ML22088A135

Guidelines for Characterizing the Safety Impact of Issues, Revision 2

ML22088A135
Person / Time
Issue date:	05/10/2022
From:	Michelle Kichline, Antonios Zoulis NRC/NRR/DRA
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Lingam S
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Download: ML22088A135 (29)

Text

U.S. NUCLEAR REGULATORY COMMISSION

GUIDELINES FOR CHARACTERIZING THE SAFETY IMPACT OF ISSUES

Revision 2 May 2022

Prepared by:

Antonios M. Zoulis Michelle Kichline Office of Nuclear Reactor Regulation Division of Risk Assessment

Summary: This revision includes guidance with the removal of l imitation on applying RIPE to license amendments involving changes to the Technical Specifica tions

TABLE OF CONTENTS

1.0 INTRODUCTION

............................................................................................................... 1 1.1 Purpose................................................................................................................ 1 1.2 Applicability.......................................................................................................... 2 1.3 Scope................................................................................................................... 3 1.4 Content of this Guidance Document.................................................................... 4

2.0 DEFINING THE ISSUE..................................................................................................... 7 2.1 Assessing the Preliminary Screening Questions................................................. 7 2.2 Assessing the Preliminary Risk Impact using Quantitative Analys is.................... 9 2.3 Important Considerations................................................................................... 10 2.4 Documentation................................................................................................... 11

3.0 EXPLORING THE IMPACT OF THE ISSUE.................................................................. 12 3.1 Generic Assessment Expert Team.................................................................... 12 3.2 Plant Integrated Decision Making Panel............................................................ 13 3.3 Documentation................................................................................................... 14

4.0 FINALIZING THE SAFETY IMPACT CHARACTERIZATION........................................ 15 4.1 Step 1 - Screening for No Impact....................................................................... 17 4.2 Step 2 - Screening for Minimal Impact............................................................... 17 4.3 Step 3 - Determining Safety Impact Using Quantitative Analyses..................... 21 4.4 Step 4 - Assess Need for Risk Management Actions........................................ 22

5.0 ASSESSING CUMULATIVE RISK................................................................................. 23

6.0 PAPERWORK REDUCTION ACT.................................................................................. 23 6.1 Paperwork Reduction Act.................................................................................. 23 6.2 Public Protection Notification............................................................................. 23

7.0 CONGRESSIONAL REVIEW ACT................................................................................. 23

8.0 REFERENCES

................................................................................................................ 24

i LIST OF FIGURES AND TABLES

Figure 1-1: Risk-Informed Process for Evaluations Applicabilit y Flowchart................................. 5 Figure 1-2: Safety Impact Characterization Process Overview.................................................... 6 Figure 4-1: Safety Impact Characterization Detailed Process Ov erview................................... 16

ii ABBREVIATIONS AND ACRONYMS

ADAMS Agencywide Documents Access and Management System CDF core damage frequency CFR Code of Federal Regulations F&Os Facts and Observations FLEX Diverse and Flexible Coping Strategy for Extended Loss of Power GAET Generic Assessment Expert Team IDP Integrated Decision-Making Panel LERF large early release frequency LOCA loss-of-coolant accident NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission NUMARC Nuclear Management and Resource Council PRA probabilistic risk assessment RCP reactor coolant pump RIPE Risk-Informed Process for Evaluations RG Regulatory Guide RITSTF Risk-Informed Technical Specifications Task Force RMA risk management action RMTS risk-managed technical specifications SGTR steam generator tube rupture SME subject matter expert SSC structure, system, and component TSG Temporary Staff Guidance TSTF Technical Specifications Task Force

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1.0 INTRODUCTION

1.1 Purpose

This guidance document describes an approach that is acceptable to the staff of the U.S.

Nuclear Regulatory Commission (NRC) for developing a risk-infor med application for an exemption request or license amendment request that applies ris k insights, consistent with the guidance in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3 (Reference 1) and as applicable RG 1.177, Plant-Spe cific, Risk-Informed Decisionmaking: Technical Specifications, Revision 2 (Referen ce 9). It provides general guidance concerning how to characterize that the safety impact of proposed changes in plant design and operation have a minimal impact on safety. This gui dance document does not have the force of law and does not contain any legal requirements.

This guidance document supports the Risk-Informed Process for E valuations (RIPE). RIPE establishes a streamlined NRC review process for risk-informed exemption and amendment requests that have a minimal safety significance. If a license e elects to use RIPE, it would characterize the risk associated with the proposed exemption or amendment request using this guidance and submit its request to the NRC. The NRC staff woul d review the request using a streamlined process outlined in Temporary Staff Guidance (TSG) TSG-2021-01, Revision 2, Risk-Informed Process for Evaluations (Reference 2).

RIPE is intended to build on the expanded use of probabilistic risk assessment (PRA) models for risk-informed initiatives and to benefit from the use of in tegrated decision-making panels (IDPs) that were developed as part of the implementation of Tit le 10 of the Code of the Federal Regulations (10 CFR) Section 50.69, Risk-informed categorization and trea tment of structures, systems and components for nuclear power plants (10 CFR 50.69). RIPE is available to licensees that have demonstrated they have a technically accept able PRA and a robust IDP by the adoption of certain risk-informed initiatives. Section 1.2 and Figure 1-1 of this document provides additional details on what constitutes a technically a cceptable PRA. This document provides guidance that the IDP can use to characterize the safe ty impact of the proposed change using both quantitative risk information from the PRA an d qualitative risk insights consistent with RG 1.174 and as applicable RG 1.177. RIPE was originally developed to be used by licensees that could demonstrate that they have a techn ically acceptable PRA by having implemented an approved license amendment to adopt Techn ical Specifications Task Force (TSTF) Traveler TSTF-505, Provide Risk Informed Extended Completion Times -

RITSTF [Risk-Informed TSTF] Initiative 4b. 1 The original implementation of RIPE was approved by memo dated January 7, 2021 (Reference 3). This rev ision expands RIPE to allow licensees to demonstrate that they have a technically acceptabl e PRA by having implemented an approved license amendment to adopt TSTF-425, Relocate Surv eillance Frequencies to Licensee Control-RITSTF Initiative 5b. 2

1 NRC has approved some licensee programs for risk-informed completion times consistent with Nuclear Energy Institute (NEI) 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, which can be used in lieu of TSTF-505 to characterize the safety impact of issues. Any references in this document to TSTF-505 includes applications approved consistent with NEI 06-09 (Reference 4).

2 NRC has approved some licensee programs for relocating surveillance frequencies consistent with NEI 04-10 Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of

1 1.2 Applicability

Use of this guidance is limited to proposed changes to faciliti es for which the safety impact associated with the issue can be modeled using PRA. This guida nce is not applicable for holders of combined licenses under 10 CFR Part 52.

RIPE may be used by licensees that have a technically acceptabl e PRA model and a robust IDP. For the purposes of RIPE, a licensee may demonstrate that they have a technically acceptable PRA model by having implemented an approved TSTF-505 or TSTF-425 license amendment and having completed all of the license conditions an d implementation items associated with the amendment.

Licensees that rely on a TSTF-425 amendment to demonstrate PRA acceptability in lieu of a TSTF-505 amendment must provide additional information relative to PRA technical acceptability in their submittals. Specifically, licensees tha t rely on their TSTF-425 program to demonstrate PRA technical acceptability will need to justify th at the issue being analyzed is limited to internal events by providing technical justification for the exclusion of external hazards that are not addressed, or identify which additional NRC-approv ed applications address any relevant external hazards beyond internal events. Technical ju stification for exclusion of hazards not addressed in the assessment should include a descri ption and disposition of hazard-specific spatial considerations and a sensitivity study involving the hazards relevant to the site. In addition, licensees relying on TSTF-425 will need to describe any open facts and observations (F&Os) for the PRA models applicable to the issue being analyzed. In order to support a streamlined NRC review, licensees should make every e ffort to close F&Os in advance of submitting a RIPE application, typically using the f inding closure process described in Nuclear Energy Institute (NEI) 17-07, Performance of PRA Pe er Reviews using the ASME/ANS [American Society of Mechanical Engineers/American Nuc lear Society] PRA Standard (Reference 6). The description of these F&Os should include an assessment of the relevance, or lack thereof, of the F&O to the decision being so ught. If an issue involves a hazard (e.g., external events) that is not covered by a previou sly approved NRC application, a licensee that only has an approved TSTF-425 program may not use this process. Further, licensees relying on TSTF-425 will need to describe their PRA m aintenance and peer review process subsequent to approval of their TSTF-425 license amendm ent.

For the purposes of RIPE, a licensee may demonstrate that it ha s a robust IDP by having implemented an approved 10 CFR 50.69 license amendment and havi ng completed all of the license conditions and implementation items associated with the amendment. Licensees do not need to have categorized any structures, systems, and component s (SSCs) in accordance with 10 CFR 50.69 to use their IDP for this process. Licensees that have not implemented an IDP under 10 CFR 50.69 may choose to apply a 10 CFR 50.69 equivalen t IDP as documented in NEI guidance, NEI Guidelines for the Implementation of the Ris k-Informed Process for Evaluations Integrated Decision-Making Panel (Reference 7), to use this process.

Surveillance Frequencies, which can be used in lieu of TSTF-425 to characterize the safety impact of issues. Any references in this document to TSTF-425 includes applications approved consistent with NEI 04-10 (Reference 5).

2 1.3 Scope

Figure 1-1 illustrates how licensees can evaluate whether the t echnical acceptability of their PRA supports the use of RIPE.

Figure 1-2 provides a high-level overview of the process to cha racterize the safety impact of issues.

For the purposes of this guidance document, all the following m ust apply in order to characterize an issue as having a minimal safety impact:

• The issue contributes less than 1 x 10 -7/year to core damage frequency (CDF).

• The issue contributes less than 1 x 10 -8/year to large early release frequency (LERF).

• The issue screens to no impact (per Step 1, Section 4.1) or mi nimal impact (per Step 2, Section 4.2).

• Cumulative risk is acceptable using the guidelines in Section 5.

If any of the criteria above are not met, then the proposed cha nge cannot be characterized as having a minimal impact on safety in accordance with this guida nce document.

The process described in this guidance document does not replac e or affect the NRCs use of the Reactor Oversight Process Significance Determination Proces s for assessing the safety significance of more-than-minor performance deficiencies.

This process is anticipated to be useful when the actions neede d to address an issue would result in a minimal safety impact. This process may also be us eful for issues in which there is a safety benefit to not implementing costly or burdensome actions. Examples of issues for which this process may be used include, but are not limited to, the f ollowing:

• Actions needed to address inspection findings,

• Resolution of issues identified through other regulatory or li censee processes,

• Responses to orders requiring changes or modifications to the plant, and

• Generic issues requiring changes or modifications to the plant.

For issues having generic implications, a generic safety charac terization could, for example, be performed by a Generic Assessment Expert Team (GAET). This generic assessment could then be used to inform a plant-specific assessment of the gener ic issue which accounts for plant-specific risk contributors, such as seismic or flooding r isk, through a licensees multi-disciplinary plant IDP.

This process may not be used for:

• Any immediate actions necessary for continued safe operation ( e.g., to support an NRC finding of adequate protection, to restore compliance with a te chnical specification, to resolve an environmental compliance issue with an adverse effec t on public health and safety, or to remove a threat to personnel safety).

• Any immediate repairs necessary for continued power production (e.g., replacing a damaged main transformer).

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• Any issues for which the safety impact cannot be directly asse ssed using PRA (e.g., fuel changes, changes to emergency planning programs, or changes to security).

1.4 Content of this Guidance Document

Section 2 presents guidance for defining the issue being assess ed.

Section 3 presents guidance for exploring the issue in detail u sing the GAET and/or IDP.

Section 4 presents guidance for finalizing the safety impact ch aracterization.

Section 5 presents guidance for assessing the cumulative risk i mpact.

Section 6 presents information about the Paperwork Reduction Ac t.

Section 7 presents information about the Congressional Review A ct.

Section 8 presents a list of references.

4 Figure 1-1: Risk-Informed Process for Evaluations Applicabilit y Flowchart

5 Figure 1-2: Safety Impact Characterization Process Overview

6 2.0 DEFINING THE ISSUE

This guidance is applied after identifying an issue that requir es NRC review for resolution. Once it is identified that some action is needed to address the issu e, then the licensee would need to define the range of possible resolutions, choose a path forward, and determine the safety impact of that resolution. This guidance can also be applied a fter the partial resolution of an issue initially having a more than minimal safety impact result s in the remaining unresolved aspects of the issue having a minimal safety impact.

The safety impact characterization process starts with defining the specific issue for which the safety impact is being assessed. This should be done by a subj ect matter expert (SME) who is knowledgeable about the issue. The SME collects any available NRC and industry information.

When evaluating an issue, the safety impact being characterized is the difference between plant risk with the proposed change and without the proposed change. For compliance issues, the change in risk is the difference between risk if the plant were fully compliant with its licensing basis, and risk with the plant in the non-compliant configurati on requested in the submittal. For licensee-identified issues that do not involve a compliance iss ue, the change in risk is the difference between risk with the plant in the current configura tion and with the plant in the configuration requested in the submittal.

Defining the issue may begin at a generic or plant-specific lev el. A generic evaluation characterizes the importance of the regulatory issue at a gener ic level and provides an overall assessment and important attributes for consideration in the pl ant-specific evaluation. The generic evaluation may be carried out by an SME or team of expe rts. The generic SME evaluation is then reviewed by the GAET for implementation at a pplicable plants. The licensees SME will revise the generic evaluation as needed to address the plant-specific considerations identified by the GAET and any plant-specific di fferences from the information provided by the GAET. The plant-specific process is carried ou t by the licensee using a plant IDP, which reviews the generic characterization provided by the GAET and the plant-specific evaluation provided by the licensees SME. If the issue does n ot apply generically, then the issue is only defined at the plant-specific level by the licens ees SME and reviewed by the plant IDP.

The SME should define the issue in enough detail for the GAET o r plant IDP to review the issue and make a final determination about the safety impact. Howeve r, the process can be iterative if needed. While the IDP is primarily a reviewing body, the ID P may provide input to the SME to arrive at a final product that supports the decision. The SME should collect any readily available information for the GAET or plant IDP to review but may identif y unknowns for the GAET or plant IDP to consider further. The GAET or plant IDP may decid e they need additional information in order to complete their review and direct the SM E to obtain additional information.

Completely defining the issue includes two essential activities :

1. Performing a detailed assessment of the preliminary screenin g questions.

2. Performing a preliminary risk assessment using a PRA model.

2.1 Assessing the Preliminary Screening Questions

The SME should document the initial assessment of the prelimina ry screening questions. This phase of the process involves screening the issue for any impac t on safety, regardless of whether the impact is adverse or beneficial. The plant IDP wil l develop final responses to similar screening questions.

7 The preliminary screening for any safety impact involves addres sing the following set of questions:

Does the issue:

1. YES NO Result in any impact on the frequency of occurrence of an a ccident initiator or result in a new accident initiator?

2. YES NO Result in any impact on the availability, reliability, or c apability of SSCs or personnel relied upon to mitigate a transient, accident, or natural hazard?

3. YES NO Result in any impact on the consequences of an accident sequence?

4. YES NO Result in any impact on the capability of a fission product barrier?

5. YES NO Result in any impact on defense-in-depth capability or impa ct in safety margin?

Although the answers to the questions are either yes or no, all answers must be explained in detail for consideration by the GAET and/or IDP. If any of the questions are answered YES, then the SME should discuss whether the impact is adverse or be neficial. The SME should discuss any adverse impacts with the risk analyst who will be p erforming the preliminary risk evaluation and have the risk analyst quantify the risk impact, if possible.

In determining whether there is any impact on safety, the first step is to determine what SSCs and human actions are affected by the issue. Next, the effects of the issue should be determined. This evaluation should include both direct and ind irect effects. Direct effects are those where the issue (e.g., changing the motor on a pump or ch anging the mounting of an electrical cabinet) changes the performance of the SSC directly, such as by decreasing its reliability or decreasing its margin to failure under accident conditions. One can directly attribute the overall impact on how the SSC performs by quantitative analysis, operating experience, or engineering judgment. Indirect effects are those where the iss ue could affect other risk contributors.

In addressing the preliminary screening questions, the followin g should be noted:

• The term capability in Questions 2 and 4 addresses the capac ity of SSCs or personnel.

Consider the following examples:

o The flow capacity of a system could be decreased by replacing a pump with a lower capacity pump.

o The tornado resistance of a wall could be decreased by removin g supports.

o The seismic capacity of a relay could be decreased by replacin g the relay with a lower capacity relay.

o The human error probability of an action could increase by dec reasing the amount of time the operator has to perform the action.

• For Screening Question 3 above, consequence is intended to m ean radiological dose from risk-significant accident sequences. The impact could be direct, such as an improved containment spray system that could reduce radiologica l releases in a core

8 damage accident, or indirect, such as an increase in containmen t bypass events.

Reducing the frequency of core damage is addressed elsewhere an d is not the intent of this question.

2.2 Assessing the Preliminary Risk Impact using Quantitative An alysis

The quantitative evaluation of risk impact is an important fact or in determining that the total safety impact of an issue is low enough to characterize the saf ety impact of the issue as minimal. Therefore, if all of the following conditions apply, licensees can leverage their PRA models to perform quantitative risk assessments to support usin g this process:

• The issue is completely within the scope of the licensees PRA model or can be bounded using surrogates. For this process, surrogates are limited to the use of basic events associated with the initiating event, mitigating system, or fun ction of the issue being investigated.

• The licensee has implemented an IDP consistent with risk-infor med initiative 10 CFR 50.69 or equivalent.

• The licensee has implemented risk-informed initiative TSTF-505 or TSTF-425 and has completed all implementation items and license conditions from the safety evaluation. If the licensee is relying on its TSTF-425 program to demonstrate PRA technical acceptability, the licensee must justify that the issue being a nalyzed is limited to internal events by providing technical justification for the exclusion o f external hazards that are not addressed, or identify which additional NRC-approved applic ations address any relevant external hazards beyond internal events.

• The licensees PRA model was found acceptable to support appro val of the relevant risk-informed applications by the NRC.

• The issue is within the scope of the portion(s) of the PRA mod el that was found acceptable by the NRC (e.g., if seismic was screened out of acc eptability, then seismic issues cannot be addressed using this process).

The PRA model must include the capability to assess the change in CDF and LERF, and the risk evaluation must include a quantified assessment of all sig nificant sources of risk (i.e., external events, internal flooding, and fires) that can be impacted by the issue being assessed. Where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact (e.g., external events, l ow power and shutdown).

A risk analyst must use an acceptable PRA model to calculate th e change in CDF and LERF.

The change in CDF and LERF must be calculated as the difference in the risk to the plant with the existing issue and the risk to the plant if there were no i ssue (i.e., if the plant were fully compliant). The risk analysis may not include any credit for p roposed risk management actions (RMAs), compensatory actions, or any other activities implement ed to reduce the risk impact associated with the issue. The risk analyst should document wh ether there are any beneficial safety impacts associated with the issue.

The preliminary risk evaluation may initially be performed on a generic level. For a generic assessment, the risk analyst may need to perform multiple risk calculations using a representative sample of plant PRA models. The representative sample of plants will depend

9 on the issue being addressed and what plants have acceptable PR A models. For example, if the issue applies to a certain plant design or vendor, then the risk evaluation should be performed using a sample of plants of that design or vendor, re spectively. Once the generic risk evaluation is reviewed by the GAET, a plant-specific risk evalu ation must be completed in order to apply this process on a plant-specific level. The plant-spe cific risk evaluation for a generic issue must address any considerations identified by the GAET. If the issue does not apply generically, then the risk is only calculated at the plant-spec ific level by a plant risk analyst and reviewed by the plant IDP.

The risk analyst should document any assumptions made when perf orming the risk evaluation, whether the issue was within the scope of the licensees PRA, a nd whether any surrogates were used to account for the impact of the issue. The impact of unc ertainty on the evaluation should be considered. For any initial screening questions that were a nswered YES, the risk analyst should quantify the risk impact associated with the adverse imp act.

2.3 Important Considerations

In order to fully understand the safety impact of an issue and account for relevant insights in an integrated manner, the assessmen t should consider the following important common elements:

• Ensuring the issue is well-defined: Although the goal of the overall process is to have clearly defined issues prior to evaluation by the GAET or IDP, the actual assessment may indicate that additional definition is appropriate. As the assessment progresses to subsequent steps, the actual conduct of the assessment may iden tify additional considerations not identified in the initial definition(s). Th us, it is critical that the specific issue is appropriately defined and communicated in order to ill ustrate the safety impact due to the issue.

• Being realistic as to not bias the assessment: The level of r ealism and analyses will vary depending on the issue, but in order to avoid bias, realistic a nalysis is the objective. The process should include sensitivity analyses to address the key assumptions and sources of uncertainty that are driving the results. The key assumptio ns, details, and results of the sensitivity studies should be documented for consideration by the GAET or IDP. If the risk impact is exceedingly small, or clearly large, then a bounding evaluation may suffice.

• Considering uncertainty: Both the GAET and IDP need to be awa re of any specific issues, including external events, for which there is uncertain ty. Sensitivity analysis should be performed, commensurate with the impact of the issue, to address any key assumptions and sources of uncertainty that may influence the r esults. The key assumptions, details, and results of the sensitivity studies sh ould be documented for consideration by the GAET or IDP.

• Evaluating the overall nature of the risk impact of a potentia l action: Both beneficial and adverse effects should be considered (e.g., replacing a small p ump with a large pump could reduce the available margin of an emergency diesel genera tor, or closing and depowering pressurizer power operated relief valve block valves to prevent spurious operation could reduce effectiveness of feed and bleed operatio ns).

• Identifying the extent of the impact: The specific intended i mpact of the issue, as well as other related or indirect effects, should be considered (e.g., FLEX provides mitigation for

10 more than external hazards even though that is its fundamental intended purpose). In other words, one specific issue could impact the specific funct ion under consideration as well as multiple other separate plant functions. As discussed above, this could include both positive and negative impacts that may not be immediately evident if the impacts of issue are considered independently.

2.4 Documentation

The issue should be documented in enough detail so that a perso n who is not familiar with the issue can understand the issue and how the safety impact charac terization was made.

Documentation should include:

• A detailed description of the specific regulatory issue.

• Related and publicly available references used to make the saf ety impact conclusion.

• Screening question results, including explanations.

• Quantitative safety impact characterization results and associ ated discussions, including sensitivity analyses, key assumptions, and sources of uncertain ty.

• Technical bases for conclusions regarding safety impact.

• Description of the scope of the risk evaluations used to suppo rt evaluation of the issue (e.g., internal events PRA, fire PRA, etc.).

• Licensees relying on TSTF-425 to demonstrate PRA technical acc eptability must include the following additional information:

o Justification for limiting the evaluation to internal events o r identification of additional NRC-approved applications that address any relevant external hazards beyond internal events.

o Justification for exclusion of any external hazards that are n ot applicable.

o References to NRC-approved applications that include evaluatio n of PRA information from risk evaluations used to support evaluation of the issue.

o Description of the maintenance process of the PRA model, inclu ding any updates, peer reviews, and independent assessments performed si nce the PRA was reviewed as part of an approved licensing action by the NRC.

o Description of any open F&Os against the PRA model(s) used for the application, including justification for whether the open F&Os are applicabl e to the application. If the issue involves an external hazard covered by a previously approved NRC application, justify that the associated PRA does not have any applicable open F&Os.

11 3.0 EXPLORING THE IMPACT OF THE ISSUE

After the issue has been defined by the SME, the potential impa ct of the issue is explored in depth by a multi-disciplinary team of experts (i.e., GAET and/or IDP). This team of experts is responsible for ensuring the issue is fully defined and all the potential safety impacts have been identified. If the team identifies that it needs additional in formation in order to make a final recommendation regarding the safety impact, additional experts should be consulted. The goal of this phase of the review is to identify and review all the a vailable information regarding the issue and characterize its safety impact.

This review may be performed on a generic or plant-specific level. The generic and plant-specific processes involve similar steps. The generic pr ocess starts with a generic evaluation performed by an SME that is reviewed by a GAET and is used to inform the plant-specific evaluation that will be reviewed by the plant ID P. The plant-specific process starts with a plant-specific evaluation by an SME that is reviewed by the plant IDP. The generic process is intended to address issues that impact multiple plan ts, where generic evaluation would simplify or otherwise inform the plant-specific review pr ocess. For the generic process, the GAET characterizes the importance of the regulatory issue a t a generic level and provides an overall assessment and important attributes for consideratio n in the plant-specific evaluation.

When a generic evaluation is performed, a plant-specific evalua tion must also be performed for each plant that plans to use this process to characterize the s afety impact of an issue as minimal. If a generic evaluation is not necessary, then a GAET is not performed, and the issue is only reviewed on a plant-specific level. The plant-specific process is carried out with the use of a plant IDP, which reviews the generic characterization prov ided by the GAET (if performed) and the plant-specific evaluation provided by a plant SME, to a rrive at plant-specific safety impact characterization. This safety impact is characterized a s having either no impact or minimal impact.

The GAET can provide generic importance characterization inform ation and attributes to the industry. Using this information in conjunction with a plant-s pecific evaluation, the plant IDP is responsible for making the plant-specific safety impact charact erization. Both the GAET and plant IDP are multi-disciplinary teams of experts. The followi ng guidance is provided relative to the makeup of these two panels.

3.1 Generic Assessment Expert Team

The GAET is comprised of industry experts with relevant experti se about the issue being evaluated. The GAET composition will vary depending upon the i ssue. Generally, the GAET is composed of knowledgeable personnel whose expertise represents the important process and functional elements of the industry and regulatory processes, s uch as operations, engineering, nuclear risk management, industry operating experience, and lic ensing. The GAET members are expected to have the essential understanding of the issues safety impact, and familiarity with the safety impact characterization process guidance and ap proach. The team can call upon additional personnel, SMEs, or external consultants, as ne cessary, to assist in the characterization of issues. Experience, plant knowledge, and f amiliarity with current regulatory issues are important elements in the selection of GAET members. Members may be experts in more than one field; however, excessive reliance on any one mem bers judgment should be avoided. In general, there should be at least five experts des ignated as members of the GAET with joint expertise in the following fields:

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• plant operations

• design and systems engineering

• safety analysis

• PRA and risk-informed decision-making

• licensing

An SME knowledgeable in the technical discipline or disciplines relevant to the issue being evaluated should function as the lead presenter of the regulato ry issue to the GAET. The SME should provide its evaluation and present the results of the pr eliminary screening questions and preliminary risk evaluation to the GAET. The SME should take r esponsibility to ensure that all relevant documents are available to the GAET. The SME should a lso ensure that the results of the GAET deliberation are documented and records are maintained.

A consensus process should be us ed for decision-making for the GAET. Differing opinions should be documented and considered. However, a simple majorit y of the panel is enough for final decisions regarding the safety impact of the issues. The GAET should apply objective criteria and minimize subjectivity.

3.2 Plant Integrated Decision-Making Panel

The composition of the plant IDP is the same as for the GAET, e xcept that the members of the plant IDP and the SME for the plant IDP should have plant-speci fic knowledge and experience.

The IDP discussed here is intended to be consistent with the ID P implemented as part of 10 CFR 50.69 or equivalent. The IDP is composed of knowledgeab le plant personnel whose expertise represents the important process and functional eleme nts of the plant organization, such as operations, engineering, nuclear risk management, indus try operating experience, licensing and maintenance. The plant IDP can call upon additio nal plant personnel or external consultants, as necessary, to assist in the evaluation of issue s. The precise makeup of the plant IDP is determined by the licensee. Experience and plant knowledge are important elements in the selection of plant IDP members. Members may be experts in more than one field; however, excessive reliance on any one members judgment should be avoided. In general, consistent with other licensee expert panels, there sh ould be experts designated as members of the plant IDP with joint expertise in the following fields:

• plant operations

• design and systems engineering

• safety analysis

• PRA and risk-informed decision-making

• licensing

An SME knowledgeable in the technical disciplines relevant to t he issue being evaluated should function as the lead presenter of the regulatory issue to the p lant IDP. If a generic assessment is available, this assessment is used by the SME as a key input into the plant-specific assessment, along with relevant plant-specific information. Th e SME should provide its evaluation and present the results of the preliminary screening questions and preliminary risk evaluation to the plant IDP. The SME should take responsibilit y to ensure that all relevant generic and plant-specific documents are available to the plant IDP. The SME should ensure that the results of the plant IDP deliberation are documented a nd records are maintained.

13 The plant IDP should be aware of the benefits and limitations o f the plant-specific PRA and other analyses, and, where necessary, should receive training o n the plant-specific PRA, its assumptions, and appropriate implementation. This training fac ilitates making well-supported technical assumptions whether quantitative or qualitative infor mation is used. The plant IDP should be familiar with the technical issue and the safety impa ct characterization process. In order to have a full understanding of the issue being character ized, all questions in each applicable step of the guidance should be answered, even if an initial yes response has already determined the outcome of that step.

A consensus process should be us ed for decision-making for the plant IDP. Differing opinions should be documented and considered. However, a simple majorit y of the panel is enough for final decisions regarding the safety impact of the issues. The plant IDP should apply objective criteria and minimize subjectivity. The plant IDP should be de scribed in a plant administrative procedure that includes the designated chairman, panel members, and panel alternates; required training and expectations for the chairman, members, a nd alternates; requirements for a quorum; attendance records; agendas; and meeting minutes.

3.3 Documentation

GAET: The GAET evaluation results, including a description of any i mportant considerations that should be addressed in the plant-specific assessment, will be documented and provided to the industry and the NRC. Documentation will be maintained to facilitate any subsequent generic update or re-evaluation of the issue, as appropriate.

The GAET should document any considerations and characteristics that may affect the plant-specific assessment, particularly for safety. For exampl e, the GAET may determine that based on reactor fleet considerations, the existing level of ri sk of an external initiator is 1 x 10 -5 to 1 x 10-4/year CDF on average. If information is available, the GAET wo uld convey what attributes could make the plant-specific assessment higher or l ower.

IDP: The IDP evaluation results, including a summary of the basis for each decision will be documented and provided to the NRC. In particular, the assessm ent of any GAET-identified important considerations and how they apply to the plant and a basis for any plant-specific departures from the GAET assessment must be noted. The level o f documentation should be such that a sufficient basis is provided for a knowledgeable in dividual to independently review the information and reach the same conclusion. The basis for a ny engineering judgment and the logic used in the assessment should be documented to the ex tent practicable and to a degree commensurate with the safety impact and complexity of th e issue. The items considered by the GAET, SME, and IDP must be clearly stated.

For each issue, licensees should maintain:

• a copy of the generic package, if applicable;

• a copy of the plant-specific package the SME submits to the pl ant IDP;

• a summary of the plant IDP discussion on the issue;

• a revised copy of the package, if applicable; and

• the final safety impact characterization assigned to the issue.

14 4.0 FINALIZING THE SAFETY IMPACT CHARACTERIZATION

After the plant IDP has reviewed the initial characterization o f the issue provided by the SME, the plant IDP is responsible for providing the final safety imp act characterization. The final safety impact characterization consists of assessing:

1. the final screening questions, and

2. the final risk impact using a PRA.

Both of these activities are essential to characterizing the sa fety impact of the issue. The final screening questions are similar to the preliminary screening qu estions. The information presented for reviewing the preliminary screening questions als o applies to reviewing the final screening questions. Assessing the final screening questions i s progressive and includes two basic steps: (1) a series of screening questions to address wh ether there is any adverse impact to safety, and (2) a series of similar screening questions to a ddress whether the impact to safety is minimal.

Screening determinations are made based on the technical inform ation supporting the issue.

Technical or engineering information that demonstrates that the issue has no adverse effect on functions, or methods of performing or controlling functions ma y be used as a basis for screening the issue.

The plant IDP reviews the issue until it has confidence that th e safety impact characterization results would not change if additional information was obtained or developed. If the plant IDP does not have confidence in the safety impact characterization results, the plant IDP should develop a plan to obtain the information needed to have confide nce in the results of the review.

For example, the plan could include conduct of additional analy ses.

In addressing the screening questions, the following should be noted:

• The term risk-significant in the screening questions refers to SSCs performing risk-significant functions, including nonsafety-related and saf ety-related SSCs and human performance. Nuclear Managem ent and Resource Council (NU MARC) 93-01, Industry Guideline for Monitoring the Effectiveness of Mainten ance at Nuclear Power Plants (Reference 8), provides specific guidance on risk-signi ficant criteria.

NUMARC 93-01 was developed to determine the risk significance o f components scoped into the maintenance rule. However, the guidance in NUM ARC 93-01 can be applied to determine the risk significance of all events, inclu ding initiating events and human actions, relevant to this characterization process by inc luding all events in the assessment of risk-significance.

• Risk impact should be based on the relative change in risk ass ociated with baseline CDF and LERF. Generally, items that are not risk-significant are t hose that contribute less than 1 x 10-7/year and 1 x 10-8/year for CDF and LERF, respectively.

Figure 4-1 on the following page provides a detailed overview o f the safety impact characterization process.

15 Figure 4-1: Safety Impact Characterization Detailed Process Ov erview

16 4.1 Step 1 - Screening for No Impact

Step 1 involves screening the issue for any adverse impact on s afety. The Step 1 screening process is not intended to be resource intensive and is not con cerned with the magnitude of the adverse or beneficial effects that are identified. Any change that adversely affects risk is screened in and must be evaluated in Step 2. The screening for no impact involves addressing the following set of questions:

Does the issue:

1. YES NO Result in an adverse impact on the frequency of occurrence of an accident initiator or result in a new accident initiator?

2. YES NO Result in an adverse impact on the availability, reliabilit y, or capability of SSCs or personnel relied upon to mitigate a trans ient, accident, or natural hazard?

3. YES NO Result in an adverse impact on the consequences of an accid ent sequence?

4. YES NO Result in an adverse impact on the capability of a fission product barrier?

5. YES NO Result in an adverse impact on defense-in-depth capability or impact in safety margin?

If ALL the responses are NO, the issue screens to NO IMPACT. C ontinue to Step 3.

If ANY response is YES, continue to Step 2.

Although the answers to the questions are either yes or no, the answers to all questions must be explained in detail. Beneficial safety impacts should be no ted in the responses to each question. If the issue is only associated with beneficial safe ty impacts, then the Step 1 screening questions would be answered NO, and the issue would s creen to no impact.

4.2 Step 2 - Screening for Minimal Impact

Step 2 involves screening the issue to determine if the magnitu de of the adverse impact on safety identified in Step 1 is minimal. Step 2 should be perfo rmed in conjunction with Step 3, as risk-significance information from the risk analysis is necessa ry to answer the Step 2 questions.

This step involves addressing the following set of questions, w hich are modified versions of the Step 1 questions:

Does the issue:

1. YES NO Result in more than a minimal increase in frequency of occu rrence of a risk significant accident initiator or result in a new ris k significant accident initiator?

2. YES NO Result in more than a minimal decrease in the availability, reliability, or capability of SSCs or personnel relied upon to mitigate a ri sk significant transient, accident, or natural hazard?

3. YES NO Result in more than a minimal increase in the consequences of a risk significant accident sequence?

4. YES NO Result in more than a minimal decrease in the capability of a fission product barrier?

5. YES NO Result in more than a minimal decrease in defense-in-depth capability or safety margin?

17 If ALL the responses are NO, the issue screens to MINIMAL IMPAC T. Continue to Step 3.

If ANY response is YES, stop. The issue has a more than minima l impact on safety.

Although the answers to the questions are either yes or no, the answers to all questions must be explained in detail. Responses must include a discussion as to whether the identified impacts were addressed by the risk analysis. Any question that is answered NO in Step 1, will also be answered NO in Step 2. Guidance on addressing the abov e questions is provided below.

Question 1: Does the issue result in more than a minimal incre ase in the frequency of a risk-significant accident initiator or result in a new risk sig nificant accident initiator?

In answering this question, the first step is to identify the r isk significant accident initiators that have been evaluated that could be affected by the issue. Then a determination should be made as to whether the frequency of these accident initiators occurr ing would be more than minimally increased. Finally, the licensee should determine if any new r isk significant accident initiators have been created. This could be a result of an increase in th e risk significance of an accident initiator that was previously not risk significant. The table below shows an example of typical accident initiators and operating modes (e.g., at power, low po wer, or shutdown conditions) that should be considered:

Accident Initiator Categories (Representative) Risk Significant? More than Minimal Increase?

Transients initiated by frontline systems Transients initiated by support systems Primary system integrity loss (e.g.,

SGTR, RCP seal LOCA, LOCA)

Secondary system integrity loss Internal flooding Internal fires Earthquakes External flooding Tornados and High Winds Other External Hazards Spent Fuel Pool Low power and shutdown conditions

External hazards: External hazard frequencies cannot be reduce d or increased by a plant-initiated or NRC-initiated change. However, the frequency and severity might be changed for certain external hazards (such as external flooding) with chang es beyond the nuclear power plant site. For example, damage to a nearby dam could increase the frequency and severity of an external flood that could affect the nuclear power plant sit e. Such changes can be considered in this process if under the control of the licensee. Otherwise changes related to external hazards will be considered in the second question.

18 The table below shows several ways that the frequency of accide nt initiators can be changed.

Accident Initiator Frequency Considerations Potential Effect? More than Minimal Increase?

Changes in maintenance, training Changes in specific SSCs (e.g., installing a more reliable component)

Changes in materials Equipment replacements to address age related degradation Changes in redundancy or diversity Addition of equipment Changes in operating practices

Reasonable engineering practices, engineering judgment, and PRA techniques should be used in determining whether the frequency of occurrence of a risk-si gnificant accident initiator would more than minimally increase as a result of the issue. A large body of knowledge has been developed in the area of accident frequency and risk-significan t sequences through plant-specific and generic studies. This knowledge should be u sed in determining what constitutes more than a minimal increase in the frequency of oc currence.

Question 2: Does the issue result in more than a minimal decre ase in the availability, reliability or capability of SSCs or personnel relied upon to m itigate a risk-significant transient, accident or natural hazard?

In answering this question, the first step is to identify the r isk significant SSCs and human actions that could be affected by the issue. This question add resses the reactivity control function, including anticipated transients without scram. Anti cipated transients without scram is not an accident initiator, it is an accident sequence. Next, a determination should be made as to whether availability, reliability, or capability of SSCs or per sonnel relied upon to mitigate a risk-significant transient, accident or natural hazard would be more than minimally decreased.

Similar to accident initiators, the availability, reliability, or capability of SSCs or personnel can be changed in several ways, such as those described in the table b elow:

Availability, Reliability, or Capability Considerations Potential Effect? More than Minimal Decrease?

Changes in maintenance, testing, training Changes in specific SSCs (e.g., installing a more reliable component)

Changes in materials Equipment replacements to address age related degradation Changes in redundancy and diversity Addition of equipment Strengthening of equipment

19 Moving equipment (to reduce the impacts of spatial events)

Eliminating the need for recovery action Improving performance shaping factor related to human performance Changes in operating practices

An appropriate calculation can be used to demonstrate the chang e in likelihood in a quantitative sense, if available and practical. An issue is considered to h ave a negligible effect on the likelihood of failure when a change in likelihood is so small o r the uncertainties in determining whether a change in likelihood has occurred are such that it ca nnot be reasonably concluded that the likelihood has actually changed (i.e., there is no cle ar trend toward decreasing the likelihood).

Question 3: Does the issue result in more than a minimal increa se in the consequences of a risk-significant accident sequence?

In answering this question, the first step is to identify the r isk significant sequences that have been evaluated that could be affected by the issue. The follow ing questions can assist in determining which accidents could have their radiological conse quences affected as a direct result of the issue:

• Will the issue change the effectiveness of an action?

• Will the issue play a direct role in mitigating the radiologic al consequences?

Next, a determination should be made as to whether the conseque nces would be more than minimally increased. In addressing the definition of what cons titutes a more than minimal increase in consequences, an increase of greater than 10 percen t in dose for risk-significant sequences is used as the criterion. An increase of less than 1 0 percent in calculated consequence is small enough that it cannot be reasonably conclu ded that the consequences have changed. Small changes in inputs and assumptions could ea sily have more of an effect than a calculated change of less than 10 percent in offsite dos e from a severe accident sequence.

SSCs, which indirectly affect dose, should also be considered, such as the following:

• containment bypass

• containment isolation and capacity

• hydrogen control

• long-term containment integrity

Question 4: Does the issue result in more than a minimal decre ase in the capability of a fission product barrier?

This question focuses on the fission product barriersfuel clad ding, reactor coolant system boundary and containment. The prior question also indirectly a ddresses containment. Each barrier is associated with specific design basis parameters suc h as fuel cladding temperature, reactor coolant system cool-down rate, and containment pressure. It is expected to be rare that

20 an issue will result in an impact on the design basis parameter s that can be directly calculated.

Rather, judgment is required here in ascertaining whether the d ecrease in capability of a fission product barrier is more than minimal.

Question 5: Does the issue result in more than a minimal decre ase in defense in depth capability or safety margin?

RG 1.174 (Reference 1) and if applicable, RG 1.177 (Reference 9 ), provide additional guidance.

4.3 Step 3 - Determining Safety Impact Using Quantitative Anal yses

A preliminary risk evaluation was completed before the IDP. In Step 3, the preliminary risk evaluation is revised to incorporate any new information and an alyses (e.g., focused scope analyses as needed) from the GAET or IDP in order to estimate t he final risk impact associated with the issue. Information from the final risk analysis shoul d be used to assist in answering the final screening questions in Step 2. The final risk analysis m ust identify whether the impacts documented in Step 2 were included in the risk analysis.

As discussed earlier, only those licensees with an acceptable P RA model can leverage their PRA models to perform quantitative risk assessments to support using this process, if all of the following conditions apply:

• The issue is completely within the scope of the licensees PRA model or can be bounded using surrogates. For this process, surrogates are limited to the use of basic events associated with the initiating event, mitigating system, or fun ction of the issue being investigated.

• The licensee has implemented an IDP consistent with risk-infor med initiative 10 CFR 50.69 or equivalent.

• The licensee has implemented risk-informed initiative TSTF-505 or TSTF-425 and has completed all implementation items and license conditions of th e safety evaluation. If the licensee is relying on its TSTF-425 program to demonstrate PRA technical acceptability, the licensee must justify that the issue being a nalyzed is limited to internal events by providing technical justification for the exclusion o f external hazards that are not addressed, or identify which additional NRC-approved applic ations address any relevant external hazards beyond internal events.

• The licensees PRA model was found acceptable to support appro val of relevant risk-informed applications by the NRC.

• The issue is within the scope of the portion(s) of the PRA mod el that was found acceptable by the NRC (e.g., if seismic was screened out of acc eptability, then seismic issues cannot be addressed using this process).

The plant-specific PRA must include the capability to assess CD F and LERF, and the risk evaluation must include a quantified assessment of all signific ant sources of risk (i.e., external events, internal flooding, and fires) that can be impacted by t he issue being assessed. Where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact (e.g., external events, low power and shutdown).

21 A risk analyst will use the licensees acceptable PRA model to calculate the change in CDF and LERF. The change in CDF and LERF will be calculated as the dif ference in plant risk with the proposed change and without the proposed change. For complianc e issues, the change in risk is the difference between risk if the plant were fully complian t with its licensing basis, and risk with the plant in the non-compliant configuration requested in the submittal. For licensee-identified issues that do not involve a compliance issue, the c hange in risk is the difference between risk with the plant in the current configuration and wi th the plant in the configuration requested in the submittal. The risk analysis may not include any credit for proposed RMAs or other activities implemented to reduce the risk impact associat ed with the issue.

The risk analyst must document any assumptions made when perfor ming the risk evaluation, any uncertainties associated with the analysis, whether any par ts of the issue were outside the scope of the licensees PRA, and whether any surrogates were us ed to account for the impact of the issue. The final quantitative risk analysis must includ e an evaluation of the impact on internal events risk, as well as the impact on any relevant ext ernal events. The risk analysis, including documentation of any influential assumptions and unce rtainties, shall be maintained for inspection by NRC personnel.

The PRA results will be compared to the relative change in risk of the licensees overall CDF and LERF. An issue is not risk-significant (i.e., minimal or l ess than minimal) if both of the following apply:

• the issue contributes less than 1 x 10 -7/year to CDF, and

• the issue contributes less than 1 x 10 -8/year to LERF.

If the risk results are less than the criteria above, the issue is considered to have a minimal impact on safety.

4.4 Step 4 - Assess Need for Risk Management Actions

Based on the assessment of the screening questions in Steps 1 a nd 2, and the outcome of the final quantitative risk evaluation in Step 3, a final safety im pact is determined. If the result of Step 1 indicates that there is no impact on safety, and the res ult of Step 3 indicates that there is minimal impact on safety, then the issue is characterized as ha ving a minimal impact on safety and RMAs do not need to be considered. If the results of Steps 2 and 3 both indicate that there is a minimal impact on safety, then the issue is characterized as having a minimal impact on safety and RMAs must be considered to offset the risk increase due to the issue. RMAs may not be used to alter the quantitative risk evaluation calculati on.

RMAs are typically associated with managing configuration risk when equipment is out of service or for temporary non-compliances. However, in this cas e, the proposed change will become the permanent plant configuration if the licensing actio n is approved. Therefore, only long-term actions to reduce risk associated with the new config uration need to be considered, such as permanent procedure changes or simple plant modificatio ns. For example, if an automatic interlock is defeated permanently, procedure changes to verify proper manual operation of the equipment may be appropriate to reduce the ris k associated with removal of the automatic interlock.

22 5.0 ASSESSING CUMULATIVE RISK

Once an issue has been characterized as having a minimal impact on safety, the cumulative risk impact of permanent changes to the risk profile of the pla nt must be evaluated consistent with the principles discussed in RG 1.174 and as applicable RG 1.177. As part of the evaluation of risk, licensees should understand the effects of the current application considering past applications. The PRA used for the current application should already model the effects of past applications. However, qualitative and synergistic effects are sometimes difficult to model.

Tracking changes in risk (both quantifiable and nonquantifiable ) that result from plant changes provides a mechanism to account for the cumulative and synergis tic effects of these plant changes and helps demonstrate that the licensee has a risk mana gement philosophy in which PRA is not just used to systematically increase risk, but is al so used to help reduce risk where appropriate and where it is shown to be cost effective.

Increases in CDF and LERF resulting from proposed licensing bas is changes should be limited to small increments. The decision process should track and con sider the cumulative effect of such changes, whether they result in an increase or a decrease in risk.

The cumulative risk impact is evaluated based on plant-specific CDF and LERF. Cumulative risk is acceptable for the purposes of this guidance if baselin e risk remains less than 1 x 10-4/year for CDF and less than 1 x 10 -5/year for LERF once the impact of the proposed change is incorporated into baseline risk.

6.0 PAPERWORK REDUCTION ACT

6.1 Paperwork Reduction Act

This document provides voluntary guidance for implementing the mandatory information collections in 10 CFR Part 50 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 number 3150-0011. Se nd comments regarding this information collection to the FOIA, Library, and Informati on Collections Branch, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0011), Office of Management and Budge t, Washington, DC 20503.

6.2 Public Protection Notification

The NRC may not conduct or sponsor, and a person is not require d to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number.

7.0 CONGRESSIONAL REVIEW ACT

These guidelines are a rule as defined in the Congressional Rev iew Act (5 U.S.C. 801-808).

However, the Office of Management and Budget has not found it t o be a major rule as defined in the Congressional Review Act.

23

8.0 REFERENCES

1. U.S. Nuclear Regulatory Commission, An Approach for Using P robabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018, Agenc ywide Documents Access and Management System (ADAMS) Accession No. ML17317A256.

2. U.S. Nuclear Regulatory Commission, NRR Temporary Staff Gui dance, Risk-Informed Process for Evaluations, TSG-2021-01, Revision 1, June 30, 202 1, ADAMS Accession No. ML21180A013.

3. Nieh, Ho K., memorandum to Craig G. Erlanger and Michael Fra novich, Approval of the Risk Informed Process for Evaluations, January 7, 2021, ADAMS A ccession No. ML21006A324.

4. Nuclear Energy Institute, Risk-Informed Technical Specifica tions Initiative 4b, Risk-Managed Technical Specificati ons (RMTS) Guidelines, NEI 0 6-09, Revision 0-A, November 2006, ADAMS Accession No. ML12286A322.

5. Nuclear Energy Institute, Risk-Informed Technical Specifica tions Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, NEI 04-10, Revision 1, April 2007, ADAMS Accession No. ML071360456.

6. Nuclear Energy Institute, Performance of PRA Peer Reviews u sing the ASME/ANS PRA Standard NEI 17-07, Revision A, December 2017, ADAMS Acces sion No. ML17341A548.

7. Nuclear Energy Institute, NEI Guidelines for the Implementa tion of the Risk-Informed Process for Evaluations Integrated Decision-Making Panel, Augu st 2020, ADAMS Accession No. ML20245E147.

8. Nuclear Energy Institute, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, NUMARC 93-01, Revision 4F, April 2018, ADAMS Accession No. ML18120A069.

9. U.S. Nuclear Regulatory Commission, Plant-Specific, Risk-In formed Decisionmaking:

Technical Specifications, Regulatory Guide 1.177, Revision 2, January 2021, ADAMS Accession No. ML20164A034.

Enclosure:

1. Appendix A: Change History

24 Appendix A - Change History

RIPE Guidance Document Change History - Page 1 of 1

Date Description of Changes Method Used to Training Announce &

Distribute

1/5/21 Initial issuance (ML20261H462) E-mail to NRR staff Recommended reading for DORL PMs and technical staff supporting license amendments and exemptions 6/30/21 Revised to include guidance for E-mail to NRR staff Recommended applying RIPE for licensees with an reading for DORL NRC-approved TSTF-425, PMs and technical Relocate Surveillance staff supporting Frequencies to Licensee Control-license RITSTF Initiative 5b, license amendments and amendment (ML21180A014) exemptions 5/10/22 Revised to remove limitation on E-mail to NRR staff Recommended applying RIPE to LARs involving reading for DORL changes to the Technical PMs and technical Specifications and added reference staff supporting to RG 1.177. license amendments and exemptions

25