ML25003A061

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NRC Staff Presentation Slides for January 14, 2025, ACRS Regulatory Rulemaking, Policies and Practices Subcommittee Meeting: Update - 10 CFR Part 53, Risk - Informed, Technology- Inclusive Regulatory Framework for Commercial Nuclear Plants
ML25003A061
Person / Time
Issue date: 01/13/2025
From: Robert Beall
NRC/NMSS/DREFS/RRPB
To:
References
RIN 3150-AK31, NRC-2019-0062, 10 CFR Part 53
Download: ML25003A061 (1)
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ACRS Subcommittee on Regulatory Rulemaking, Policies, and Practices:

I nfo rm at io n B rief in g o n Up d ates to P ro p o s ed 1 0 CF R Part 5 3, "Ris k-I nfo rm ed,

Tec h n o lo gy-I n c lu sive Regu lato r y F ram ewo rk fo r Ad van c ed Reac to rs "

J a n u a r y 1 4, 2 0 2 5

List of Speakers Anders Gilbertson - NRR Technical Lead Bill Reckley-NRR Technical Lead Jesse Seymour - NRR Plant Operator Licensing & Human Factors 2

ADAMS Accession No. ML25003A061

Proposed Rule

Past ACRS Interactions on Part 53 16 ACRS meetings during development of draft proposed rule in 2020, 2021, and 2022 4 Interim Letter Reports Final Letter Report dated November 22, 2022 (ML22319A104)

NRC Staff Response dated February 10, 2023 (ML22341A047) 3

Key Rulemaking Documents SECY-23-0021, Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors, dated March 1, 2023 (ADAMS ML21162A093)

In SRM-SECY-23-0021, dated March 4, 2024 (ADAMS ML24064A047), the Commission approved, in part, the NRC staffs draft proposed rule with exceptions and clarifications 4

Part 53 Final Rule Milestones

  • Close of public comment period: February 28, 2025
  • ACRS Interactions: will work with ACRS staff to set up dates for late 2025 and early 2026
  • Final Rule to the Commission: May 2026
  • Final Rule Published: NEIMA Deadline - December 2027 5

Recent & Ongoing Activities 6

Part 53 Licensing Frameworks Framework A o PRA in foundational role o Uses risk metrics o Functional design criteria for SSCs Framework B o Traditional use of risk insights o Principal design criteria o Includes AERI approach Subpart A - General Provisions Subpart X - Enforcement Subpart B - Safety Requirements Subpart C - Design Requirements Subpart D - Siting Subpart E - Construction/Manufacturing Subpart F - Operations Subpart G - Decommissioning Subpart H - Application Requirements Subpart I - License Maintenance Subpart J - Reporting Subpart K - Quality Assurance Subpart N - Siting Subpart O - Construction/Manufacturing Subpart P - Operations Subpart Q - Decommissioning Subpart R - Application Requirements Subpart S - License Maintenance Subpart T - Reporting Subpart U - Quality Assurance Draft Part 53 Proposed Rule (SECY-23-0021) 7

Sections 53.000 and 53.010 1)

Remove Framework B from Part 53 and provide new options 2)

Replace references to QHOs with comprehensive risk metrics 3)

Allow flexibility in PRA acceptability determinations 4)

Revise requirements related to as low as reasonably achievable (ALARA) 5)

Remove facility safety program 6)

Explain process for ongoing evaluations of external hazards in preamble 7)

Include requirement for design experience program 8)

Include provisions for factory fuel loading and engage stakeholders on possible operational testing of fueled manufactured reactors 9)

Harmonize consideration of security-related events within security and emergency preparedness requirements

10) Replace Subpart K QA requirements with reference to Appendix B to Part 50
11) Remove safety objectives section
12) Include question on processes for similar designs at multiple sites
13) Consider suggested edits in Commission vote sheets
14) Provide final version of Federal Register Notice within six months
15) Consider administrative rulemaking for potential errors in Parts 50 and 52 8

Staff Requirements Memorandum Summary

Proposed Part 53 Licensing Framework Part 53 Organization Subpart A General Provisions Subpart B Technology-Inclusive Safety Requirements Subpart C Design and Analysis Requirements Subpart D Siting Requirements Subpart E Construction and Manufacturing Requirements Subpart F Requirements for Operation Subpart G Decommissioning Requirements Subpart H Licenses, Certifications, and Approvals Subpart I Maintaining and Revising Licensing-Basis Information Subpart J Reporting and Other Administrative Requirements Subpart M Enforcement Framework B and related references removed Subpart K removed (added references to Appendix B to Part 50)

SRM-RELATED DELTA 9

Part 53 Structure - Project Life Cycle Plant Documents (Systems, Procedures, etc.)

Analyses (Prevention, Mitigation, Compare to Criteria)

Plant/Site (Design, Construction, Configuration Control)

LB Documents (Applications, SAR, TS, etc.)

Project Life Cycle Subpart B Subparts H & I Requirements Definition Safety Criteria Safety Functions Defense in Depth Other Subpart J Admin & Reporting Clarify Controls and Distinctions Between Subpart A General Provisions Subpart M Enforcement Retirement Staffing &

Human Factors Configuration Control Surveillance Maintenance Operation Construction/

Manufacturing Construction Siting Design and Analysis Design Features Analysis Requirements Subpart C Subpart D Subpart E Subpart G Special Treatment External Hazards Site Characteristics Population Considerations Ensuring Capabilities Factory Manufacturing Programs (Security, EP)

Subpart F Decom Funding License Termination 10

Subpart A General provisions

§ 53.015 Scope.

§ 53.020 Definitions.

§ 53.040 Written communications.

§ 53.050 Deliberate misconduct.

§ 53.060 Employee protection.

§ 53.070 Completeness and accuracy of information.

§ 53.080 Specific exemptions.

§ 53.090 Standards for review.

§ 53.100 Jurisdictional limits.

§ 53.110 Attacks and destructive acts.

§ 53.115 Rights related to special nuclear material.

§ 53.117 License suspension and rights of recapture.

§ 53.120 Information collection requirements: OMB approval.

11

New or Revised Terminology Compared to Parts 50 & 52 (§ 53.020)

  • Event categories & related terms
  • Commercial nuclear plant/reactor
  • Consensus code or standard
  • Construction
  • Defense in depth
  • Functional design criteria
  • Licensing basis information
  • Safety classification categories
  • Programmatic controls
  • Special treatment 12 Subpart A General provisions

§ 53.210 Safety criteria for design-basis accidents.

§ 53.220 Safety criteria for licensing-basis events other than design-basis accidents.

§ 53.230 Safety functions.

§ 53.240 Licensing basis events.

§ 53.250 Defense-in-depth.

§ 53.260 Normal operations.

§ 53.270 Protection of plant workers.

13 Subpart B Technology-inclusive safety requirements

§ 53.210 Safety criteria for design-basis accidents.

  • Design features and programmatic controls provided such that the identification and analyses of design-basis accidents (DBAs) demonstrate that the calculated offsite doses are below established reference values

§ 53.450(f) Analysis of design-basis accidents.

  • DBAs address possible challenges to the safety functions required to be identified by § 53.230 and include events that, if not terminated, have the potential for exceeding the safety criteria in

§ 53.210.

  • DBAs analyzed using deterministic methods that address event sequences from initiation to a safe stable end state and assume only the SR SSCs and human actions addressed by the requirements of Subpart F to perform the safety functions
  • The analysis must conservatively demonstrate compliance with the safety criteria in § 53.210.

15 Subpart B Technology-inclusive safety requirements

§ 53.220 Safety criteria for licensing-basis events other than design-basis accidents.

Design features and programmatic controls provided such that the identification and analysis of licensing-basis events (LBEs) other than DBAs demonstrate the following:

a)

Plant SSCs, personnel, and programs provide the necessary capabilities and maintain the necessary reliability to address LBEs other than DBAs and provide measures for defense in depth, and b)

The analysis of risks to public health and safety resulting from LBEs other than DBAs under § 53.450(e) includes comprehensive risk metrics that satisfy associated risk performance objectives that are acceptable to the NRC and provide an appropriate level of safety.

SRM-RELATED DELTA 16 Subpart B Technology-inclusive safety requirements

§ 53.450(e)

Analysis of licensing-basis events other than design-basis accidents.

  • The analyses must use insights from a PRA in combination with other generally accepted approaches for systematically evaluating engineered systems to identify and analyze equipment failures and human errors.
  • The analysis of LBEs other than DBAs must include definition of evaluation criteria for each event or specific categories of LBEs to determine the acceptability of the plant response to the challenges posed by internal and external hazards to provide an appropriate level of safety.
  • The analyses of LBEs other than DBAs must address event sequences from initiation to a defined end state and be used in combination with other engineering analyses to demonstrate that the functional design criteria required by § 53.420 provide sufficient barriers to the unplanned release of radionuclides to satisfy the evaluation criteria defined for each LBE other than DBAs, to satisfy the safety criteria specified in accordance with § 53.220 and provide defense in depth as required by § 53.250.
  • The methodology used to identify, categorize, and analyze LBEs must include a means to identify event sequences deemed significant for controlling the risks posed to public health and safety.

17 Subpart B Technology-inclusive safety requirements

Comprehensive risk metrics and associated risk performance objectives

  • Consist of proposed plant risk metric or set of proposed risk metrics that approximate the total, overall risk from the facility and that address the range of possible plant configurations and associated internal and external hazards to the extent practicable.
  • The associated risk performance objectives are preestablished, indicative values of the comprehensive risk metrics that are used as part of risk-informed decision-making.
  • The methodology for developing and using proposed comprehensive risk metrics and associated risk performance objectives is defined by the proposed requirements for analyses in § 53.450.

SRM-RELATED DELTA 18 Subpart B Technology-inclusive safety requirements

§ 53.230 Safety functions.

(a) The primary safety function is limiting the release of radioactive materials from the facility and must be maintained during normal operation and for LBEs over the life of the plant.

(b) Additional safety functions needed to support the retention of radioactive materials during LBEssuch as controlling reactivity, heat generation, heat removal, and chemical interactionsmust be identified for each commercial nuclear plant.

(c) The primary and additional safety functions are required to satisfy the safety criteria defined in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470, and must be fulfilled by the design features, human actions, and programmatic controls specified throughout this part.

19 Subpart B Technology-inclusive safety requirements

§ 53.240 Licensing-basis events.

(a) Licensing-basis events must be identified for each commercial nuclear plant and analyzed under § 53.450 to demonstrate that the safety requirements in this subpart have been satisfied.

(b) The identified LBEs, ranging from anticipated event sequences to very unlikely event sequences, must collectively address combinations of malfunctions of plant SSCs, human errors, facility hazards, and the effects of external hazards.

(c) The analysis of LBEs must (1) Include analysis of one or more DBAs under § 53.450(f);

(2) Confirm the adequacy of design features and programmatic controls needed to satisfy the safety criteria defined in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470; and (3) Establish related functional requirements for plant SSCs, personnel, and programs.

§ 53.020 Definitions.

Licensing-basis events means a collection of event sequences considered in the design and licensing of the commercial nuclear plant. Licensing-basis events are unplanned events and include anticipated event sequences, unlikely event sequences, very unlikely event sequences, and DBAs.

20 Subpart B Technology-inclusive safety requirements

§ 53.250 Defense in depth.

(a) Measures must be taken for each commercial nuclear plant to ensure appropriate defense in depth is provided to compensate for uncertainties in the analysis of the safety criteria such that there is reasonable assurance that the safety criteria in this subpart are met over the life of the plant.

(b) The uncertainties that must be addressed under paragraph (a) of this section include those related to the state of knowledge and modeling capabilities, the ability of barriers to limit the release of radioactive materials from the facility during LBEs other than DBAs, the reliability and performance of plant SSCs and personnel, and the effectiveness of programmatic controls.

(c) The safety analysis may not rely upon a single engineered design feature, human action, or programmatic control, no matter how robust, to address the range of LBEs other than DBAs.

21 Subpart B Technology-inclusive safety requirements

§ 53.260 Normal operations.

Holders of licenses to operate commercial nuclear plants under this part must control public doses and dose rates in unrestricted areas from normal plant operations to meet the requirements in 10 CFR part 20.

§ 53.270 Protection of plant workers.

Holders of licenses to operate commercial nuclear plants under this part must control occupational doses to meet the requirements in 10 CFR part 20.

SRM-RELATED DELTA 22 Subpart B Technology-inclusive safety requirements

SRM-RELATED DELTA Requirements related to radiation protection programs 53.260 OL/COL holders meet 10 CFR part 20 (public doses) 53.270 OL/COL holders meet 10 CFR part 20 (plant workers) 53.425 Define design features and functional design criteria ALARA design objective of 10 mrem TEDE annual dose 53.430 Define design features and functional design criteria 53.450(g)(3)

Analysis of expected releases and doses to the public 53.850 Radiation protection program 53.1645 Reports of radiation exposure to the public 53.1239(a)

(DC)

Design features supporting normal operations How programmatic controls support meeting requirements Design features supporting the protection of plant workers How programmatic controls support meeting requirements 53.1209(b)

(SDA) 53.1279(a)

(ML) 53.1309(a)

(CP) 53.1369 (OL)

Design features supporting normal operations Radiation protection program Design features supporting the protection of plant workers Radiation protection program 53.1416(a)

(COL) 23 Subpart B Technology-inclusive safety requirements

§ 53.400 Design features for licensing-basis events.

§ 53.410 Functional design criteria for design-basis accidents.

§ 53.415 Protection against external hazards.

§ 53.420 Functional design criteria for licensing-basis events other than design-basis accidents.

§ 53.425 Design features and functional design criteria for normal operations.

§ 53.430 Design features and functional design criteria for protection of plant workers.

§ 53.440 Design requirements.

§ 53.450 Analysis requirements.

§ 53.460 Safety categorization and special treatments.

§ 53.470 Maintaining analytical safety margins used to justify operational flexibilities.

§ 53.480 Earthquake engineering.

24 Subpart C Design and analysis requirements

Part 53 Hierarchy 25 Subpart C Design and analysis requirements

§ 53.400 Design features for licensing-basis events.

  • Design features must be provided such that, when combined with corresponding human actions and programmatic controls, the plant will satisfy the safety criteria and ensure that safety functions are fulfilled during LBEs.

§ 53.410 Functional design criteria for design-basis accidents.

§ 53.415 Protection against external hazards.

  • Safety-related (SR) SSCs must be protected against or must be designed to withstand the effects of external hazards up to the design-basis external hazard levels.

§ 53.420 Functional design criteria for licensing-basis events other than design-basis accidents.

26 Subpart C Design and analysis requirements

§ 53.440 Design requirements.

(a)

  • Demonstrate functional design criteria via analysis, test, etc.;
  • Evaluate operating, design and construction experience (b)

Consensus codes and standards acceptable to NRC (c)

Materials qualified for conditions (d)

Evaluate possible degradation mechanisms (e)

Design and locate to minimize probability and effects of fires and explosions (f)

Consider safety and security together during design process (g)

Subcritical condition during normal operations and after LBE (h)

Long-term cooling during normal operations and after LBE (i)

Design, analysis, staffing and programs cover all units, inventories (j)

Physical barrier(s) maintained assuming aircraft impact (k)

Control risk from chemical hazards of licensed material (l)

Minimize contamination to facilitate eventual decommissioning (m)

Criticality monitoring (alternative to § 70.24)

(n)

Consider human factors, functional analysis and function allocation SRM-RELATED DELTA 27 Subpart C Design and analysis requirements

§ 53.450 Analysis requirements.

(a)

Requirement to have a probabilistic risk assessment (PRA)

(b)

Specific uses of analyses using PRA in combination with other generally accepted approaches for systematically evaluating engineered systems (LBEs, classification, defense in depth)

(c)

Maintenance and upgrade of analyses (d)

Qualification of analytical codes.

(e)

Analyses of licensing-basis events other than design-basis accidents.

  • Evaluation criteria for each event or specific categories of LBEs
  • Means to identify event sequences significant for controlling risks (f)

Analysis of design-basis accidents.

  • Deterministic methods from initiation to a safe stable end state (g)

Other required analyses.

  • Fire protection
  • Aircraft impact
  • Doses to members of the public 28 Subpart C Design and analysis requirements

PRA Acceptability

  • Development, use, and maintenance of a PRA would be a key component in the proposed analysis requirements.
  • The PRA, together with other techniques, would have required uses such as -

o identify and categorize LBEs, o

classify SSCs, and o

evaluate defense in depth.

  • Consistent with the current state of practice, acceptability of a PRA would be assessed based on the required uses of the PRA and the needs and scope of the application.

o Consensus PRA standards would not be applied as a strict checklist of requirements for PRA acceptability determinations under the Part 53 proposed rule.

  • NRC guidance on non-LWR PRA acceptability is currently available, which includes NRC-endorsed processes on the use of consensus PRA standards and PRA peer review.

SRM-RELATED DELTA/CLARIFICATION 29 Subpart C Design and analysis requirements

§ 53.500 General siting and siting assessment.

§ 53.510 External hazards.

§ 53.520 Site characteristics.

§ 53.530 Population-related considerations.

§ 53.540 Siting interfaces.

30 Subpart D Siting requirements

§ 53.600 Construction and manufacturing - scope and purpose.

§ 53.605 Reporting of defects and noncompliance.

§ 53.610 Construction.

§ 53.620 Manufacturing.

31 Subpart E Construction and manufacturing requirements

§ 53.620 Manufacturing.

  • Management and control o

Provides programmatic and organizational requirements o

Supports compliance with the design and analysis requirements in subpart C

  • Manufacturing activities o

ML holder has the authority to establish controls at facility(s) o Manufacturing processes must be performed in accordance with the ML and the referenced codes and standards o

A post-manufacturing inspection and acceptance process

  • Control of radioactive materials
  • Fuel loading
  • Transportation
  • Acceptance and installation at final place of operation SRM-RELATED DELTA*
  • Delta between ACRS Review and SECY-23-0021 32 Subpart E Construction and manufacturing requirements

§ 53.620(d) Fuel loading.

  • A manufacturing license may include authorizing the loading of fresh (unirradiated) fuel into a manufactured reactor under Part 70
  • Specifies required protections to prevent criticality o At least two independent physical mechanisms in place, each of which is sufficient to prevent criticality assuming optimum neutron moderation and neutron reflection conditions
  • Commission finding that a manufactured reactor module in required configuration is not in operation SRM-RELATED DELTA 33 Subpart E Fuel loading for manufactured reactor module

§ 53.620(d) Fuel loading.

  • Holders of these Part 70 licenses must comply with the requirements of Subpart H to Part 70
  • Procedures, equipment, and personnel required by the Part 70 license must be in place before the receipt of SNM at the manufacturing facility
  • The loading or unloading of fresh fuel into or from a manufactured reactor and any changes to the configuration of reactivity control and prevention systems for the fueled manufactured reactor must be performed by a certified fuel handler meeting the requirements in Subpart F of this part
  • For a manufactured reactor that is to be loaded with fresh fuel before transport to the place of operation, the ML must specify that transportation will be in accordance with Parts 71 and 73 of this chapter
  • Security requirements 34 Subpart E Fuel loading for manufactured reactor module

Proposed § 53.620(d) would allow and establish requirements for the loading of fuel into a manufactured reactor at the manufacturing facility for transport to a site with a combined license Included question in Federal Register Notice

  • Prepared and released preliminary draft material (i.e., not complete NRC management or legal review) to support discussions ADAMS Accession No. ML24344A037 Public meeting - January 8, 2025 Consideration of comments received Factory Testing of Fueled Manufactured Reactors Staff Requirements Memorandum (SRM)-SECY-23-0021 8.

The staff should include factory fuel load provisions in the proposed rule. The staff should work with stakeholders following publication of the proposed rule to develop regulatory text that would also allow a holder of a manufacturing license to accomplish operational testing on a fueled manufactured reactor at the factory prior to delivery to the site where it will ultimately be used.

SRM-RELATED DELTA 35

  • Should Part 53 include provisions for the testing of fueled manufactured reactors in the manufacturing facility?

o What would be both practical and safe?

o What tests are expected to collect data on fuel or other structures, systems, and components (SSCs)?

  • What would be appropriate limits on operations?

o Power levels o Durations (limit creation of byproduct material)

  • What requirements could be revised given limitations on operation?

o Licensing basis events, aircraft impact assessments, external hazards (seismic)

Question in Federal Register Notice 36

  • What regulations would be appropriate for manufacturing facility?

o Construction (proposed § 53.610) o Operations (proposed §§ 53.710 and 53.715) o Personnel (proposed § 53.730)

  • What licensing mechanism(s) should be considered for in-factory testing of manufactured reactors?

o License for each manufactured reactor o License for manufacturing facility/multiple manufactured reactors o Inspections, tests, analyses, and acceptance criteria (ITAAC)

Question in Federal Register Notice 37

§53.700 Operational objectives.

(1) Each holder of an OL or COL under this part must maintain the capabilities, availability, and reliability of plant SSCs to ensure that the safety functions identified in §53.230 will be performed if called upon during licensing-basis events (LBEs).

(2) Each holder of an OL or COL under this part must ensure that plant personnel have adequate knowledge and skills to perform their assigned duties that support the performance of the safety functions identified in §53.230.

(3) Each holder of an OL or COL under this part must implement plant programs sufficient to ensure that the safety functions identified in

§53.230 will be performed if called upon during normal operations and LBEs.

Subpart F Organization of Sections

§ 53.710 - § 53.720 SSCs

§ 53.725 - § 53.830 Personnel

§ 53.845 - § 53.910 Programs 38 Subpart F Requirements for operation

§§ 53.725 - 53.830: General staffing, training, personnel qualifications, and human factors requirements Sections 53.725 - 53.830 include the following key areas:

Content of application requirements (§ 53.730) o Human factors engineering (HFE) has a safety function focus (versus generic application to a control room) o Facility-specific staffing plans and engineering expertise Conditions of license for facility licensees (§ 53.740) o Allows for automatic load following o

Addresses online refueling oversight Operator licensing requirements for specifically-licensed Senior Reactor Operators (SRO) and Reactor Operators (ROs) (§§ 53.760-53.795) o Addresses use of customized operator licensing programs o

Allows facility licensees to administer license exams Requirements for Generally Licensed Reactor Operators (GLROs)

(§§ 53.800- 53.820) o Establishes criteria for self-reliant-mitigation facilities.

o Contains the general license for GLROs Plant staff training requirements (§ 53.830) 39 Subpart F Requirements for operation

Other highlights from §§ 53.725 - 53.830 Load following is permitted under 53.740 if one of the following is immediately capable of refusing unsafe demands:

1) an automatic protection system that utilizes setpoints more conservative than those otherwise credited for the purposes of reactor protection; or 2) an automated control system; or 3) intervention of an RO, SRO, or GLRO.

Prescriptive timeframes for establishing training programs are no longer used; 53.830 requirement is based on needs.

Staffing plans are proposed by applicants under 53.730, with HFE analyses and performance-based tests being used to determine operator numbers, qualifications, and locations (approved plan then becomes license condition).

o A flexible requirement for engineering expertise is used in lieu of traditional Shift Technical Advisor staffing.

o A location-neutral approach is taken to operator staffing; for example, control room staffing is not prescribed.

o Self-reliant-mitigation facilities do not require these HFE-based staffing analyses and, instead, only have GLRO oversight and engineering expertise requirements.

40 Subpart F Requirements for operation

Self-reliant-mitigation facilities and GLROs

§ 53.800 - Criteria for self-reliant-mitigation facilities o

No human actions to meet radiological consequence criteria, address LBEs, or provide for adequate DID o

Safety functions not allocated to human action o

Reliance upon robust and highly reliable safety features

§ 53.805 - Facility licensee requirements for GLROs o

Facilities must continue to meet the criteria of 53.800 (failure would be a reportable unanalyzed condition)

§ 53.810 - General license for GLROs o

Grants similar level of administrative authority as an SRO o

No application needs to be submitted for GLRO licensing o

Individuals operating under license subject to conditions o

License can still be suspended on an individual basis

§ 53.815 - GLRO training, exams, & proficiency o

SAT-based training program is required o

Uses customized, Commission-approved exam programs o

After approval, GLRO programs are facility-administered o

Facilities determine requalification exam periodicity o

Simulation facilities do not require Commission-approval

§ 53.820 - Cessation of individual applicability 41 Subpart F Requirements for operation

Subparts G, H, I, J, and M

  • Subpart G Decommissioning Requirements
  • Subpart H - Licenses, Certifications, and Approvals
  • Subpart I Maintaining and Revising Licensing-Basis Information
  • Subpart J Reporting and Other Administrative Requirements
  • Subpart M Enforcement 42

Wrap Up Discussion and Questions 43

ACRS Advisory Committee on Reactor Safeguards ADAMS Agencywide Documents Access and Management System AEA Atomic Energy Act of 1954 AERI Alternative Evaluation for Risk Insights ALARA as low as is reasonably achievable ARCAP Advanced Reactor Content of Application Project CFR Code of Federal Regulations COL combined license COL-TMR combined license for testing of manufactured reactors CP construction permit DBA design-basis accident DC design certification DID defense in depth EP emergency planning FR Federal Register FRN Federal Register Notice GLRO generally licensed reactor operator GEIS Generic Environmental Impact Statement HFE human factors engineering ITAAC Inspections, Tests, Analyses, and Acceptance Criteria LB licensing basis LBE licensing-basis event LWR light-water reactor ML manufacturing license mrem millirem NEIMA Nuclear Energy Innovation and Modernization Act NRC U.S. Nuclear Regulatory Commission 44 Acronyms

NRR Office of Nuclear Reactor Regulation NUREG U.S. Nuclear Regulatory Commission technical report designation OL operating license OMB Office of Management and Budget PRA probabilistic risk assessment QA quality assurance QHO quantitative health objectives RG Regulatory Guide RO Reactor Operator SAR safety analysis report SAT systems approach to training SDA standard design approval SECY Office of the Secretary SNM special nuclear material SR safety-related SRM Staff Requirements Memorandum SRO Senior Reactor Operator SSC structure, system, or component TICAP Technology-Inclusive Content of Application Project TEDE total effective dose equivalent TS technical specifications 45 Acronyms

Backup Slides 46

  • White Paper organized to provide:

o Description o Draft preliminary rule text (§ 53.1480)

Combined license for testing manufactured reactors (COL-TMR)

Commission findings on operating states*

  • See also FRN Question 7. under Part 53, Subparts E and HManufacturing Licenses
7. Some stakeholders have suggested that a fueled manufactured reactor with appropriate protections against criticality should not be categorized as a utilization facility under NRC regulations or Section 11cc. of the AEA.

The NRC is seeking comment on possible approaches where the NRC could find that a fueled manufactured reactor would not be a utilization facility, the basis for such a finding, and the potential benefits of and potential issues with such a finding.

White Paper (ML24344A037)

Provided to support discussions

Should not be interpreted as official agency positions 47

  • White Paper basic approach o Building from proposed § 53.620(d)

Unirradiated fuel loaded (manufacturing license; Part 70) o Limit introduction of byproduct material Radioactive inventory, decay heat Assume in-factory conditions for licensing-basis events Limited consequences assumed in categorizing hazards o Consideration of various regulations and licenses Part 53 (Manufacturing license, combined license)

Part 70 (Special nuclear material)

Part 30 (Byproduct material)

Parts 71, 73, 74 and others as needed White Paper

Provided to support discussions

Should not be interpreted as official agency positions 48

  • Selected White Paper examples (technical requirements) o Limit power level ( 5% rated thermal power (commercial))

o Limit inventory (indirectly via defining restrictive safety criteria (Part 20 annual dose))

o Licensing-basis events Identified for reactor as tested (e.g., fresh fuel)

Mitigated without reliance on human actions Consistent with use of generally licensed reactor operators (GLROs)

Design features of manufacturing facility and manufactured reactor o Holder of manufacturing license ensures testing does not adversely affect downstream activities (storage, transport, deployment)

White Paper

Provided to support discussions

Should not be interpreted as official agency positions 49

Selected White Paper examples (technical requirements) o Possible alternatives mentioned in draft paper:

§ 53.440(j) (aircraft impact) would not apply

§§ 53.415, 53.480, and 53.510 (external hazards) would not apply Based on limited consequences, commercial codes

§ 53.610 (construction) would apply to portions of manufacturing facility

§§ 53.710 and 53.715 (SSC configuration control) would apply for testing

§§ 53.730(a) through (e) (human factors) would apply

§ 53.730(f) (staffing plan) would be supplemented Test Engineer, Reactor Engineer, GLRO

§§ 53.870 and 53.880 (ISI/IST, Integrity assessment) would not apply

Alternate decommissioning funding requirements (such as Parts 70 and 30) might apply White Paper

Provided to support discussions

Should not be interpreted as official agency positions 50

Selected White Paper examples (licensing construct) o COL-TMR Applicable to portions of manufacturing facility and each manufactured reactor (1 through n)

Updates to the ITAAC schedule under § 53.1449(a) and ITAAC closure notifications under § 53.1449(c) may address multiple manufactured reactors that are under fabrication or planned to be fabricated under the ML and tested under the COL-TMR Conforming changes (e.g., § 53.620(d))

Testing criteria for first reactor Testing criteria for subsequent reactors Criteria for final place of operation Manufacturing facility ITAAC (COL-TMR)

§§ 53.710 and 53.715 n/a Manufactured reactor ITAAC (COL-TMR (incl ML))

ITAAC (COL-TMR (incl ML))

ITAAC (COL (incl ML))

White Paper

Provided to support discussions

Should not be interpreted as official agency positions 51

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