ML21348A102

12/16-17/2021 - Advisory Committee on Reactor Safeguards (ACRS) Future Plant Designs Subcommittee, 10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors

ML21348A102
Person / Time
Issue date:	12/17/2021
From:	Robert Beall NRC/NMSS/DREFS/RRPB
To:
Beall, Robert
References
NRC-2019-0062, RIN 3150-AK31
Download: ML21348A102 (122)

Text

Advisory Committee on Reactor Safeguards (ACRS)

Future Plant Designs Subcommittee 10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors December 16-17, 2021 1

December 16th Agenda 9:30am - 9:40am Opening Remarks & Staff Introductions 9:40am - 10:00am Update on Part 53 Rulemaking Schedule 10:00am - 1:00pm Graded PRA and Possible Licensing Pathways 1:00pm - 2:00pm Lunch Break 2:00pm - 5:00pm Staffing, Operator Certification, Simulators

[Subpart F - Requirements for Operations]

5:00pm - 5:30pm Discussion 2

NRC Staff Plan to Develop Part 53 Subpart B Subpart C Subpart D Subpart E Subpart F Subpart G Project Life Cycle Design and Siting Construction Operation Decommission Requirements Definition Analysis

• Safety Objectives External Configuration Funding

• Safety Criteria System Hazards Construction/ Control

& Component Manufacturing Termination

• Safety Functions Design Site Surveillance Characteristics Ensuring Maintenance Analysis Capabilities/

Requirements Environmental Reliabilities Staffing &

Considerations Human Factors Safety Change Control Categorization

& Special Environmental Programs Treatment Considerations Security, EP Facility Safety Program Other Plant/Site (Design, Construction, Configuration Control)

Clarify Subpart A General Provisions Analyses (Prevention, Mitigation, Compare to Criteria) Controls and Subpart J Admin & Reporting Plant Documents (Systems, Procedures, etc.) Distinctions Between Other 10 CFR Parts LB Documents (Applications, SAR, TS, etc.) Subparts H & I 3

Current Status Subpart Description Status Subpart A General Requirements Released 1st iteration, including initial definitions (April 2021)

B Safety Criteria Released 3rd iteration (August 2021)

C Design and Analysis Released 3rd iteration (August 2021)

D Siting Released 1st iteration (April 2021)

E Construction Released 1st iteration (April 2021)

Manufacturing Released 1st iteration (April 2021)

F SSCs Released 1st iteration (April 2021)

Personnel Released 1st iteration (October 2021)

Programs Released 1st iteration (April 2021)

G Decommissioning Under development (Planned release December 2021)

H Licensing (LWA, ESP, SDA, Released 1st iteration (August 2021)

DC)

Licensing (ML, CP/OL, COL) Released 1st iteration (October 2021)

I Maintaining Licensing Basis Released 1st iteration (August 2021)

J Reporting & Financial Released 1st iteration (August 2021)

Part 5X Deterministic Alternative Released 1st iteration (October 2021)

Part 73 Physical Security Released 2nd iteration (November 2021)

Cyber Security Released 2nd iteration (November 2021)

Access Authorization Released 2nd iteration (November 2021)

Part 26 Fitness-for-duty Under development (Planned release December 2021)

Other Technology-Inclusive, Risk- Under development Informed Maximum Accident Approach Conforming Changes Under development Statements of Consideration Under development Regulatory Analysis Under development 4

Revised Part 53 Rulemaking Schedule Key Rulemaking Milestones Activity Date(s)

Public Outreach & Generation of Proposed Rule Package July 2020 to August 2022

• NGO Public Meeting February 2022 Submit Draft Proposed Rule to Commission February 2023 Commission Review March 2023 to May 2023 OMB and OFR Processing May 2023 Publish Proposed Rule June 2023 Public Outreach & Generation of Final Rule Package June 2023 to May 2024 Submit Draft Final Rule to Commission December 2024 Commission Review January 2025 to April 2025 OMB and OFR Processing April 2025 to June 2025 Publish Final Rule July 2025 5

Other Key Milestones &

Activities

• Continuation of topical public meetings, as needed.

• Continuation of frequent ACRS meetings o Focus of the meetings will be topics ACRS members want to discuss in more detail

• Continuing to publicly release preliminary proposed rule text o In January 2022, the staff will release a section of the preliminary FRN with all of the Part 53 subparts in one document

Coordination with Other Rulemakings

• Emergency Planning for Small Modular Reactors (SMRs) and Other New Technologies (ONTs)

• Limited Scope Physical Security

• Decommissioning

• Part 50-52 Lessons Learned o Severe Accidents (add to Part 50) o Probabilistic Risk Assessments (PRAs) (add to Part 50) o Three Mile Island (TMI) Requirements (add to Part 50) o Fire Protection Requirements (align Parts 50 and 52) o Operating Licensing (revise simulation, walkthrough requirements) o Licensing Processes (eliminate need to renew design certifications (DCs) and standard design approvals (SDAs))

o Environmental (allow reference to DC environmental assessment)

Graded PRA and Possible Licensing Pathways Marty Stutzke Division of Advanced Reactors and Non-Power and Utilization Facilities Office of Nuclear Reactor Regulation December 16, 2021 8

Overview Technology-Inclusive, Risk-Informed Maximum Accident (TIRIMA) Approach Three Possible Licensing Pathways A Change in Direction Identification and Review of PRA Uses in Initial Licensing Graded PRA Initiative 9

Graded PRA Working Group

• In the spring of 2021, a working group was formed to develop viable options to grade (right-size or customize) the PRA developed to support initial license applications (construction permits (CPs), operating licenses (OLs), DCs, SDAs, manufacturing licenses (MLs),

combined licenses (COLs)).

• The staff originally envisioned a three-phase process:

Phase 1 Develop a graded PRA approach Phase 2 Craft guidance Phase 3 Explore alternatives to PRA 10

Spring 2021 Working Definitions Graded PRA approach

• A process that uses bounding, conservative, and/or qualitative assessments to establish a PRAs scope, level of detail, degree of plant representation, and/or level of peer review commensurate with the licensing stage (which dictates the level of detail and finality of the information used to develop the PRA) and how the PRA will be used in risk-informed decision-making.

Graded PRA

• A PRA of appropriate degree of scope, level of detail, plant representation, and technical adequacy to support a specific advanced reactor licensing application.

• Note: Graded should not imply that a design is not yet complete -acceptance of a graded PRA could only be considered if a design is well understood and conservatively modeled.

Dose/consequence-based criterion

• A potential entry condition to enable a graded PRA that uses bounding, conservative, and/or qualitative assessments of the doses or consequences arising from potential unplanned release scenarios, without consideration of the release scenario likelihood.

This approach is being considered as a specific criterion for developing a graded PRA to adequately demonstrate that an applicant meets the intent of the Commissions Severe Accident Policy in an efficient and effective manner.

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A Change in Direction

• Based on feedback during the Advanced Reactor Stakeholders public meeting held 5/27/2021, the staff learned that industry concerns were largely directed at grading how PRA was used in the licensing process, rather than grading the technical content of the PRA itself.

• There was general recognition from industry that the non-light water reactor (non-LWR) PRA standard already offers opportunities to grade the content of the PRA.

Uses of Technical the PRA Content 12

Grading PRA Technical Content:

Use of Qualifiers

• The non-LWR PRA standard (ASME/ANS RA-S-1.4-2021) uses qualifiers to indicate when certain supporting requirements (SRs) apply.

• Examples:

o For operating plants, o For PRAs performed during the pre-operational stage,

• SRs without qualifiers apply to all life cycle stages.

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Grading PRA Technical Content:

The Risk Application Process Section 3 of the non-LWR PRA standard (ASME/ANS RA-S-1.4-2021) provides a process to grade the PRA:

• Describes required activities to establish the capability of a PRA to support a particular risk-informed application.

• Designed to be used during all life cycle stages.

Stage A Stage B Stage C Stage D Stage E

• Characterize plant

• Determine if PRA

• Determine if PRA

• Determine if PRA

• Use PRA to life cycle stage scope, level of standard has sufficient support the and PRA detail, and risk requirements are capability to application application metrics are sufficient for support application

• As needed, use

• Define site sufficient to application.

• If not, determine if supplementary characteristics support the

• Determine if deviations are analyses and

• Select PRA scope application insufficient significant to the requirements to and level of detail

• Upgrade PRA as requirements are applicant support the

• Select risk needed relevant to the

• Upgrade the PRA application significance application. as needed criteria

• Determine Capability Staff will provide PRA-related guidance for initial Category needed to for each part of licensing in TICAP and/or ARCAP guidance.

the PRA needed to support the 14 application

Focus on Understanding How and Why PRA is Used

• Since the May 2021 Advanced Reactor Stakeholders public meeting , the staff has further explored the scope of the PRA and how it is used in initial licensing.

• Significant effort has been invested in thoroughly understanding:

o The uses and role of the PRA in the licensing process, o Whether those uses and role could be adequately addressed with other tools/techniques/bounding assessments, and o How that information fits into the overall approach to licensing under Part 50, Part 52, and preliminary Part 53.

Required Uses Expected Uses

• Regulations

• Regulatory guides (RGs)

• Rulemakings

• Commission policy statements

• Commission staff requirement memoranda

• Standard review plans

• IAEA SSR-2/1 15

Why PRA? Post-TMI Recommendations

• ACRS letter* (May 16, 1979) o The ACRS believes that it is time to place the discussion of risk, nuclear and non-nuclear, on as quantitative basis as possible.

• Kemeny Report* (October 30, 1979) Recommendation #4:

o The [Presidential] Commission recommends that continuing in-depth studies should be initiated on the probabilities and consequences (on-site and off-site) of nuclear power plant accidents, including the consequences of meltdown.

• Rogovin Report* (NUREG/CR-1250, January 1980),

Recommendation #8:

o The best way to improve the existing design review process is by relying in a major way upon quantitative risk analyses, and by emphasizing those accident sequences that contribute significantly to risk.

• Available from the Idaho National Laboratory Knowledge Management Library for the Three Mile Island Unit 2 Accident of 1979 at https://tmi2kml.inl.gov/HTML/Page1.html 16

Advanced Reactors and PRA Regulations and Rulemakings Policy Statements 2019: Part 53 rulemaking 2009: Parts 50/52 lessons learned rulemaking; propose to add PRA requirements for CP and OL applications 10/14/2008: Rev. 2 to ARPS; cites SAPS, SGPS, and PRAPS 8/28/2007: Part 52 revised with requirement to submit a description of the PRA and its results 8/16/1995: PRA policy statement (PRAPS) 7/12/1994: Rev. 1 to ARPS; added reference to 4/18/1989: Part 52 issued with requirements metrification policy statement to meet § 50.34(f) and submit PRA 8/21/1986: Safety goal policy statement (SGPS) 7/8/1986: Advanced reactor policy statement (ARPS); cites SAPS and the forthcoming SGPS 8/8/1985: Severe accident policy statement (SAPS);

1/15/1982: TMI requirements added in § 50.34(f); use PRA to search for severe accident vulnerabilities perform PRA to seek improvements in the reliability of core and containment heat removal systems 17

Advanced Reactor Policy Statement (73 FR 60612; October 14, 2008)

• Comment labeled Toshiba-3 (pp. 60613-60614):

o Comment:

• Policy statement makes no mention of the use of PRA.

• Helpful to provide advanced reactor designers with interim guidance regarding NRC efforts for a risk informed, technology neutral licensing framework.

o Response:

• The NRC has established specific requirements related to the use of PRA in licensing new nuclear power plants, such as 10 CFR 52.47 and 10 CFR 50.71(h).

• The Commission has also issued policy statements on:

- Use of PRA in regulatory activities (PRAPS)

- Severe accidents regarding future designs and existing plants (SAPS)

• Page 60616: The Commission also expects that advanced reactor designs will comply with the Commissions SGPS.

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PRA Policy Statement (60 FR 42622; August 16, 1995)

• The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

• PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices.

• PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

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

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Severe Accident Policy Statement (50 FR 32138; August 8, 1985)

• Describes the policy the Commission intends to use to resolve safety issues related to reactor accidents more severe than design basis accidents (DBAs).

o Main focus is on the criteria and procedures the Commission intends to use to certify new designs for nuclear power plants.

o Severe nuclear accidents are those in which substantial damage is done to the reactor core whether or not there are serious offsite consequences.

• Policy for new plant applications:

o Comply with current regulations, including the TMI requirements in 50.34(f) o Demonstrate technical resolution of all applicable Unresolved Safety Issues and the medium- and high priority Generic Safety Issues, including a special focus on assuring the reliability of decay heat removal systems and the reliability of both AC and DC electrical supply systems.

o Complete a PRA and consider the severe accident vulnerabilities the PRA exposes along with the insights that it may add to the assurance of no undue risk to public health and safety.

o Complete a staff review of the design with a conclusion of safety acceptability using an approach that stresses deterministic engineering analysis and judgment complemented by PRA.

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Safety Goal Policy Statement (51 FR 28044; August 4, 1986 as corrected and republished at 51 FR 30028; August 21, 1986)

• Qualitative goals o Individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health.

o Societal risks to life and health from nuclear power plant operation should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risks.

• Quantitative objectives o The risk to an average individual in the vicinity of a nuclear power plant [one mile] of prompt fatalities that might result from reactor accidents should not exceed one-tenth of one percent (0.1 percent) of the sum of prompt fatality risks resulting from other accidents to which members of the U.S. population are generally exposed.

o The risk to the population in the area near a nuclear power plant [10 miles] of cancer fatalities that might result from nuclear power plant operation should not exceed one-tenth of one percent (0.1 percent) of the sum of cancer fatality risks resulting from all other causes.

• Proposed general performance guideline (large release frequency - LRF):

o Consistent with the traditional defense-in depth approach and the accident mitigation philosophy requiring reliable performance of containment systems, the overall mean frequency of a large release of radioactive materials to the environment from a reactor accident should be less than 1 in 1,000,000 per year of reactor operation.

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Large LWR Risk Surrogates

• NUREG-0880, Rev. 1, May 1983 (ML071770230),

provides numerical interpretation of the quantitative health objectives (QHOs):

o Individual early fatality risk (IEFR) 5E-7/ry o Individual latent cancer fatality risk (ILCFR) 2E-6/ry

• NUREG-1860, Vol. 2, December 2007 (ML080440215),

derives surrogate risk metrics for large light water reactors (LWRs):

o IEFR: Large early release frequency (LERF) 1E-5/ry o ILCFR: Core-damage frequency (CDF) 1E-4/ry The non-LWR PRA standard (ASME/ANS RA-S-1.4-2021) does not use risk surrogates such as CDF, LRF, or LERF!!!

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Large Release Frequency

• SRM-SECY-89-144, June 5, 1990, ML051660712:

o Within a particular design class (e.g., LWRs, LMRs, HTGRs) the same subsidiary objectives [risk surrogates] should apply to both current as well as future designsHowever, the Large Release Guideline relates to all current as well as future designs.

o The Commission believes that "adequate protection" is a case by case finding based on evaluating a plant and site combination and considering the body of our regulations.

• According to Forrest Remick (former Director of the Office of Policy Evaluation, ACRS member, and Commissioner), LRF was proposed to break a deadlock between the ACRS and the staff over the use of CDF and conditional containment failure probability (CCFP) as safety goals (see ML051660709):

o Staff wanted to only include CDF; ACRS wanted to also include CCFP.

o Dr. Remick received a call from the Executive Assistant to the then Commission Chairman [Palladino]. [H]e asked me what I thought of getting out of the deadlock by eliminating both the CDF and the idea of a CCFP by substituting a Large Release Guideline of 10-6 per reactor year as a surrogate. In response, I laid out the following steps: if one assumes a CDF of 10-4, a conditional probability of "core on the floor" of 10-1 (a probability still not accurately known), and a CCFP of 10-1 (which was bandied about at the time), then a Large Release Guideline of 10-6 per reactor year appeared to be a reasonable surrogate.

• SRM-SECY-12-0081, October 22, 2012, ML12296A158: The Commission has approved the staffs recommendation (Option 2C) to transition from large release frequency to large early release frequency (LERF) at or before initial fuel load and discontinue regulatory use of large release frequency (LRF) and conditional containment failure probability thereafter.

• SECY-13-0029, March 22, 2013, ML13022A207:

o The Commission acknowledged that analyses indicated that the cancer fatality QHO was not the more controlling objective and that, if the prompt fatality QHO is met, the cancer fatality risk generally would be much lower than the cancer fatality QHO.

o Recognizing that the prompt fatality QHO is more controlling, the staff worked to develop a large release definition that focused on LRF being a surrogate for the prompt fatality QHO.

o In 1993, the staff concluded that defining large release beyond a simple qualitative statement related to its 106 per reactor year release frequency (such as is currently contained in the safety goal policy statement) was neither practical nor required for regulatory or design purposes.

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Large Early Release Frequency

• The concept of LERF was originally developed by the Electric Power Research Institute, PSA Applications Guide, TR-105396, August 1, 1995.

o Use qualitative characteristics to identify large early release sequences o Avoids the need to perform source term and radiological consequence calculations

• Adopted by the NRC when RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, was developed in the late-1990s.

• Current definition provided in RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Rev. 3, December 2020:

o Large early release frequency (LERF) is defined as the sum of the frequencies of those accidents leading to rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of offsite emergency response and protective actions such that there is the potential for early health effects. (Such accidents generally include unscrubbed releases associated with early containment failure shortly after vessel breach, containment bypass events, and loss of containment isolation.)

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Risk-Informed Regulation Definitions (1 of 2)

• Risk triplet (Kaplan and Garrick*, SRM-SECY-98-144**):

o What can go wrong?

o How likely is it? , ,

o What are the consequences?

• Risk assessment (SRM-SECY-98-144): A systematic method for addressing the risk triplet as it relates to the performance of a particular system (which may include a human component) to understand likely outcomes, sensitivities, areas of importance, system interactions and areas of uncertainty.

• Risk insights (SRM-SECY-98-144): The results and findings that come from risk assessments.

• Risk-informed approach (SRM-SECY-98-144): A philosophy whereby risk insights are considered together with other factors to establish requirements that better focus licensee and regulatory attention on design and operational issues commensurate with their importance to public health and safety.

• Kaplan, S. and Garrick, B. J., On the Quantitative Definition of Risk, Risk Analysis, Vol. 1, Issue 1, March 1981.

• • NRC, Staff Requirements - SECY-98-144 - White Paper on Risk-Informed and Performance-Based Regulation, February 24, 1998, ML003752593.

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Risk-Informed Regulation Definitions (2 of 2)

• PRA o NRC online glossary*: A systematic method for assessing three questions that the NRC uses to define "risk." These questions consider (1) what can go wrong, (2) how likely it is, and (3) what its consequences might be. These questions allow the NRC to understand likely outcomes, sensitivities, areas of importance, system interactions, and areas of uncertainty, which the staff can use to identify risk-significant scenarios. The NRC uses PRA to determine a numeric estimate of risk to provide insights into the strengths and weaknesses of the design and operation of a nuclear power plant.

o RG 1.200**: An approach is considered to be a PRA when it (1) provides a quantitative assessment of the identified risk in terms of scenarios that result in undesired consequences (e.g., core damage or a large early release) and their frequencies and (2) comprises specific technical elements in performing the quantification.

o Draft RG 1.247: A risk assessment approach is considered to be a PRA when it (1) provides a quantitative assessment of the identified risk in terms of scenarios that result in undesired consequences (e.g., releases of radioactive material, radiological consequences) and their frequencies and (2) comprises specific PRA elements for quantifying risk.

• https://www.nrc.gov/reading-rm/basic-ref/glossary/probabilistic-risk-assessment-pra.html

• • NRC, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, RG 1.200, Rev. 3, December 2020, ML20238B871.

NRC, Acceptability of Probabilistic risk Assessment Results for Advanced Non-Light Water Reactor Risk-Informed Activities, draft trial use RG 1.247, September 3, 2021, ML21246A216. 26

Role of the PRA in Initial Licensing

• Traditional role o Consistent with previous DC and COL applications o Includes, but not limited to:

• Searching for severe accident vulnerabilities (SAPS)

• TMI requirement § 50.34(f)(1)(i), which under Part 52 requires LWR applicants to Perform a plant/site specific probabilistic risk assessment, the aim of which is to seek such improvements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant.

• Demonstrating that the QHOs are met (SGPS)

• Using PRA in the design process (PRAPS) o Previously referred to as PRA in a supporting role

• Enhanced role o Any use of PRA beyond its traditional role o Includes, but not limited to:

• Certain proposed required uses of PRA in preliminary 10 CFR Part 53 rule text (e.g.,

identifying licensing basis events (LBEs); classifying systems, structures, and components (SSCs); evaluating defense-in-depth (DID))

• Voluntary risk-informed applications (e.g., risk-managed technical specifications (TS), risk-informed fire protection) o Previously referred to as PRA in a leading role 27

28 PRA Requirements for Part 50 CP and OL Applicants

• PRA not currently required for CP or OL applicants.

• SRM-SECY-15-0002, ML15266A023: The Commission has approved the staffs recommendation to confirm that the Commissions guidance given in the "Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants" and other Commission direction identified by staff apply to new 10 CFR Part 50 power reactor applications in a manner consistent with 10 CFR Part 52 design and license applications.

• The Part 50/52 lessons learned rulemaking (NRC-2009-0196; RIN 3150-AI66) proposes to add PRA-related requirements for CP applicants, OL applicants, and OL holders like the PRA-related requirements for Part 52 applicants and COL holders.

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PRA Requirements for Part 52 Applicants

• § 52.47(a)(27) requires DC applicants to describe the design-specific PRA and its results.

• § 52.79(a)(46) requires COL applicants to describe the plant-specific PRA and its results.

o § 52.79(c)(1) requires COL applicants that reference an SDA to use and update the PRA information for the SDA to account for site-specific design information and any design changes or departures.

o § 52.79(d)(1) requires COL applicants that reference a DC to use and update the PRA information for the standard DC to account for site-specific design information and any design changes or departures.

o § 52.79(e)(1) requires that COL applicants that reference an ML to use and update the PRA information for the ML to account for site-specific design information and any design changes or departures.

• § 52.137(a)(25) requires SDA applicants to describe the design-specific PRA and its results.

• § 52.157(a)(31) requires ML applicants to describe the design-specific PRA and its results.

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PRA Requirements for Part 52 COL Holders

• § 50.71(h)(1) requires that:

o Each COL holder shall develop a Level 1 and a Level 2 PRA, and o The PRA must cover those initiating events and modes for which NRC-endorsed consensus standards on PRA exist 1 year before the scheduled date for initial loading of fuel.

• § 50.71(h)(2) requires that:

o Each COL holder shall maintain and upgrade the PRA o The upgraded PRA must cover initiating events and modes of operation contained in NRC-endorsed consensus standards on PRA in effect 1 year before each required upgrade, and o The PRA must be upgraded every 4 years until the permanent cessation of operations under 10 CFR 52.110(a).

• § 50.71(h)(3) requires that each COL holder shall, no later than the date on which the licensee submits an application for a renewed license, upgrade the PRA required by 10 CFR 50.71(h)(1) to cover all modes and all initiating events.

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Subpart C-Design and Analysis Requirements

§ 53.450 Analysis Requirements (3rd iteration preliminary rule text)

(a) Requirement to have a probabilistic risk assessment. A probabilistic risk Traditional use assessment (PRA) of each commercial of the PRA nuclear plant must be performed to identify potential failures, susceptibility to internal Severe accident and external hazards, and other contributing policy statement factors to event sequences that might challenge the safety functions identified in 53.230 and to support demonstrating that Safety goal policy each commercial nuclear plant meets the statement safety criteria of 53.220.

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Subpart C-Design and Analysis Requirements

§ 53.450 Analysis Requirements (3rd iteration preliminary rule text)

(b) Specific uses of analyses. The PRA, other generally accepted risk-informed approach for systematically evaluating engineered systems, or combination thereof must be used:

(1) In determining the licensing basis events, as described in 53.240, which must be considered in the design to determine compliance with the safety Enhanced criteria in Subpart B of this part.

use of the (2) For classifying SSCs and human actions according to their safety significance in accordance with 53.460 and for identifying the PRA environmental conditions under which the SSCs and operating staff must perform their safety functions.

(3) In evaluating the adequacy of defense-in-depth measures required in accordance with 53.250.

(4) To identify and assess all plant operating states where there is the PRA potential for the uncontrolled release of radioactive material to the scope environment.

• All sources (5) To identify and assess events that challenge plant control and safety

• All hazards systems whose failure could lead to the uncontrolled release of

• All modes radioactive material to the environment. These include internal events, such as human errors and equipment failures, and external events, such as earthquakes, identified in accordance with Subpart D of this part.

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Uses of the PRA 34

Observations

• A risk-informed approach may be based on risk insights developed from:

o A PRA (i.e., quantitative), or o A qualitative risk assessment

• PRA not used to support non-power production and utilization facility licensing:

o Not addressed in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors o Discussion with NRR/DANU staff

• Integrated safety analysis is required by 10 CFR Part 70 for certain licensees:

o Risk-informed process, but does not require development of a PRA

• The current preliminary rule text for Part 53:

o Builds on the traditional role of the PRA o Adds requirements that use PRA in an enhanced role Three potential licensing pathways

• Enhanced PRA approach

• Traditional PRA approach

• TIRIMA approach 35

Three Licensing Pathways 36

Development of How-To Guidance for the TIRIMA Approach (1 of 2)

• The staff intends to develop guidance for:

o Box A: Comprehensive search for initiating events o Box B: Comprehensive definition of event sequences o Box C: Decision guidance and entry conditions o Box D: Licensing basis event selection o Box E: Licensing basis event consequence analysis o Box F: Maximum accident identification o Box G: Conservative risk estimation

• Leverage existing guidance and studies such as, but not limited to:

o NUREG-1513, Integrated Safety Analysis Guidance Document o NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications o NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors o Occupational Safety and Health Administration regulations (29 CFR 1910.119),

standards, handbooks, and guidance o EPRI TR 3002011801, Program on Technology Innovation: Early Integration of Safety Assessment into Advanced Reactor Design - Preliminary Body of Knowledge and Methodology 37

Development of How-To Guidance for the TIRIMA Approach (2 of 2)

• Initial thoughts:

o Start with a blank sheet of paper o Use a combination of inductive and deductive methods o How much searching is enough? How do you know when you are finished?

o Focus on how plant design actually works vs. how plant design is supposed to work o Consolidate/group similar items o Be (very) careful when screening

• Process:

o Multi-disciplinary team effort o Independent review o Documentation

• Tell the story

• Capture assumptions and decisions 38

Initiating Event and Scenario Search (Mitigation Scenarios vs. Prevention Scenarios; MCA or MHA)

Step 1 Initiating Event Identification Step 1A Inductive Methods Quality Control Well-Established (AOO, DBA, BDBA, DBE, BDBE)

Methods Scenario Analysis HAZOP Categorization FMEA Step 2 Step 3 Step 4 Step 1B Verification Candidate Initiating Deductive Methods Operating Initiating Event Well-Established Experience Events Grouping Methods Analysis MLD FTA Step 1C Design/

Human-Induced Events Configuration Control Well-Established Method HRA 39

Initiating Events Analysis Preliminary Table of Contents

1. Introduction 4. Completeness of Initiating Event Lists 1.1 Background 4.1 Operating Experience Analysis 1.2 Purpose 4.2 Review of Generic Initiating Event Lists 1.3 Scope 4.3 Discussions with Knowledgeable

2. Acceptable Approach for Searching Design Personnel for Initiating Events 5. Interfacing with Configuration Management,

3. Identification of Initiating Events Quality Assurance, and Control Processes 3.1 Internal Hazards 5.1 Analysis Team Composition 3.2 External Hazards and Qualification 3.3 Concurrent Hazards 5.2 Independent Review 3.4 Definition of Initiating Events 5.3 Documentation 3.5 Categories of Initiating Events 6. Grouping of Initiating Events 3.5.1 Internal Plant Caused Initiators 6.1 Principle for Grouping 3.5.2 External Plant Caused Initiators 7. Determination of Initiating Event Frequencies 3.5.3 Human-Caused Initiators 7.1 Approaches to Quantification 3.5.4 Special Common-Cause Initiators 3.6 Methodology for Identifying Initiating Events 3.6.1 Systematic Approach 3.6.1.1 Inductive Methodologies 3.6.1.2 Deductive Methodologies 3.6.2 Human Reliability Analysis 40

Path Forward

• Revise preliminary rule text developed for the deterministic option to allow use of the TIRIMA approach

• Develop guidance to implement the TIRIMA approach:

- Leverage existing guidance to the extent possible

- Outline of guidance to conduct the systematic and comprehensive search for initiators (Box A) has been developed

- Developing outline of the entire guidance document

• Challenging technical issues:

- Technology-inclusive definition of severe accident

- Providing guidance to help applicants decide up-front if TIRIMA is a viable licensing pathway (TIRIMA entry conditions - Box C)

- Using TIRIMA to support SAMDA analysis 41

TIRIMA Approach Discussion 42

MEETING BREAK Meeting to resume in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 43

Subpart F - Requirements for Operation 44

Subpart F - Requirements for Operation

• Operational objectives

• Transition from construction/manufacturing to operation

• Maintaining capabilities and availability of SSCs

• Maintenance, repair, and inspection programs

• Design control

• Staffing, training, personnel qualifications, human factors requirements

• Programs 45

Subpart F - Programs

• Radiation protection

• Emergency preparedness (EP)

• Security programs

• Quality assurance

• Integrity assessment programs

• Fire protection

• Inservice inspection/inservice testing (ISI/IST)

• Criticality safety program

• Facility safety program (FSP) 46

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Overview of Topics for Discussion

• Shift Technical Advisor (STA) Position

• Load Following

• Certified Operators versus Licensed Operators

• Simulator Scope

• Training Program Review Guidance 47

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Shift Technical Advisor Position - Background

§ 53.753(f) requires a staffing plan describing numbers, positions, and qualifications of licensed operators and senior licensed operators, or certified operators, across all modes

• Must describe personnel providing support in plant operations, equipment surveillance and maintenance, radiological protection, chemistry control, fire brigades, engineering, security, and emergency response.

• Must also describe how the proposed licensed operator staffing, for plants requiring licensed operators, will be sufficient to assure that plant safety functions can be maintained; this must be supported by human factors engineering (HFE) analyses and assessments.

o Guidance for evaluating these staffing plans is being developed by the staff in the form of interim staff guidance (ISG) to be used in conjunction with NUREG-1791 48

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Shift Technical Advisor Position

• Staff have considered various approaches to the STA role under Part 53 and several alternatives will be discussed here

• Staff are receptive to feedback on different approaches as further iterations of preliminary rule language are developed

• In developing preliminary Part 53 requirements, the staff considered that the 1985 Policy Statement on engineering expertise on shift (50 FR 43621) stated that the STA was an interim measure until goals that included upgrading human-system interfaces (HSIs) and operator training were achieved o Current perspective is that the upgrades to HSIs and operator training envisioned within this Policy Statement will be the norm under Part 53 and driven by regulatory requirements

• Staff recognize that this represents a policy issue and intend to use the Part 53 rulemaking process for Commission engagement 49

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Shift Technical Advisor Position (contd)

Generic Part 53 elimination of STA could be justified on following bases:

• Licensed operator training requirements on knowledge and abilities to maintain plant safety functions; review criteria would confirm testing of reactor theory, thermodynamics, systems, and emergency operations

• State-of-the-art HFE required in all settings where operators are fulfilling plant safety functions; design requirements for HSIs requiring operators be provided with information on safety parameters, safety system status, important component status, and core damage states

• HFE-based analyses and assessments required to demonstrate how licensed operator staffing levels will maintain safety functions and support full range of tasks needed for safety (irrespective of an STA)

• Part 53 codifies DID principles under § 53.250 and requires (in part) DID use compensates for uncertainties in state of knowledge and modeling capabilities, and for personnel reliability and performance 50

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Shift Technical Advisor Position (contd)

Staff have also considered alternative of codification of an STA staffing requirement, with provision for justifying case-by-case position omission:

• In such an approach, the following definition would be provided in § 53.750(b):

Shift technical advisor means an individual possessing at least a baccalaureate in physical science, engineering, or engineering technology (or, alternately, a Professional Engineering license) and whose function is to provide independent on-shift engineering expertise, accident assessment, and technical advice to licensed operators at nuclear power plants.

• The staffing plan requirements of § 53.753(f) would be modified to include the following additional requirement for plants requiring licensed operator staffing:

A description of how the shift technical advisor position, as defined by § 53.750(b),

will be implemented during all plant conditions other than cold shutdown or refueling while shutdown or, alternatively, shall provide a justification for omission of the Shift Technical Advisor position that is supported by relevant human factors engineering-based analyses and assessments.

51

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Shift Technical Advisor Position (contd)

• Another alternative considered by staff would require that proposed staffing plans for facilities with licensed operators account for how on-shift engineering expertise will be provided o Such an approach would likely be accomplished by modifying the requirements of § 53.753(f) such that facilities requiring licensed operators would need to describe (in part) within staffing plans:

• how the numbers and positions of licensed operators provide assurance that plant safety functions can be maintained, and

• how on-shift engineering expertise will be provided for the on-shift crew

• Overall, the staff perspective currently remains that any STA position requirement would only apply to plants that require licensed operators 52

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Load Following - Background

§ 53.755, Conditions for Operations Staffing for Operating or Combined Licenses under this Part

• § 53.755(c) restricts control manipulations to licensed or certified operators.

• § 53.755(e) requires that operations (other than control manipulations) affecting reactor power level only occur while plant conditions are being monitored by a licensed or certified operator.

o However, load-following is permitted if one of the following can immediately refuse demands from the grid operator when they could challenge safe operation or if precluded by equipment conditions:

the actuation of an automatic protection system, an automated control system; or a licensed or certified operator.

53

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Load Following (contd)

• Current intent is to supplement rule with guidance (e.g., ARCAP ISG)

• While preliminary, the following illustrates general staff perspectives:

o Load following should be restricted to power levels where automation supports needed plant operations in order to avoid transients (e.g., if one feedwater train must be secured as part of reducing power <50% and automation cannot accomplish this, then 50% becomes the bottom of the load following envelope) o If the actuation of an automatic protection system will be relied upon to truncate/terminate load following, then any such protective actions should not be the same as those credited for core protection and should use more restrictive setpoints than those credited for safety purposes (avoids challenging limits) o Crediting operators requires an immediate capacity to take control o No restriction envisioned on facility types if an adequate design is demonstrated; usage of certified operators could be acceptable 54

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Certified Operators - Background

• § 53.755(a) requires facility licensees to have licensed operators unless they can meet criteria of § 53.755(b) to use certified operators

• § 53.755(b) contains the requirements that must be met in order to justify not using any licensed operators as a part of facility staffing o Two current proposals for criteria o The first proposal would require the following:

No human actions for event mitigation required to meet safety criteria, achieve safety functions, or provide DID PRA demonstrating the evaluation criteria for each event sequence can be met without human action for mitigation; LBE response not needing human action for SSCs to perform o The second proposal would require the design-basis accident safety criteria to be met without mitigation by human actions, active engineered features, or passive design features (except for only those passive features that can survive LBEs and not be defeated by credible human errors) 55

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Certified Operators - Background (contd)

• § 53.755(i) contains specific requirements for plants using certified operators o Certified operator are responsible for specified administrative functions.

o Certified operator staffing must always ensure continuity of responsibility for facility operations during the operating phase.

o Continuous monitoring of fueled units with the following capabilities:

• receiving plant operating data and parameters

• ability to immediately initiate a reactor shutdown

• ability to promptly dispatch ops/maintenance personnel

• the ability to implement any emergency plan responsibilities

• conducting reactivity manipulations that require human action 56

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Certified Operators - Background (contd)

• § 53.774, Issuance of Certificates (for Certified Operators) o Requires that facility licensees ensure that individuals to be certified:

complete either a high school diploma or GED complete the approved initial training program pass an initial operator certification examination demonstrate competence in conducting control manipulations meet medical condition requirements 57

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Certified Operators - Background (contd)

• § 53.775, Conditions of Certificates (for Certified Operators) o Requires facility licensees to ensure that certified operators:

only perform duties at facilities for which they are certified complete a continuing training program pass periodic continuing training examinations Complete biennial medical examinations maintain proficiency in accordance with the facility program 58

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Purpose of Certified Operator Alternative

• If a facility lacks an operator role in safety (e.g., an autonomous reactor design), then a key driver warranting federal licensing of individuals is removed (i.e., operator performance would not have a credible influence on public health and safety outcomes within that context) o Regardless of whether the operators were licensed, the facility itself would still be licensed by the NRC

• Important administrative job tasks that would remain still need to be accomplished by adequately qualified personnel.

o Precedent shows that similar administrative tasks have been fulfilled by non-licensed personnel, such as Certified Fuel Handlers 59

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Purpose of Certified Operator Alternative (contd)

• Staff are endeavoring to create a durable rule; part of this is accounting for the possibility of future advancements in safety

• While the framework for certified operators parallels that for licensed operators, there is less regulatory interface and more flexibilities; this brings with it the potential for cost savings on the part of the industry o Current industry burden estimates for staff fee-billable hours suggest such differences may reduce costs 60

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Certified Operators versus Licensed Operators Comparison of licensed and certified operator program components:

Program Component Licensed Operator Certified Operator NRC approval of training & Yes Yes exam programs required?

NRC approval of exams Yes No prior to administration?

NRC approval of operator Yes No applications & medical?

NRC approval of simulators Yes No for use in training & exams?

Required submittal of Yes No renewals & terminations?

NRC approval of Yes No examination waivers?

Flexibility for requalification No Yes training & exam periodicity?

61

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Certified Operators versus Licensed Operators (contd)

• Certified operators would be trained to conduct reactivity manipulations as part of their initial training program

• Due to the non-licensed nature of certified operators, the facility licensee retains ultimate accountability for operations o In contrast, licensed operators are individually accountable o Certified operators have a reduced operator safety role due to facility safety characteristics

• § 53.753(e) requires an operating experience program; staff envision guidance for licensed and certified operator training programs will include the incorporation of operating experience

• There are no known industry initiatives for maintaining Part 53 programs consistent within facility classes; staff are working to ensure compliance with Atomic Energy Act (AEA) requirements

• If a facility licensed with certified operators was determined to later need licensed operators (e.g., the § 53.755(b) criteria were no longer met), a safety issue may exist; NRC would have authority to modify the facilitys license as necessary 62

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Simulator Scope - Background

• § 53.765(e) establishes simulation facility requirements for plants with licensed operators and § 53.773(e) establishes less stringent simulation facility requirements for plants with certified operators o Full-scope simulators are not mandated; partial scope simulators may be acceptable, provided that the scope is adequate to meet intended usage; alternatives to simulators are possible as well o Simulation facilities for plants with licensed operators must be approved by the Commission if the facility licensee will rely upon them for training, experience requirements, or for initial or requalification examinations

• Equivalent approval not required for certified operator facilities o Must demonstrate that adequate simulator scope is provided to support HFE analyses/assessments in order to use a simulation facility for conducting these analyses/assessments 63

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Simulator Scope

• In developing preliminary rule language, staff reviewed Section 306 of the Nuclear Waste Policy Act (NWPA) and 52 FR 9453 which discussed implementation of the Acts simulator-related provisions:

o Flexibilities were historically provided to allow for potential use of the plant itself, and/or a plant-referenced simulator, and/or some other type of simulation device (such as a part-task or basic-principles simulator) for the conduct of the simulator portion of the operating test o The NRCs stated intent was not to permit the initiation of transients on the plant itself if used as a simulation facility; rather, the use of the plant was envisioned as an option that might be used in conjunction with another simulation device or devices, in lieu of a plant-referenced simulator

• Current perspective is that NWPA does not mandate NRC to require that plants have simulators, but instead requires regulations address the use of simulations in training; flexibility exists to allow the use of the actual plant to simulate tasks for training and operating test purposes without having a separate simulation facility (e.g., simulator) 64

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Simulator Scope (contd)

• Philosophical basis behind preliminary rule language is:

o Plant-referenced, full-scope simulators remain the preferred approach and would represent the best route for meeting Part 53 requirements

• Staff expect majority of Part 53 applicants will have them due to regulatory certainty and technology lowering the associated costs o Existing regulations do not strictly mandate plant-referenced, full-scope simulators either, but still adopted by all current power reactors o Part 53 rule language leaves alternatives to simulator usage (full-scope or otherwise), but the burden will be on the applicant to demonstrate how the following are supported:

• Licensed or certified operator training and exams; simulators used require sufficient scope and fidelity for operators to acquire and demonstrate knowledge and abilities needed for job duties.

• Experience requirements (i.e., reactivity manipulations)

• HFE analyses/assessments and HSI design testbed needs 65

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Training Program Review Guidance - Background

• § 53.765(a) requires initial licensed operator training programs to:

o Be based upon a systems approach to training (SAT) o Ensure that license applicants at the facility will possess the knowledge, skills, and abilities necessary to:

• protect the public health, and

• maintain design-specific plant safety functions o Be approved by the Commission prior to use

• § 53.765(c) requires facilities to establish requalification training programs for licensed operators. These programs must:

o Be based on SAT o Ensure that reactor operators and senior reactor operators maintain knowledge, skills, and abilities necessary to protect the public health and maintain those plant safety functions specific to the facility design.

o Be conducted for continuous period not to exceed 24 months.

o Be approved by the Commission.

66

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Training Program Review Guidance (contd)

• § 53.773(a) requires initial operator certification training programs to:

o Be based upon SAT o Ensure that certified operator trainees will possess the knowledge, skills, and abilities necessary to protect the public health o Be approved by the Commission prior to use

• § 53.773(c), requires continuing training programs for certified operators to:

o Be based upon SAT o Ensure that certified operators maintain the knowledge, skills, and abilities necessary to protect the public health o Be approved by the Commission prior to use

• §§ 53.780-781 addresses training requirements for other plant personnel o § 53.781, Training and Qualification Requirements Requires use of SAT Requires the training and qualification of supervisors, technicians, and other appropriate operating personnel to be provided for 67

Subpart F - Staffing, Training, Personnel Qualifications, and Human Factors Requirements Training Program Review Guidance (contd)

• Applicants might forgo training program accreditation, requiring:

o Staff to determine acceptability of proposed training programs o Staff assessment of ongoing conformance of facility licensee training programs with applicable regulatory requirements; o Staff inspections of other training programs required by § 53.781

• Staff will require guidance to support determinations regarding whether SAT is adequately applied; existing guidance (e.g., NUREG-1220) is dated o Working group established to develop this guidance for both initial and continuing training with objective to have an ISG developed no later than 2024 to support Part 53 rulemaking

• Effort underway to accelerate development to support near-term applicants applying under Parts 50/52 if needed 68

Subpart F - Requirements for Operation Discussion 69

Final Discussion and Questions 70

December 17th Agenda 9:30am - 9:35am Opening Remarks 9:35am - 11:30am Update on TICAP/ARCAP Guidance Document Developments 11:30am - 11:45am Break 11:45am - 1:00pm U.S. Nuclear Energy Institute / U.S. NIC Presentation: Letter of November 5, 2021 and Attachments 1:00pm - 2:00pm Lunch Break 2:00pm - 3:50pm U.S. Nuclear Energy Institute / U.S. NIC Presentation (continued) 3:50pm - 4:30pm Discussion 71

Advanced Reactors Overview of ARCAP Roadmap ISG and TICAP DG White Papers

Purpose

• Provide the ACRS Future Plant Designs Subcommittee an update on the Advanced Reactor Content of Application Project (ARCAP) and Technology Inclusive Content of Application Project (TICAP) guidance document developments since the last Subcommittee briefing in July of 2021

• Highlight some of the key draft white paper guidance with particular attention to portions of the guidance that maybe of interest to the ACRS

• Key documents associated with this presentation are available on NRCs Advanced Reactor ARCAP/TICAP public webpage (see:

https://www.nrc.gov/reactors/new-reactors/advanced/details.html#advRxContentAppProj) 73

ARCAP/TICAP

Background

• Previous ACRS Subcommittee briefings provided:

March 17, 2021, and July 21, 2021

• ARCAP/TICAP o Purpose is to develop technology-inclusive, risk-informed and performance-based application guidance o Developed to support 10 CFR Part 50, Part 52, and Part 53 applications Near-term need to develop guidance to support expected advanced reactor Part 50/52 applications using the licensing modernization project (LMP) process LMP process endorsed in 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 ARCAP/TICAP guidance will be revised as Part 53 proposed rule language is adjusted 74

ARCAP/TICAP

Background

• TICAP guidance o LMP process is used to define guidance for the content of major portions of the Safety Analysis Report (SAR)

LMP process uses risk-informed, performance-based approach to select LBEs, develop SSC categorization, identify special treatments for SSCs, and ensure DID adequacy o Industry developing key portions of TICAP guidance for NRC endorsement o Industry TICAP guidance will be supplemented by NRC staff-developed guidance as necessary

• ARCAP guidance o Broader in nature than TICAP and intended to cover guidance for SMR and non-LWR applications for a COL, CP, OL, DC, SDA, or ML.

o Encompasses TICAP guidance and provides supplemental and additional guidance for SAR and application requirements beyond the SAR.

75

ARCAP and TICAP - Nexus Outline SAR - Based on TICAP Guidance Additional Portions of Application

1. General Plant Information, Site

• Technical Specifications Description, and Overview of the Safety

• Technical Requirements Manual Case

• Quality Assurance Plan (design)

2. Methodologies and Analyses

• Fire Protection Program (design)

3. LBE Analysis

• Quality Assurance Plan

4. Integrated Evaluations (construction and operations)

5. Safety Functions, Design Criteria, and

• Emergency Plan SSC Safety Classification

• Physical Security Plan Audit/inspection of Applicant Records

6. Safety-Related (SR) SSC Criteria and

• Special nuclear material (SNM)

• Calculations Capabilities physical protection program

• Analyses

7. Non-safety-related with special treatment

• SNM material control and

• P&IDs (NSRST) SSC Criteria and Capabilities accounting plan

• System Descriptions

8. Plant Programs

• Design Drawings

• Cyber Security Plan

• Design Specs

• Fire Protection Program Additional SAR Content -Outside the Scope

• Procurement Specs (operational) of TICAP

• Probabilistic Risk Assessment

• Radiation Protection Program

9. Control of Routine Plant Radioactive

• Offsite Dose Calculation Manual Effluents, Plant Contamination, and Solid

• Inservice inspection/Inservice Waste testing (ISI/IST) Program

10. Control of Occupational Doses

• Environmental Report

11. Organization and Human-System

• Site Redress Plan Considerations

• Exemptions, Departures, and

12. Post-construction Inspection, Testing and Variances Analysis Programs

• FSP (under consideration for Part 53 applications)

• SAR structure based on clean sheet approach

• Additional contents of application outside of SAR are still under discussion. The above list is draft and for illustration purposes only. 76

ARCAP/TICAP

Background

• Status of ARCAP ISG Draft White Papers ARCAP ISG Title Date Accession No.

Review of Risk-Informed, Technology-Inclusive Dec 2, 2021 ML21336A702 Advanced Reactor Applications - Roadmap Chapter 2, Site Information July 6, 2021 ML21189A031 Chapter 9, Control of Routine Plant Radioactive July 6, 2021 ML21189A033 Effluents, Plant Contamination and Solid Waste Chapter 10, Control of Occupational Doses July 6, 2021 ML21189A035 Chapter 11, Organization and Human-System Nov 5, 2021 ML21309A020 Consideration Chapter 12, Post Construction Inspection, Testing and Oct 21, 2021 ML21294A266 Analysis Program Licensing Modernization Project-based Approach for May 10, 2021 ML21133A490 Developing Technical Specifications Risk-Informed, Performance-Based Fire Protection Sept 10, 2021 ML21253A134 Program (for Operations)

Risk-Informed ISI/IST Programs Aug 4, 2021 ML21216A051 77

ARCAP/TICAP

Background

• Status of TICAP Guidance Documents TICAP Title Date Accession No.

NEI 21-07, Revision 0, Technology Inclusive Aug 30, 2021 ML21250A378 Guidance for Non-Light Water Reactors Safety Analysis Report Content for Applicants Utilizing NEI 18-04 Methodology RG Draft White Paper, Guidance for a Technology- Dec 2, 2021 ML21336A697 Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Advanced Reactors

• TICAP guidance documents being revised based on continuing interactions with stakeholders

• December 14, 2021, public workshop

• During this meeting, NEI 21-07, Revision 0-B was discussed (ADAMS Accession No. ML21343A292)

• Planning for another public workshop in mid-January 2022 78

Advanced Reactor Content of Application Project - Overview 79

ARCAP - High Level Overview

• ARCAP Roadmap ISG

• Proposes a 12-chapter SAR guidance structure

• Guidance for first 8 SAR chapters references the TICAP guidance

• Chapters 9, 10, 11, and 12 point to individual ISGs developed for each SAR chapter

• Includes pointers to draft white papers, guidance under development or to be developed for portions of the application outside the SAR

• Examples of guidance that the staff has developed: TS, Risk Informed ISI/IST, and Fire Protection for Operations

• Examples of high-level guidance embedded in the Roadmap ISG:

Technical Requirements Manual, Quality Assurance Plan, Fire Protection (design), and Offsite Dose Calculation Model

• Examples of guidance being considered: security, emergency planning, material control and accountability, financial qualification and cyber security 80

ARCAP - High Level Overview

• ARCAP Roadmap ISG (continued)

• Includes several appendices:

• Appendix C on preapplication engagement guidance

• Based on white paper discussed extensively during advanced reactor public stakeholder meetings

• Purpose of Appendix C is to capture the white paper guidance in a durable product that will have the benefit of a formal public comment period

• Appendix D on Analysis of Applicability of NRC Regulations to non-LWRs (plan to include later)

• Based on white paper discussed extensively during advanced reactor public stakeholder meetings (ADAMS Accession No. ML21175A287)

• Purpose of Appendix D is to capture the white paper guidance in a durable product that will have the benefit of a formal public comment period 81

ARCAP - High Level Overview

• ARCAP Roadmap ISG (continued)

• Includes several appendices (continued):

• Appendix E on CP guidance

• Three parts to draft CP guidance

• Common portion applicable to both LWRs and non-LWRs

• Will be updated to be consistent with LWR CP ISG when it is issued

• Portion applicable to LMP based approach - point to TICAP CP guidance

• Portion applicable to CP guidance outside the scope of TICAP 82

ARCAP - High Level Overview

• ARCAP Chapter 2: Site Information

• Supplements information in the SAR that is outside the scope of LMP.

• Intent is to limit the amount of material in SAR Chapter 2 to what is necessary for establishing safety significant design parameters and for performing the safety analysis, along with its supporting bases.

• If necessary, additional supporting information (e.g., historical records, geological data, etc.) could be documented in a separate report available for audit.

• Section 2.6 includes a process for establishing the ground motion response spectrum using the Senior Seismic Hazard Analysis Committee guidance

• ARCAP Chapters 9 and 10: Normal Effluents and Occupational Dose

• Applies a performance-based approach for level of detail of information provided in the SAR 83

ARCAP - High Level Overview

• ARCAP Chapter 11: Organization and Human-System Consideration

• Developed to support near-term Part 50 and 52 applications with a more traditional concept of operations

• Guidance will be updated later to include concepts discussed as part of the Part 53 proposed rulemaking effort

• Human Factors Engineering

• NRC staff identified a need to provide guidance in this area to supplement LMP and the associated pending TICAP guidance

• LMP provides human factors insights but provides limited guidance on how to develop a HFE program

• ARCAP Chapter 11 ISG covers HFE information that would support NRC findings 84

ARCAP - High Level Overview

• ARCAP Chapter 11 - Organization and Human-System Consideration (continued)

• Operator Licensing

• Proposed guidance extends beyond what would be expected in an application

• Centrally located guidance to provide a holistic approach to operator licensing

• May eventually split out guidance not specifically associated with the content of an application

• Guidance includes areas such as:

• Description and qualification of simulator used to administer initial operator licensing examinations

• Use of simulator for operations training experience and examinations during construction

• Operator license issuance prior to fuel load

• Operator Staffing guidance includes areas such as:

• Option of providing technical basis for control room staffing in conjunction with control room configuration that would support capturing the requirements necessary in a DC rulemaking

• Provide technical basis that could support a future exemption from §§ 50.54(m) and/or 50.54(k) requirements 85

ARCAP - High Level Overview

• ARCAP Chapter 12 - Post Construction Inspection, Testing and Analysis Program

• Intended to provide guidance to the NRC staff regarding application content that would support making the finding that the applicant has met the applicable Part 50 and Part 52 regulations

• ISG differentiates between 10 CFR Part 52 applicants that must include inspections, tests, analysis and acceptance criteria (ITAAC) and 10 CFR Part 50 applications that are not required to include ITAAC.

• Requirements to describe preoperational testing and initial operations in OL and COL applications are contained in 50.34(b)(6)(iii) and 52.79(a)(28), respectively.

• Provides guidance for:

• post-construction inspection, preoperational testing (i.e., tests conducted following construction and construction-related testing, but prior to initial fuel load), analysis verification, and

• initial startup testing (i.e., tests conducted during and after initial fuel load, up to and including initial power ascension).

86

ARCAP - High Level Overview

• ARCAP TS Guidance

• The text in the 10 CFR 50.36 regulations for TS content require adaptation to correlate to the analysis and outputs of the risk-informed LMP approach described in NEI 18-04.

• 10 CFR 50.36 requirements for safety limits, limited safety system settings, and limiting condition for operations Criteria 1 through 3 involve challenges to the integrity of a fission product barrier.

• To evaluate the acceptability of risk-informed TS for advanced reactors, this ISG correlates the 10 CFR 50.36 text with appropriate NEI 18-04 process analysis/outputs. These analysis/outputs include:

• required safety functions

• SR SSCs

• frequency-consequence (F-C) target

• 10 CFR 50.34 dose limits 87

ARCAP - High Level Overview

• ARCAP Fire Protection for Operations

• 10 CFR 50.48(a) requires that each operating nuclear power plant have a fire protection plan that meets the requirements of either 10 CFR Part 50, Appendix A, Criterion 3 for LWRs or the applicants proposed principal design criteria (PDC) that have been deemed acceptable by the NRC.

• Although 10 CFR 50.48(c) - NFPA 805 - does not apply to non-LWRs concepts associated with this risk-informed approach are included in the draft ISG

• The scope of this ISG addresses the review of the application content regarding the fire protection program for operations including application descriptions of:

• Management policy and program direction and the responsibilities of those individuals responsible for the program/plans implementation.

• The integrated combination of procedures and personnel that will implement fire protection program activities.

88

ARCAP - High Level Overview

• ARCAP Risk-Informed ISI/IST

• ISG purpose is to facilitate the review of advanced reactor applications that use a risk-informed approach to developing their ISI/IST programs

• ISG guidance requires the use of risk information from a plant-specific PRA that is in conformance with an NRC endorsed PRA standard

• For advanced LWRs, guidance on how to risk-inform ISI/IST already exists (RG 1.175, and 1.178) and is used in the ISG

• 10 CFR Part 50 contains only general requirements (e.g., 50.34(b)(6)(iv))

related to non-LWR ISI/IST programs, although ASME has recently issued Section XI, Division 2 - NRC has reviewed and issued DG-1383 for public comment on endorsement of ASME Section XI, Division 2

• For non-LWR ISI, applicants are expected to use the risk information from their plant-specific PRA to identify the piping, reactor coolant boundary, pressure retaining and passive components and their supports to be included in the program, along with other components whose failure could prevent a safety function from being accomplished 89

ARCAP - High Level Overview

• ARCAP Risk-Informed ISI/IST (continued)

• For non-LWR IST, applicants are expected to use the risk information from their plant-specific PRA and associated design reviews to identify the active valves, pumps and dynamic restraint devices and the passive components with active safety functions to be included in the program.

• For non-LWR ISI, the ISG is based upon the applicant using the requirements in ASME Section XI, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants (which is the subject of draft RG-1383).

90

Technology Inclusive Content of Application Project - Overview 91

TICAP - High Level Overview

• TICAP guidance

• Goal is to develop technology-inclusive guidance that proposes an optional formulation of advanced reactor application content that is based on a risk-informed, performance-based approach for demonstrating that plant safety meets the underlying intent of the current requirements

• Guidance is intended to increase efficiency of developing and reviewing an application

• Scope is governed by the LMP process to facilitate a systematic, technically acceptable, and predictable approach for developing key portions of a designs SAR

• LMP provides process for identifying LBEs, determining SSC categorization, establishing special treatments for SSCs, and ensuring DID adequacy 92

TICAP - High Level Overview

• TICAP SAR Structure

• Chapter 1 - General Plant and Site Description, And Overview of The Safety Case

• The information in this chapter should allow the reviewer to obtain a basic understanding of the overall facility, such as the type of permit, license, certification or approval requested, the number of plant units, a brief description of the proposed plant location, and the type of advanced reactor being proposed.

• Chapter 2 - Methodologies and Analyses

• An important part of the design process for reactor designs is the identification of events that could challenge key safety functions and layers of defense against the release of radioactive materials.

Therefore, a key part of the review of an advanced reactor application is the selection of LBEs.

93

TICAP - High Level Overview

• TICAP SAR Structure

• Chapter 3 - Licensing Basis Events

• The information in this chapter should describe the systematic and reproducible process and methodology used to select the LBEs, and the specific analysis and evaluation of the selected LBEs for the proposed design. The analysis in this section is primarily described in terms of event sequences comprised of an initiating event, the plant response to the initiating event (which includes a sequence of successes and failures of mitigating systems) and a well-defined end state. The consequences from LBEs are expressed as dose at the exclusionary boundary and compared to the F-C curve in LMP documentation.

• Chapter 4 - Integrated Evaluations

• The information in this chapter should describe the integrated risk of all LBEs selected for the proposed design and evaluated against three cumulative risk targets and the DID evaluation.

94

TICAP - High Level Overview

• TICAP SAR Structure

• Chapter 5 - Safety Functions, Design Criteria, and Systems, Structures, and Components Classification

• As part of the LMP process, LBEs are generally defined in terms of successes and failures of SSCs that perform safety functions and are modeled in the PRA. Therefore, the PRA safety functions are those functions responsible for the prevention and mitigation of an unplanned radiological release from any source within the plant.

• The information in this chapter should describe the proposed PDC necessary to ensure that the important to safety SSCs provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public

• The information in this chapter should describe the approach for designating SSC classifications 95

TICAP - High Level Overview

• TICAP SAR Structure

• Chapter 6 - Safety-Related Systems, Structures, and Components Criteria and Capabilities

• The information in this chapter should leverage the analysis performed for the SR SSCs in Chapter 5 of NEI 21-07 and describe in further detail the criteria, capabilities and special treatment of all SR SSCs.

• Chapter 7 - Non-Safety-Related Special Treatment (NSRST) Systems, Structures, and Components Criteria and Capabilities

• The information in this chapter should describe the regulatory design and special treatment requirements for SSCs classified as NSRST. NSRST SSCs are relied upon to perform risk-significant functions. 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. SSCs considered necessary for ensuring DID adequacy are also categorized as NSRST.

96

TICAP - High Level Overview

• TICAP SAR Structure

• Chapter 8 - Plant Programs

• The information in this chapter should provide information on those plant programs relied upon to provide special treatment to SR and NSRST SSCs that are part of the affirmative safety case. The information should provide an overview of the special treatment programs, addressing the purpose, scope, and performance objectives as well as applicability to SSCs. The information for the programs should provide reasonable assurance that 1) reliability and performance targets are met, and 2) safety-significant uncertainties are addressed. Program areas could include human factors, quality assurance, startup testing, and equipment qualification, among others.

97

TICAP - High Level Overview

• TICAP - PDC

• Development of an LMP-based approach for developing proposed PDCs has been a subject of ongoing stakeholder interactions with the most recent occurring at the 12/14 TICAP public workshop

• Issues discussed include:

• Applicability of General Design Criteria (GDC) in 10 CFR part 50, Appendix A, including their scope, to non-LWR advanced reactor applicants

• Additional guidance available to advanced reactor applicants for developing proposed PDC

• The possible need for exemptions to applicable regulations for proposed PDC developed using the LMP methodology 98

TICAP - High Level Overview

• Summary of TICAP - PDC Discussions

• PDC are required to be proposed by applicants for the following:

10 CFR 50.34(a)(3) for CPs 10 CFR 52.79(a)(4) for COLs 10 CFR 52.47(a)(3) for DCs 10 CFR 52.137(a)(3) for SDAs 10 CFR 52.157(a) for MLs

• PDC are a means to meet the requirements of the AEA, Section 182 for inclusion in license applications of 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 or production of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public.

99

TICAP - High Level Overview

• Summary of TICAP - PDC Discussions (continued)

• GDC are applicable to LWRs (minimum requirements) and provide guidance to applicants for construction permits in establishing principal design criteria for other types of nuclear power units.

• Advanced Reactor Design Criteria (ARDC) developed by the NRC in RG 1.232 are intended to provide insight into the staffs views on how the underlying safety bases for the GDC could be applied to address non-LWR design features. As noted in RG 1.232, the development of the ARDC was an important first step to address the unique characteristics of non-LWR technology but the NRC recognizes the future benefits to risk-informing the non-LWR design criteria and determining the role of such criteria within a new regulatory framework.

100

TICAP - High Level Overview

• Summary of TICAP - PDC Discussions (continued)

• The NRC position on the requirement for proposed PDC is that it includes the scope of PDC described in the regulations as well as in the regulatory and judicial history.

The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

• PDC are particularly important for CP applications since CP applicants are required to provide less information, comparatively speaking, and the information that is provided is preliminary.

• Proposed PDC play a significant role in supporting the NRCs finding that there is reasonable assurance that safety questions will be satisfactorily resolved and that the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public. 101

TICAP - High Level Overview

• Summary of TICAP - PDC Discussions (continued)

• Proposed PDC determined to be necessary for a non-LWR design and submitted in an application under 10 CFR Part 50 or Part 52 should be as comprehensive in scope as the GDC and ARDC (i.e., establish the necessary design, fabrication, construction, testing, and performance requirements).

• Non-LWR applicants proposing PDC that do not fully address the scope of PDC (i.e., design, fabrication, construction, testing, and performance requirements) will need to request exemptions from the applicable regulations.

• Non-LWR applicants must provide supporting information that justifies to the NRC how their design meets their proposed PDC and how their proposed PDC demonstrate reasonable assurance of safety.

• NRC believes that it is feasible for applicants for CPs, COLs, DCs, SDAs and MLs to provide justification for an exemption by ensuring that the elements of the PDC scope not fully addressed in their proposed PDC are included in their application.

102

TICAP - High Level Overview

• Summary of TICAP - PDC Discussions (continued)

• TICAP guidance document NEI 21-07 proposed an approach to supplement PDC focused on SR SSCs with proposed Complementary Design Criteria (CDC) that focus on NSRST SSCs.

• In fitting the LMP approach to developing PDC and CDC into the Part 50 and Part 52 regulatory framework, the NRC concluded that both PDC and CDC would need to be relied on the NRC to make its regulatory finding.

• NRC suggested that a two-tiered PDC approach would comply with the regulations (i.e., PDC Type A for functions performed by SR SSCs and PDC Type B for functions performed by NSRST SSCs).

• NRC expects further stakeholder interactions at the next TICAP public workshop in January 2022.

103

Next Steps - Future Milestones TICAP Near-Term Milestones Target Date Update of NRC Draft Guidance Early December 2021 Documents Continuation of Discussion of NRC TBD draft Exceptions, Clarifications, and Additions (possibility of future draft industry or staff documents)

NEI 21-07, Revision 1 TBD Issuance of TICAP draft RG and Early Calendar ARCAP ISG for public comment Year 2022 104

TICAP/ARCAP Guidance Document Development Discussion 105

MEETING BREAK Meeting to resume in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 106

Nuclear Energy Institute /

U.S. Nuclear Industry Council Presentation 107

Final Discussion and Questions 108

Acronyms and Abbreviations ACRS Advisory Committee for Reactor Safeguards Division of Advanced Reactors and Non-DANU Power Production and Utilization Facilities Agencywide Documents Access and ADAMS Management System DBA Design basis accident AEA Atomic Energy Act DBE Design basis event ANS American Nuclear Society DC Design certification AOO Anticipated operational occurrence DG Draft regulatory guide Advanced Reactor Content of Application DID Defense-in-depth ARCAP Project D-RAP Design reliability assurance program ARDC Advanced reactor design criteria EP Emergency preparedness Advanced reactor policy statement (Policy EPRI Electric Power Research Institute ARPS Statement on the Regulation of Advanced Reactors; 73 FR 60612; October 14, 2008) ESP Early site permit ASME American Society of Mechanical Engineers F-C Frequency-consequence BDBA Beyond design basis accident FMEA Failure modes and effects analysis BDBE Beyond design basis event FR Federal Register CCFP Conditional containment failure probability FRN Federal Register Notice CDC Complementary design criteria FSAR Final safety analysis report CDF Core damage frequency FSP Facility safety program CFR Code of Federal Regulations FTA Fault tree analysis COL Combined license GDC General design criteria CP Construction permit 109

Acronyms and Abbreviations HAZOP Hazard and operability LWA Limited work authorization HFE Human factors engineering LWR Light water reactor HRA Human reliability analysis MCA Maximum credible accident HSI Human-system interface MHA Maximum hypothetical accident HTGR High temperature gas cooled reactor ML Manufacturing license IAEA International Atomic Energy Agency MLD Master logic diagram IEFR Individual early fatality risk NEI Nuclear Energy Institute ILCFR Individual latent cancer fatality risk NFPA National Fire Protection Association ISG Interim staff guidance NGO Non-governmental organization ISI Inservice inspection non-LWR Non-light water reactor IST Inservice testing NRC Nuclear Regulatory Commission Inspections, tests, and acceptance ITAAC NRR Office of Nuclear Reactor Regulation criteria LB Licensing basis NSRST Non-safety-related special treatment LBE Licensing basis event NUREG U.S. NRC technical report designation LERF Large early release frequency NWPA Nuclear Waste Policy Act LMP Licensing Modernization Project OFR Office of the Federal Register LMR Liquid metal cooled reactor OL Operating license LRF Large release frequency OMB Office of Management and Budget 110

Acronyms and Abbreviations ONT Other new technology SECY Office of the Secretary P&ID Piping and instrumentation diagrams Safety goal policy statement (Safety Goals for the Operation of Nuclear Power Plants; PDC Principal design criteria SGPS 51 FR 28044; August 4, 1986; as corrected PRA Probabilistic risk assessment and republished at 51 FR 30028; August 21, 1986)

PRA policy statement (Use of Probabilistic Risk Assessment Methods SMR Small modular reactor PRAPS in Nuclear Regulatory Activities; 60 FR SNM Special nuclear material 42622; August 16, 1995)

Supporting requirement (NLWR PRA PSA Probabilistic safety assessment SR standard)

QHO Quantitative health objective SR Safety-related RG Regulatory guide SRM Staff requirements memorandum RIM Reliability and Integrity Management SSC Systems, structures, and components RIN Regulation identifier number SSR Specific safety requirement (IAEA)

Severe accident mitigation design STA Shift technical advisor SAMDA alternative Technology Inclusive Content of Severe accident policy statement TICAP Application Project (Severe Reactor Accidents Regarding SAPS Technology-inclusive, risk-informed, Future Designs and Existing Plants; 50 TIRIMA FR 32138; August 8, 1985) maximum accident SAR Safety analysis report TMI Three Mile Island SAT Systems approach to training TR Technical report SDA Standard design approval TS Technical specifications 111

Background Slides 112

First Principles See: SECY-18-0096, Functional Containment Performance Criteria for Non-Light-Water-Reactors, and INL/EXT-20-58717, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities 113

Integrated Approach Siting near densely populated Functional areas EP for SMRs Containment and ONTs Licensing Modernization Project Insurance and Liability Environmental Reviews Consequence Based Security 114

Background

• Nuclear Energy Innovation and Modernization Act (NEIMA; Public Law 115-439) signed into law in January 2019 requires the NRC to complete a rulemaking to establish a technology-inclusive, regulatory framework for optional use for commercial advanced nuclear reactors no later than December 2027 o (1) ADVANCED NUCLEAR REACTORThe term advanced nuclear reactor means a nuclear fission or fusion reactor, including a prototype plant with significant improvements compared to commercial nuclear reactors under construction as of the date of enactment of this Act, 115

Severe Accidents

• Severe Accident Policy Statement o Although in the licensing of existing plants the Commission has determined that these plants pose no undue risk to public health and safety, this should not be viewed as implying a Commission policy that safety improvements in new plant designs should not be actively sought. The Commission fully expects that vendors engaged in designing new standard (or custom) plants will achieve a higher standard of severe accident safety performance than their prior designs.

• 10 CFR 52.47(a)(23) o For light-water reactor designs, a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass

• NUREG-1226 (Development and Utilization of the NRC Policy Statement on the Regulation of Advanced Nuclear Power Plants; Executive Summary) o (4) While the Final Policy Statement encourages innovative reactor designs and safety criteria, the review of advanced reactor designs will still require satisfactory consideration of the Commission's regulations, regulatory guides and other guidelines, such established and developing criteria as the defense-in-depth philosophy, standardization, the Commission's safety goal and severe accident policies, and applicable industry codes and standards.

116

Backup Slides

Technology-Inclusive Content of Application (TICAP) and Advanced Reactors Content of Application (ARCAP)- Nexus to Part 53 Part 53- Proposed Structure Subpart H - Licenses, Cert, and Approvals This subpart is envisioned to address requirements for initial applications for licenses, certifications, or approvals. The A- General Provisions subpart will support the licensing processes available under the Part 50 and Part 52 frameworks. Assessment and update of manufacturing licenses is possible.

B- Tech-Incl Safety Requirements C- Design and Analysis Req.

D- Siting ARCAP-Guidance for Content of Application Guidance (Roadmap)

E- Const. and Manufacturing TICAP - LMP-based Guidance under F- Operations portions of FSAR that are Development examples:

related to:

• Fuel Qualification

• LBEs

• G- Decommissioning

• Safety classification ASME Section III, Division 5

• Defense in Depth

• EP rulemaking H- Licenses, Cert, and Approvals

• Security ARCAP specific chapter Rulemaking guidance - examples:

I- Maintaining/Revising LB Info

• Site information Existing

• ARCAP chapters 9, Regulatory 10, 11, and 12 Guidance J- Administrative requirements Note: The illustrated content structure for Part 53 (including Subpart H) is part of ongoing work and subject to change. 118

Industry-led TICAP NRC-led TICAP

• Focused on portions of the

• NRC plans to issue a RG license application Safety endorsing TICAP that also Chapters 1-8 Analysis Report (SAR) for non- focuses on providing exceptions LWR designs related to the and/or clarifications on TICAP.

TICAP Licensing Modernization Project (LMP)-based affirmative safety

• Include supplemental TICAP case. guidance for areas outside of the LMP for the first 8 SAR chapters. Examples includes FSAR site information, ASME Section III, Division 5 Chapters 9-12 NRC-led SAR Guidance ARCAP*

• Focused on remaining portions of the license application Safety Analysis Report (SAR) not related to LMP.

• ARCAP ISGs under development that include an overall roadmap ISG and separate ISGs for FSAR Chapters 2, 9, 10, 11, and 12 Additional

• For example:

Contents of o Technical Specifications, o QA Plan, o Fire Protection, etc.

Application

• Staff plans to issue an ARCAP Roadmap ISG that would provide pointers to various guidance documents developed/issued.

119

Key Part 53 Guidance by Subpart Subpart A: General Provisions Existing / Ongoing Guidance Additional Guidance N/A Subpart B: Safety Criteria Existing / Ongoing Guidance Additional Guidance Further explanation of criteria and N/A structure in the Statements of Consideration Subpart C: Design and Analysis Existing / Ongoing Guidance Additional Guidance NEI 18-04 & RG 1.233 (LMP)

ANS/ASME-RA-S-1.4 (Non-LWR ISG on PRA for Initial Licensing PRA Standard)

RG 1.247 Endorsing Non-LWR PRA Industry PRA Peer Review Guidance Standard and NEI Peer Review for Non-LWRs (NEI 20-09)

Guidance ANS/ASME Standards (ASME III-5, Application of Analytical Margins ASME XI-II)

Treatment of Chemical Hazards Fuel Qualification (NUREG-2246)

RG 1.232 (ARDCs)

Subpart D: Siting Requirements Existing / Ongoing Guidance Additional Guidance SECY-20-0045/RG 4.7 External Hazard Updates N/A Risk-Informed Seismic Design; ANS 2.26 120

Key Part 53 Guidance by Subpart Subpart E: Construction and Manufacturing Existing / Ongoing Guidance Additional Guidance Manufacturing Guidance N/A QA Alternatives Subpart F: Operations SSCs Existing / Ongoing Guidance Additional Guidance Technical Specifications Special Treatment NEI 18-04 & RG 1.233 (LMP)

Maintenance, Repair & Inspection Facility Safety Program Personnel Existing / Ongoing Guidance Additional Guidance DRO Paper/preliminary ISG Concept of Operations Programs Existing / Ongoing Guidance Additional Guidance Emergency Preparedness EPZ Draft Final Rule, RG 1.242 Security Programs (e.g., FFD, Access Radiation Protection (ARCAP) Authorization, Cyber Security)

Integrity Assessment Program 121

Key Part 53 Guidance by Subpart Subpart G: Decommissioning Existing / Ongoing Guidance Additional Guidance N/A N/A Subpart H: Licensing Existing / Ongoing Guidance Additional Guidance TICAP Manufacturing Licenses ARCAP Subpart I: Maintaining Licensing Basis Existing / Ongoing Guidance Additional Guidance 50.59 Equivalent N/A FSAR/PRA Updates Subpart J: Administrative/Misc.

Existing / Ongoing Guidance Additional Guidance Reporting Requirements N/A Financial/Liability 122