ML25010A417

DG-1426 (RG 1.225 Rev 0) an Approach for Riep Supporting 50.46a - ACRS Version Rev 1

ML25010A417
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Issue date:	01/13/2025
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DR-1426 RG-1.225
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U.S. NUCLEAR REGULATORY COMMISSION DRAFT REGULATORY GUIDE DG-1426 Proposed new Regulatory Guide 1.225 Issue Date: Month 202#

Technical Lead: Michelle Kichline This RG is being issued in draft form to involve the public in the development of regulatory guidance in this area. It has not received final staff review or approval and does not represent an NRC final staff position. Public comments are being solicited on this DG and its associated regulatory analysis. Comments should be accompanied by appropriate supporting data. Comments may be submitted through the Federal rulemaking website, http://www.regulations.gov, by searching for draft regulatory guide DG-1426. Alternatively, comments may be submitted to Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.

Comments must be submitted by the date indicated in the Federal Register notice.

Electronic copies of this DG, previous versions of DGs, and other recently issued guides are available through the NRCs public website under the Regulatory Guides document collection of the NRC Library at https://nrc.gov/reading-rm/doc-collections/reg-guides. The DG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML24284A342. The regulatory analysis is associated with a rulemaking and may be found in ADAMS under Accession No. ML24239A776.

Pre-Decisional/Public version for meetings with the Advisory Committee on Reactor Safeguards AN APPROACH FOR A RISK-INFORMED EVALUATION PROCESS SUPPORTING ALTERNATIVE ACCEPTANCE CRITERIA FOR EMERGENCY CORE COOLING SYSTEMS FOR LIGHT-WATER REACTORS A. INTRODUCTION Purpose This regulatory guide (RG) describes an approach that is acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) to demonstrate that proposed facility changes will satisfy the regulatory requirements in section 50.46a(h) of Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 1), in Title 10 of the Code of Federal Regulations (10 CFR). Specifically, this RG provides guidance for the risk-informed evaluation specified in 10 CFR 50.46a(c)(1)(iv) and the risk-informed evaluation process (RIEP) specified in 10 CFR 50.46a(c)(1)(v). It includes all enabled changes to the facility, technical specifications, and procedures that satisfy the alternative emergency core cooling system (ECCS) analysis requirements under 10 CFR 50.46a, Alternative acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, but do not satisfy the ECCS requirements under 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Applicability This RG applies to operating light-water nuclear power reactor licensees subject to 10 CFR Part 50, or 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 2), that receive NRC approval to use the alternative risk-informed process for analyzing ECCS performance during loss-of-coolant accidents (LOCAs) in 10 CFR 50.46a. This RG also applies to applicants for or holders of licenses under either 10 CFR Part 50 or 10 CFR Part 52 for new light-water nuclear power reactors or small modular reactors that receive NRC approval to use the aforementioned risk-informed process in 10 CFR 50.46a.

Pre-Decisional/Public version for meetings with the Advisory Committee on Reactor Safeguards

DG-1426, Page 2 Applicable Regulations 10 CFR Part 50 provides for the licensing of production and utilization facilities pursuant to the Atomic Energy Act of 1954, as amended (Ref. 3), and Title II of the Energy Reorganization Act of 1974, as amended (Ref. 4).

o 10 CFR 50.46a(a)(1) defines changes enabled by this section, as used in 10 CFR 50.46a, as changes to the facility, technical specifications, and procedures that satisfy the alternative ECCS analysis requirements in 10 CFR 50.46a but do not satisfy the ECCS requirements under 10 CFR 50.46.

o 10 CFR 50.46a(a)(4) defines entity, as used in 10 CFR 50.46a and this guide, as an applicant for or a holder of a construction permit, operating license, combined license, standard design approval, or manufacturing license, or an applicant for a standard design certification rule (including such applicant after NRC issuance of a final standard design certification rule).

o 10 CFR 50.46a(c)(1)(vi)(A) requires that a description of the risk-informed evaluation used to demonstrate that the proposed changes to the facility meet the requirements in 10 CFR 50.46a(h) be provided in an application for NRC approval to use 10 CFR 50.46a.

o 10 CFR 50.46a(c)(1)(vii) requires an entity other than a design certification applicant or a holder of a manufacturing license that wishes to make changes enabled by this section without prior NRC review and approval, to submit in the application for NRC approval to use 10 CFR 50.46a a process to be used for evaluating the acceptability of these changes.

o 10 CFR 50.46a(d)(3) requires, in part, that changes made under 10 CFR 50.46a be evaluated by a risk-informed evaluation demonstrating that the acceptance criteria in 10 CFR 50.46a(h) are met.

o 10 CFR 50.46a(d)(4) requires the entity to periodically maintain and upgrade, as necessary, its risk assessments to meet the requirements in 10 CFR 50.46a(h)(4)-(5). This maintenance and upgrading must be in conformance with methods, standards, and practices that have been endorsed or otherwise found acceptable by the NRC and must be completed in a timely manner, at least once within five years of NRC approval of the entitys application and at least once every five years thereafter. Based upon a reevaluation of the risk assessments after the periodic maintenance and upgrading are completed, the entity must take appropriate action to ensure that the acceptance criteria in 10 CFR 50.46a(h), as applicable, are met.

o 10 CFR 50.46a(h) requires, in part, that an entity that wishes to make changes enabled by this section must perform a risk-informed evaluation.

o 10 CFR 50.59, Changes, tests and experiments, establishes a screening process that licensees may use to determine whether facility changes require prior review and approval by the NRC. 10 CFR 50.59(c)(1) states that a licensee may make changes in the facility as described in the final safety analysis report (FSAR) (as updated), make changes in the procedures as described in the FSAR (as updated), and conduct tests or experiments not described in the FSAR (as updated) without obtaining a license amendment pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, only if a change to the technical specifications incorporated in the license is not

DG-1426, Page 3 required and the change, test, or experiment does not meet any of the criteria in 10 CFR 50.59(c)(2).

o 10 CFR 50.90 requires that, whenever a holder of a license (including a construction permit and operating license under 10 CFR Part 50 and an early site permit, combined license, or manufacturing license under 10 CFR Part 52) wishes to amend the license or permit, it must file an application for a license amendment with the Commission that fully describes the changes desired and follows, as far as applicable, the form prescribed for the original application.

10 CFR Part 52 governs the issuance of early site permits, standard design certifications, combined licenses, standard design approvals, and manufacturing licenses for nuclear power facilities pursuant to the Atomic Energy Act of 1954, as amended, and Title II of the Energy Reorganization Act of 1974, as amended.

o 10 CFR 52.98(b) requires a holder of a combined license that does not reference a design certification or a reactor manufactured under a manufacturing license to make changes in the facility as described in the FSAR (as updated), make changes in the procedures as described in the FSAR (as updated), and conduct tests or experiments not described in the FSAR (as updated) under the applicable change processes in 10 CFR Part 50.

Related Guidance NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 19.1, Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests after Initial Fuel Load (Ref. 5), is designed to guide the NRC staff in evaluating licensee requests for changes to the licensing basis that apply risk insights. Guidance developed in selected application-specific RGs and the corresponding chapters of NUREG-0800 also applies to these types of licensing basis changes.

NUREG-0800, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance (Ref. 6), addresses the review of risk information used to support permanent plant-specific changes to the licensing basis.

RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Ref. 7), describes an approach acceptable to the NRC for developing risk-informed applications for licensing basis changes that considers engineering issues and applies risk insights.

RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (Ref. 8), provides an approach acceptable to the NRC for determining whether the base probabilistic risk assessment (PRA) (in total or the parts that are used to support an application) is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors. RG 1.200 currently endorses a PRA standard developed by the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS), ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Ref. 9), which addresses core damage frequency (CDF) and large early release frequency (LERF) for internal and external hazard groups during at-power operations.

DG-1426, Page 4 Design Certification/Combined License (DC/COL) Interim Staff Guidance (ISG)

DC/COL-ISG-028, Assessing the Technical Adequacy of the Advanced Light-Water Reactor Probabilistic Risk Assessment for the Design Certification Application and Combined License Application, issued November 2016 (Ref. 10), provides guidance for assessing the technical adequacy of PRAs needed for applications for both DCs and COLs for advanced light-water reactors under 10 CFR Part 52.

Purpose of Regulatory Guides The NRC issues RGs to describe methods that are acceptable to the staff for implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific issues or postulated events, and to describe information that the staff needs in its review of applications for permits and licenses. Regulatory guides are not NRC regulations and compliance with them is not required. Methods and solutions that differ from those set forth in RGs are acceptable if the applicant provides sufficient basis and information for the NRC staff to verify that the alternative methods comply with the applicable NRC regulations.

Paperwork Reduction Act This RG provides voluntary guidance for implementing the mandatory information collections in 10 CFR Parts 50 and 52 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et. seq.). These information collections were approved by the Office of Management and Budget (OMB), under control numbers 3150-0011 and 3150-0151, respectively. Send comments regarding this information collection to the FOIA, Library, and Information Collections Branch, Office of the Chief Information Officer, Mail Stop: T6-A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or to the OMB reviewer at: OMB Office of Information and Regulatory Affairs, (3150-0011 and 3150-0151), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street, NW, Washington, DC 20503.

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 valid OMB control number.

DG-1426, Page 5 B. DISCUSSION Reason for Issuance The regulation at 10 CFR 50.46 requires certain entities to design the ECCS so that its calculated cooling performance can mitigate the full spectrum of LOCAs up to the double-ended guillotine break (DEGB) of the largest reactor coolant system pipe. The NRC amended 10 CFR 50.46a to establish an alternative to the conventional regulatory framework of 10 CFR 50.46. The alternative requirements enable operational as well as design changes to operating power reactor facilities. The alternative option provided in 10 CFR 50.46a also requires entities to design the ECCS so that its calculated cooling performance can mitigate the full spectrum of LOCAs up to the DEGB of the largest reactor coolant system pipe. However, commensurate with the lower probability of breaks larger than the transition break size (TBS), 10 CFR 50.46a(e)(3) specifies less conservatism for the analyses and associated acceptance criteria for breaks larger than the TBS. The alternative option allows best-estimate modeling and more realistic assumptions to be used for larger, lower likelihood LOCAs.

This RG provides guidance for performing the risk-informed evaluation in 10 CFR 50.46a that must be used to evaluate compliance with the acceptance criteria in 10 CFR 50.46a(h) for every proposed change enabled by this section at the plant, both those that affect the licensing basis and those that do not. The regulation in 10 CFR 50.46a uses the term changes enabled by this section. For simplicity, this RG uses the phrase proposed change, which is synonymous with changes enabled by this section.

To support the use of risk-informed decision-making, the NRC developed RG 1.174, which is an approach that the NRC has determined acceptably integrates deterministic and risk insights into a risk-informed decision-making process. This RG supplements the guidance in RG 1.174 by providing guidance for evaluating changes enabled under 50.46a. This RG is structured to follow the key principles discussed in RG 1.174 and uses the general four-element approach for risk-informed applications discussed in RG 1.174 to ensure that all the key principles are addressed.

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Background===

The NRC recognizes that PRA technology has evolved to the point that it can be used increasingly as a tool in regulatory decision-making. In August 1995, the Commission issued a policy statement on the use of PRA methods in nuclear regulatory activities titled, Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities: Final Policy Statement (Ref. 11). The statement adopted the following policy:

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 complements the NRCs deterministic approach and supports the agencys traditional defense-in-depth (DID) philosophy.

PRAs and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state of the art, to reduce unnecessary conservatism associated with current regulatory requirements, RGs, license commitments, and staff practices. Where appropriate, a PRA should be used to support a proposal for additional regulatory requirements in accordance with 10 CFR 50.109, Backfitting.

Appropriate procedures for including a PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations will be complied with unless they are revised.

DG-1426, Page 6 PRA evaluations in support of regulatory decisions should be as realistic as practicable, and appropriate supporting data should be publicly available for review.

The Commissions safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power licensees.

In its approval of the policy statement, the Commission stated its expectation that implementing the policy statement will improve the regulatory process in three ways: (1) foremost, through safety decision-making enhanced by the use of PRA insights, (2) through more efficient use of agency resources, and (3) through a reduction in unnecessary burdens on licensees.

This RG also uses the Commissions Safety Goal Policy Statement, dated August 4, 1986 (Ref. 12). As described in section C of this RG, one key principle in risk-informed regulation is that proposed increases in risk are small and are consistent with the intent of the Commissions Safety Goal Policy Statement. The safety goals and associated quantitative health objectives define an acceptable level of risk that is a small percentage (0.1 percent) of other risks to which the public is exposed. The risk acceptance guidelines in section C.2.2.3 of this RG are defined for light-water reactors in terms of CDF, LERF, and changes in CDF and LERF (i.e., CDF and LERF) risk metrics. These risk metrics are based on subsidiary objectives derived from the safety goals and their quantitative health objectives. In particular, the CDF risk metric is used as a surrogate for the individual latent cancer fatality risk, and the LERF risk metric is used as a surrogate for the individual early fatality risk.

The NRC has established a set of regulatory requirements for commercial nuclear power reactors to ensure that a reactor facility does not impose an undue risk to public health and safety (i.e., the NRC has reasonable assurance of adequate protection of public health and safety). The current body of NRC regulations and their implementation are largely based on a deterministic approach. While the regulatory framework of 10 CFR 50.46, based on traditional engineering criteria, continues to serve its purpose in ensuring no undue risk to public health and safety, the current information base contains insights gained from many years of plant operating experience. This information, combined with modern risk assessment techniques and associated data, can be used to develop a more effective approach to providing reasonable assurance that public health and safety are protected from the effects of LOCAs.

Nuclear plant licensees and other entities are required to have an evaluation of ECCS cooling performance that meets either 10 CFR 50.46 or 10 CFR 50.46a requirements. An entity may choose to meet the requirements in 10 CFR 50.46 for all breaks up to and including the DEGB of the largest reactor coolant system pipe, or it may choose to meet the 10 CFR 50.46a requirements. The regulation in 10 CFR 50.46a divides the current spectrum of LOCA break sizes into two regions. The division between the two regions is determined by the TBS. The first region includes small breaks, up to and including the TBS. The second region includes breaks larger than the TBS, up to and including the DEGB of the largest reactor coolant system pipe. These larger breaks are considered to have a much lower likelihood than the smaller breaks in the first region. For breaks at or below the TBS, 10 CFR 50.46a(e)(2) specifies that there must be a high level of probability that the calculated ECCS cooling performance would not exceed the criteria set forth in 10 CFR 50.46a(e)(1), just as currently required for all breaks. However, commensurate with the lower probability of breaks larger than the TBS, 10 CFR 50.46a(e)(3) specifies more flexible requirements associated with the rigor and conservatism of the analyses and associated acceptance criteria for breaks larger than the TBS. NUREG-1903, Seismic Considerations For the Transition Break Size, issued February 2008 (Ref. 13); Continued Applicability of NUREG-1903, issued August 2024 (Ref. 14); NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA)

Frequencies Through the Elicitation Process, issued March 2008 (Ref. 15); Continued Applicability of NUREG-1829, issued August 2024 (Ref. 16); and DG-1428, Plant-Specific Applicability of Transition

DG-1426, Page 7 Break Size (Ref. 17), contain further information and staff guidance concerning the plant-specific applicability of TBS.

10 CFR 50.46a allows all future proposed changes1 to a facility, technical specifications,2 or operating procedures to be evaluated by an NRC-approved RIEP. The RIEP may include qualitative risk insights using non-PRA risk methods. A RIEP assessment should include quantitative and qualitative risk analysis tools (as appropriate, a framework for evaluating the DID implications of the changes, a framework for evaluating safety margins, and performance measurement programs that monitor the facility and provide feedback information for timely corrective actions. The RIEP ensures that all proposed changes to the plant involve an acceptable change in risk and are consistent with the criteria from RG 1.174 to ensure adequate DID, safety margins, and performance measurement.

Entities3 with an approved RIEP may make certain facility changes without prior NRC review if they meet the requirements in 10 CFR 50.59 and 10 CFR 50.46a, including the criterion that risk increases cannot exceed a minimal level. Section C.2.2.3.2 of this RG gives the acceptance guidelines for minimal change in risk. Entities can make other facility changes after NRC approval if they meet the 10 CFR 50.90 requirements for license amendments and the criteria in 10 CFR 50.46a, including the criterion that the changes to be implemented are themselves very small and that plant baseline risk remains small.

This RG describes an acceptable approach for assessing the nature and impact of proposed changes by considering engineering issues and applying risk insights. These assessments should consider relevant safety margins and DID attributes, including success criteria and equipment functionality, reliability, and availability. The analyses should reflect the actual design, construction, and operational practices of the plant. Consideration of the Commissions Safety Goal Policy Statement is an important element in regulatory decision-making. Consequently, this RG provides acceptance guidelines for evaluating the results of assessments that are consistent with this policy statement. This guide also addresses implementation strategies and performance monitoring plans associated with changes that will help to ensure that assumptions and analyses supporting the change are verified.

Consideration of International Standards The International Atomic Energy Agency (IAEA) works with member states and other partners to promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops Safety Requirements and Safety Guides for protecting people and the environment from harmful effects of ionizing radiation. This system of safety fundamentals, safety requirements, safety guides, and other relevant reports, reflects an international perspective on what constitutes a high level of safety. To inform its development of this RG, the NRC considered IAEA Safety Requirements and Safety Guides pursuant to the Commissions International Policy Statement (Ref. 18) and Management Directive and Handbook 6.6, Regulatory Guides (Ref. 19).

The following IAEA Safety Requirements and Guides were considered in the development of the Regulatory Guide:

1 The scope of changes subject to the change criteria in 10 CFR 50.46a(h) is greater than the changes currently subject to 10 CFR 50.59, which applies only to the facility as described in the FSAR. The change criteria in 10 CFR 50.46a apply to all facility and procedure changes, regardless of whether they are described in the FSAR.

2 Under the Atomic Energy Act of 1954, as amended, technical specifications are part of the license. Therefore, plant-specific technical specifications must be modified by a license amendment.

3 Entities other than a design certification applicant or a holder of a manufacturing license.

DG-1426, Page 8 IAEA Safety Standards Series No. SSG-3, Revision 1, Development and Application of Level 1 Probabilistic Safety Assessment for Nuclear Power Plants, issued 2024 (Ref. 20)

IAEA Safety Standards Series No. SSG-4, Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants, issued 2010 (Ref. 21)

DG-1426, Page 9 C. STAFF REGULATORY GUIDANCE 1

Element 1: Define the Proposed Change

a.

Element 1 involves three primary activities.

(1)

The entity should determine whether each proposed change affects the licensing bases. If a proposed change affects the licensing bases, the entity should identify those aspects of the plants licensing bases that may be affected, including but not limited to rules and regulations, the FSAR, technical specifications, licensing conditions, and licensing commitments. If a proposed change affects the licensing bases, the entity should determine whether the proposed change should be implemented as a license amendment in accordance with 10 CFR 50.90 or may be implemented without NRC review. If the proposed change does not affect the licensing bases, the entity should identify those aspects of the entity-controlled documentation that may be affected.

(2)

The entity should identify all structures, systems, and components (SSCs), as well as procedures and activities that are covered by the proposed change(s) being evaluated and should consider the original reasons for including each program requirement. When considering a proposed change(s), an entity may identify regulatory requirements or commitments in its licensing bases that it believes are overly restrictive or unnecessary to ensure safety at the plant. The corollary is also true; that is, entities may identify design and operational aspects of the plant that should be enhanced consistent with an improved understanding of their safety significance. Such enhancements should be embodied in appropriate engineering analyses.

(3)

The entity should identify available engineering studies, methods, codes, plant-specific and industry data and operational experience, PRA findings, and research and analysis results relevant to the proposed change(s). Regarding plant-specific PRAs, the entity should assess the capability to use, refine, augment, and update system models as needed to support a risk assessment of the proposed change(s).

b.

The above information should be used collectively to describe the proposed change(s) and to outline the method of analysis. The entity should describe the proposed change(s) and how it meets the objectives of the Commissions PRA Policy Statement, including enhanced decision-making, more efficient use of resources, and reduction of unnecessary burden. In addition to improvements in reactor safety, this assessment may consider benefits from the proposed change(s), such as reduced fiscal and personnel resources and radiation exposure. The entity should affirm that every proposed change meets the current regulations unless the proposed change is explicitly related to a proposed exemption.

2 Element 2: Perform Engineering Analysis

a.

The scope, level of detail, and technical acceptability of the engineering analyses conducted to justify the proposed change(s) should be appropriate for the nature and scope of the proposed change(s). Entities that implement 10 CFR 50.46a may use an integrated risk-informed evaluation to demonstrate the acceptability of proposed changes. This integrated risk-informed evaluation must demonstrate that (1) increases in plant risk (if any) meet appropriate risk acceptance criteria, (2)

DG-1426, Page 10 adequate DID is maintained, (3) adequate safety margins are maintained, and (4) adequate performance measurement programs are implemented.

b.

The entity should appropriately consider uncertainty in the analysis and interpretation of findings. In selecting appropriate engineering analyses to support regulatory decision-making, the entity should judge the complexity and difficulty of implementing the proposed change(s). Thus, the entity should consider the appropriateness of qualitative and quantitative analyses, as well as analyses using traditional engineering approaches, and those techniques associated with the use of PRA findings.

Regardless of the analysis methods chosen, the entity should show that it has met the four principles set forth above, as described in part C of this RG, by using scrutable acceptance guidelines established for making that determination.

2.1 Integrated Risk-Informed Evaluation Process Assessment

a.

For entities wishing to make changes without NRC approval, using 10 CFR 50.46a requires that all future proposed change(s) to a facility, technical specifications, or operating procedures be evaluated by an NRC-approved RIEP. A RIEP assessment of every proposed change should include quantitative and qualitative (if used) risk analyses, an evaluation of impact on DID, and an evaluation of impact on safety margins. As discussed further below, the RIEP assessment should also include performance-measurement programs that monitor the facility and provide feedback of information for timely corrective actions. The RIEP assessment should ensure that each proposed change involves an acceptable change in risk and is consistent with RG 1.174 guidelines to ensure adequate DID, safety margins, and performance measurement.

b.

Certain entities with an approved RIEP may make certain proposed changes without NRC review if they meet 10 CFR 50.59 and 10 CFR 50.46a requirements, including the criterion that risk increases, if any, cannot exceed a minimal level. Section C.2.2.3.2 of this RG gives the acceptance guidelines for minimal change in risk. Entities can make other proposed changes after NRC approval if they meet the 10 CFR 50.90 requirements for license amendments and the criteria in 10 CFR 50.46a, including the criterion that the changes to be implemented are themselves very small, and that plant baseline risk remains small. Section C.2.2.3.1 of this RG contains the acceptance guidelines for very small change in risk.

2.2 Risk Assessment

a.

A risk assessment is to be performed as part of the risk-informed evaluation. The risk-informed evaluation may include qualitative risk insights using non-PRA risk methods, which is discussed in section C.2.2.2 of this RG. To satisfy the requirements in 10 CFR 50.46a(h)(2)(ii) that the total increases in CDF and LERF are very small and overall plant risk remains small, the total risk from all proposed changes to the plant since the adoption of 10 CFR 50.46a4 must be estimated. It is important to estimate the total change in risk from the changes to the facility, technical specifications, and procedures to ensure that these changes, when evaluated in total as they are implemented over time, do not contribute more than a small increase in risk. The quantitative guidelines in RG 1.174 provide an indication, in numerical terms, of what is considered acceptable. As such, the numerical guideline values are approximate, and therefore there is no compelling reason to require a high degree of precision in the total change in risk estimates that are compared to the guideline. The total change in risk estimates need only be sufficiently precise to provide reasonable assurance that the overall plant baseline risk increase is 4

The entity is considered to have adopted 10 CFR 50.46a into its licensing bases upon NRC approval of the entitys request made under 10 CFR 50.46a(c).

DG-1426, Page 11 small. Section C.2.2.3.1 of this RG gives the acceptance guidelines for very small changes in risk and small changes in overall plant baseline risk.

b.

Before implementing 10 CFR 50.46a, entities should clarify and document the scope of their baseline PRA models and estimate the overall CDF and LERF from these models. If the entity possesses a full-scope PRA that includes all initiators (both internal and external) to the plant and for all modes of operation, including low-power and shutdown modes, the CDF and LERF values at the time an entity implements 10 CFR 50.46a will become the baseline values from which the total increases in CDF and LERF from all proposed changes can be estimated. An entity may, however, implement 10 CFR 50.46a without a full-scope PRA. In this case, the entity should estimate baseline CDF and LERF values for the in-scope modes and initiators for which there are NRC-endorsed PRA standards or methods (or both) and establish baseline CDF and LERF values for the remaining modes and initiators. RG 1.200 endorses an ASME/ANS PRA standard that addresses the base PRA for CDF and LERF for internal and external hazard groups at-power. Other standards (e.g., for low-power and shutdown modes of operation and Level 2 PRAs) are under development.

c.

Once the baseline CDF and LERF estimates have been developed, the total change in risk from all proposed changes to the plant that affect the in-scope modes and initiators could be estimated as the difference between the baseline and the current risk estimates (after accounting for the PRA improvements discussed below). Entities may exclude the change in risk from all out-of-scope modes and initiators from the total change in risk estimates if the entity can demonstrate that any change in risk from the out-of-scope modes and initiators is not a significant contributor to the total change in risk. Otherwise, the potential risk increases from the out-of-scope modes and initiators should be estimated and added to the total change in risk estimates.

d.

Entities should periodically update (i.e., maintain and upgrade5) their PRA in a timely manner, but no less often than once every 5 years, so that the current design and operation of the plant are reflected. RG 1.200 contains the staffs guidance on the characteristics and attributes needed for an acceptable process for maintaining and upgrading the base PRA. In addition, RG 1.200 contains the staffs guidance on the process for determining whether a change to a PRA is classified as PRA maintenance or a PRA upgrade. Any change(s) to the PRA caused by plant changes that were analyzed with, but not put into, the PRA should be incorporated into it during these updates.

e.

Entities should demonstrate that any proposed change does not significantly increase LOCA frequencies or invalidate the evaluation demonstrating the applicability of the TBS to the facility, or used to establish the plant-specific TBS. NUREG-1903, NUREG-1829, and DG-1428 contain further information and the staffs guidance concerning the seismic considerations and plant-specific applicability of the TBS and should be used in conjunction with this RG to evaluate the proposed change.

f.

The regulation in 10 CFR 50.46a(h)(3) requires, in part, that the risk-informed evaluation demonstrate that DID is maintained and adequate safety margins are retained to account for uncertainties.

Sections C.2.2.4 and C.2.2.5 present guidance on assessing whether a proposed change remains consistent with the DID philosophy and maintains adequate safety margins.

2.2.1 Probabilistic Risk Assessments Used for Risk Assessment

a.

RG 1.200 describes one approach acceptable to the NRC staff for determining whether a base PRA, in total or in the portions that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water 5

RG 1.200 defines the terms PRA maintenance and PRA upgrade.

DG-1426, Page 12 reactors. When used in support of a 10 CFR 50.46a application, RG 1.200 will obviate the need for an in-depth review of the base PRA by NRC reviewers, allowing them to focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application.

b.

Four areas collectively determine the acceptability of a PRA:

(1) scope of a PRA, (2) technical elements of a PRA, (3) level of detail of a PRA, and (4) plant representation and PRA configuration control.

Probabilistic Risk Assessment Scope

c.

Under 10 CFR 50.46a(h)(4)(i), risk assessments must address initiating events, from both internal and external sources and for all plant operating modes, that would affect the regulatory decision in a substantial manner. The risk-informed evaluation, including the risk assessment, must also be used to demonstrate the acceptability of all future facility changes performed under 10 CFR 50.46a such that, eventually, most modes and initiators would be applicable to one or more proposed changes. Therefore, entities may find it necessary to expand the scope of the PRA to take full advantage of 10 CFR 50.46a.

d.

A full-scope PRA model that includes all modes and initiators would be helpful in quantitatively assessing the risk of the proposed facility, technical specification, or procedure change(s). It is possible, and in some cases may be necessary, to resolve difficulties associated with inadequate PRA scope by excluding some proposed changes. However, all risk contributors that cannot be shown to be insignificant to the decision should be assessed through quantitative risk assessment methods to support the risk-informed evaluation.

e.

Where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact. The risk assessment should document sources of risk shown to be insignificant or unaffected by the proposed change(s). This ensures that the risk assessment appropriately considers all potentially significant sources of risk. RG 1.200 contains the staffs guidance concerning the scope of the base PRA.

f.

Two types of evaluations may be used to satisfy the 10 CFR 50.46a requirement that the risk assessment considers all modes and initiating events that would affect the regulatory decision in a substantial manner:

(1)

Where NRC-endorsed PRA methods exist, a PRA of sufficient technical acceptability is used to estimate the change in CDF and LERF.

(2)

Where NRC-endorsed PRA methods do not exist, a non-PRA conservative or bounding risk-assessment method may be used to evaluate the resulting change in risk either qualitatively or quantitatively (that should not be significant and therefore would not affect the decision in a substantive manner).

Technical Elements and Level of Detail of a Probabilistic Risk Assessment

g.

The PRA should be capable of calculating both CDF and LERF. The PRA should have sufficient technical acceptability to provide confidence that the overall CDF and LERF are reasonable estimates to demonstrate that the risk increase guidelines are met. The non-PRA risk assessment or other evaluation used to demonstrate that the out-of-scope PRA modes and initiators do not significantly

DG-1426, Page 13 contribute to the change in risk should also have sufficient technical acceptability to support the conclusion.

h.

As stated in RG 1.200, the level of detail of a base PRA is defined in terms of the resolution of the modeling used to represent the behavior and operations of the plant. A minimal level of detail is necessary to ensure that the impacts of designed-in dependencies (e.g., support system dependencies, functional dependencies, and dependencies on operator actions) are correctly captured.

This minimal level of detail is implicit in the technical elements comprising the base PRA and their associated characteristics and attributes. The level of detail required of the PRA is that which is sufficient to adequately model the impact of the proposed change(s). If the impacts of any proposed change to the plant cannot be modeled with the existing technical elements of the PRA, the PRA should be modified accordingly, or the impact of the proposed change(s) should be evaluated using non-PRA risk assessments as discussed below in this RG.

i.

For entities that wish to make changes enabled by 10 CFR 50.46a without prior NRC review and approval, the initial application requesting NRC approval to implement 10 CFR 50.46a must include a complete description of the RIEP and the supporting PRA model and non-PRA risk assessment methods that will be used in the RIEP assessment. The NRC approval of the RIEP will reflect its scope, technical acceptability, and supporting risk assessments for demonstrating that self-approved changes satisfy the change criteria under 10 CFR 50.59. Therefore, NRC approval may limit the scope of changes the entity may implement under 10 CFR 50.46a to those for which the available risk assessments are adequate to demonstrate that any increase in risk will be minimal. If an entity expands the PRA risk assessments to address additional modes or initiators, the entity should submit a request that the NRC approval recognize the expanded risk assessments.

j.

The regulation in 10 CFR 50.46a(h)(4)(iv) requires that the PRA be determined, through peer review, to meet industry standards for PRA quality endorsed or otherwise found acceptable by the NRC. RG 1.200 discusses the various national consensus PRA standards and industry documents that provide guidance on the development, performance, and peer reviews of PRAs. The entity should address and resolve all peer reviewer comments about the PRA. Resolution may take the form of a discussion of how the PRA model has been changed, or a justification in the form of a sensitivity study that demonstrates that the significant accident sequences or contributors were not impacted (i.e., remained the same) by the proposed change(s). If the option to use sensitivity studies will be applied to future changes made under 10 CFR 50.46a, the description of the RIEP included in the application to implement 10 CFR 50.46a should describe this process.

k.

The risk assessment of proposed changes to the facility, technical specifications, and procedures should discuss the analysis uncertainties and the potential impact of uncertainty on the analysis results. Key sources of uncertainty6 should be identified and their impact on the results analyzed.

The sensitivity of the model results to model boundary conditions and other key assumptions7 should be evaluated using sensitivity analyses to look at key assumptions both individually and in logical combinations, if applicable. The combinations analyzed should be chosen to account for interactions among the variables. The uncertainty evaluation and conclusion for each proposed change should be considered together with the results from all previous changes to provide reasonable assurance that appropriate consideration of uncertainty does not call into question the acceptability of the results.

RG 1.200 contains the staffs guidance concerning PRA technical acceptability, level of detail, and peer review.

6 Key source of uncertainty is defined in RG 1.200.

7 Key assumption is defined in RG 1.200.

DG-1426, Page 14 Plant Representation and Probabilistic Risk Assessment Configuration Control

l.

The PRA model should reasonably represent the current configuration and operating practices at the plant. Plant representation is defined in terms of how closely the base PRA represents the as-built and as-operated plant. In general, PRA results used to support a 10 CFR 50.46a application must be derived from a base PRA model that represents the as-built and as-operated plant to the extent needed to support the application. Consequently, the base PRA is maintained and upgraded, where necessary, to ensure it represents the as-built and as-operated plant. RG 1.200 contains the staffs guidance concerning plant representation and PRA configuration control.

2.2.2 Risk Assessments Other Than Probabilistic Risk Assessment

a.

The impact of plant proposed changes on operating modes and initiators out of scope from the PRA should be evaluated and estimated with other risk assessment methods. An entity using risk assessment methods other than PRAs to provide numerical representations on estimates of risk for some modes and initiators should justify that the methods used produce realistic results. The results should be either a quantitative estimate of the change in CDF and LERF or a demonstration that the proposed changes will not significantly affect the risk associated with the out-of-scope modes and initiators. The total change in CDF and LERF after implementation of 10 CFR 50.46a from changes in these non-PRA risk assessments should be tracked and combined with the total change estimates from the PRA assessments. The potential for greater impact on uncertainty from these non-PRA risk assessment methods should be appropriately considered.

b.

Even with a PRA that addresses all the relevant initiators and operating modes, many proposed facility changes may affect equipment that is not explicitly modeled in the PRA. For example, containment leak detection systems are relied upon to provide vital information to the operator following some accidents and transients. The leak detection system may be credited while evaluating the operator error probabilities, but the PRA model does not explicitly include the success or failure of the equipment.

The risk-informed evaluation should include a methodology and a decision framework that will permit the evaluation of the risk significance of equipment that is not explicitly modeled in the PRA and the potential change in risk associated with any associated proposed change.

c.

Entities choosing to implement 10 CFR 50.46a should develop and obtain NRC approval of a framework for determining the risk significance of equipment effected by the proposed change that is not explicitly modeled in the PRA. Processes developed to determine the risk significance of equipment that is not explicitly modeled in the PRA but is scoped into the Maintenance Rule (10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants) could potentially provide an adequate framework for determining the risk significance of such equipment for 10 CFR 50.46a. Proposed changes to equipment not explicitly modeled in the PRA may be risk significant and should be assessed to determine whether it can be reasonably concluded that any such change would not increase risk. If an assessment concludes that a proposed change to the plant might increase risk, the PRA should be modified to become capable of reflecting the impact of the proposed change so that the change in risk may be quantitatively estimated. In any case, the effects of the proposed changes on the reliability and unavailability of SSCs or on operator actions that are modeled in the PRA should be accounted for appropriately. For 10 CFR 50.46a submittals that contain non-PRA type analyses, entities should document the bases for why the method employed is technically adequate for this application.

DG-1426, Page 15 2.2.3 Risk Metrics The regulation in 10 CFR 50.46a(h)(2)(ii) requires, in part, that the total increases in CDF and LERF due to facility, facility design, technical specifications, and procedure changes that have been and are proposed to be implemented are themselves very small, and that plant baseline risk remains small.

10 CFR 50.46a(h)(1)(ii) requires, in part, that any increase in CDF and LERF due to facility, facility design, and procedure changes to be minimal for an entity to make the change under 10 CFR 50.46a without prior NRC approval. Therefore, the increase in and total CDF and LERF should be evaluated. If they can be calculated, these metrics should be estimated and compared to the acceptance guidelines provided below.

2.2.3.1 Acceptance Guidelines for Risk-Informed Evaluations Requiring Prior NRC Approval

a.

10 CFR 50.46a(h)(2)(ii) requires, in part, that the total increases in CDF and LERF due to facility design, technical specifications, and procedure changes that have been completed and are proposed are very small, and the overall risk remains small. Entities are required to submit changes evaluated under 10 CFR 50.46a(h)(2) for NRC staff review and approval.

b.

The risk acceptance guidelines presented in this RG are based on the principles and expectations for risk-informed regulation discussed in part C of RG 1.174, as modified by part C of this RG. The guidelines are structured like those in RG 1.174. Regions are established in the two planes generated by a measure of the base risk metric (CDF or LERF) along the x-axis and the change in those metrics (CDF or LERF) along the y-axis (see figures 4 and 5 in RG 1.174). Acceptance guidelines are established for the regions as discussed below in this RG. These guidelines should be used for comparison with a full-scope (including internal and external hazards, at-power, low-power, and shutdown) assessment of the change in risk metric and, when necessary, as discussed below, the base value of the risk metric (CDF or LERF). However, the NRC recognizes that many PRAs are not full scope, and PRA information of less than full scope may be acceptable, as discussed in section C.2.2.1 of this RG.

c.

The two sets of acceptance guidelines, one for CDF and one for LERF, should both be used:

Applications that result in increases of CDF above 10-6 per reactor year (i.e., the increase in CDF falls within Region I or II of figure 4 in RG 1.174) will not be considered. These resultant increases of CDF are outside the acceptance criteria stated in 10 CFR 50.46a(h)(2)(ii).

When the calculated increase in CDF is very small (i.e., the increase in CDF falls within Region III of figure 4 in RG 1.174), which means the increase is less than 10-6 per reactor year, the proposed change should only be considered if it can be shown that the total CDF is less than 10-4 per reactor year.

AND Applications that result in increases to LERF above 10-7 per reactor year (i.e., the increase in LERF falls within Region I or II of figure 5 in RG 1.174) will not be considered. These resultant increases to LERF are outside the acceptance criteria stated in 10 CFR 50.46a(h)(2)(ii).

DG-1426, Page 16 When the calculated increase in LERF is very small (i.e., the increase in LERF falls within Region III of figure 5 in RG 1.174), which means the increase is less than 10-7 per reactor year, the proposed change should only be considered if it can be shown that the total LERF is less than 10-5 per reactor year.

d.

Following these guidelines provides reasonable assurance that proposed increases in CDF and LERF are small and consistent with the Commissions Safety Goal Policy Statement. The guidelines discussed above are applicable for at-power, low-power, and shutdown operations.

e.

Section 6.3.2 of RG 1.174 contains the staffs guidance on cumulative risk. As part of the risk assessment, the entity should address cumulative risks. If previous changes have been made under 10 CFR 50.46a, the entity should provide the cumulative effect on risk of the proposed change and all previous changes made under 10 CFR 50.46a in the application. If multiple plant changes, including plant changes not enabled by 10 CFR 50.46a, are combined into a group for the purpose of evaluating acceptable risk increases, the entity should evaluate each individual change and separately evaluate the combined changes. The acceptance guidelines above apply to both individual changes and combined changes.

2.2.3.2 Acceptance Guidelines for Self-Approved Risk-Informed Evaluations

a.

Under 10 CFR 50.46a(h)(1)(ii), an entity is precluded from making changes enabled by 10 CFR 50.46a without NRC review and approval if the RIEP assessment, considered individually, does not constitute a minimal increase in risk compared to the overall plant risk profile.

b.

Minimal impacts on risk should not be risk significant. A minimal increase in risk is an increase less than 10 percent of the risk increases that would be very small for any entity. The acceptance guidelines in RG 1.174 state that the calculated increase in CDF is very small when the increase in CDF falls within Region III of figure 4, which means the increase is less than 10-6 per reactor year. Similarly, the calculated increase in LERF is very small when the increase in LERF falls within Region III of figure 5, which means the increase is less than 10-7 per reactor year. Calculated increases in CDF less than 10-7 per reactor year and calculated increases in LERF less than 10-8 per reactor year are less than the very small region as shown in figures 4 and 5 in RG 1.174, respectively. Therefore, a minimal risk increase is an increase of less than 10-7 per reactor year for CDF and an increase in LERF less than 10-8 per reactor year. These values are two orders of magnitude below the maximum allowed risk increase guidelines in RG 1.174 and one order of magnitude less than the very small criterion.

c.

Identifying an increase as a minimal increase does not free the entity from ensuring that the PRA used to calculate future proposed changes reflects the current design and operation of the plant.

Facility and procedure changes that affect the PRA models should be incorporated into the PRA as part of the PRA update process, regardless of the quantitative impact of the individual change on risk. Therefore, tracking the total change as the difference between the baseline risk and the current risk will include the cumulative impact of all the minimal changes without requiring any additional processing or tracking. If PRA scope limitations require that the total change is being partially tracked independently of the overall risk estimates, the minimal increases that might affect the risk estimates beyond the scope of the PRA should be tracked as part of the non-PRA cumulative risk increase tracking process.

d.

Regardless of whether an entity makes changes under 10 CFR 50.46a(h)(1) instead of 10 CFR 50.46a(h)(2), the total cumulative risk (including all the individually minimal risk increases, as well as any increases approved by the NRC under 10 CFR 50.46a(h)(2)) should be considered.

DG-1426, Page 17 2.2.4 Defense-in-Depth Evaluation

a.

The engineering evaluation should demonstrate whether the implementation of the proposed change is consistent with the DID philosophy. The purpose of this key principle of risk-informed decision-making is to ensure that any impact of the proposed change on DID is fully understood and addressed and that consistency with the DID philosophy is maintained. The purpose is not to prevent changes in the way DID is achieved. The entity should fully understand how the proposed change impacts plant design and operation from both risk and traditional engineering perspectives.

b.

The entity must assess whether the proposed facility, technical specification, and procedure changes (individually and cumulatively) meet the DID principle. Even when a comprehensive risk assessment of the proposed change has been conducted, consideration of the impact of the change on DID should still be conducted of sufficient depth to ensure that unacceptable erosion of DID has not occurred. When a comprehensive risk analysis is not or cannot be done, additional attention should be given to traditional DID considerations, which should be used or maintained to account for uncertainties.

Every DID evaluation should consider the general design criteria, national standards, and engineering principles such as the single failure criterion. Further, the evaluation should consider the impact of the proposed change on barriers (both preventive and mitigative) to core damage, containment failure or bypass, and the balance among DID attributes. As implied above, a comprehensive risk assessment will often provide much, but not necessarily all, of the insights necessary to conclude that the DID philosophy is maintained.

c.

RG 1.174 contains the staffs guidance concerning DID. DID consists of several considerations summarized below and discussed in RG 1.174. Consistency with the DID philosophy is maintained if all the following apply:

A reasonable balance is preserved among the layers of defense.

Adequate capability of design features is preserved without overreliance on programmatic activities as compensatory measures.

System redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainties, are preserved.

Defenses against potential common-cause failures are preserved and the potential for the introduction of new common-cause failure mechanisms is assessed.

Multiple fission product barriers are maintained.

Sufficient defense against human errors is preserved.

The intent of the plants design criteria continues to be met.

d.

The regulation at 10 CFR 50.46a enables a wide variety of containment-related changes, including some that may affect the frequency of late containment failure without affecting either CDF or LERF. Entities should, however, retain a level of mitigation to ensure that mitigation capabilities are maintained for accident sequences that lead to relatively late containment failure and result in late radiological releases to the public. Therefore, demonstration of reasonable balance between prevention of core damage and prevention of containment failure requires, in part, that any increase in the probability of

DG-1426, Page 18 containment failure (early and late) does not significantly increase the frequency of a significant fission product release.

e.

As stated above, maintaining multiple fission product barriers is one of the several considerations of DID. DG-1434, Addressing the Consequences of Fuel Dispersal in Light-Water Reactor Loss-of-Coolant Accidents (Ref. 22), contains the staffs guidance concerning fuel fragmentation, relocation, and dispersal. The entity should address any impact on fission product barriers from fuel fragmentation, relocation, and dispersal in the DID evaluation.

2.2.5 Safety Margins

a.

The engineering evaluation should assess whether the impact of the proposed change is consistent with the principle that sufficient safety margins are maintained. The entity should choose the method of engineering analysis appropriate for evaluating whether sufficient safety margins would be maintained if the proposed change were implemented. An example of maintaining sufficient safety margins occurs when the margin between the analysis results and the performance criteria compensates for the uncertainties associated with the analysis and data. Another way that safety margins are maintained is through the application of consensus codes and standards. Consensus codes and standards are typically designed to ensure that sufficient margins exist between the licensing basis performance criteria and the actual capacity, such that no additional margin is needed between the analysis results and the performance criteria. This section summarizes an acceptable set of guidelines for making that assessment, though entities may use other equivalent guidelines.

b.

With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met and (2) safety analysis acceptance criteria in the plant licensing basis (e.g., FSAR, supporting analyses) are met or proposed revisions provide sufficient margin to account for uncertainty in the analysis and data.

3 Element 3: Define Implementation and Monitoring Program

a.

RG 1.174 contains the staffs guidance concerning implementation and the associated performance monitoring strategies for the proposed changes.

b.

Key components of risk-informed regulation are the feedback of information, monitoring of changes in plant risk, and updating the risk assessment or plant design activities and processes that are the subject of the risk assessment, or both. Under 10 CFR 50.46a, the implemented monitoring programs should be designed, so that the evaluation of the acceptability of proposed changes to the plant and the subsequent monitoring and feedback of the plant safety after the changes have been implemented are accomplished with an integrated set of tools. The program should be structured such that (1) degradation in SSC performance is detected and corrected before plant safety can be compromised, (2) feedback of information and corrective actions are accomplished in a timely manner, and (3) SSCs are monitored commensurate with their safety significance (i.e., monitoring for SSCs categorized as having low safety significance may be less rigorous than that for SSCs of high safety significance).

c.

Entities should propose monitoring aspects that include a means to adequately track the performance of equipment that, when degraded, can affect the conclusions of the entitys engineering evaluation and integrated decision-making. The program should be capable of trending equipment performance after a proposed change has been implemented, to demonstrate that performance is consistent with that assumed in the traditional engineering and probabilistic analyses conducted to justify the change. The program should also be capable of evaluating new information and modifications to

DG-1426, Page 19 assumptions that could affect the conclusions of the risk-informed evaluation that supports the proposed change.

d.

In support of the monitoring, 10 CFR 50.46a(h)(3)(iii) requires that adequate performance measurement programs and feedback strategies are implemented to ensure that the risk-informed evaluation continues to reflect actual plant design and operation. The risk-informed evaluation includes the risk assessment, maintenance of DID, and adequate safety margins. An important component in the implementation of the risk-informed evaluation is the risk assessment. A plants risk may vary as its configuration or procedures are modified. Therefore, an entity that implements any facility, technical specifications, or procedure change under 10 CFR 50.46a should update the risk assessment (both PRA and associated non-PRA). The updating is to address changes to the plant, operational practices, equipment performance, and plant operational experience. In addition, when errors, nonconformances, degraded conditions, or conditions adverse to quality are discovered, the entity should ensure that any impact on the analyses used to support changes to the facility, technical specifications, or procedures is determined. The updated risk assessments should continue to meet the minimum quality requirements specified in section C.2.2.1 of this RG.

e.

Licensees should integrate the performance measuring programs required by 10 CFR 50.46a with existing programs for monitoring equipment performance and other operating experience on their site and throughout industry. Monitoring that is performed in conformance with the Maintenance Rule for equipment scoped into the Maintenance Rule could be used when such monitoring is sufficient to meet the requirements in 10 CFR 50.46a(h)(3)(iii). Licensees that have implemented previous risk-informed regulatory actions have normally also been required to implement risk-informed monitoring and feedback programs; for example, licensees that adopt 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, will need to develop relatively extensive risk-informed monitoring and feedback programs. These programs should be integrated into the proposed 10 CFR 50.46a(h)(3)(iii) performance measuring programs to the extent practicable.

f.

The updating of the risk assessments should be conducted no less often than once every 5 years. More frequent updates should be performed when events (e.g., plant modifications) or circumstances (e.g., discovery of one or more errors in the PRA model) could be expected to change the risk assessment to a significant extent. Based upon updated risk assessments or changes to the TBS, the entity should take appropriate action to ensure that all changes accomplished under 10 CFR 50.46a continue to meet all applicable acceptance criteria.

g.

Entities implementing 10 CFR 50.46a are required under 10 CFR 50.46a(j) to estimate the effect of any change to or error in evaluation models or analysis methods or in the application of such models or methods to determine whether the change or error is significant. Significant errors must be reported to the NRC. Under 10 CFR 50.46a(j)(3), entities are also required to submit, every 24 months, as specified in 10 CFR 50.4, Written communications, and 10 CFR 52.3, Written communications, a short description of each implemented change involving minimal changes in risk since the last report.

This report periodicity is the same as the 10 CFR 50.59 reporting requirement, so an entity may include both types of changes in one report.

4 Element 4: Submit the License Amendment Request

a.

An entity that is complying with 10 CFR 50.46 but wishes to implement 10 CFR 50.46a requirements should submit a license amendment request under 10 CFR 50.90 for NRC approval to implement the alternative requirements. The license amendment request should include a description of the method(s) and results of the analyses to demonstrate compliance with the ECCS acceptance criteria in

DG-1426, Page 20 10 CFR 50.46a(e) and must include a description of the risk-informed evaluation used in evaluating whether proposed changes to the facility, technical specifications, or procedures meet the requirements in 10 CFR 50.46a(h). For the risk-informed evaluation, 10 CFR 50.46a(c)(1)(v)(B) requires a description of the entitys PRA model and/or non-PRA risk assessment methods used to demonstrate compliance with 10 CFR 50.46a(h)(4)-(5). 10 CFR 50.46a(c)(1)(v)(A) requires a description of the approach, methods, and decision-making process to be used for evaluating compliance with the acceptance criteria in 10 CFR 50.46a(h)(1), (2) and (3), which includes the risk criteria, DID criteria, safety margin criteria, and performance measurement criteria.

b.

The information required to be submitted in the application forms the basis for the NRCs determination of whether the entitys method(s) for demonstrating compliance with the ECCS acceptance criteria in 10 CFR 50.46a(e)(1) meets the requirements in 10 CFR 50.46a(e)(2)-(3); whether the entitys risk-informed evaluation, including any PRA model and other risk assessment methods, meets the requirements in 10 CFR 50.46a(h); and whether the entitys RIEP ensures that changes made pursuant to 10 CFR 50.46a(h)(1) are permitted under 10 CFR 50.59. Upon approval of the license amendment by NRC, the entity will use the RIEP for all pending and subsequent proposed changes. In the case of changes that can be performed without prior NRC review under 10 CFR 50.46a(h)(1), the entitys RIEP will be how the entity determines that the proposed change results in no more than a minimal increase in risk and meets 10 CFR 50.59. Therefore, the entity should provide this information to a level of detail sufficient to provide the NRC staff with confidence that the requirements of 10 CFR 50.46a will be met.

c.

To support the NRC staffs conclusion that the proposed change is consistent with key principles of risk-informed regulation, the entity should submit the information described below as part of its request to adopt 10 CFR 50.46a into its licensing bases.

Risk-Informed Evaluation Process

d.

The entity should describe in detail its process for risk-informed decision-making that will be used to support changes after incorporating 10 CFR 50.46a into the plants licensing bases. The entity should describe existing performance measurement strategies (e.g., equipment scoped into the Maintenance Rule program) and its process for developing additional, change-specific monitoring mechanisms that might be required to ensure no adverse safety degradation occurs following future plant changes. The entity should provide a discussion of how the RIEP will evaluate proposed changes to the plant against the risk criteria, including CDF and LERF, and the criteria on maintaining an appropriate level of DID and sufficient safety margins.

e.

The entity should emphasize how the RIEP will demonstrate the acceptability of proposed changes that may be made without prior NRC review and approval. This should include the technical process to be used in evaluating such changes. The submittal of the technical process should include the methodology for performing the risk assessment and criteria for identifying changes that are both minimal from a risk standpoint and do not significantly affect DID or plant physical security. The entity should identify how the RIEP will be integrated with 10 CFR 50.59. The entity should include a discussion identifying in general terms the types of future proposed changes that its models and processes are capable of evaluating and provide the required confidence that the acceptance criteria in 10 CFR 50.46a(h) are met without prior NRC review and approval.

Probabilistic Risk Assessment Models and Methods

f.

To generate confidence in the risk assessment used to support the proposed change, the entity should submit a summary of the PRA model and methods used. Consistent with its current practice, the NRC will make publicly available any information submitted to the NRC for its consideration in

DG-1426, Page 21 making risk-informed regulatory decisions, unless such information is properly identified as proprietary in accordance with the regulations. Entities should submit the following information to show that the engineering analyses conducted to justify the proposed change are appropriate to the nature and scope of the change:

a description of the PRA that will be used in the risk assessment (this description should include the plant operating modes and the initiating events modeled),

documentation showing that the base PRA is acceptable, a description of the process for ensuring PRA acceptability and a discussion of why the PRA is acceptable to support the application (e.g., the description should discuss the technical acceptability and level of detail of the PRA model and provide the results of any peer review or other evaluation of the PRA model to consensus standards, including a discussion of the resolution of all review comments),

the quality assurance practices for the PRA model, including a description of the process for model update, qualification of risk assessment personnel, use and extent of written procedures to control PRA activities, software quality assurance, and the entitys corrective action program, the key modeling assumptions necessary to support the analysis or that affect the application,8 information related to consideration of uncertainty in the analyses used to support the application (NUREG-1855 provides acceptable guidance for the treatment of uncertainties in risk-informed decision-making),

the event trees and fault trees that require modification to support analyses of the proposed change with a description of their modification, and a list of operator actions modeled in the PRA that affect the application and their error probabilities.

g.

The submitted information summarizing the results of the risk assessment should include the following:

the effects of the proposed change on the more significant sequences (e.g., sequences that contribute more than 5 percent to the risk) to show that the proposed change does not create risk outliers and does not exacerbate existing risk outliers; an assessment of the change to CDF and LERF, including a description of the significant contributors to the change; 8

A modeling assumption is related to a model uncertainty and is made with the knowledge that a different reasonable alternative assumption exists. A reasonable alternative assumption is one that has broad acceptance within the technical community and for which the technical basis for consideration is at least as sound as that of the assumption being made. An assumption is considered key when it may influence (i.e., have the potential to change) the decision being made. NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, issued March 2017 (Ref. 23), provides useful insights related to these concepts.

DG-1426, Page 22 information related to the assessment of the full-scope base CDF and LERF, including quantification details such as truncation level, demonstration of convergence of results, the results of a parametric uncertainty study, and the dominant sequences that comprise the plants risk profile; results of sensitivity analyses showing that the conclusions as to the impact of the proposed change on plant risk do not vary significantly under a different set of plausible assumptions; and information related to the issues identified in section C.2.6 of RG 1.174 if the risk metrics approach the acceptance guidelines in section C.2.2.3 of this RG.

Risk Assessments Other Than the Probabilistic Risk Assessment

h.

The entity should submit detailed information on any non-PRA risk methods that may be used in the risk-informed evaluation. This description should include the plant operating modes and the internal and external initiating events that are evaluated. The entity should (1) discuss the technical acceptability and level of detail of the evaluation, (2) provide the results of any peer review or other evaluations of the risk assessment against generally accepted methods, standards, and practices that have been endorsed or otherwise found acceptable by the NRC, and (3) discuss the resolution of all review comments. The entity should provide its quality assurance practices, including a description of the process for updating the evaluation. The entity should include a justification that the risk assessment will produce realistic available best estimates of CDF and LERF, including identifying key assumptions in the evaluation.

5 Quality Assurance

a.

As stated in section C.2 of this RG, the engineering analyses conducted should justify that proposed changes are appropriate for the nature of the change. For traditional engineering analyses (e.g., deterministic engineering calculations), existing provisions for quality assurance (e.g., Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50 for safety-related SSCs) should apply and provide the appropriate quality needed. Likewise, when a risk assessment of the plant is used to provide insights into the decision-making process, the PRA should be subject to quality control.

b.

To the extent that an entity elects to use PRA information to enhance or modify activities affecting the safety-related functions of SSCs, the following (in conjunction with the other guidance presented in this RG) describes methods acceptable to the NRC staff to ensure that the pertinent quality assurance requirements of Appendix B to 10 CFR Part 50 are met and that the PRA is sufficient for use in regulatory decisions:

Use personnel qualified for the analysis.

Use procedures that ensure control of documentation, including revisions, and provide for independent review, verification, or checking of calculations and information used in the analyses. (An independent peer review or certification program can be an important element in this process.)

DG-1426, Page 23 Provide documentation and maintain records in accordance with the guidelines in section C.6 of this guide.

Use procedures to ensure appropriate attention and corrective actions if assumptions, analyses, or information used in previous decision-making are changed (e.g., entity voluntary action) or determined to be in error.

c.

When performance monitoring programs are used in the implementation of proposed changes to the licensing basis, those programs should include quality assurance provisions commensurate with the safety significance of affected SSCs. An existing PRA or analysis can be used to support a proposed change, if it can be shown that the appropriate quality provisions are met.

6 Documentation The entity should document each plant change risk-informed evaluation consistent with section 4 of RG 1.200 and must retain the documentation in accordance with 10 CFR 50.46a(l).

DG-1426, Page 24 D. IMPLEMENTATION Licensees generally are not required to comply with the guidance in this regulatory guide. If the NRC proposes to use this regulatory guide in an action that would constitute backfitting, as that term is defined in 10 CFR 50.109, Backfitting, and as described in NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests (Ref. 24); affect the issue finality of an approval issued under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants; or constitute forward fitting, as that term is defined in Management Directive 8.4, then the NRC staff will apply the applicable policy in Management Directive 8.4 to justify the action.

If a licensee believes that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may inform the NRC staff in accordance with Management Directive 8.4.

DG-1426, Page 25 REFERENCES9

1.

U.S. Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, Energy.

2.

CFR, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Chapter 1, Title 10, Energy.

3.

Atomic Energy Act of 1954, as amended, Section 42, United States Code (U.S.C.) § 2161, et. seq.

4.

Energy Reorganization Act of 1974, as amended, Section 42, U.S.C. § 5801, et. seq.

5.

U.S. Nuclear Regulatory Commission (NRC), NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 19.1, Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed License Amendment Requests after Initial Fuel Load, Washington, DC.

6.

NRC, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, Washington, DC.

7.

NRC, Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Washington, DC.

8.

NRC, RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Washington, DC.

9.

American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS),

ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, New York, New York, and La Grange Park, Illinois, February 2009.10

10.

NRC, Design Certification/Combined License (DC/COL) Interim Staff Guidance (ISG)

DC/COL-ISG-028, Assessing the Technical Adequacy of the Advanced Light-Water Reactor Probabilistic Risk Assessment for the Design Certification Application and Combined License Application, Washington, DC, November 2016.

9 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

10 Copies of International Atomic Energy Agency (IAEA) documents may be obtained through their website: www.iaea.org/ or by writing the International Atomic Energy Agency, P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria.

Copies of American Society of Mechanical Engineers (ASME) standards may be purchased from ASME, Two Park Avenue, N

DG-1426, Page 26

11.

NRC, Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement, Federal Register, Vol. 60, No. 158: pp. 42622-42629 (60 FR 42622),

Washington, DC, August 16, 1995.

12.

NRC, Safety Goals for the Operations of Nuclear Power Plants; Policy Statement, Federal Register, Volume 51, pp. 30028-30033 (51 FR 30028), Washington, DC, August 4, 1986.

13.

NRC, NUREG-1903, Seismic Considerations For the Transition Break Size, Washington, DC, February 2008.

14.

NRC, Continued Applicability of NUREG-1903, Washington, DC, August 2024.

(ML24207A140).

15.

NRC, NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, Washington, DC, March 2008.

16.

NRC, Continued Applicability of NUREG-1829, Washington, DC, August 2024 (ML24205A015).

17.

NRC, DG-1428 (proposed new RG 1.258), Plant-Specific Applicability of Transition Break Size, Washington, DC.

18.

NRC, Nuclear Regulatory Commission International Policy Statement, Federal Register, Vol. 79, No. 132: pp. 39415-39418 (79 FR 39415), Washington, DC, July 10, 2014.

19.

NRC, Management Directive (MD) 6.6, Regulatory Guides, Washington, DC.

20.

International Atomic Energy Agency (IAEA) Safety Standard Series No. SSG-3, Revision 1, Development and Application of Level 1 Probabilistic Safety Assessment for Nuclear Power Plants, Vienna, Austria, 2024.11

21.

IAEA Safety Standard Series No. SSG-4, Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants, Vienna, Austria, 2010.

22.

NRC, DG-1434 (proposed new RG 1.259), Addressing the Consequences of Fuel Dispersal in Light-Water Reactor Loss-of-Coolant Accidents, Washington, DC.

23.

NRC, NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Washington, DC, March 2017.

24.

NRC, MD 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, Washington, DC.

11 Copies of International Atomic Energy Agency (IAEA) documents may be obtained through their website: www.iaea.org/ or by writing the International Atomic Energy Agency, P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria.