain the 

‘‘credited’’ SSCs per § 50.69. 

In § 50.69(e)(3) the rule requires the 

licensee or applicant to consider the 

performance data collected in 

§ 50.69(d)(2)(i) for RISC–3 SSCs to 

determine whether there are any 

adverse changes in performance such 

that the SSC unreliability values 

approach or exceed the values used in 

the evaluations conducted to meet 

§ 50.69(c)(iv) and to make adjustments 

as necessary to either the categorization 

or treatment processes so the 

categorization process and results are 

maintained valid. Based on the review 

of this information, if SSC reliability 

degrades so as not to support the 

categorization process assumptions, the 

licensee or applicant must adjust the 

treatment to improve SSC reliability or 

make appropriate changes to the 

categorization of SSCs. 

V.7.0 Section 50.69(f) Program 

Documentation and Change Control and 

Records 

Section 50.69(f) contains 

administrative requirements for keeping 

information current, handling planned 

changes to programs and processes, and 

records. Each requirement is discussed 

below. 

Section 50.69(f)(1) states that the 

licensee or applicant shall document the 

basis for categorization of SSCs in 

accordance with this section before 

removing any requirements. The 

documentation must address why a 

component was determined to be either 

safety significant or low safety 

significant based upon the requirements 

in § 50.69(c). 

Section 50.69(f)(2) specifies that the 

licensee must update its FSAR to reflect 

which systems have been categorized 

using the provisions of § 50.69. Systems 

that are categorized by § 50.69 will have 

their treatment revised consistent with 

the RISC category into which the SSC is 

categorized and the associated treatment 

requirements of § 50.69(d). This 

provision is included to maintain clear 

information, at a minimum level of 

detail, about which requirements a 

licensee is satisfying. However, detailed 

information about particular SSCs is not 

required to be submitted to the NRC. For 

an applicant, this updating would be 

expected to be either part of the original 

68044 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

application or as a supplement to the 

FSAR under § 50.34(b). For licensees, 

the updating must be in accordance 

with the provisions of § 50.71(e). 

Once the NRC has completed its 

review of a § 50.69 application, the 

licensee can adjust its treatment 

processes provided that the 

requirements of § 50.69 are met. NRC 

does not plan to perform a preimplementation review of the revised 

treatment requirements under 

§ 50.69(d). However, the Commission 

recognizes that existing information in 

the quality assurance (QA) plan or in 

the FSAR may need to be revised to 

reflect the changes to treatment that are 

made as a result of implementation of 

§ 50.69. Any revisions to these 

documents are to be submitted to NRC 

in accordance with the existing 

requirements of § 50.54(a)(2) and 

§ 50.71(e), respectively. 

Section 50.69(f)(3) specifies that for 

initial implementation of the rule, 

changes to the FSAR for implementation 

of this rule need not include a 

supporting § 50.59 evaluation of 

changes directly related to 

implementation. Future changes to the 

treatment processes and procedures for 

§ 50.69 implementation may be made, 

provided the requirements of the rule 

and § 50.59 continue to be met. While 

the licensee is to update its programs to 

reflect implementation of § 50.69, the 

Commission concluded that no 

additional review under § 50.59 is 

necessary for such changes to these 

parts of the FSAR that might occur. 

Section 50.69(f)(4) specifies that for 

initial implementation of the rule, 

changes to the quality assurance plan 

directly related to implementation of 

this rule need not be considered a 

reduction in commitment for the 

purposes of § 50.54(a). Future changes 

to the treatment processes and 

procedures for § 50.69 implementation 

may also be made, provided the 

requirements of the rule and § 50.54(a) 

continue to be met. While the licensee 

is to update its programs to reflect 

implementation of § 50.69, the 

Commission concluded that no 

additional NRC staff review under 

§ 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 § 50.69(c). At this 

time, the NRC is unable to determine 

generic criteria for the control of 

changes to the categorization process 

during its implementation that could be 

included in § 50.69. As a result, the NRC 

will review and approve a license 

amendment submittal containing the 

licensee or applicant’s categorization 

process and intends to impose a license 

condition upon which the 

categorization process approval is based 

to control categorization process 

changes. The license condition will 

require the licensee to notify the NRC in 

advance of implementing changes with 

respect to specific aspects of the 

categorization process. With experience 

in the application of § 50.69, the NRC 

might modify the rule to specify generic 

criteria for the control of changes to the 

categorization process during 

implementation of the rule. 

No explicit requirements are included 

in § 50.69 for the period for retention of 

records. The rule specifies only a few 

specific types of records that must be 

prepared (e.g., those for the basis for 

categorization in § 50.69(f)(1)). In 

accordance with § 50.71(c), these 

records are to be maintained until the 

Commission terminates the facility 

license.

V.8.0 Section 50.69(g) Reporting 

Section 50.69(g) provides a new 

reporting requirement applicable to 

events or conditions that prevented, or 

would have prevented, a RISC–1 or 

RISC–2 SSC from performing a safety 

significant function. Most events 

involving these SSCs will meet existing 

§ 50.72 and § 50.73 reporting criteria. 

However, it is possible for events and 

conditions to arise that impact whether 

RISC–1 or RISC–2 SSCs would perform 

beyond design basis functions 

consistent with the performance 

capability credited in the categorization 

process. This reporting requirement is 

intended to capture these situations. 

The reporting requirement is contained 

in § 50.69, rather than as a revision of 

§ 50.73, so that its applicability only to 

those facilities that have implemented 

§ 50.69 is clear. The existing reporting 

requirements in § 50.72 and § 50.73 are 

removed for RISC–3 (and RISC–4) SSCs 

under § 50.69(b)(vii) and (viii). 

V.9.0 Inspection of 10 CFR 50.69 

Implementation 

The NRC will review and update, as 

appropriate, the current inspection 

procedures under the NRC Reactor 

Oversight Process to incorporate 

inspection guidance for monitoring the 

implementation of § 50.69 at nuclear 

power plants. The NRC intends to 

conduct sample inspections of plants 

implementing § 50.69 in a manner that 

is sensitive to conditions that could 

significantly increase risk. These sample 

inspections are intended to gather 

information that will enable the NRC to 

assess whether modifications are 

needed to the ongoing baseline 

inspection program. The sample 

inspections will focus on the 

implementation of the categorization 

process approved as part of the NRC 

review of the § 50.69 license 

amendment request. The sample 

inspections will also evaluate the 

treatment established under § 50.69 

with primary attention directed to 

programmatic and common-cause 

issues; including those associated with 

known degradation mechanisms. The 

inspections might help provide 

operating experience information on 

RISC–3 SSCs that can also be provided 

to other licensees. 

VI. Guidance 

VI.1 Regulatory Guide and 

Implementation Guidance for § 50.69 

NEI submitted a proposed 

implementation guide for this 

rulemaking in the form of NEI 00–04, 

‘‘10 CFR 50.69 SSC Categorization 

Guideline’’. As part of the effort to 

develop the rule, the NRC staff reviewed 

drafts of this document and in addition, 

NEI 00–04 was used in the pilot 

programs discussed earlier. The 

objective of the staff’s review was to 

determine the acceptability of the 

proposed implementing guidance, with 

the intent that the NEI guidance could 

be endorsed in an NRC regulatory guide. 

The revision of NEI 00–04 submitted on 

April 14, 2004 forms the basis for the 

NRC RG ‘‘Guidelines for Categorizing 

Structures, Systems and Components in 

Nuclear Power Plants According to 

Their Safety Significance.’’ Availability 

of this document is noted in Section IX. 

The NRC staff’s review of NEI 00–04 

resulted in several areas where the staff 

finds it necessary to identify 

clarifications, limitations, and 

conditions to the NEI guidance or to 

include further guidance to supplement 

the document, as it is currently written. 

These clarifications, limitations, and 

conditions, and the reasons therefore, 

are set forth in Section C of RG 1.201. 

These issues are best resolved by testing 

the guide against actual applications. 

Therefore, this RG is being issued for 

trial use. This RG does not establish any 

final staff positions, and may be revised 

in response to experience with its use. 

As such, this trial regulatory guide does 

not establish a staff position for 

purposes of the Backfit Rule, 10 CFR 

50.109, and any changes to this RG prior 

to staff adoption in final form will not 

be considered to be backfits as defined 

in 10 CFR 50.109(a)(1). This will ensure 

that the lessons learned from regulatory 

review of pilot and follow-on 

applications are adequately addressed 

in this document and that the guidance 

is sufficient to enhance regulatory 

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68045

stability in the review, approval, and 

implementation in the use of PRAs and 

their results in the risk informed 

categorization process required by 10 

CFR 50.69. 

The NRC staff and NEI continue to 

interact on the implementation 

guidance. Consequently, it is expected 

that NEI will submit an improved 

revision to NEI 00–04 that will enable 

the NRC to issue a RG with fewer 

clarifications, limitations, and 

conditions, and as a consequence, the 

NRC is delaying issuance of the RG. 

VI.2 Review Guidance Concerning 

PRA Quality and Peer Review 

RG 1.200, ‘‘An Approach for 

Determining the Technical Adequacy of 

Probabilistic Risk Assessment Results 

for Risk-Informed Activities,’’ provides 

guidance on the NRC position on 

voluntary consensus standards for PRA 

(in particular on the ASME standard for 

internal events PRAs) and associated 

industry PRA documents (e.g., NEI 00–

02, ‘‘Probabilistic Risk Assessment Peer 

Review Process Guideline’’). Further, 

this guide will be modified to address 

PRA standards on fire, external events, 

and low power and shutdown modes, as 

they become available. The NRC has 

also developed a draft supporting 

Standard Review Plan, SRP 19.1, to 

provide guidance to the staff on how to 

determine if a PRA providing results 

being used in a decision is technically 

acceptable. 

In a letter dated April 24, 2000, NEI 

requested that the NRC staff review the 

suitability of the peer review process 

described in NEI 00–02 to address PRA 

quality issues for this application. NRC 

issued a request for additional 

information on September 19, 2000, to 

which NEI responded by letter dated 

January 18, 2001. By letter dated April 

2, 2002 (ADAMS accession number 

ML020930632), the NRC staff sent to 

NEI, draft staff review guidance that was 

developed as a result of its review of 

NEI 00–02, for intended use for § 50.69 

applications. 

The draft staff review guidance is for 

a focused review of the plant-specific 

PRA based on a review of NEI 00–02 

and NEI 00–04. To reach the conclusion 

that the PRA results support the 

proposed categorization, the review 

guidance is structured to lead the staff 

reviewer to look for evidence that the 

impact of a given peer review issue on 

PRA results has been adequately 

addressed in the peer review report and, 

when necessary, has been identified for 

consideration by the IDP, or to request 

further information from the licensee. 

VII. Criminal Penalties 

For the purposes of Section 223 of the 

Atomic Energy Act, as amended, the 

Commission is issuing a rule to add 

§ 50.69 under one or more of Sections 

161b, 161i, or 161o of the AEA. Willful 

violations of the rule are subject to 

criminal enforcement. Criminal 

penalties, as they apply to regulations in 

Part 50, are discussed in § 50.111. 

VIII. Compatibility of Agreement State 

Regulations 

Under the ‘‘Policy Statement on 

Adequacy and Compatibility of 

Agreement States Programs,’’ approved 

by the Commission on June 20, 1997, 

and published in the Federal Register

(62 FR 46517, September 3, 1997), this 

rule is classified as compatibility 

‘‘NRC.’’ Compatibility is not required for 

Category ‘‘NRC’’ regulations. The NRC 

program elements in this category are 

those that relate directly to areas of 

regulation reserved to the NRC by the 

AEA or the provisions of Title 10 of the 

Code of Federal Regulations and, 

although an Agreement State may not 

adopt program elements reserved to 

NRC, it may wish to inform its licensees 

of certain requirements via a mechanism 

that is consistent with the particular 

State’s administrative procedure laws, 

but does not confer regulatory authority 

on the State.

IX. Availability of Documents 

The NRC is making the documents 

identified below available to interested 

persons through one or more of the 

following methods as indicated. 

Public Document Room (PDR). The 

NRC Public Document Room is located 

at 11555 Rockville Pike, Rockville, 

Maryland. 

Rulemaking Web site (Web). The 

NRC’s interactive rulemaking Web site 

is located at http://ruleforum.llnl.gov.

These documents may be viewed and 

downloaded electronically via this Web 

site. 

NRC’s Public Electronic Reading 

Room (PERR). The NRC’s public 

electronic reading room is located at 

http://www.nrc.gov/reading-rm.html.

Note: Public access to documents, 

including access via ADAMS and the PDR, 

has been temporarily suspended so that 

security reviews of publicly available 

documents may be performed and potentially 

sensitive information removed. However, 

access to the documents identified in this 

rule continues to be available through the 

rulemaking web site at http://

ruleforum.llnl.gov, which was not affected by 

the ADAMS shutdown. Please check with the 

listed NRC contact concerning any issues 

related to document availability.

Document PDR Web PERR 

Response to Public Comments ................................................................................................ X X ML042990011

Environmental Assessment ...................................................................................................... X X ML041040236

Regulatory Analysis .................................................................................................................. X X ML041000474

Industry Implementation Guidance ........................................................................................... X X ML041120253

Regulatory Guide ...................................................................................................................... X X ML041340087

Final Rule SRM ........................................................................................................................ X X ML042810516

SRM on PRA Quality ................................................................................................................ X X ML033520457

X. Voluntary Consensus Standards 

The National Technology Transfer 

and Advancement Act of 1995, Pub. L. 

104–113, requires that Federal agencies 

use technical standards that are 

developed or adopted by voluntary 

consensus standards bodies unless 

using such a standard is inconsistent 

with applicable law or is otherwise 

impractical. In this rule, the NRC is 

using the following Government-unique 

standard (RG 1.201, June 2004). The 

Commission notes the development of 

voluntary consensus standards on PRAs, 

such as an ASME Standard on 

Probabilistic Risk Assessment for 

Nuclear Power Plant Applications. RG 

1.201 and RG 1.200 (PRA Technical 

Adequacy) discuss how this standard 

could be used for the purpose of the 

internal events, full-power PRA. 

In addition, the Commission 

acknowledges development of riskinformed code cases by the ASME on 

categorization of certain components, 

particularly with respect to pressure 

boundary considerations. RG 1.201 

explicitly notes these code cases and 

that they could be proposed by a 

licensee or applicant as part of the 

means for satisfying the rule 

requirements. The government 

standards allow use of these voluntary 

consensus standards, but do not require 

68046 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

their use. The Commission does not 

believe that these other standards are 

sufficient to provide the overall 

construct for the alternative approach to 

categorization and treatment of SSCs 

that is the goal of this rulemaking. For 

example, the current standards do not 

address all types of components that 

might be categorized, nor do standards 

currently exist for addressing the PRA 

requirements for all initiating events 

and modes of operation. Additionally, 

there are no voluntary consensus 

standards that can address other parts of 

the approach laid out such as 

determining the basis for the 

evaluations to show an acceptably small 

increase in risk. The NRC is not aware 

of any voluntary consensus standard 

that could be used instead of the 

Government-unique standards. 

XI. Finding of No Significant 

Environmental Impact: Environmental 

Assessment: Availability 

The Commission has determined 

under the National Environmental 

Policy Act of 1969, as amended, and the 

Commission’s regulations in Subpart A 

of 10 CFR part 51, that this rule is not 

a major Federal action significantly 

affecting the quality of the human 

environment and, therefore, an 

environmental impact statement is not 

required. As set forth in the final 

environmental assessment, this action 

will not have a significant 

environmental impact principally 

because it is structured to maintain the 

design basis functional requirements for 

the SSCs in the facility, because the rule 

contains feedback and process 

adjustment requirements to maintain 

the validity of the categorization process 

over time, and because the standards 

and requirements applicable to 

radiological releases and effluents are 

not affected by this rulemaking. 

The NRC requested public comments 

on any aspect of the environmental 

assessment. No public comments were 

received. The NRC requested the views 

of the States on the environmental 

assessment for this rule. No State 

comments were received. Availability of 

the final environmental assessment is 

provided in Section IX. 

XII. Paperwork Reduction Act 

Statement 

This rule contains information 

collection requirements that are subject 

to the Paperwork Reduction Act of 1995 

(44 U.S.C. 3501 et seq.). These 

requirements were approved by the 

Office of Management and Budget, 

approval number 3150–0011. 

The burden to the public for these 

information collections is estimated to 

average 1,032 hours per response, 

including the time for reviewing 

instructions, searching existing data 

sources, gathering and maintaining the 

data needed, and completing and 

reviewing the information collection. 

Send comments on any aspect of these 

information collections, including 

suggestions for reducing the burden, to 

the Records and FOIA/Privacy Services 

Branch (T–5 F52), U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001, or by Internet 

electronic mail to 

INFOCOLLECTS@NRC.GOV; and to the 

Desk Officer, John A. Asalone, Office of 

Information and Regulatory Affairs, 

NEOB–10202, (3150–0011), Office of 

Management and Budget, Washington, 

DC 20503. You may also e-mail 

comments to 

John_A._Asalone@omb.eop.gov or 

comment by telephone at (202) 395–

4650. 

Public Protection Notification 

The NRC may not conduct or sponsor, 

and a person is not required to respond 

to, a request for information or an 

information collection requirement 

unless the requesting document 

displays a currently valid OMB control 

number. 

XIII. Regulatory Analysis 

The Commission has prepared a 

regulatory analysis on this regulation. 

The analysis examines the costs and 

benefits of the alternatives considered 

by the Commission. Availability of the 

regulatory analysis is provided in 

Section IX. 

XIV. Regulatory Flexibility Certification 

In accordance with the Regulatory 

Flexibility Act (5 U.S.C. 605(b)), the 

Commission certifies that this rule does 

not have a significant economic impact 

on a substantial number of small 

entities. This rule affects only the 

licensing and operation of nuclear 

power plants. The companies that own 

these plants do not fall within the scope 

of the definition of ‘‘small entities’’ set 

forth in the Regulatory Flexibility Act or 

the size standards established by the 

NRC (10 CFR 2.810).

XV. Backfit Analysis 

The NRC has determined that the 

Backfit Rule does not apply to this rule; 

therefore, a backfit analysis is not 

required for this rule. As a voluntary 

alternative to existing requirements, the 

final rule does not impose different or 

new requirements on 10 CFR part 50 

licensees or applicants and thus does 

not constitute a backfit pursuant to 

§ 50.109. 

XVI. Small Business Regulatory 

Enforcement Fairness Act 

In accordance with the Small 

Business Regulatory Enforcement 

Fairness Act of 1996, the NRC has 

determined that this action is not a 

major rule and has verified this 

determination with the Office of 

Information and Regulatory Affairs of 

OMB.

List of Subjects in 10 CFR Part 50 

Antitrust, Classified information, 

Criminal penalties, Fire protection, 

Intergovernmental relations, Nuclear 

power plants and reactors, Radiation 

protection, Reactor siting criteria, 

Reporting and recordkeeping 

requirements.

■ For the reasons set out in the preamble 

and under the authority of the Atomic 

Energy Act of 1954, as amended; the 

Energy Reorganization Act of 1974, as 

amended; and 5 U.S.C. 552 and 553, the 

NRC is adopting the following 

amendments to 10 CFR part 50.

PART 50—DOMESTIC LICENSING OF 

PRODUCTION AND UTILIZATION 

FACILITIES

■ 1. The authority citation for Part 50 

continues to read as follows:

Authority: Secs. 102, 103, 104, 105, 161, 

182, 183, 186, 189, 68 Stat. 936, 937, 938, 

948, 953, 954, 955, 956, as amended, sec. 

234, 83 Stat. 444, as amended (42 U.S.C. 

2132, 2133, 2134, 2135, 2201, 2232, 2233, 

2236, 2239, 2282); secs. 201, as amended, 

202, 206, 88 Stat.1242, as amended, 1244, 

1246 (42 U.S.C. 5841, 5842, 5846); See 1704, 

112 Stat. 2750 (44 U.S.C. 3504 note).

Section 50.7 also issued under Pub. L. 95–

601, sec. 10, 92 Stat. (42 U.S.C. 5841). 

Sections 50.10 also issued under secs. 101, 

185, 68 Stat. 955, as amended (42 U.S.C. 

2131, 2235); sec. 102, Pub. L. 91–190, 83 Stat. 

853 (42 U.S.C. 4332). Sections 50.13, 

50.54(dd), and 50.103 also issued under sec. 

108, 68 Stat. 939, as amended (42 U.S.C. 

2138). Sections 50.23, 50.35, 50.55, and 50.56 

also issued under sec. 185, 68 Stat. 955 (42 

U.S.C. 2235). Sections 50.33a, 50.55a, and 

Appendix Q also issued under sec. 102, Pub. 

L. 91–190, 83 Stat. 853 (42 U.S.C. 4332). 

Sections 50.34 and 50.54 also issued under 

Sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). 

Sections 50.58, 50.91, and 50.92 also issued 

under Pub. L. 97–415, 96 Stat. 2073 (42 

U.S.C. 2239). Section 50.78 also issued under 

sec. 122, 68 Stat. 939 (42 U.S.C. 2152). 

Sections 50.80, 50.81 also issued under sec. 

184, 68 Stat. 954, as amended (42 U.S.C. 

2234). Appendix F also issued under sec. 

187, 68 Stat. 955 (42 U.S.C. 2237).

■ 2. In § 50.8 paragraph (b) is revised to 

read as follows:

§ 50.8 Information collection 

requirements: OMB approval.

* * * * *

Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 68047

(b) The approved information 

collection requirements contained in 

this part appear in §§ 50.30, 50.33, 

50.33a, 50.34, 50.34a, 50.35, 50.36, 

50.36a, 50.36b, 50.44, 50.46, 50.47, 

50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 

50.60, 50.61, 50.62, 50.63, 50.64, 50.65, 

50.66, 50.68, 50.69, 50.70, 50.71, 50.72, 

50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 

50.120, and appendices A, B, E, G, H, 

I, J, K, M, N,O, Q, R, and S to this part.

* * * * *

■ 3. A new § 50.69 is added under center 

heading ‘‘Issuance, Limitations, and 

Conditions of Licenses and Construction 

Permits’’ to read as follows:

§ 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 riskinformed treatment of SSCs and 

submittal/approval process. (1) A holder 

of a license to operate a light water 

reactor (LWR) nuclear power plant 

under this part; a holder of a renewed 

LWR license under part 54 of this 

chapter; an applicant for a construction 

permit or operating license under this 

part; or an applicant for a design 

approval, a combined license, or 

manufacturing license under part 52 of 

this chapter; may voluntarily comply 

with the requirements in this section as 

an alternative to compliance with the 

following requirements for RISC–3 and 

RISC–4 SSCs: 

(i) 10 CFR part 21. 

(ii) The portion of 10 CFR 50.46a(b) 

that imposes requirements to conform to 

Appendix B to 10 CFR part 50. 

(iii) 10 CFR 50.49. 

(iv) 10 CFR 50.55(e). 

(v) The inservice testing requirements 

in 10 CFR 50.55a(f); the inservice 

inspection, and repair and replacement 

(with the exception of fracture 

toughness), requirements for ASME 

Class 2 and Class 3 SSCs in 10 CFR 

50.55a(g); and the electrical component 

quality and qualification requirements 

in Section 4.3 and 4.4 of IEEE 279, and 

Sections 5.3 and 5.4 of IEEE 603–1991, 

as incorporated by reference in 10 CFR 

50.55a(h). 

(vi) 10 CFR 50.65, except for 

paragraph (a)(4). 

(vii) 10 CFR 50.72. 

(viii) 10 CFR 50.73. 

(ix) Appendix B to 10 CFR part 50. 

(x) The Type B and Type C leakage 

testing requirements in both Options A 

and B of Appendix J to 10 CFR part 50, 

for penetrations and valves meeting the 

following criteria: 

(A) Containment penetrations that are 

either 1-inch nominal size or less, or 

continuously pressurized. 

(B) Containment isolation valves that 

meet one or more of the following 

criteria: 

(1) The valve is required to be open 

under accident conditions to prevent or 

mitigate core damage events; 

(2) The valve is normally closed and 

in a physically closed, water-filled 

system;

(3) The valve is in a physically closed 

system whose piping pressure rating 

exceeds the containment design 

pressure rating and is not connected to 

the reactor coolant pressure boundary; 

or 

(4) The valve is 1-inch nominal size 

or less. 

(xi) Appendix A to part 100, Sections 

VI(a)(1) and VI(a)(2), to the extent that 

these regulations require qualification 

testing and specific engineering 

methods to demonstrate that SSCs are 

designed to withstand the Safe 

Shutdown Earthquake and Operating 

Basis Earthquake. 

(2) A licensee voluntarily choosing to 

implement this section shall submit an 

application for license amendment 

under § 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 § 50.69(c)(1)(i). 

(iv) A description of, and basis for 

acceptability of, the evaluations to be 

conducted to satisfy § 50.69(c)(1)(iv). 

The evaluations must include the effects 

of common cause interaction 

susceptibility, and the potential impacts 

from known degradation mechanisms 

for both active and passive functions, 

and address internally and externally 

initiated events and plant operating 

modes (e.g., full power and shutdown 

conditions). 

(3) The Commission will approve a 

licensee’s implementation of this 

section if it determines that the process 

for categorization of RISC–1, RISC–2, 

RISC–3, and RISC–4 SSCs satisfies the 

requirements of § 50.69(c) by issuing a 

license amendment approving the 

licensee’s use of this section. 

(4) An applicant choosing to 

implement this section shall include the 

information in § 50.69(b)(2) as part of 

application. The Commission will 

approve an applicant’s implementation 

of this section if it determines that the 

process for categorization of RISC–1, 

RISC–2, RISC–3, and RISC–4 SSCs 

satisfies the requirements of § 50.69(c). 

(c) SSC Categorization Process. (1) 

SSCs must be categorized as RISC–1, 

RISC–2, RISC–3, or RISC–4 SSCs using 

a categorization process that determines 

if an SSC performs one or more safety 

significant functions and identifies 

those functions. The process must: 

(i) Consider results and insights from 

the plant-specific PRA. This PRA must 

at a minimum model severe accident 

scenarios resulting from internal 

initiating events occurring at full power 

operation. The PRA must be of 

sufficient quality and level of detail to 

support the categorization process, and 

must be subjected to a peer review 

process assessed against a standard or 

set of acceptance criteria that is 

endorsed by the NRC. 

(ii) Determine SSC functional 

importance using an integrated, 

systematic process for addressing 

initiating events (internal and external), 

SSCs, and plant operating modes, 

including those not modeled in the 

plant-specific PRA. The functions to be 

identified and considered include 

design bases functions and functions 

credited for mitigation and prevention 

of severe accidents. All aspects of the 

integrated, systematic process used to 

characterize SSC importance must 

reasonably reflect the current plant 

configuration and operating practices, 

and applicable plant and industry 

operational experience. 

(iii) Maintain defense-in-depth. 

(iv) Include evaluations that provide 

reasonable confidence that for SSCs 

categorized as RISC–3, sufficient safety 

68048 Federal Register / Vol. 69, No. 224 / Monday, November 22, 2004 / Rules and Regulations 

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 §§ 50.69(b)(1) and 

(d)(2) are small. 

(v) Be performed for entire systems 

and structures, not for selected 

components within a system or 

structure. 

(2) The SSCs must be categorized by 

an Integrated Decision-Making Panel 

(IDP) staffed with expert, plantknowledgeable members whose 

expertise includes, at a minimum, PRA, 

safety analysis, plant operation, design 

engineering, and system engineering. 

(d) Alternative treatment 

requirements.—(1) RISC–1 and RISC 2 

SSCs. The licensee or applicant shall 

ensure that RISC–1 and RISC–2 SSCs 

perform their functions consistent with 

the categorization process assumptions 

by evaluating treatment being applied to 

these SSCs to ensure that it supports the 

key assumptions in the categorization 

process that relate to their assumed 

performance. 

(2) RISC–3 SSCs. The licensee or 

applicant shall ensure, with reasonable 

confidence, that RISC–3 SSCs remain 

capable of performing their safetyrelated functions under design basis 

conditions, including seismic 

conditions and environmental 

conditions and effects throughout their 

service life. The treatment of RISC–3 

SSCs must be consistent with the 

categorization process. Inspection and 

testing, and corrective action shall be 

provided for RISC–3 SSCs. 

(i) Inspection and testing. Periodic 

inspection and testing activities must be 

conducted to determine that RISC–3 

SSCs will remain capable of performing 

their safety-related functions under 

design basis conditions; and 

(ii) Corrective action. Conditions that 

would prevent a RISC–3 SSC from 

performing its safety-related functions 

under design basis conditions must be 

corrected in a timely manner. For 

significant conditions adverse to 

quality, measures must be taken to 

provide reasonable confidence that the 

cause of the condition is determined 

and corrective action taken to preclude 

repetition. 

(e) Feedback and process 

adjustment.—(1) RISC–1, RISC–2, RISC–

3 and RISC–4 SSCs. The licensee shall 

review changes to the plant, operational 

practices, applicable plant and industry 

operational experience, and, as 

appropriate, update the PRA and SSC 

categorization and treatment processes. 

The licensee shall perform this review 

in a timely manner but no longer than 

once every two refueling outages. 

(2) RISC–1 and RISC–2 SSCs. The 

licensee shall monitor the performance 

of RISC–1 and RISC–2 SSCs. The 

licensee shall make adjustments as 

necessary to either the categorization or 

treatment processes so that the 

categorization process and results are 

maintained valid. 

(3) RISC–3 SSCs. The licensee shall 

consider data collected in 

§ 50.69(d)(2)(i) for RISC–3 SSCs to 

determine if there are any adverse 

changes in performance such that the 

SSC unreliability values approach or 

exceed the values used in the 

evaluations conducted to satisfy 

§ 50.69(c)(1)(iv). The licensee shall 

make adjustments as necessary to the 

categorization or treatment processes so 

that the categorization process and 

results are maintained valid. 

(f) Program documentation, change 

control and records. (1) The licensee or 

applicant shall document the basis for 

its categorization of any SSC under 

paragraph (c) of this section before 

removing any requirements under 

§ 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 

§ 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 § 50.69(d) 

need not include a supporting § 50.59 

evaluation of the changes directly 

related to implementation. Thereafter, 

changes to the programs and procedures 

for implementation of § 50.69(d), as 

described in the FSAR, may be made if 

the requirements of this section and 

§ 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 § 50.69(d) need not 

include a supporting § 50.54(a) review 

of the changes directly related to 

implementation. Thereafter, changes to 

the programs and procedures for 

implementation of § 50.69(d), as 

described in the quality assurance plan 

may be made if the requirements of this 

section and § 50.54(a) continue to be 

met. 

(g) Reporting. The licensee shall 

submit a licensee event report under 

§ 50.73(b) for any event or condition 

that prevented, or would have 

prevented, a RISC–1 or RISC–2 SSC 

from performing a safety significant 

function.

Dated in Rockville, Maryland this 15th day 

of November 2004.

For the Nuclear Regulatory Commission. 

Annette L. Vietti-Cook, 

Secretary of the Commission.

[FR Doc. 04–25665 Filed 11–19–04; 8:45 am] 

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