ML22048B549

Draft Interim Staff Guidance - DANU-ISG-2022-07, Advanced Reactor Content of Application Project, Risk-Informed Inservice Inspection/Inservice Testing

ML22048B549
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Issue date:	05/31/2023
From:	Joseph Sebrosky NRC/NRR/DANU/UARP
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Joseph Sebrosky, 301-415-1132
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ML22048A520:ML22048A520	List: ML22048B549
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DANU-ISG-2022-07
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DANU-ISG-2022-07

Advanced Reactor Content of Application Project

Risk-Informed Inservice Inspection/Inservice Testing Programs for Non-LWRs

Draft Interim Staff Guidance

May 2023

ML22048B549 OFFICE QTE NRO/DRO/IRAB NRR/DANU/UTB1 BC NRR/DANU/UTB2 BC NAME Keith Azaria-Kribbs CCauffman MHayes SPhilpott DATE 3/25/22 4/6/22 4/6/22 4/20/22 OFFICE NRR/DEX/EMIB BC NRR/DANU/UARP PM OGC NRR/DANU/UARP BC(A)

NAME ITseng MOrenak RWeisman (NLO) SLynch DATE 4/21/22 4/17/23 5/12/23 4/26/23 OFFICE NRR/DANU NAME MShams DATE 5/12/23 DRAFT INTERIM STAFF GUIDANCE

ADVANCED REACTOR CONTENT OF APPLICATION PROJECT

RISK-INFORMED INSERVICE INSPECTION/INSERVICE TESTING PROGRAMS FOR NON-LWRS

DANU-ISG-2022-07

PURPOSE

The U.S. Nuclear Regulatory Commission (NRC or Commission) staf f is providing this interim staff guidance (ISG) for two reasons. First, this ISG provides guidance on the contents of applications to an applicant submitting a risk-informed, perfor mance-based application for a construction permit (CP) or operating license (OL) under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utiliz ation Facilities (Ref.1), or for a combined license (COL), a manufacturing licen se (ML), a standard design approval (SDA), or a design certification (DC) under 10 CFR Par t 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 2), for a non-lig ht-water reactor (non-LWR). The application guidance found in this ISG supports the development of the portion of a non-LWR application associated with an applicants risk-informed inserv ice inspection (ISI) and inservice testing (IST) programs.1 Second, this ISG provides guidance to NRC staff on how to revi ew such an application.

As of the date of this ISG, the NRC is developing a rule to ame nd 10 CFR Parts 50 and 52 (RIN 3150-Al66). The NRC staff notes this guidance may need to be up dated to conform to changes to 10 CFR Parts 50 and 52, if any, adopted through that rulemak ing. Further, as of the date of this ISG, the NRC is developing an optional performance-based, technology-inclusive regulatory framework for licensing nuclear power plants designated as 10 C FR Part 53, Licensing and Regulation of Advanced Nuclear Reactors, (RIN 3150-AK31). Afte r promulgation of those regulations, the NRC staff anticipates that this guidance will be updated and incorporated into the NRCs Regulatory Guide (RG) series or a NUREG series docume nt to address content of application considerations specific to the licensing processes in this document.

BACKGROUND

This ISG is based on the advanced reactor content of applicatio n project (ARCAP), whose purpose is to develop technology-inclusive, risk-informed, and performance-based application guidance for non-LWRs. The ARCAP is broader than, and encompasses, the industry-led technology-inclusive content of application project (TICAP). Th e guidance in this ISG supplements the guidance found in Division of Advanced Reactors and Non-power Production and Utilization Facilities (DANU)-ISG-2022-01, Review of Risk-Informed, Technology-Inclusive

1 The NRC is issuing this ISG to describe methods that are acceptable to the NRC staff for implementing specific parts of the agencys regulations, to explain techniques that the NRC staff uses in evaluating specific issues or postulated events, and to describe information that the NRC staff needs in its review of applications for permits and licenses. The guidance in this ISG that pertains to applicants is not part of NRC regulations and compliance with it is not required. Methods and solutions that differ from those set forth in this ISG are acceptable if supported by a basis for the issuance or continuance of a permit or license by the Commission.

DANU-ISG-2022-07 Page 2 of 18

Advanced Reactor Applications - Roadmap, issued in May 2023 (R ef. 3), which provides a roadmap for developing all portions of an application for non-L WRs. The guidance in this ISG is limited to the portion of a non-LWR application associated with the risk-informed ISI and IST programs for the nuclear reacto r plant applicant and the NRC staff review of that portion of the application.

Following approval of the 10 CFR Part 53 final rule, this ISG g uidance will be supplemented, as necessary, to provide guidance for developing risk-informed ISI and IST programs to reflect any differences between current requirements in 10 CFR Parts 50 and 52 and new requirements in 10 CFR Part 53. The 10 CFR Part 53 rulemaking would revise the NRC's regulations by adding a risk-informed, performance-based, technology-inclusive regula tory framework for commercial nuclear reactors in response to the related requirements of the Nuclear Energy Innovation and Modernization Act (NEIMA; Public Law 115-439), as amended by th e Energy Act of 2020. Key documents related to the 10 CFR Part 53 rulemaking, including p reliminary and draft proposed rule language and stakeholder comments, can be found at Regulat ions.gov under Docket ID NRC-2019-0062.

RATIONALE

The current application guidance related to risk-informed ISI a nd IST programs is directly applicable only to light water reactors (LWRs) and may not full y identify the information to be included in a non-LWR application or efficiently provide a tech nology-inclusive, risk-informed, and performance-based review approach for non-LWR technologies. This ISG serves as the non-LWR application guidance for risk-informed ISI and IST prog rams. This ISG provides both applicant content of application and NRC staff review guidance.

APPLICABILITY

This ISG is applicable to applicants for non-LWR 2 permits and licenses that submit risk-informed, performance-based applications for CPs or OLs un der 10 CFR Part 50 or for COLs, SDAs, DCs, or MLs under 10 CFR Part 52. This ISG is also applicable to the NRC staff reviewers of these applications.

PAPERWORK REDUCTION ACT

This ISG provides voluntary guidance for implementing the manda tory information collections in 10 CFR Parts 50 and 52 that are subject to the Paperwork Reduct ion Act of 1995 (44 U.S.C.

3501 et. seq.). These information collections were approved by the Office of Management and Budget (OMB), approval numbers 3150-0011 and 3150-0151. Send co mments regarding this information collection to the FOIA, Library, and Information Co llections Branch (T6-A10M),

U.S. Nuclear Regulatory Commission, Washington, DC 20555 0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-0011 and 3150-0151), Attn: Desk Office r for the Nuclear Regulatory Commission, 725 17th Street, NW Washington, DC 20503; e-mail:

oira_submission@omb.eop.gov.

2 Applicants desiring to use this ISG for a light water reactor application should contact the NRC staff to hold pre-application discussions on their proposed approach.

DANU-ISG-2022-07 Page 3 of 18

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.

GUIDANCE

This ISG describes the methods acceptable to the NRC staff for preparing and reviewing descriptions of risk-informed ISI and IST programs submitted by non-LWR applicants as part of licensing applications. Currently, the requirements for ISI and IST programs are described in 10 CFR 50.55a. These requirements apply only to LWRs and are ba sed upon requirements developed by the American Society of Mechanical Engineers (ASME ). With the increased use of probabilistic risk information in the design and regulation of nuclear power plants, the staff anticipates that new applications for nuclear power plants will include risk-informed ISI and IST programs. (The NRC is currently developing guidance for alterna tive approaches that use risk information to focus traditional ISI and IST programs on the mo st important structures, systems, and components (SSCs) and adjust their inspection and testing f requencies accordingly.) ASME is also considering development o f a new Code for Operations an d Maintenance of Nuclear Power Plants (OM) Code (referred to as OM-2) that would provide IST provisions for fluid flow and control devices in non-LWR reactors. The new ASME OM-2 Code may be available when non-LWR applicants are preparing to develop their plant-specifi c IST programs. For this ISG, the NRC staff is assuming that applicants will be designing and qualifying their equipment to the latest applicable guidance and r equirements developed by ASME and accepted by the NRC, such as ASME Boiler and Pressure Vessel Code,Section III, Rul es for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors ( Ref. 4), which is endorsed by Regulatory Guide (RG) 1.87, Revision 2, Acceptability of ASME Code Section III, Division 5,

'High Temperature Reactors, (Ref. 5), and the QME-1 standard, Qualification of Active Mechanical Equipment Used in Nuclear Facilities. (Ref. 6), whi ch is endorsed by Regulatory Guide (RG) 1.100, Seismic Qualification of Electric and Mechan ical Equipment for Nuclear Power Plants (Ref. 7).

The purpose of a risk-informed ISI program is to periodically m onitor and track degradation (defects, corrosion, erosion) in welds and base metal of compon ents and component supports within the programs scope to determine their suitability for c ontinued operation, consistent with the plant-specific probabilistic risk assessment (PRA). The cur rent ISI programs for LWRs include inspections of ASME Boiler and Pressure Vessel Code (BP V Code) Class 1, 2, and 3 piping; safety-related pressure-retaining components; and compo nent supports. Non-safety-related but safety signi ficant components are typically inspected as part of reliability assurance or maintenance programs.

The purpose of a risk-informed IST program is to periodically m easure, assess, and track the performance of components within the programs scope to confirm that their performance remains consistent with the plant-specific PRA. The current IST programs for water-cooled reactors include components cons isting of pumps, valves, and dynamic restraints (snubbers) that perform safety functions. As discussed below, some non-LWR s might rely on types of components other than pumps, valves, and dynamic restraints (snubbers), such as fluid control devices.

DANU-ISG-2022-07 Page 4 of 18

Components that Control Fluid without Mechanically Interacting with the Fluid

In addition to conventional components, non-LWRs might include components that perform active safety functions to control fluid flow without mechanica lly interacting with the controlled fluid. In the context of this guidance, such a component perfor ms an active safety function insofar as it performs a function to move fluid, stop the movem ent of fluid, or transfer energy as part of a safety function. Examples of such components include electromagnetic or magnetic flux pumps, which move fluid without mechanical interaction wit h the fluid, and heat pipes, which transfer energy from one location to another in performing a sa fety function to cool a reactor core. The scope of the risk-informed ISI and IST programs for a non-LWR might also include activities to assess degradation of the components with active safety functions to control fluid flow without mechanically contacti ng the controlled fluid by such means included for example condition monitoring, surveillance, testing, or inspection.

Currently operating nuclear power plants do not include compone nts that perform active safety functions to control fluid flow without mechanically interactin g with the controlled fluid.

Therefore, ISI and/or IST programs at currently operating nucle ar power plants do not address the degradation that could adversely affect the ability of such components to perform their safety functions. Non-LWR applicants will need to develop and j ustify periodic condition monitoring, surveillance, testing, or inspection plans for such components.

ISI/IST Personnel Hazards for Some Non-LWR Designs

Some non-LWR designs may have unique characteristics that resul t in additional hazards to plant personnel who perform ISI and IST activities; therefore, the performance of ISI and IST activities is an important consideration for those designs. One example of a technology with such additional hazards is the liquid-fueled molten salt reacto r technology. Due to the high radiation contained throughout reactor coolant system piping, a n extreme thermal environment, and maintenance activities that may involve draining and flushi ng the fuel salt, some level of remote ISI/IST capability may be necessary to mitigate the haza rds to ISI/IST personnel.

Use of Risk Information

The use of risk information in formulating an ISI/IST program h elps focus the program on the most risk-significant SSCs, conditions, and failure modes. It a lso helps ensure that the inspection and testing frequency is sufficient to detect reliab ility and performance degradation that could affect continued safe operation.

Application Guidance

The development of an acceptable risk-informed ISI/IST program depends on having a high-quality, plant-specific PRA. Accordingly, the applicant sh ould use a plant-specific PRA in the risk-informing process and develop the PRA using an NRC-end orsed PRA consensus standard. For non-LWRs, ASME/ANS RA-S-1.4-2021, Probabilisti c Risk Assessment Standard for Advanced Non-Light Water Reactor Nuclear Power Pla nts (Ref. 8), is available, as endorsed with exceptions and clarifications in RG 1.247, Accep tability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor Risk-Informed Ac tivities (for trial use), issued March 2022 (Ref. 9).

Applications for an OL, COL, DC, SDA, or ML may describe other programs in addition to risk-informed ISI/IST programs (e.g., programs for maintenance, reliability assurance, and aging DANU-ISG-2022-07 Page 5 of 18

management). If the ISI or IST program is being used to satisfy any of these other operational requirements, the applicant should state this and explain how t he ISI or IST activities will satisfy such requirements. This may be done either in the ISI/IST progr am or in the application.

A nuclear power plant applicant may request implementation of 1 0 CFR 50.69, Risk-informed categorization and treatment of structures, systems, and compon ents for nuclear power plants, for risk-informed treatment of SSCs as an alternative to certai n special treatment requirements (STRs) in the NRC regulations. The regulations in 10 CFR 50.69 define four risk-informed safety classes (RISCs) of SSCs: (1) RISC-1 SSCs are safety-related SS Cs that perform safety-significant functions, (2) RISC-2 SSCs are nonsafety-rel ated SSCs that perform safety-significant functions, (3) RISC-3 SSCs are safety-relate d SSCs that perform functions of low safety significance, and (4) RISC-4 SSCs are nonsafety-rela ted SSCs that perform functions of low safety significance. The regulations in 10 CFR 50.69 indicate that, if approved, an applicant or licensee may voluntarily comply with the requir ements in 10 CFR 50.69 as an alternative to specific special treatment requirements for RISC -3 and RISC-4 SSCs. The special treatment requirements that the 10 CFR 50.69 requirements may r eplace include certain ISI/IST requirements in 10 CFR 50.55a. While the requirements for speci al treatment in the regulation in 10 CFR 50.69 were developed based on LWR technology, the sco pe of 10 CFR 50.69 includes all applicants for a CP, an OL, a COL, an SDA, or an M L. For non-LWR applicants that propose to use 10 CFR 50.69 to risk-inform their ISI/IST progra ms, justification must be provided showing how the resulting RISC-3 and 4 SSCs were deriv ed from the PRA.3

Limited performance data may be available for SSCs of new desig ns or technologies; therefore, an OL or COL applicant will need to describe the ISI/IST measur es and their bases for those components in the application.

The application does not need to describe the risk-informed ISI and IST programs in as much detail as the program plans. However, it should describe the me thods for using the PRA to determine the reliability targets and performance assumptions f or the individual components.

The application should also describe the methodology to be used to verify, based on the proposed inspections and inspection frequencies, whether the re liability targets and performance assumptions are met.

Staff Review Guidance

Before beginning a detailed review of an application, the revie wer should confirm that the applicant has used a plant-specific PRA in the risk-informing p rocess and that the applicant developed the PRA using an NRC-endorsed PRA consensus standard.

The reviewer should determine which other programs (e.g., progr ams for maintenance, reliability assurance, and aging management) are being coordina ted with the risk-informed ISI/IST programs and whether any parts of the risk-informed ISI/IST programs are being

3 Non-LWR applicants may use the Licensing Modernization Project (LMP) found in NEI 18-04, Revision 1, "Risk-Informed Performance-Based Guidance for Non Light Water Reactor Licensing Basis Development" (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19241A336), as endorsed by RG 1.233, "Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light Water Reactors" (ML200091L698). Use of the LMP may obviate the need for a non-LWR applicant to implement 10 CFR 50.69 to risk-inform the categorization of SSCs because the LMP includes a risk-informed process for SSC classification.

DANU-ISG-2022-07 Page 6 of 18

incorporated into another program. If so, the reviewer should c onfirm that all of the programs, taken together, cover the proposed scope of the risk-informed I SI/IST programs.

If the application includes a request to implement 10 CFR 50.69, the reviewer should consult the NRCs Statements of Consideration (SOC) on the 10 CFR 50.69 fin al rule in Volume 69 of the Federal Register (FR), page 68008 (69 FR 68008; November 22, 2004), and the gui dance and training materials prepared by the NRC staff for the review and evaluation of 10 CFR 50.69 programs. However, the SOC states that the 10 CFR 50.69 rule do es not include 10 CFR 50.36 in the list of special treatment requirements that may be repla ced by the alternative 10 CFR 50.69 requirements for RISC-3 and RISC-4 SSCs when imple menting a 10 CFR 50.69 license amendment.

The reviewer should consider the topics discussed in RIS 2012-0 8 when evaluating risk-informed ISI/IST programs for a non-LWR application. Note: the reviewer should consider component snubbers with other components described in RIS 201 2-08.

Organization of this ISG

The guidance in this ISG is divided into three parts: one for I SI application and review guidance, one for IST application and review guidance, and one for organi zational responsibilities and review guidance. This ISG provides guidance for ISI/IST program s that are entirely risk-informed. Partially risk-in formed ISI or IST programs will need to be justified and will be reviewed on a case-by-case basis.

This ISG applies to risk-informed applications for OLs, COLs, D Cs, SDAs, and MLs. An application for a CP may contain less detail than one for an OL ; however, as a minimum, it should identify the associated regulations, RGs, NUREGs, standa rds and other guidance the applicant intends to follow at the OL stage.

PART 1Risk-Informed Inservice Inspection for Non-Light Water R eactors

Application Guidance

The regulations in 10 CFR 50.55a do not contain requirements fo r non-LWR ISI programs.

However, the following general design criteria (GDC) in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, are generally applic able to and provide guidance in establishing the principal design criteria for types of reactor s other than water-cooled reactors governed by the GDC. The guidance in the below GDC indicates th at SSCs should be designed to permit inservice inspection in various areas:

• GDC 32, Inspection of reactor coolant pressure boundary

• GDC 36, Inspection of emergency core cooling system

• GDC 39, Inspection of containment heat removal system

• GDC 42, Inspection of containment atmosphere cleanup system

• GDC 45, Inspection of cooling water system

• GDC 53, Provisions for containment testing and inspection

In addition, as described in RG 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors, (Ref. 10), the NRC has developed a s et of non-LWR design criteria (ARDC), based on the GDC, for application to and insights for t he development of principal DANU-ISG-2022-07 Page 7 of 18

design criteria for non-LWRs. The guidance in the following ARD C also indicate that SSCs should be designed to permit in-service inspection in various a reas:

• ARDC 32, Inspection of reactor coolant boundary

• ARDC 36, Inspection of emergency core cooling system

• ARDC 39, Inspection of containment heat removal system

• ARDC 42, Inspection of containment atmosphere cleanup system

• ARDC 45, Inspection of structural and equipment cooling systems

• ARDC 53, Provisions for containment testing and inspection

These ARDC call for consideration of ISI activities in the desi gn of non-LWR SSCs. The areas covered in the above GDC and ARDC correspond to the basic safet y functions described in the ARCAP documents related to control of heat removal and release of radioactive material.

Applications for an OL or a COL for a non-LWR must include prov isions for ISI, among other things, in accordance with 10 CFR 50.34(b)(6)(iv) or 10 CFR 52. 79(a)(29)(i), respectively.

Applications for a CP may contain less detail than one for an O L, however, as a minimum, it should identify the associated regulations, RGs, NUREGs, standa rds and other guidance the applicant intends to follow.

The scope of a risk-informed ISI program includes all piping, p ressure-retaining components, and component supports that perform safety-significant function s, as well as piping or other components whose failure could prevent SSCs from performing the ir safety functions.

Therefore, the application should describe the scope of the pro posed risk-informed ISI program.

The scope of the program should include all safety-related and safety-significant piping and components (including supports and snubbers), consistent with t he results of the plant-specific PRA.

As discussed earlier, non-LWRs may contain components that perf orm active safety functions to control fluid flow without mechanically interacting with the co ntrolled fluid. The applicant should describe the condition monitoring, surveillance, and inspection activities for such safety-significant components; provide the basis for the inspec tion activities, including inspection intervals; and ensure that they are part of the ISI program.

The scope of a risk-informed ISI program for non-LWRs needs to be based on a plant-specific PRA. The piping should be modeled in segments to better allow f or identification of the most risk-significant locations and welds. Accordingly, the applicat ion should describe how the PRA models the SSCs that are part of the ISI program. Specifically, the application should describe how risk information is used to guide (1) the selection of the inspection locations during each inspection interval, (2) the inspection frequency for each loca tion, (3) the inspection technique to be used, and (4) how the selection process varies from one insp ection interval to the next to cover all components of interest. Although intended for LWRs, t he ASME BPV Code,Section XI, Division 1, might provide useful information on ins pection techniques and frequencies within the conditions for which the ASME Code speci fies their use.

In addition, the application should describe the process to be followed when the ISI program identifies that degradation has occurred. This process should i nclude tracking of the degradation over time. If necessary, it should also include actions such as expanding the inspections to other similar components or locations, reducing the time interv al to the next inspection, or taking corrective action. The application should include the criteria for deciding what additional actions to take to allow continued operation consistent with the licens ing basis.

DANU-ISG-2022-07 Page 8 of 18

In 2019, ASME issued BPV Code,Section XI, Division 2, Require ments for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants ( Ref. 11). In RG 1.246, Acceptability of ASME Code,Section XI, Division 2, Requireme nts for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants, for Non-Li ght Water Reactors, (Ref. 12), the NRC endorsed (with conditions) the use of ASME B PV Code,Section XI, Division 2, for non-LWR ISI.

ASME BPV Code,Section XI, Division 2, does not call for a spec ific risk-informed ISI program to be implemented but rather allows the applicant to propose a pro gram specific to the design and technology of the non-LWR, based on input from expert panels an d considering the degradation mechanisms relevant to the materials and the operating conditio ns of the design. In following ASME BPV Code,Section XI, Division 2, the application needs to describe how component reliability targets are derived from the PRA, how the component s in the program and the corresponding inspection intervals are selected, and how the ap plicants reliability and integrity management strategies will be able to demonstrate that the reli ability targets are met.

Applicants may refer to ASME BPV Code,Section XI, Division 1, for information on inspection methods and frequencies within the conditions for which the Div ision 1 specifies their use, although each applicant is ultimately responsible for proposing a risk-informed ISI program appropriate for its own design and technology.

ASME BPV Code,Section XI, Division 2, also contains acceptance criteria for the inspections; however, these apply only in the temperature range allowed in A SME BPV Code,Section III, Rules for Construction of Nuclear Facility Components, Divisi on 1 (Ref. 13). Appropriate justification for flaw evaluation acceptance criteria for any c omponents that will be used in applications in which the temperature exceeds the temperature l imits specified in the ASME BPV Code,Section III, Division 1 should be provided as part of the information to be included in an application.

Staff Review Guidance

The scope of a risk-informed ISI program should include all pip ing, pressure-retaining components, and component supports that perform safety-related and safety-significant functions, as well as piping or other components whose failure could prevent SSCs from performing their safety functions. Therefore, the reviewer shou ld confirm that the scope of the applicants proposed risk-informed ISI program includes all saf ety-related and safety-significant piping and components (including supports and snubbers).

The reviewer should confirm that the PRA models all of the SSCs that are part of the ISI program and models the piping in segments to identify the most risk significant piping sections and welds. In addition, the reviewer should evaluate how risk i nformation is used to guide (1) the selection of the inspection locations during each inspectio n interval, (2) the inspection frequency for each location, (3) the inspection technique to be used, and (4) how the selection process varies from one inspection interval to the next to cove r all components of interest.

The reviewer should confirm that the application clearly descri bes the applicable regulations, codes and standards and other guidance documents used in the de velopment of the ISI program and provides justificati on for any alternatives or exemptions.

DANU-ISG-2022-07 Page 9 of 18

In reviewing a non-LWR application that describes a risk-inform ed ISI program, the reviewer should consider the following:

(1) Is the application based on the use of an NRC-endorsed PRA standard?

(2) Does the application describe the degradation mechanisms (e.g., corrosion, high temperature creep, thermal cycling) applicable to the design an d technology (e.g.,

materials, coolant, service conditions) and list the SSCs to wh ich they apply?

(3) Does the application describe how risk information was used to determine and justify (a) the components included in the ISI program, (b) the inspections to be conducted for each component in the program, and (c) the inspection frequency for each component in the program, including how components will be selected at each inspection interval so that, over time, all components will receive testing?

(4) Are the proposed inspection techniques capable of detecting the degradation types of interest, and does the application provide the basis for that d etermination (e.g., previous experience)?

(5) Does the application identify a multi-disciplinary process for reviewing inspection results and determining what, if any, corrective action should be taken as a result of any degradation identified? The reviewer should verify that the des cription of the process includes the criteria for deciding what additional actions to t ake to allow continued operation consistent with the licensing basis.

(6) Does the application describe and justify acceptance criter ia for evaluating inspection results? Are the acceptance criteria consistent with the values used in the PRA for the frequency of leaks, the size of leaks, the location of leaks, a llowable corrosion and erosion, and the structural integrity of components and support s? Do the acceptance criteria limit any degradation to within the uncertainty of the values used in the PRA for the parameters that determine component integrity and reliabili ty?

(7) Are the frequencies proposed for each inspection, as well a s their bases, consistent with the PRA results on SSC reliability? Inspections should be frequ ent enough to ensure that SSC integrity and reliability remain within the uncertaint ies in values of the parameters used in the PRA.

(8) Does the application describe a program for retention of IS I records?

(9) Is the QA to be applied to the program in accordance with 1 0 CFR Part 50, Appendix B, or is an exemption from these requirements justified?

(10) Has ASME BPV Code,Section XI, Division 2, along with RG 1.246, been used in developing the provisions for the ISI?

The reviewer should confirm that, to establish a baseline for c omparison to future inspection results, the preservice inspection program described in the app lication will include condition monitoring, surveillance, and inspections using the same techni ques and equipment proposed for use in the risk-informed ISI program. The reviewer should a lso determine how the baseline inspection results will be used to determine what constitutes u nacceptable degradation (i.e., how are acceptance criteria defined). In addition, the r eviewer should confirm that the DANU-ISG-2022-07 Page 10 of 18

design provides space for accessibility of equipment, shielding, and personnel to conduct the inspections.

The reviewer should determine how the applicant will monitor th e effectiveness of the risk-informed ISI program. Monitoring could include looking at trends in degradation detected to determine whether to change inspection frequencies or technique s.

The reviewer should verify that the applicant has described the ISI activities applicable to components that perform active safety functions to control flui d flow without mechanically interacting with the controlled fluid, and that these ISI activ ities are capable of assessing (through direct measurement or analysis) the degradation of suc h components.

The review should also verify whether risk information was used to determine the plant conditions under which inspections are best performed (e.g., fu ll power or shutdown). The goal is to conduct inspections under the plant conditions that, give n the inspection techniques and related constraints, provide the necessary performance informat ion while minimizing risks to workers and the general public. The reviewer should determine w hether the application has addressed this aspect of the program or has justified not doing so.

Based on the above, the reviewer should be able to determine wh ether the risk-informed ISI program will provide data sufficient to detect degradation affe cting each subject components ability to perform its safety functions consistent with the PRA and whether the application complies with the applicable requirements for a CP, OL, COL, DC, SDA or ML.

Part 2 - Risk-Informed Inservice Testing for Non-Light-Water Re actors

Application Guidance

The regulations in 10 CFR 50.55a do not contain requirements fo r non-LWR IST programs. The ASME OM Code, Division 1, as incorporated by reference in 10 CF R 50.55a, applies to water-cooled reactors. ASME has created a task group to draft a new division of the ASME OM Code that will provide high-level requirements for IST activiti es for components in non-LWRs.

However, the following GDC in 10 CFR Part 50, Appendix A, are g enerally applicable to and provide guidance in establishing the principal design criteria for types of reactors other than water-cooled reactors governed by the GDC. The guidance in the below GDC indicates that SSCs should be designed to permit in-service testing in various areas:

• GDC 1, Quality standards and records

• GDC 4, Environmental and dynamic effects design bases

• GDC 30, Quality of reactor coolant pressure boundary

• GDC 37, Testing of emergency core cooling system

• GDC 40, Testing of containment heat removal system

• GDC 43, Testing of containment atmosphere cleanup system

• GDC 46, Testing of cooling water system

• GDC 53, Provisions for containment testing and inspection

• GDC 61, Fuel storage and handling and radioactivity control

DANU-ISG-2022-07 Page 11 of 18

In addition, as described in RG 1.232, the NRC has developed a set of ARDC, based on the GDC, for non-LWRs. The guidance in the following ARDC also indi cate that SSCs should be designed to permit in-service testing in various areas:

• ARDC 1, Quality standards and records

• ARDC 4, Environmental and dynamic effects design bases

• ARDC 30, Quality of reactor coolant boundary

• ARDC 37, Testing of emergency core cooling system

• ARDC 40, Testing of containment heat removal system

• ARDC 43, Testing of containment atmosphere cleanup system

• ARDC 46, Testing of structural and equipment cooling systems

• ARDC 53, Provisions for containment testing and inspection

• ARDC 61, Fuel storage and handling and radioactivity control

These ARDC call for consideration of IST activities in the desi gn of non-LWR SSCs. The areas covered in the above GDC and ARDC correspond to the basic safet y functions covered in the ARCAP documents related to control of heat generation, control of heat removal, and release of radioactive material. Applications for an OL or a COL for a non -LWR must include provisions for IST, among other things, in accordance with 10 CFR 50.34(b)(6)( iv) or 10 CFR 52.79(a)(29)(i),

respectively. An application for a CP may contain less detail t han one for an OL, however, as a minimum, it should identify the associated regulations, RGs, NU REGs, standards, and other guidance the applicant intends to follow at the OL stage.

The scope of a risk-informed IST program includes safety-relate d and safety-significant components that perform an active safety function; these may in clude the following:

• motor-operated valves

• air-operated valves

• hydraulic-operated valves

• solenoid-operated valves

• check valves

• manually operated valves

• explosively actuated valves

• rupture disks

• safety and relief valves

• pumps (motor-and turbine-operated)

• dynamic restraints (snubbers)

Non-LWR reactor applicants should identify safety-related and s afety-significant components that perform functions similar to those of the components liste d above but have different names (such as fluid-moving or fluid isolation components) as within the scope of the IST program.

In addition, non-LWR reactor designs may rely on new types of s afety features to accomplish active safety functions. The applicant will need to incorporate IST activities for components that perform active safety functions to control fluid flow without m echanically interacting with the controlled fluid (such as electromagnetic pumps and heat pipes) by expanding the scope of the IST program. The application should describe the IST activities applicable to such components and how these IST activities are capable of assessing (through direct measurement or analysis) the performance of such components.

DANU-ISG-2022-07 Page 12 of 18

The scope of the risk-informed IST program needs to be based on a plant-specific PRA. Based on component function and importance to safety, the application should describe how risk information is used to guide (a) the selection of components fo r IST activities, (b) the specific IST activity to be performed for each component, (c) the IST fr equency for each component, and (d) how the selection process varies from one IST interval to the next to cover all components of interest. In addition, the application needs to d escribe how component reliability targets and assumptions on component performance are derived fr om the PRA. Although applicable to water-cooled reactors, the ASME OM Code as incorp orated by reference in 10 CFR 50.55a may provide helpful information on component test ing for other fluid media within the conditions under which it specifies testing. ASME is currently considering the development of OM-2 for non-LWRs.

The application should explain why the proposed risk-informed I ST program is adequate to assess the operational readiness of the components within the s cope of the program. The program may include various activities, such as valve actuation (opening and closing times),

relief valve actuation (opening and closing pressure), pump act uation (start time, flow rate, pressure, speed, differential pressure, discharge pressure, and vibration), check valve (opening and closing, and leakage), and dynamic restraint (snubber) oper ation (examination, functional testing, and service life monitoring), as applicable to the non -LWR reactor design. It may also include testing of components other than those associated with LWRs. To simulate actual operating conditions, the application should propose IST condit ions (e.g., pressure and temperature) that are as realistic as practical. The applicatio n should describe practical IST techniques (e.g., bench testing of relief valves) if in-place o r at-power IST activities are not feasible.

For a non-LWR reactor design with components that perform activ e safety functions to control fluid flow without mechanically interacting with the controlled fluid, the risk-informed IST program needs to include IST activities capable of assessing th e operational readiness of those components to perform their active safety functions. For a comp onent in standby, the application should describe how the IST program provides for te sting the component at the conditions (e.g., pressure, fluid level, and temperature) neces sary to activate the feature and ensuring that it performs its safety function when so tested. F or a component in operation, the application should describe how the IST program verifies that t he performance of the component (e.g., heat transfer and reactivity insertion) aligns with predicted performance. This may involve measuring inlet and outlet temperatures, flow rates, changes in power level, or other appropriate parameters and analyzing these to determine o verall performance. For valves that maintain their obturator positions and do not need to chan ge their state to accomplish their safety functions (e.g., isolation valves that remain closed dur ing plant operation), the application should describe how the IST program verifies that seat leakage and position indication (remote and local) are consistent with the plant safety analysis.

The application should also describe the process to be followed when the IST program identifies that degradation or misalignment has occurred. This process sho uld include tracking of the degradation over time. It should also include actions such as e xpanding the IST activities to other similar components, reducing the time interval between IS T activities, or taking corrective action to improve the components performance if the testing re sults warrant such action.

DANU-ISG-2022-07 Page 13 of 18

Staff Review Guidance

The scope of a risk-informed IST program should include all saf ety-related and safety significant SSCs that perform active safety functions. Therefore, the revie wer should confirm that the scope of the proposed risk-informed IST program includes all sa fety-related and safety significant SSCs identified by the PRA that perform an active safety function.

The reviewer should also confirm that the PRA models all of the SSCs that are part of the IST program. In addition, the reviewer should evaluate how the risk information from the PRA and a risk-informed decisionmaking process was used to guide (1) the selection of the SSCs included in the program, (2) the testing frequency for each SSC included in the program, (3) the testing technique to be used for each SSC, and (4) how the selection of the SSCs for testing varies from one testing interval to the next to cover all SSCs of inte rest.

The reviewer should confirm that the application clearly descri bes the applicable regulations, codes and standards, and other guidance documents used in devel oping the IST program and provides justification for any a lternatives or exemptions.

In reviewing a non-LWR application that describes a risk-inform ed IST program, the reviewer should also consider the following:

(1) Is the application based on the use of an NRC-endorsed PRA standard?

(2) Does the application describe the important types of degrad ation for the design and technology being reviewed?

(3) Does the application define the IST conditions and testing techniques for each component included in the risk-informed IST program? Are the pr oposed IST testing techniques capable of detecting (directly or through analysis) the relevant degradation in performance for each component?

(4) Does the application describe how risk information was used to determine and justify (a) the components included in the IST program, (b) the IST activit ies to be conducted for each component in the program, and (c) the frequency of IST act ivities for each component, including how components will be selected for testin g at each testing interval so that, over time, all components will receive testin g? For components of a new or unique design, sampling may not be sufficient to confirm the ir performance and additional testing may be warranted.

(5) Does the application describe a multidisciplinary process f or reviewing IST results and determining what, if any, corrective action is needed. The rev iewer should verify that the description of the process includes the criteria for deciding w hat additional actions to take to allow continued operation consistent with the licensing basis.

(6) Does the application describe and justify acceptance criter ia for evaluating IST results for each component? Are these acceptance criteria consistent wi th the values used in the PRA for parameters that determine component reliability and performance? Are the acceptance criteria based on the performance uncertainties in t he PRA and do they limit any degradation to within these levels of uncertainty?

DANU-ISG-2022-07 Page 14 of 18

(7) Are the frequencies of the IST activities for each componen t within the scope of the risk-informed IST program consistent with maintaining the component s reliability and performance within the PRA results? The IST intervals should pr ovide assurance that SSC reliability and performance remain within the uncertainties in the PRA.

(8) Does the application describe a program for retention of IS T records?

(9) Is the QA to be applied to the program in accordance with 1 0 CFR Part 50, Appendix B or is an exemption to these requirements justified?

(10) Has ASME developed provisions for IST activities in non-LW Rs? The reviewer should determine the status of the ASME effort for guidance in the rev iew of risk-informed IST programs in non-LWR applications.

The reviewer should confirm that, to establish a baseline for c omparison to future IST results, the preservice testing program described in the application wil l use the same techniques and equipment proposed for use in the risk-informed IST program. Th e reviewer should determine how the baseline results will be used to determine what constit utes unacceptable degradation in performance (i.e., acceptance criteria). The reviewer also shou ld confirm that the design provides space for accessibility of equipment, shielding, and p ersonnel to conduct the testing.

The reviewer should determine how the applicant will monitor th e long-term effectiveness of the risk-informed IST program. Monitoring could include looking at trends in degradation detected to determine whether to change the IST intervals or techniques.

The reviewer should verify that the applicant has described the IST activities applicable to components that perform active safety functions to control flui d flow without mechanically interacting with the controlled fluid, and that these IST activ ities are capable of assessing (through direct measurement or analysis) the performance of suc h components.

The reviewer should also verify whether risk information was us ed to determine the plant conditions under which IST activities are best performed (e.g., full power or shutdown). The goal is to conduct IST activities under the plant conditions that, g iven the IST techniques and related constraints, provide the necessary performance information whil e minimizing risks to workers and the general public. The reviewer should determine whether t he application has addressed this aspect of the program or has justified not doing so.

Based on the above, the reviewer should be able to determine wh ether the risk-informed IST program will provide data sufficient to detect degradation affe cting each subject components ability to perform its safety functions consistent with the PRA and whether the application complies with the applicable requirements for a CP, OL, COL, DC, SDA or ML.

DANU-ISG-2022-07 Page 15 of 18

PART 3Organizational Responsibilities

Application Guidance

A nuclear power plants organizational responsibility for risk-informed ISI/IST programs are broadly the same regardless of whether the program is for ISI o r IST activities. The application should describe an organizational structure that satisfies the NRC regulations for conducting risk-informed ISI/IST programs or should identify and justify a lternatives or exemptions consistent with the processes specified in the NRC regulations. In general, the nuclear power plant organization is responsible for all aspects of the progra ms, although other parties (e.g., contractors) may conduct some of the inspections or test ing under appropriate supervision. The application should describe organizational res ponsibilities for risk-informed ISI/IST programs include the following:

• defining the qualifications of the personnel managing, conduct ing, and reviewing the program results, consistent with the codes and standards being used

• providing training, as necessary, to ensure that personnel are qualified to perform their functions

• developing the schedule, sequence, prerequisites, procedures, safety precautions, and acceptance criteria for conducting the programs

• managing the programs, including coordination with other eleme nts of the plant organization (e.g., operations, engineering) and other operatio nal programs (e.g., the reliability assurance program)

• ensuring that QA is in accordance with 10 CFR Part 50, Appendi x B, or that an exemption to these requirements is justified

• providing a multidisciplinary review team to evaluate and disp osition inspection and testing results, initiate corrective action as necessary, and e valuate any safety implications for continued plant operation

• preparing, approving, and retaining ISI/IST reports

• monitoring the long-term effectiveness of the risk-informed IS I/IST programs

Staff Review Guidance

The reviewer should confirm that the application clearly descri bes the applicants organizational responsibilities and that they are consistent with the applicat ion guidance in Part 3 above.

The reviewer should confirm that if the application commits to using one or more consensus codes and standards in the development of the ISI or IST progra ms, and any organizational responsibilities contained in those codes and standards are als o described in the application.

DANU-ISG-2022-07 Page 16 of 18

IMPLEMENTATION

The NRC staff will use the information discussed in this ISG to review non-LWR applications for CPs, OLs, COLs, SDAs, DCs, and MLs under 10 CFR Part 50 and 10 CFR Part 52. The NRC staff intends to incorporate this guidance in updated form in t he RG or NUREG series, as appropriate.

BACKFITTING AND ISSUE FINALITY DISCUSSION

The NRC staff may use DANU-ISG-2022-07 as a reference in its re gulatory processes, such as licensing, inspection, or enforcement. However, the NRC staff d oes not intend to use the guidance in this ISG to support NRC staff actions in a manner t hat 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, Iss ue Finality, and Information Requests (Ref. 14), nor does the NRC staff intend to use the guidance to affect the issue finality of an approval under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The staff also does not intend to use the guidan ce to support NRC staff actions in a manner that constitutes forward fitting as that term is de fined and described in Management Directive 8.4. If a licensee believes that the NRC is using thi s ISG in a manner inconsistent with the discussion in this paragraph, then the licensee may file a backfitting or forward fitting appeal with the NRC in accordance with the process in Management Direc tive 8.4.

CONGRESSIONAL REVIEW ACT

Discussion to be provided in the final ISG.

FINAL RESOLUTION

The NRC staff will transition the information and guidance in t his ISG into the RG or NUREG series, as appropriate. Following the transition of all pertine nt information and guidance in this document into the RG or NUREG series, or other appropriate guid ance, this ISG will be closed.

ACRONYMS

ARCAP advanced reactor content of application project ARDC advanced reactor design criterion/a ASME American Society of Mechanical Engineers BPV Code Boiler and Pressure Vessel Code CFR Code of Federal Regulations COL combined license CP construction permit DC design certification GDC general design criterion/a ISG interim staff guidance ISI inservice inspection IST inservice testing LWR light-water reactor ML manufacturing license NRC U.S. Nuclear Regulatory Commission OL operating license OM Code Operation and Maintenance of Nuclear Power Plants DANU-ISG-2022-07 Page 17 of 18

PRA probabilistic risk assessment QA quality assurance RIM reliability and integrity management RISC risk-informed safety class RG regulatory guide SDA standard design approval SOC statement of considerations SSC structure, system, or component STR special treatment requirements

REFERENCES

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

2 10 CFR Part 52, Licenses, Certifications, and Approvals f or Nuclear Power Plants.

3 U.S. Nuclear Regulatory Commission, DANU-ISG-2022-01, Revie w of Risk-Informed, Technology-Inclusive Advanced Reactor Applications - Roadmap, May 2023 (Agencywide Documents Access and Management System (ADAMS) Acce ssion No. ML22048B546).

4 American Society of Mechanical Engineers (ASME),Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature R eactors, July 1, 2021.

5 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.87, R evision 2, Acceptability of ASME Code Section III, Division 5, High Temperature Reactor s, Washington, DC

6 ASME, QME-1, Qualification of Active Mechanical Equipment U sed in Nuclear Facilities, January 2017.

7 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.100, Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants, Wa shington, DC

8 ASME/American Nuclear Society, ASME/ANS RA-S-1.4-2021, Prob abilistic Risk Assessment Standard for Advanced Non-Light-Water Reactor Nuclea r Power Plants, February 2021.

9 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.247 ( for trial use),

Acceptability of Probabilistic Risk Assessment Results for Non -Light-Water Reactor Risk-Informed Activities, Revision 0, March 2022 (ADAMS Access ion No. ML21235A008).

10 U.S. Nuclear Regulatory Commission Regulatory Guide 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reacto rs, Revision 0, April 2018 (ADAMS Accession No. ML17325A611).

11 ASME Boiler and Pressure Vessel Code, 2019 Edition,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 2, Req uirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants.

DANU-ISG-2022-07 Page 18 of 18

12 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.246, Acceptability of ASME Code,Section XI, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants for Non-Light-Water Rea ctors, Washington, DC

13 ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Facility Components, Division 1.

14 U.S. Nuclear Regulatory Commission, Management Directive 8. 4, Management of Backfitting, Forward Fitting, Issue Finality, and Information R equests.