ML23297A165

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Redline Strikeout for Interim Staff Guidance - DANU-ISG-2022-08, Advanced Reactor Content of Application Project, Risk-Informed Technical Specifications
ML23297A165
Person / Time
Issue date: 03/31/2024
From: Joseph Sebrosky
NRC/NRR/DANU/UARP
To:
Shared Package
ML23277A105 List:
References
DANU-ISG-2022-08, NRC-2022-0081, ML23277A105
Download: ML23297A165 (22)
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Text

DANU-ISG-2022-08

Advanced Reactor Content of Application Project

Risk-Informed Technical Specifications

Interim Staff Guidance March 2024

DANU-ISG-2022-08 Advanced Reactor Content of Application Project Risk-Informed Technical Specifications Interim Staff Guidance

AD AMS Accession No.: Package - ML23277A105; ISG - ML23277A146; Enclosure - ML23277A155; FRN -ML23277A232; CRA Summary - ML23277A272 OFFICE OCIO/GEMSD/FLICB QTE NRR/DRO/IRAB (PM) NRR/DANU/UTB1 (BC)

/ICT NAME DCullison KAziria-Kribbs CCaufman GOberson DATE 2/9/2024 3/11/2022 3/12/2024 11/16/2023 OFFICE NRR/DANU/UTB2 NRR/DNRL/STSB NRR/DANU/UARP NRR/DANU/UARP (BC)

(BC) (BC) (PM)

NAME CdeMessiers MShivani JSebrosky SLynch DATE 12/28/2023 11/3/2023 10/26/2023 12/14/2023 OFFICE OGC (NLO) NRR/DANU (D)

NAME RWeisman MShams DATE 3/21/2024 2/24/2024 OFFICIAL RECORD COPY

INTERIM STAFF GUIDANCE

ADVANCED REACTOR CONTENT OF APPLICATION PROJECT

RISK-INFORMED TECHNICAL SPECIFICATIONS

DANU-ISG-2022-08

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 p rov id es g ui da nce o n the co nte nt s of applications to an applicant submitting a risk-informed, performance-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 Utilization Facilities (Ref. 1), or for a combined license (COL), a manufacturing license (ML), or a design certification (DC) under 10 CFR Part 52, Licenses, Certifications, and Appro vals for Nuclear Power Plants (Ref. 2), for a nonlight-water reactor (non-LWR). The applicati on guidance found in this ISG supports the development of the portion of a non-LWR application associated with an applicants technical specifications (TS).1 Second, this ISG provides guidance to NRC staff on how to review 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 application project (ARCAP), whose purpose is to develop technology-inclusive, risk-informed, and performance-based application guidance. The ARCAP is broader than, and encompasses, the indus try-led technology-inclusive content of application project (TICAP). The 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 Advanced Reactor Applications - Roadmap, issued in March 2024 (Ref. 3), which p rovides a roadmap for developing all portions of an application. The guidance in this ISG is limited to the portion of

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 t he NRC staff uses in evaluating specific issues or postulated events, and to describe information that the NRC sta ff needs in its review of applications for permits and licenses. The guidance in this ISG that pertains to applicants is not NRC regulations and compliance with it is not required. Methods and solutions that differ from those set fort h in this ISG are acceptable if supported by a basis for the issuance or continuance of a permit or license by the Commission.

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non-LWR application associated with the development of technical specifications for the nuclear reactor plant applicant and the NRC staff review of that portio n of the application.

RATIONALE

The current application guidance related to technical specifications is directly applicable only to light water reactors (LWRs) and may not fully identify the info rmation to be included in a non-LWR application or efficiently provide a technology-inclusive, risk-informed, and performance-based review approach for non-LWR technologies. This ISG serves as the non-LWR application guidance for technical specifications. This ISG provides both applicant content of application and NRC staff review guidance.

AP PLIC ABI LITY

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

PAPERWORK REDUCTION ACT

This ISG provides voluntary guidance for implementing the mandatory 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 Collections 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.

PUBLIC PROTECTION NOTIFICATION

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

GUIDANCE

Section 182a. of the Atomic Energy Act of 1954, as amended, requires applicants for nuclear power plant operating licenses to provide the following:

[S]uch technical specifications, including information of the a mount, kind, and source of special nuclear material required, the place of the u se, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization...of special nuclear material will be in accord with the common defense

2 An applicant desiring to use this ISG for a light water reacto r application should contact the NRC staff to hold pre-application discussions on its proposed approach.

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and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued.

In 10 CFR 50.36, Technical specifications, the Commission established its regulatory requirements related to the content of TS. In doing this, the Commission emphasized matters related to the prevention of accidents and the mitigation of ac cident consequences; the Commission noted that applicants were expected to incorporate into their TS those items that are directly related to maintaining the integrity of the physic al barriers designed to contain radioactivity.3

According to 10 CFR 50.36, TS for operating nuclear power react ors are required to include items in the following categorie s: (1) safety limits and limiting safety system settings (LSSSs),

(2) limiting conditions for operation (LCOs), (3) surveillance requirements, (4) design features, and (5) administrative controls.

In its Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated July 22, 1993 (Ref. 5), the Commission stated that it

[e]xpects that licensees, in preparing their Technical Specification related submittals, will utilize any plant-specific PSA [probabilistic safety assessment] or risk survey and any available literature on risk insights and PSAs. Similarly, the NRC staff will also employ risk insights and PSAs in evaluating Technical Specifications related submittals. Further, as a part of the Commissions ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements.

This ISG describes methods acceptable to the NRC staff for an a pplicant to prepare proposed TS using a risk-informed evaluation process, such as the process described in Regulatory Guide (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 Rea ctors, issued June 2020 (Ref. 6). For risk-informed application that does not use the R G 1.233 methodology, an applicant should discuss with the NRC staff in preapplication interactions how their TS approach differs from that proposed in this ISG. This ISG also includes guidance for the NRC staff to review risk-informed TS.

Application Guidance

RG 1.233 provides the NRC staffs guidance on using a technology-inclusive, risk-informed, and performance-based methodology to inform the licensing basis and content of applications for non-LWRs, including, but not limited to, molten salt reactors, high-temperature gas-cooled reactors, and a variety of fast reactors at different thermal c apacities. This RG may be used by non-LWR applicants applying for permits, licenses, certificatio ns, and approvals under 10 CFR Part 50 and 10 CFR Part 52. RG 1.233 endorses Nuclear En ergy Institute (NEI) 18-04,

3 The Commission adopted its current approach to § 50.36, Techn ical Specifications, which implements section 182a. of the Atomic Energy Act on December 17, 1968 (33 FR 18610). The Commission first promulgated § 50.36 in 1962 (27 FR 5492), but the rule took a different approach than the 1968 amendment and currently employed in the regulations.

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Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Revision 1, issued August 2019 (Ref. 7), as one acceptable method for non-LWR designers to use when carrying out these activities and preparing their applications. The methodology in NEI 18-04 provides a process by which the content of applications will permit understanding of the system designs and their relationship to safety evaluations for a variety of non-LWR designs.4 Figure 1 is taken from NEI 18-04, Revision 1, and illustrates the concepts used to classify safety-related structures, systems, and components (SSCs), risk-significant SSCs, and safety-significan t SSCs.

Figure 1 - Definition of Safety Related, Risk Significant, and Safety Significant Structures, Systems and Components from NEI 18-04, Revision 1

NEI 18-04 states that the safety classification for an SSC requ ires that an assessment be performed of the risk significance of SSCs and the licensing-basis events (LBEs). The assessment should describe the probabilistic risk assessment (PRA) safety functions (PSFs)5 of the SSCs credited in the prevention and mitigation of events. Information from the PRA is used as input to the selection of reliability targets and performance requirements6 for SSCs that set the stage for the selection of special treatment requirements. NEI 18-04, Task 16, Specify Special Treatment Requirements for SR [Safety-Related] and NSRST [Non-Safety-Related with

4 As noted in NEI 18-04, the plant on which the TSs are based includes the collection of site, buildings, radionuclide sources, and SSCs seeking a single design certification or one or more operating licenses under the LMP framework.

The plant may include a single reactor unit or more than one reactor modules as well as non-reactor radionuclide sources 5 According to NEI 18-04, PSFs are reactor design-specific SSC functions modeled in a PRA that serve to prevent or mitigate a release of radioactive material or to protect one or more barriers to release. They are a broader set of safety functions than those defined in NEI 18-04 by the term r equired safety function (RSF), which only applies to safety functions performed by safety-related SSCs.

6 Performance requirements, as referenced in NEI 18-04, should be understood as recommendations that the NRC staff considers adequate to satisfy portions of NRC regulatory requirements but that are not the only acceptable methods of compliance.

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Special Treatment] SSCs, states the following:

All safety-significant SSCs that are distributed between SR and NSRST are subject to special treatment requirements. These requirements always include specific performance requirements to provide adequate assurance that the SSCs will be capable of performing their PSFs with significant margins and with appropriate degrees of reliability. These include numerical tar gets for SSC reliability and availability, design margins for performance of the PSFs, and monitoring of performance against these targets with appropriate corrective actions when targets are not fully realized.

NEI 18-04 specifies special trea tments, including TS, to address programmatic defense-in-depth (DID) attributes. Considerations specified in NEI 18-04 involvi ng TS include the following:

  • Are all risk-significant LBE LCOs reflected in TS?
  • Are allowable outage (LCO action completion) times in TS consistent with assumed functional reliability levels for risk-significant LBEs?
  • Are the TS for risk-significant SSCs consistent with achieving the necessary safety function outcomes for the risk-significant LBEs?

RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Ref. 8), descri bes a general approach to developing risk-informed regulatory applications for licensing basis changes and discusses specific topics common to risk-informed regulatory applications. RG 1.177, Plant-Specific, Risk-Informed Decision-making: Technical Specifications (Ref. 9) also supports this ISG. While RG 1.177 focuses on methods acceptable to the NRC staff for assessing the use of risk analysis of proposed changes to TS, its guidance is also useful in evaluating certain aspects of initial TS development.

In 10 CFR 50.34(a)(5), the NRC requires an applicant for a CP u nder 10 CFR Part 50 to include, in the preliminary safety analysis report, an identif ication and justification for the selection of those variables, conditions, or other items which are determined as the result of preliminary safety analysis and evaluation to be probable subjects of technical specifications for the facility, with special attention given to those items which may significantly influence the final design. As an option, a CP applicant may propose preliminary T S and include them in the preliminary safety analysis report or in a separate application document. Under 10 CFR 52.47(a)(11), a DC application must include proposed generic technical specifications,7 as required by 10 CFR 50.36(a)(2), which should be derived from the analyses and evaluations included in the proposed DC FSAR.

At the CP, DC, or ML application stage, some numerical values, graphs, and other data are not as complete as necessary for plant operation because determination of specific numerical values is pending future decisions by the OL or COL applicant on selection and procurement of hardware after issuance of the CP, DC, or ML. A DC application may describe COL action items related to the generic technical specifications to be denoted b y square brackets in the proposed

7 For DCs, the TS required under 10 CFR 52.47(a)(11) that have been incorporated by reference in the DC rulemaking Appendices A, B, C, D, E, F, and G of 10 CFR Part 52 are referred to as generic technical specifications.

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generic technical specifications and associated bases with appropriate guidance to COL applicants for completing COL action items. At the OL or COL application stage, as-procured or site-specific information (denoted by brackets in the reference DC (i.e., generic design control document (DCD)) or ML TS) must be replaced with the final operational information, which must be in conformance with the final safety analysis report. For a COL application referencing a DC, this information is in the plant-specific DCD.

Content for a CP application is limited to whether the values r easonably agree with the anticipated operational capability of the plant. For preliminar y safety analysis report (PSAR) technical specifications, information which may significantly influence the final design should be provided in preliminary LCOs, a preliminary list of the types o f surveillance tests being considered, and a preliminary description of important design features. The PSAR technical specifications need not include surveillance requirement freque ncies or administrative controls although inclusion of such information, if available, would assist the staff in understanding the design. Furthermore, the PSAR should include a preliminary Technical Specification Bases document to summarize the information in the PSAR on which the preliminary technical specifications are based. Commented [A1]: NRC-2022-0074-DRAFT-0006-3 For a DC application, the applicant should provide generic TSs to confirm that they will preserve NRC-2022-0075-DRAFT-0004-34 the validity of the plant design, as described in the DCD, by e nsuring that the plant will be operated (1) within the required conditions bounded by the DCD and (2) with operable equipment that is essential to prevent postulated design-basis events or mitigate their consequences. For an ML application, the applicant should propose TSs in a similar manner to those provided in a DC application. For an OL or COL application, the applicant should propose TSs to ensure compliance with the applicable acceptance criteria below. For COL applications referencing a DC or ML, the applicant should also ensure that bracketed information is replaced with site specific information or final operational information, as applicable, in conformance with the final safety analysis report for the application.

Acceptance criteria are based on meeting the relevant requirements of the following Commission regulations:

  • 10 CFR 50.36a, Technical specifications on effluents from nuclear power reactors

Contents of Technical Specifications

In 10 CFR 50.36, the NRC requires proposed TS for nuclear reactors to include the following:

(1) 10 CFR 50.36(c)(1)(i)(A) safety limitsSafety limits apply to important process variables necessary for an appropriate level of protection for the integrity of certain physical barriers that guard against the uncontrolled release of radioactive material.

(2) 10 CFR 50.36(c)(1)(ii)(A) Limiting safety system settings (LSSSs)LSSSs are for automatic protective devices affecting variables with significant safety functions.

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(3) 10 CFR 50.36(c)(2) Limiting conditions for operation (LCOs)LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee must shut down the reactor or follow any remedial action permitted by the TS until the condition can be met. A TS LCO of a nuclear reactor must be established for eac h item meeting one or more of the following 10 CFR 50.36(c)(2)(ii) criteria:

a. Criterion 1. Installed instrumentation that is used to dete ct, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
b. Criterion 2. A process variable, design feature, or operati ng restriction that is an initial condition of a design-basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
c. Criterion 3. An SSC that is part of the primary success pat h and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barr ier.
d. Criterion 4. An SSC that operating experience or PRA has sh own to be significant to public health and safety.

(4) 10 CFR 50.36(c)(3) surveillance requirementsSurveillance r equirements are requirements relating to test, calibration, or inspection to as sure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

(5) 10 CFR 50.36(c)(4) design featuresDesign features include aspects of the facility (e.g., construction materials and geometric arrangements) not covered in the categories described above that, if altered or modified, would have a significant effect on safety.

(6) 10 CFR 50.36(c)(5) administrative controlsAdministrative c ontrols are provisions for organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure safe operation of the facility.

Also, 10 CFR 50.36 requires that a summary statement of the bas es or reasons for the TS, other than those covering administrative controls, be included in the application but shall not become part of the TS. In addition, the TS should be consistent with the applicants principal design criteria (PDC) in that all safety-related features specified in the PDC should be addressed in the TS.

To provide suitable guidance on risk-informed TS for advanced reactors, this ISG correlates the text in 10 CFR 50.36 with the analysis and outputs of the risk-informed approach described in NEI 18-04 and with the PDC applicable to the design (A risk-inf ormed application not using NEI 18-04 may need to consider an approach that is modified in comparison to the corresponding guidance in NEI 18-04.). In some cases, this corr elation may be inconsistent with the regulation text as applied to a particular design, in which case the applicant should include an exemption request as part of its application. Commented [A2]: NRC-2022-0074-DRAFT-0006-4 Safety Limits NRC-2022-0075-DRAFT-0004-35 DANU-ISG-2022-08 Page 8 of 20

In the definition of safety limits in 10 CFR 50.36(c)(1)(i)(A), the text important process variables that are necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity is compatible with NEI 18-04 analysis and outputs. The NEI 18-04 process provides insights on identification of barriers that guard against the release of radioactivity. Specifically, NEI 18-04 calls for the identification of reactor design-specific functional criteria t hat are necessary and sufficient to meet RSFs that maintain the consequences of one or more design-basis events (DBEs) or the frequency of one or more high-consequence beyond-design-basis e vents (BDBEs) inside the Frequency-Consequence (F-C) Target. 8 These RSFs include protecting barrier integrity to guard against release of radioactivity, therefore, the safety limit d efinition need not be changed.

Hence, for applications using the NEI 18-04 approach, 9 the TS should address safety limits as follows:

10 CFR 50.36(c)(1)(i)(A) TS Content Based on Corresponding NEI 18-04 Output Safety limits for nuclear reactors are limits Safety limits for nuclear reactors are limits upon important process variables that are upon important process variables that are found to be necessary to reasonably protect found to be necessary to reasonably protect the integrity of certain of the physical barriers the integrity of certain of the physical barriers that guard against the uncontrolled release of that guard against the uncontrolled release of radioactivity. radioactivity.

Limiting Safety System Settings

In the definition of LSSSs in 10 CFR 50.36(c)(1)(ii)(A), the phrase settings for automatic protective devices related to those variables having significant safety functions can be correlated to NEI 18-04 outputs related to reactor design-speci fic functional criteria that are necessary and sufficient to meet RSFs. 10 RSFs prevent or mitigate a release of radioactive material or protect one or more barriers to maintain the consequences of one or more DBEs or the frequency of one or more high-consequence BDBEs inside the F-C Target. The discussion above on 10 CFR 50.36(c)(1)(i)(A) and safety limits contains more information.

An applicant may propose an administrative control TS to maintain a setpoint control program to satisfy 10 CFR 50.36(c)(1)(ii)(A) in lieu of specifying explicit values for the LSSSs in the TS.

8 An applicant using a risk-informed process but not NEI 18-04 sh ould discuss its alternative risk-informed process with the NRC staff in pre-application interactions.

9 See footnote 8.

10 Reactor design-specific functional criteria that are necessary and sufficient to meet the RSFs are defined as required functional design criteria in NEI 18-04.

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Hence, for applications using the NEI 18-04 approach, 11 the TS should address LSSSs as follows:

10 CFR 50.36(c)(1)(ii)(A) TS Content Based on Corresponding NEI 18-04 Output Limiting safety system settings for nuclear Limiting safety system settings are reactors are settings for automatic protective settings for automatic protective devices devices related to those variables having related to those variables that prevent or significant safety functions. Where a limiting mitigate a release of radioactive material or safety system setting is specified for a protect one or more barriers to maintain the variable on which a safety limit has been consequences of one or more DBEs or the placed, the setting must be so chosen that frequency of one or more high-consequence automatic protective action will correct the BDBEs inside the F-C Target. Where a abnormal situation before a safety limit is limiting safety system setting is specified for a exceeded. variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

Limiting Conditions for Operation

The NEI 18-04 process specifies that the TS for risk-significant SSCs be consistent with achieving the necessary safety function outcomes for the risk-significant LBEs. Additionally, the programmatic DID process should determine allowable outage (LCO action completion) times for applicable SSCs in TS such that they are consistent with assumed functional reliability levels for risk-significant LBEs. The NEI 18-04 process refines the fundamental safety functions applicable to all reactors (controlling heat generation, controlling heat removal, and retaining radionuclides) as necessary into reactor-technology-specific safety functions (i.e., RSFs). The RSFs provide the foundation for analyzing reactor-technology-sp ecific SSCs selected to perform each function. LCOs should be specified for SSCs that (1) perform an RSF needed to mitigate the consequences of DBEs to within the F-C Target, (2) mitigate DBAs that only rely on the SR SSCs to meet the dose criteria of 10 CFR 50.34(a)(1)(ii)(D) or 52.79(a)(1)(vi), (3) maintain the frequency of one or more high-consequence BDBEs inside the F-C Target, or (4) perform risk-significant functions. Structures and physical barriers that ar e necessary to protect any SR SSCs in performing their RSFs in response to any design-basis e xternal event are also classified as SR and should be addressed in an LCO. The discussion below on each specific 10 CFR 50.36(c)(2)(ii) LCO Criterion contains further informati on.

Section 50.36(c)(2)(ii)(A)-(C) (LCO Criteria 1 through 3)

LCO Criterion 1 applies to installed instrumentation that is us ed to detect a significant abnormal degradation of the reactor coolant pressure boundary. Criterion 2 applies to a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to t he integrity of a fission product barrier. Criterion 3 pertains to SSCs that are part of the prim ary success path and that function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a

11 See footnote 8.

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challenge to the integrity of a fission product barrier.

The 10 CFR 50.36 text for these three criteria cannot be direct ly correlated to outputs for an advanced reactor using the NEI 18-04 process. Because each of these criteria involves challenges to the integrity of a fission product barrier, the RSFs are the NEI 18-04 process outputs that correlate to these criteria (as discussed above under Safety Limits). Since NEI 18-04 calls for SR SSCs to perform RSFs, LCO Criteria 1 thr ough 3 should be defined for an advanced reactor in terms of SR SSCs. In accordance with NEI 18-04, SR SSCs are selected by the designer from the SSCs that are available to perform the RSFs to mitigate the consequences of DBEs to within the LBE F-C Target and to mitiga te DBAs that only rely on the SR SSCs to meet the dose criteria of 10 CFR 50.34(a)(1)(ii)(D) or 52.79(a)(1)(vi) using conservative assumptions. Note that SR SSCs are also relied on to perform the RSFs to prevent the frequency of BDBEs with consequences greater than the 10 CFR 50.34 or 52.79 dose criteria from increasing into the DBE region and beyond the F-C Target. The discussion of LCO Criterion 4 below covers this latter function.

Hence, for applications using the NEI 18-04 approach, 12 the TS should address LCO Criteria 1 through 3 as follows:

10 CFR 50.36(c)(2) TS Content Based on Corresponding NEI 18-04 Output Limiting conditions for operation are the Limiting conditions for operation are the lowest functional capability or performance lowest functional capability or performance levels of equipment required for safe levels of equipment required for safe operation of the facility. operation of the facility.

Criterion 1. Installed instrumentation that is Criterion 1. Installed instrumentation that is used to detect, and indicate in the control used to detect, and indicate where room, a significant abnormal degradation of necessary, a significant abnormal the reactor coolant pressure boundary. degradation of barriers necessary to maintain the release of radioactive materials from the plant to within the DBE F-C Target or to mitigate DBAs that only rely on the SR SSCs to meet the dose criteria of 10 CFR 50.34 or identical criteria in 10 CFR Part 52 (i.e., 10 CFR 52.47(a)(11), 10 CFR 52.79(a)(30), and 10 CFR 52.158(f)(18).

Criterion 2. A process variable, design Criterion 2. A process variable, design feature, or operating restriction that is an feature, or operating restriction that is an initial condition of a design basis accident or initial condition of an anticipated operational transient analysis that either assumes the occurrence (AOO) or DBE and is necessary failure of or presents a challenge to the to maintain consequences to within the F-C integrity of a fission product barrier. Target or is necessary for a SR SSC to mitigate a DBA to meet the dose criteria of 10 CFR 50.34 or identical criteria in 10 CFR Part 52.

Criterion 3. A structure, system, or Criterion 3. An SSC that is part of the component that is part of the primary success primary success path and that performs an path and which functions or actuates to RSF to mitigate the consequences of DBEs mitigate a design basis accident or transient to within the F-C Target or to mitigate DBAs

12 See footnote 8.

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10 CFR 50.36(c)(2) TS Content Based on Corresponding NEI 18-04 Output that either assumes the failure of or presents that only rely on the SR SSC to meet the a challenge to the integrity of a fission dose criteria of 10 CFR 50.34 or identical product barrier. criteria in 10 CFR Part 52.

LCO Criterion 4

Criterion 4 relates to SSCs that the PRA shows to be significan t to public health and safety.13 In correlating this text to the NEI 18-04 process, it is necessary to understand the term significant to public health and safety. In the supplementary information provided in the NRCs 1995 revision to 10 CFR 50.36 (Volume 60 of the Federal Register (FR), page 36953 (60 FR 36953 (July 19, 1995))) (which codified the Final Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated J uly 22, 1993), the Commission described Criterion 4 as follows:

Criterion 4 is intended to capture those constraints that probabilistic risk assessment or operating experience show to be significant to public health and safety, consistent with the Commissions PRA Policies. The leve l of significance either would need to be such that it justified including the constraints in the technical specifications to ensure adequate protection of the public health and safety or that the addition of such constraints provides substantial additional protection to the public health and safety.

60 FR at 36955-56 (emphasis added). The Commission also discuss ed the application of Criterion 4 in the context of relocation of TS to licensee-c ontrolled documents as follows:

[With respect to relocating items from existing technical speci fications which do not meet the first three criteria,] [i]f a technical specificat ion provision does not meet any of the first three criteria, and if the current PRA kn owledge or operating experience does not identify the structure, system, or component as risk significant, the NRC staff will not preclude relocating such technical specifications.

Id. (emphasis added). The NEI 18-04 process uses PRA as one inp ut to identify RSFs that are tied to public health and safety through the F-C Target. The NE I 18-04 process identifies two groups of SSCs that are tied to public safety but are not addre ssed by Criteria 1 through 3 above:

(1) SR SSCs that perform RSFs to prevent the frequency of BDBEs with consequences greater than the 10 CFR 50.34 dos e criteria (or the identical criteria in Part 52) from increasing into the DBE region and beyond the F-C Target.

(2) Non-SR SSCs that are relied on to perform risk-significant functions (i.e., NSRST SSCs).

Risk-significant SSCs are those which perform functions that pr event or mitigate any LBE from exceeding the F-C Target or make significant contributions to the cumulative risk metrics selected for evaluating the total risk from all analyzed LBEs. The cumulative risk limit criteria address the situation in which an SSC may contribute to two or more

13 In assessing Criterion 4, operating experience is not likely a vailable for some aspects of non-LWRs.

DANU-ISG-2022-08 Page 12 of 20

LBEs that collectively may be risk significant even though each individual LBE may not be significant. All LBEs within the scope of the supporting PRA should be included when evaluating these cumulative risk limits. In such cases, the rel iability and availability of such SSCs may need to be controlled to manage the total integra ted risk over all the LBEs. Section 4.2.2 of NEI 18-04 further clarifies risk-significant SSCs.

Hence, for applications using the NEI 18-04 approach, 14 the TS should address LCO Criterion 4 as follows:

10 CFR 50.36(c)(2) TS Content Based on Corresponding NEI 18-04 Output Criterion 4. A structure, system, or Criterion 4. (a) An SR SSC relied on to component which operating experience or perform a RSF to prevent the frequency of a probabilistic risk assessment has shown to BDBE with consequences greater than the be significant to public health and safety. 10 CFR 50.34 dose criteria or the identical criteria in Part 52 from increasing into the DBE region and beyond the F-C Target.

(b) An NSRST SSC relied on to perform a risk-significant function. These risk-significant SSCs are those that perform functions that prevent or mitigate any LBE from exceeding the F-C Target or make significant contributions to the cumulative risk metrics selected for evaluating the total risk from all analyzed LBEs.

Note that Criterion 4 for the corresponding NEI 18-04 output do es not include NSRST SSCs that only perform functions credited for DID. 15

Limiting Condition for Operation Format

Applicants may determine the format for LCOs. However, in determining the format, the staff recommends that applicants use the format utilized in the Stand ard Technical Specifications (STS) [for example, NUREG-1431, Volume 1, Revision 5.0, Standa rd Technical Specifications

- Westinghouse Plants, (Ref. 10)] since the STS format has been developed jointly by the NRC and the industry over the last 30 years to be logical, concise, and clear for nuclear power plant operators. In addition, applicants may also be informed by the format used for non-power utilization facilities (e.g., SHINE technical specifications at ADAMS No. ML19211C135). At a minimum, each LCO should include the following:

(1) Describe the operable condition.

14 See Footnote 8.

15 As noted in NEI 18-04, and in the Technical Requirements Manual section of DANU-ISG-2022-01, availability controls outside of TSs similar to those approved for some SSCs of passive light water reactors under the regulatory treatment of non-safety systems (RTNSS) approach could be appropriate for the NSRST SSCs that only perform functions credited for DID.

DANU-ISG-2022-08 Page 13 of 20

(2) Include the mode(s) of applicability (i.e., the operating m odes during which the LCO must be met ).

(3) Explain the actions that must be taken when the limiting condition is not met, including any required action and the associated completion time. For determining various LCO completion times, the risk impact should be evaluated using the PRA and DID analysis.

RG 1.177, Regulatory Position 2.3.4, contains additional guidance in this area.

RG 1.177, position 2.3.4, references the risk metrics of core damage frequency and large early release frequency based on LWRs as factors in deter mining completion times. Advanced reactor applicants should use other risk metrics, such as those described in NEI 18-04, for determining completion times. NEI 1 8-04, Section 3.3.5, Selection of Risk Metrics for PRA Model Development, describes several possible risk metrics (that are different from the core damage frequency (CDF) and Large Early Release Frequency (LERF) metrics developed for LWRs). These met rics could be used by an applicant to develop LCO completion times. Applicants should discuss their proposed risk metrics for developing LCO completion times with NRC staff during preapplication discussions. Commented [A3]: NRC-2022-0074-DRAFT-0006-2 (4) Include a set of associated surveillance requirements. NRC-2022-0075-DRAFT-0004-36

Surveillance Requirements

In 10 CFR 50.36(c)(3), the NRC requires that TS include surveil lance requirements.

Surveillance requirements are requirements relating to test, ca libration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

Surveillance requirements should be determined through the development of the Special Treatments Considered for Programmatic DID task in the NEI 18- 04 process. The PRA, supplemented with additional data and analysis where necessary should provide a basis for determining the specified TS surveillance frequency. RG 1.177, Regulatory Position 2.3.4, offers additional guidance in this area.

Applicants may propose to locate time-based surveillance frequencies to a licensee-controlled program, called the surveillance frequency control program (SFC P), and add the SFCP to the administrative controls section of TS. In a letter dated Septe mber 19, 2007, (Ref. 11) NRC staff has accepted NEI 04-10, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1 (Ref. 12) as an acceptable approach to developing the SFCP. As stated in NEI 04-10, the SFCP is not applicable to surveillance frequencies that are event driven, controlled by an existing program, or condition based.

Design Features

In 10 CFR 50.36(c)(4), the NRC requires that TS describe design features, specifically those features of the facility such as materials of construction and geometric arrangements that, if altered or modified, would have a significant effect on safety and are not covered in categories described in 10 CFR 50.36(c)(1), (2), and (3). Section 50.36(c) (4) covers design features such as the natural circulation configuration of a structure or the material composition of a graphite matrix. This requirement can be correlated to the design features that provide the RSFs determined via the NEI 18-04 process.

DANU-ISG-2022-08 Page 14 of 20

Administrative Controls

In 10 CFR 50.36(c)(5), the NRC requires that TS include adminis trative controls. Administrative controls are the provisions relating to organization and manage ment, procedures, record keeping, review and audit, and reporting necessary to assure op eration of the facility in a safe manner. Administrative controls can be derived, in part, from the development of special treatment and the Application of Programmatic DID Guidelines described in the NEI 18-04 process. In addition to controls identified for special treatme nt, the TS administrative controls should include requirements that address the following areas:16

(1) a description of important responsibilities within the operations organizational structure

(2) a description of onsite and offsite organizations, including lines of authority and facility staffing

(3) a description of facility staff qualifications

(4) a requirement that procedures be established, implemented, and maintained covering the following:

a. applicable procedures recommended in RG 1.33, Quality Assur ance Program Requirements (Operation) (Ref. 13)
b. emergency operating activities
c. quality assurance for effluent and environmental monitoring
d. fire protection program implementation
e. all programs specified below

(5) a requirement that programs and reports necessary to operat e the plant in a safe manner be established, implemented, and maintained, including b ut not limited to the following:

a. safety function determination program (SFDP)This program en sures loss of safety function is detected and appropriate actions taken.17 The SFDP description should specify that the program includes the following:
i. provisions for cross train checks to ensure a loss of the ca pability to perform the safety function credited or relied upon in the accident analysis does not go undetected

16 NUREG-1431, Volume 1, Revision 5.0, Standard Technical Specifi cationsWestinghouse Plants, issued September 2021 (Ref. 10), Section 5.5, Administrative ControlsPrograms and Manuals, provides a better understanding of these terms. Note that, depending on the spec ific reactor technology, additional programs may need to be included in the section of the TS on administrative controls.

17 The SFDP identifies where a loss of safety function exists. I f a loss of safety function is determined to exist by this program, the appropriate conditions and required actions of the LCO in which the loss of safety function exists are required to be entered.

DANU-ISG-2022-08 Page 15 of 20

ii. provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists

iii. provisions to ensure that an inoperable support systems completion time is not inappropriately extended as a result of multiple support system inoperabilities

iv. other appropriate limitations and remedial or compensatory actions

b. setpoint control program (if used)This program should estab lish the requirements for ensuring that setpoints for automatic protective devices are initially within, and remain within, the bounds of the applicab le safety analyses; provide a means for processing changes to instrumentation setpoints; and identify setpoint methodologies to ensure instrumentation will function as credited.
c. surveillance frequency control program (if used)This progra m provides controls for surveillance frequencies and should ensure that surveillanc e requirements specified in the TS are performed at intervals sufficient to assure that the necessary quality of systems and components is maintained, that facility operation will be maintained within safety limits, and that the associated LCOs are met.
d. program that addresses high radiation area controls as provi ded in 10 CFR 20.1601(c) (Ref. 14)
e. Offsite Dose Calculation manual and radiological effluent control program
f. annual radiological environmental operating report and radio active effluent release report covering the operation of the plant during the previous calendar year
g. core operating limits report (or similar report for reactor cores that do not have a traditional stationary reactor core) that defines core operatin g limits before each reload cycle or before any remaining portion of a reload cycle
h. TS bases control program that addresses provisions to ensure that the bases are maintained consistent with the final safety analysis report
i. Reactor Coolant System (RCS) Pressure and Temperature Limits Report that addresses RCS pressure and temperature limits for heat up, cool down, low temperature operation, criticality, and hydrostatic testing, Low Temperature Overpressure Protection (LTOP) arming, and power operated relie f valves (PORVs) lift settings, if applicable to the specific design, as well as heatup and cooldown rates

Technical Specification Bases

Applicants should provide a TS bases document that provides the technical basis for all safety limits, LCOs, surveillance requirements, and design feature TS. This document should provide a basis for the operability and availability controls, including allowable outage times and DANU-ISG-2022-08 Page 16 of 20

surveillance testing intervals that are included in the TS. The TS bases should conform to the applicable analysis described in the safety analysis report. Th is document will be licensee controlled and updated according to the requirements in 10 CFR 50.59, Changes, tests and experiments, or a similar change process under 10 CFR Part 52.

As an alternative, an applicant may provide the appropriate TS bases within the scope of the safety analysis report and alleviate the need to provide a separate TS bases document. If this approach is used, the safety analysis report bases should clear ly address each TS, other than those covering administrative controls.

Technical Specification Use and Application Information

In addition to the information specified above, the TS should address the following: 18

(1) A description of the Use and Application rules for the tech nical specifications, including as a minimum:

a. Include a set of definitions for terms used in the TS.
b. Define the plant modes used in determining LCO applicability.
c. Describe logical connectors (if used). Logical connectors ar e used in TS to discriminate between, and yet connect, discrete conditions, required actions, completion times, surveillances, and frequencies. Logical connectors that have been generally used in TS include AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.
d. Describe the completion time conventions used in the TS and guidance for their use.

(2) A set of Surveillance Requirements that establish general requirements applicable to all specifications and apply at all times unless otherwise stated. For example, a general Surveillance Requirement includes one that states that failure to meet an individual surveillance requirement means that the associated LCO is not met.

(3) A set of LCOs that establish the general requirements applicable to all specifications and apply at all times, except when otherwise stated. For example, an LCO providing the requirements for what actions the licensee must take when an in dividual LCO is not met and the associated Required Actions are also not met.

Staff Review Guidance - Acceptance Criteria

The NRC reviewer should ensure that the application includes information sufficient to allow the NRC reviewer to understand the proposed technical specifications. The reviewer should be able to reach a safety finding and address the topic[s] in the NRCs safety evaluation report if the application includes the following information:

18 NUREG-1431, Volume 1, Revision 5. 0, Section 1.0, Use and Application, and Section 3.0, Limiting Condition for Operation Applicability, and Surveillance Requirement Applicability, provide a better understanding of the items set forth below.

DANU-ISG-2022-08 Page 17 of 20

(1) For a CP application, in accordance with 10 CFR 50.34, those variables, conditions, or other items identified through preliminary safety analysis as probable subjects for plant-specific TS and justification for their selection, with special attention to items that could significantly influence the final design.

(2) Justification that the TS preserve the validity of the plan t design, as described in the safety analysis, by ensuring that the plant will be operated (1 ) within the conditions bounded by the safety analysis, (2) with operable equipment tha t is essential to prevent accidents and to mitigate the consequences of accidents postula ted in the safety analysis, and (3) with key design features consistent with those described in the safety analysis report.

(3) An LCO for each aspect of the design that meets the criteri a in 10 CFR 50.36(c)(2)(ii) as correlated to the corresponding outputs of a risk-informed analysis.

(4) TS that reflect all risk-significant SSCs for preventing or mitigating LBEs.

(5) Completion times for LCO actions in TS that conform to func tional reliability levels for risk-significant LBEs credited or relied upon in the FSAR.

(6) TS for risk-significant SSCs sufficient to assure achieveme nt of the necessary safety function outcomes for the risk-significant LBEs.

(7) Surveillance requirements sufficient to assure that the nec essary quality of systems and components is maintained, that facility operation will be withi n safety limits, and that the LCOs will be met. Specifically, the reviewer should confirm tha t the surveillance requirements include specific performance requirements and freq uencies to provide adequate assurance that the TS SSCs will do the following:

a. Be capable of performing their RSFs with significant margins and with appropriate degrees of reliability.
b. Provide additional confidence that the risk-significant SSCs will perform as intended.

(8) TS that meet the regulations in 10 CFR 50.36 unless the departure is explicitly related to a requested exemption.

(9) TS consistent with the DID philosophy as described in NEI 18-04. (RG 1.177, Regulatory Position 2.2.1, contains additional guidance.)

(10) TS that maintain sufficient safety margins. (RG 1.177, Reg ulatory Position 2.2.2, contains additional guidance.)

(11) Administrative controls adequate to address organization a nd management, procedures, recordkeeping, review and audit, and reporting necessary to ass ure operation of the facility in a safe manner.

(12) TS bases consistent with the analysis described in the safety analysis report and justify the specified variables, conditions, or other limitations as those required by DANU-ISG-2022-08 Page 18 of 20

10 CFR 50.36 (as modified above) to be LCO subjects.

(13) TS that address all of the safety-related features specifi ed in the PDC.

(14) TS that meet the requirements of 10 CFR 50.36a in that the technical specifications include TS that require (a) operating procedures for the control of effluents and (b) annual reports of the quantity of principal radionuclides relea sed to unrestricted areas in both gaseous and liquid effluents.

IMPLEMENTATION

The NRC staff will use the information discussed in this ISG to review non-LWR applications for CPs, OLs, COLs, 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 the RG or NUREG series, as appropriate.

BACKFITTING AND ISSUE FINALITY DISCUSSION

The NRC staff may use DANU-ISG-2022-08 as a reference in its regulatory processes, such as licensing, inspection, or enforcement. However, the NRC staff does 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. 15), nor does the NRC staff intend to use the g uidance to affect the issue finality of an approval under 10 CFR Part 52, Licenses, Certif ications, and Approvals for Nuclear Power Plants. The staff also does not intend to use the guidance to support NRC staff actions in a manner that constitutes forward fitting as that te rm is defined and described in Management Directive 8.4. If a licensee believes that the NRC i s using this ISG in a manner inconsistent with the discussion in this paragraph, then the li censee may file a backfitting or forward fitting appeal with the NRC in accordance with the process in Management Directive 8.4.

CONGRESSIONAL REVIEW ACT

DANU-ISG-2022-08 is 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 foun d it to be a major rule as defined in the Congressional Review Act.

FINAL RESOLUTION

The NRC staff will transition the information and guidance in this ISG into the RG or NUREG series, as appropriate. Following the transition of all pertinent 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 BDBE beyond-design-basis event CDF core damage frequency CFR Code of Federal Regulations COL combined license DANU-ISG-2022-08 Page 19 of 20

CP construction permit DBA design-basis accident DBE design-basis event DC design certification DCD design control document DID defense in depth ISG interim staff guidance LBE licensing-basis event LCO limiting condition for operation LERF large early release frequency LSSS limiting safety system setting LWR light-water reactor ML manufacturing license NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission NSRST non-safety-related with special treatment OL operating license PDC principal design criterion/a PRA probabilistic risk assessment PSA probabilistic safety assessment PSF probabilistic risk assessment safety function RG regulatory guide RSF required safety function SFCP surveillance frequency control program SFDP safety function determination program SR safety related SSC structure, system, and component TICAP technology-inclusive content of application project TS technical specification/s

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 for Nuclear Power Plants.
3. U.S. Nuclear Regulatory Commission, DANU-ISG-2022-01, Revie w of Risk-Informed, Technology-Inclusive Advanced Reactor Applications - Roadmap, March 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23277A139).
4. U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Re view Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LW R Edition, Chapter 16, Technical Specifications.

DANU-ISG-2022-08 Page 20 of 20

5. U.S. Nuclear Regulatory Commission, Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, 58 FR 44071, July 22, 1993 (Available at https://www.nrc.gov/reading-rm/doc-collections/commission/policy/58fr39132.pdf ).
6. U.S. Nuclear Regulatory Commission, Regulatory Guide 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, Washington, DC
7. Nuclear Energy Institute (NEI) 18-04, Risk-Informed Perform ance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Revision 1, August 2019 (ADAMS Accession No. ML19241A472).
8. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Pla nt-Specific Changes to the Licensing Basis, Washington, DC
9. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.177, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Washington, DC
10. U.S. Nuclear Regulatory Commission, NUREG-1431, Standard T echnical SpecificationsWestinghouse Plants, Volume 1, Specifications, Revision 5.0, September 2021 (ADAMS Accession No. ML21259A155).
11. U.S. Nuclear Regulatory Commission, Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04-10, Revision 1, Risk-Informed Technical Specification Initiative 5B, Risk -Informed Method for Control of Surveillance Frequencies, September 19, 2007 (ADAMS Accession No. ML0725702 67)
12. NEI 04-10, Risk-Informed Method for Control of Surveillan ce Frequencies, Revision 1, April 2007 (ADAMS Accession No. ML071360456).
13. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Washington, DC
14. 10 CFR Part 20, Standards for Protection against Radiation.
15. U.S. Nuclear Regulatory Commission, Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests.
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