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NRC Presentation 10 CFR Part 53, Subpart I - Section 53.1322, Evaluating Changes to Facility as Described in Final Safety Analysis Reports
	ML21252A124
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Issue date:	09/15/2021
From:	Robert Beall; NRC/NMSS/DREFS/RRPB
To:	;
	Beall, Robert
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	NRC-2019-0062, RIN 3150-AK31
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10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors

10 CFR Part 53, Subpart I -Section 53.1322, Evaluating changes to facility as described in final safety analysis reports.

September 15, 2021

1 Agenda

1:00pm - 1:15pm Welcome / Introductions / Logistics / Goals

1:15pm - 2:30 pm Subpart I - Section 53.1322, Evaluating changes to facility as described in final safety analysis reports -

Paragraph (a)

2:30pm - 3:00pmSubpart I - Section 53.1322, Evaluating changes to facility as described in final safety analysis reports -

Paragraphs (b) & (c) and Section 53.1333 (Programs) 3:00pm - 3:15pmBreak 3:15pm - 4:30pm Discussion of Other Recently Released Iterations of Preliminary Rule Language; Subpart F - Programs

4:30pm - 5:00pm Additional Public Comments/Closing Remarks

2 Welcome/Introductions

Welcome:

John Segala, Office of Nuclear Reactor Regulation (NRR)

Speakers/Presenters:

Dennis Andrukat, Office of Nuclear Materials Safety and Safeguards -

Meeting Facilitator

Bill Reckley, NRR - Technical Lead

Nuclear Energy Institute (NEI)

Public Meeting Slides: ADAMS Accession No. ML21252A124

3 Purpose of Todays Meeting

Review preliminary proposed rule language for Subpart I -

Sections 53.1322 & 53.1333 (evaluation of plant changes).

Open discussion of previously released preliminary rule language, including Subpart F - Program-related sections

Todays meeting is a Comment -Gathering meeting, which means that public participation is actively sought in the discussion of the regulatory issues during the meeting.

o This meeting is being held in a workshop format to facilitate the discussion of todays topics.

o The meeting is being transcribed and the transcription will be available with the meeting summary by October 14, 2021.

No regulatory decisions will be made at todays meeting.

4 Subpart I - Section 53.1322, Evaluating changes to facility as described in final safety analysis reports Section 53.1333, Evaluating changes to programs included in licensing basis information

5 Subpart I - § 53.1322(a)

Licensees may make changes in the facility as described in the UFSAR and make changes in the procedures as described in the UFSAR without obtaining a license amendment only if:

o A change to the technical specifications incorporated in the license is not required.

o The change does not:

Result in a change to the frequency or consequences of an event sequence previously deemed not risk significant such that it becomes risk significant.

Result in a change to the frequency or consequences of an event sequence deemed risk significant such that it has a decrease of 10 percent or more in the calculated margins to the LBE evaluation criteria.

Result in a change to the frequency or consequences of one or more event sequences such that the margin between the calculated cumulative risks posed by the commercial nuclear plant and the safety criteria of § 53.220 decreases by 10 percent or more.

6 Subpart I - § 53.1322(a) (contd)

o The change does not (contd):

Involve a departure from a method of evaluation described in the UFSAR used in assessing margins in accordance with § 53.450(e) unless the results of the analysis are conservative or essentially the same, the revised method of evaluation has been previously approved by the NRC for the intended application, or the revised method of evaluation can be used in accordance with an NRC endorsed consensus code or standard.

For commercial nuclear plants licensed under Part 53 for which alternative evaluation criteria are applicable, result in a change to the frequency or consequences of event sequences such that the calculated margins between the results for event sequences and the alternative evaluation criteria decreases by 25 percent or more.

o In implementing this paragraph, the UFSAR is considered to include changes since submittal of the last update of the UFSAR.

o This section does not apply to changes to the facility or procedures when the applicable regulations establish more specific criteria for accomplishing such changes.

7

§ 53.1322(a)(2)(i)

(i ) Does not result in a change to the frequency or consequences of an event sequence such that an event sequence previously deemed not risk significant becomes risk significant by the analyses performed in accordance with § 53.450(e).

§ 53.450(e) NEI 18-04

The analyses must address event sequences from initiation to a defined end state and demonstrate that the functional design criteria required by § 53.420 provide sufficient barriers to the unplanned release of radionuclides to satisfy evaluation criteria defined for licensing basis events, to satisfy the safety criteria of § 53.220, and provide defense in depth as required by § 53.250. The methodology used to identify, categorize, and analyze licensing basis events must include a means to identify event sequences deemed significant for controlling the risks posed to public health and safety.

8

§ 53.1322(a)(2)(ii)

(ii) Does not result in a change to the frequency or consequences of an event sequence such that an event sequence deemed risk significant in accordance with

§ 53.450(e) has a decrease of 10 percent or more in the calculated margins to the LBE evaluation criteria required to be established in accordance with § 53.450(e).

NEI 18-04

9

§ 53.1322(a)(2)(iii)

(iii) Does not result in a change to the frequency or consequences of one or more event sequences such that the margin between the calculated cumulative risks posed by the commercial nuclear plant and the safety criteria of § 53.220 decreases by 10 percent or more.

§ 53.220 Safety Criteria for Licensing Basis Events Other Than Design Basis Accidents

(b) Maintain overall cumulative plant risk from licensing basis events such that the risk to an average individual within the vicinity of the plant receiving a radiation dose with the potential for immediate health effects remains below five in 10 million years, and the risk to such an individual receiving a radiation dose with the potential to cause latent health effects remains below two in one million years.

10

§ 53.1322(a)(2)(iv)

(iv) Does not involve a departure from a method of evaluation described in the UFSAR used in assessing margins in accordance with § 53.450(e) unless the results of the analysis are conservative or essentially the same, the revised method of evaluation has been previously approved by the NRC for the intended application, or the revised method of evaluation can be used in accordance with an NRC endorsed consensus code or standard.

§ 50.59(c)(1)(viii)

(viii) Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses..

11

§ 53.1322(a)(2)(v)

(v) For commercial nuclear plants licensed under this part for which alternative evaluation criteria are applicable in accordance with § 53.470, does not result in a change to the frequency or consequences of event sequences such that the calculated margins between the results for event sequences evaluated in accordance with

§ 53.450(e) and the alternative evaluation criteria decreases by 25 percent or more.

NEI 18-04 (w/ 1 rem goal)

§ 53.470 Application of Analytical Safety Margins to Operational Flexibilities.

Where an applicant or licensee so chooses, alternative criteria more restrictive than those defined in §§ 53.220 and 53.450(e) may be adopted to support operational flexibilities (e.g., emergency planning requirements under Subpart F of this part). In such cases, applicants and licensees must ensure that the functional design criteria of § 53.420, the analysis requirements of § 53.450(e),

and identification of special treatment of SSCs and human actions under § 53.460 reflect and support the use of alternative criteria to obtain additional analytical safety margins. Licensees must ensure that measures taken to provide the analytical margins supporting operational flexibilities are incorporated into design features and programmatic controls and are maintained within programs required in other Subparts.

12 Subpart I - § 53.1322(b)

Licensees who reference a design certification rule may make departures from the standard design, without prior Commission approval, unless the proposed departure involves a change to the design as described in the rule certifying the design, in which case the requirements of § 53.1315 are applicable.

Licensees shall maintain records of all departures from the certified design of the facility and these records must be maintained and available for audit until the date of termination of the license. L icensees will identify the location and nature of departures from licensing basis information within supporting documents for a certified design within the updates to the safety analysis report.

Licensees for which the NRC has docketed the certifications required under Subpart G of this part are not required to retain records of departures from the design of the facility associated with structures, systems, and components that have been permanently removed from service using an NRC-approved change process.

13 Subpart I - § 53.1322(c)

Licensees shall maintain records of changes in the facility and procedures made pursuant to paragraph (a) of this section. These records must include a written evaluation which provides the bases for the determination that the change does not require a license amendment pursuant to paragraph (a)(2) of this section.

Licensees shall submit, as specified in § 53.040 of this part, a report containing a brief description of any changes, including a summary of the evaluation of each. A report must be submitted at intervals not to exceed 24 months. For combined licenses, the report must be submitted at intervals not to exceed 6 months during the period from the date of application for a combined license to the date the Commission makes its findings under 10 CFR 53.[TBD]

(52.103(g)).

The records of changes in the facility must be maintained until the termination of an operating license or combined license issued under this part, or the termination of a renewed license issued under [TBD], whichever is later.

Records of changes in procedures must be maintained for a period of 5 years.

14 Subpart I - § 53.1333

Evaluating Changes to Programs in Licensing Basis Information

This iteration provides a uniform approach for program documents, which correspond to the programs required under Subpart F. The staff is interested in stakeholder views on the benefits of possibly developing a common approach versus the current practice of establishing program-specific requirements for reporting and change control. Note that this current iteration includes pointers in (a)(2) and (a)(3) that would allow program-specific change control criteria to be defined in other regulations or administrative sections of technical specifications. Where needed, the staff is seeking stakeholder views on the appropriate location and possible criteria for evaluating specific program documents.

§ 53.1333(a)

(2) An exemption from an NRC regulation is not required, (3) The change conforms to program-specific requirements included in regulations or technical specifications,

15 Subpart I - Section 53.1322

Discussion

16 Part 53 Rulemaking:

Change Control

Marc Nichol Senior Director, New Reactors

September 15, 2021

General observations

Very quantitative and should appeal to PRA intensive approaches, would not work for traditional approaches to using PRA (i.e., confirmatory)

No noticeable benefit to being different, and functionally not so different from existing

Proposed change criteria are extremely confusing and difficult to interpret

Needs significant guidance to explain how it works, and 50.59 guidance based on hard earned experience

The 10% change criterion is consistent in magnitude with 50.59 guidance, but more than minimal provides more flexibility than 10% (e.g., need for some deterministic elements)

Questions

What does a 10% decrease in margin mean? Margin to what - proximity to F-C curve?

How are margins measured and evaluated? Mean or upper uncertainty band?

How are design basis accidents and DID addressed?

Potential concerns

Criteria apply to all risk-significant licensing basis events (AOOs, DBEs, and BDBEs)

There are hundreds if not thousands of event sequences (they collapse into LBEs)

More than 10% change on very low frequencies will occur often, but have no safety significance

Margin to metrics in LWRs are totally different from margin for metrics in advanced reactors

Term used in assessing margins not in 50.59 and not consistent with definition in guidance calculational framework used for evaluating behavior or response of the facility or an SSCs ©2021 Nuclear Energy Institute 18 Discussion of Other Recently Released Iterations of Preliminary Rule Language

19 Other Recently Released Iterations of Preliminary Rule Language

Subpart B - Technology-Inclusive Safety Requirements (3rd iteration) (ML21202A162)

Subpart C - Requirements for Design and Analysis (3rd iteration) (ML21202A162)

Subpart H - Licenses, Certifications, and Approvals (ML21202A178)

Subpart I - Maintaining and Revising Licensing Basis Information (ML21202A175)

Subpart J - Reporting and Other Administrative Requirements (ML21225A224) 20 Other Recently Released Iterations of Preliminary Rule Language

Discussion

21 Part 53 Rulemaking:

Programs

Marc Nichol Senior Director, New Reactors

September 15, 2021

©2021 Nuclear Energy Institute Evaluating NRC Proposed Part 53 Programs Programs with a Part 50/52 Equivalent Required in NRC Part 53 Preliminary Language Part 50/52 Equivalent Requirements 53.710(a)* - Initial Startup Testing 50.34(b)(6(iii) 53.870 Inservice Inspection/Inservice Testing 50.55a 53.730 Maintenance, repair, and inspection programs 50.65 - although some elements may not have Part 50/52 counterpart 53.720 Maintaining capabilities and availability of SSCs 50.36 and 50.69 53.710(b)* - Training (Expected in future Subpart F requirements 50.2, Part 55, 50.120 on human actions) 53.710(c)* - Operating Plans (Expected in future Subpart F 50.34(b)(6)(iv and v) requirements on human actions) 53.860 Fire Protection 50.48 53.810 Radiation Protection Part 20 53.820 Emergency Preparedness 50.47 or 50.160 (in development) 53.830 Security Programs Part 73 (73.54, 73.55, 73.56) and Part 26 53.550 Environmental Considerations - Points to Part 51 50.36b - Points to Part 51 (if applicable)

Note that the NRC has not yet released the Subpart F regulations for human actions, which could include duplicative requirement s

Required in NRC Part 53 Preliminary Language Part 50/52 Equivalent Requirements 53.840 Quality Assurance Most of Appendix B QA Program 53.480 Design Control Quality Assurance None - Duplicates QA Program 53.610(a)(1&7) and 53.620(a)(1&6) Construction and None - Duplicates QA Program Manufacturing Quality Assurance 53.490 Design and Analyses Interfaces None - Duplicates QA Program 53.740 Design Control None - Duplicates QA Program 53.620(b)(1)(IV)(vii) - Manufacturing, Manufacturing Activities None - Duplicates QA Program

NRC should eliminate requirements that duplicate the QA Program NRC should put all of the QA Requirements together similar to Appendix B

Enable the use of ISO-9001 and other commercial QA standards

Preserve the ability to use Appendix B for those that wish to NRC does not need to specific QA requirements for non-safety -related but safety significant SSCs

©2021 Nuclear Energy Institute 24 Evaluating NRC Proposed Part 53 Programs NRC Required Programs without any Part 50/52 equivalent Required in NRC Part 53 Preliminary Language Part 50/52 Equivalent Requirements 53.700 Operational Objectives None - Duplicates most other operational programs 53.800 Operational Programs None - Duplicates most other operational programs 53.850 Integrity Assessment Programs None - Duplicates Maintenance, ISI/IST, Technical Specifications, and creates an aging management program from Day 1 53.890, 53.892, and 53.894 Facility Safety Program, Criteria and None - Duplicates other programs, codifies periodic Plan safety review, and circumvents backfit protection 53.880 Criticality Safety Program None - Not necessary to require a program for compliance with each requirement. 50.68 is a better model for Part 53 requirement.

53.610 (a)(2-5), (c&d) and 53.620(a)(2-4), Construction and None - Not necessary for NRC to approve the Manufacturing Organization and Procedures organization and plan during construction and manufacturing 53.1225 PRA Maintenance Program for 53.450(c) None - Not necessary for NRC to approve the controls for updating the PRA 53.460(c) Human Action Performance Program None - Duplicates the training and other operational programs related to performance of human actions NRC should eliminate all of these programs as they are not needed for

©2021 Nuclear Energy Institute 25 reasonable assurance of adequate protection NRC Approach to Programs in Part 53 An unstructured approach is inefficient and creates unintentional challenges NRCs approach to administrative controls results in:

Dramatic expansion of NRC regulatory footprint over licensee controls

An unclear and unbounded set of programmatic information subject to NRC approval Part 53 requires more programs and administrative controls be approved by the NRC, as compared to Parts 50/52 Part 53 requires approval of programmatic controls not required by Part 50/52

Programmatic controls mean administrative procedures that govern the actions of equipment and personnel of an advanced nuclear plant.

Required in 53.210, 53.220, 53.230, 53.240, 53.250, 53.260, 53.270, 53.400, 53.410, 53.420, 53.425, 53.430, 53.440, 53.460, 53.470, 53.490, 53.500, 53.510, 53.540, 53.610, 53.1225, etc.

Typically stated as Design features and programmatic controls must be provided for - Not performance-based, clear or predictable

©2021 Nuclear Energy Institute 26 NRC Needs a Regulatory Philosophy for Part 53 Lack of a Regulatory Philosophy leads to a patchwork approach to requirements Regulatory philosophy, starting with the Atomic Energy Act, to establish

The standard of adequate protection, in terms of radiological consequences to public health

The types of technical features and corresponding performance criteria that are necessary and sufficient to satisfy the adequate protection standard

The type, scope and level of detail of the technical information (licensing basis) that NRC needs to have a reasonable assurance that the technical features meet the adequate protection standard

Scope and level of oversight and inspection of licensee to provide reasonable assurance of compliance with license and requirements Technical features are: design features, human actions and programs

The role of programs is to provide reasonable assurance that the design features and human actions will perform the actions described in the licensing basis

Not all of the programs used by the licensee need to be required to be approved by the NRC ©2021 Nuclear Energy Institute 27 Recognize Confidence in Licensee Controls Input provided in Unified Industry Position letter The NRC imposes requirements that are effective even after the NRC issues a license for a new reactor

NRC has an oversight and inspection program to ensure compliance

NRC does not need to approve licensee controls related to compliance The licensee is competent in fulfilling their responsibility to perform administrative controls

QA Program permeates the plant at each stage

QA comprises all those planned and systematic actions necessary to provide adequate confidence that a structure, system, or component will perform satisfactorily in service

Very little need for NRC approval of other administrative controls to achieve reasonable assurance that design features and human actions will perform functions in the licensing basis ©2021 Nuclear Energy Institute 28 Performance-Based Approach to Part 53 Programs Leads to a clear, predictable and flexible regulatory framework Recognize that the QA Program provides substantial assurance that design features and human actions will perform functions in the licensing basis Establish the purpose for programs (e.g., by stage)

Design - Provide reasonable assurance that the plant design is in accordance with the license and regulations.

Manufacturing and Construction - Provide reasonable assurance that the plant is constructed and manufactured according to the license and regulations.

Maintenance - Provide reasonable assurance that the SSCs are capable of performing their intended functions described in the SAR.

Operations - Provide reasonable assurance that the plant is operated according to the license and regulations.

Establish performance criteria for each program, and entry criteria (graded)

Evaluate suitability of historical programs required by Part 50/52 Identify historical administrative controls not required to have NRC approval

Performance Criteria Part 50 Programs Programs not needing Requiring NRC Approval NRC Approval Design Provide reasonable assurance that the plant design is in

Criterion III - Design

Change Control accordance with the license and regulations. Control (Appendix B) (50.59)

1. Applicable regulatory requirements and the design

Records, reports and basis specified in the license are correctly translated

FSAR Update (50.71) into specifications, drawings and procedures.

Reliability Assurance

2. The design process used appropriate quality Program (SRM-standards, selected materials, parts and processes, SECY 132) controlled interfaces among participating

Environmental organizations, suitable to the safety significance of Qualification the SSCs, and provided for verifying the adequacy of (50.49(a))

the design.

3. Performance characteristics of SSCs that serve as the basis for the design and analyses are supported by validation data.

4. Design changes are subject to the same design control measures and approved by the same design organization used for the original design.

Performance Criteria Part 50 Programs Programs not needing Requiring NRC Approval NRC Approval Manufacturing Provide reasonable assurance that the plant is

Criteria IV, VI thru XV -* NSR SSC - Any and constructed and manufactured according to the license for safety-related SSCs commercial quality Construction and regulations (Quality Assurance - program

1. As-built SSCs are consistent with their as-designed Appendix B)

Procurement program specifications.

Defined by Applicant - for

Receipt and verification

2. The applicable regulatory requirements are non-safety related but risk programs referenced in the procurement documents. important (50.69

Turnover and routine

3. Procured material, equipment and services conform Augmented Quality) startup program to the procurement specifications.

Initial startup testing

Reporting of Defects

4. As-built SSCs, prior to operation, are capable of program (50.34(b)(6(iii)) and Nonconformances performing the functions described in the license. (Part 21)

©2021 Nuclear Energy Institute 31 Performance-Based Approach to Part 53 Programs Leads to a clear, predictable and flexible regulatory framework Performance Criteria Part 50 Programs Requiring Programs not needing NRC Approval NRC Approval Maintenance Provide reasonable assurance that the SSCs are capable

Maintenance

FLEX Equipment -

of performing their intended functions described in the Monitoring Program if applicable SAR. (50.65) (50.155)

1. SSCs, during operations, continue to be capable of

ISI/IST (50.55a)* Maintenance performing the functions described in the license.

Material Surveillance procedure

2. SSCs, for which the code or regulations require Program - if applicable development periodic inspection or testing, are confirmed to have (Part 50 Appendix H) not experienced unexpected degradation.

©2021 Nuclear Energy Institute 32 Performance-Based Approach to Part 53 Programs Leads to a clear, predictable and flexible regulatory framework Performance Criteria Part 50 Programs Requiring NRC Programs not needing Approval NRC Approval Operations Provide reasonable assurance that the plant is

Technical specifications (50.36)* Effluent release operated according to the license and

Training and Requalification program regulations. Programs for Operators, Fuel

Worker safety

1. Plant stays within the licensed conditions Handlers and Other Identified training programs of operations. Positions (50.2, Part 55, 50.120) and effectiveness

2. Administrative controls provide

Operating Plans, Normal and assessments reasonable assurance that human actions Emergency (50.34(b)(6)(iv and v))

OSHA worker credited for protection of public health and

Fire Protection Plan (50.48)safety safety will be performed when needed.

Radiation Protection (Part 20)* Procedure

3. Humans relied upon are trained and

Emergency Planning (50.47 or development for capable of performing assigned actions as 50.160) operations and described in the license.

Security (Physical, cyber, access emergencies and FFD) (Part 73, Part 26)

Event Reporting

Environmental Protection - if (50.72/50.73) applicable (51.50)

NRCs assertion that increased design and analysis requirements would lead to a reduction in operational requirements does not appear accurate NRC needs to establish a regulatory philosophy for Part 53 that defines the regulatory purpose of programs

Having clarity on why programs are needed will ensure that the program requirements are efficient NRC needs to reassess the program requirements in Part 53

11 program areas have equivalents in Part 50/52

13 program areas do not have a Part 50/52 equivalent or duplicate others

Over 20 instances of open ended requirements for programmatic controls NRC should ensure needed programs are performance-based, graded and appropriately scoped with entry criteria

Some programs (with Part 50/52 equivalents) require more than Parts 50/52

35 Part 53 Rulemaking Schedule

Milestone Schedule Major Rulemaking Activities/Milestones Schedule Public Outreach, ACRS Interactions and Present to April 2022 Generation of Proposed Rule Package (7 months)

Submit Draft Proposed Rule Package to May 2022 Commission

Publish Proposed Rule and Draft Key Guidance October 2022

Public Comment Period - 60 days November and December 2022 Public Outreach and Generation of Final Rule January 2023 to February 2024 Package (14 months)

Submit Draft Final Rule Package to Commission March 2024 Office of Management and Budget and Office of July 2024 to September 2024 the Federal Register Processing Publish Final Rule and Key Guidance October 2024

36 Future Public Meetings

The NRC staff will continue to announce public meetings to discuss and receive feedback on various regulatory topics and preliminary proposed rule text.

o Preliminary proposed rule language will be posted on regulations.gov under docket ID NRC-2019 -0062 before the public meetings.

The NRC staff is scheduled to meet with the ACRS Future Plants Subcommittee on September 23rd-24 th, 2021.

o Subpart B - Technology-Inclusive Safety Requirements (3 rd iteration) o Subpart C - Requirements for Design and Analysis (3rd iteration) o Subpart H - Licenses, Certifications, and Approvals o Subpart I - Maintaining and Revising Licensing Basis Information o Subpart J - Reporting and Other Administrative Requirements 37 Closing Remarks Rulemaking Contacts Robert.Beall@nrc.gov 301-415-3874 William.Reckley@nrc.gov 301-415-7490

Regulations.gov docket ID: NRC-2019- 0062

Please provide feedback on this public meeting using this link:

https://www.nrc.gov/public-involve/public-meetings/contactus.html

38 Acronyms and Abbreviations

ACRS Advisory Committee on Reactor Safeguards LBE Licensing basis event

ADAMS Agencywide Document Access LWR Light water reactor Management System NEI Nuclear Energy Institute AOOs Anticipated operational occurrence NRC U.S. Nuclear Regulatory Commission

BDBE Beyond design basis event NRR Office of Nuclear Reactor Regulations

CFR Code of Federal Regulations NSR Non-safety related

DBE Design basis event OSHA Occupational Safety and Health Administration DID Defense in Depth EAB Exclusion Area Boundary PRA Probabilistic Risk Assessment

F-C Frequency-consequence QA Quality Assurance

FFD Fitness for Duty REM Roentgen-equivalent man

FLEX Flexible Mitigation Capability SAR Safety analysis report

FSAR Final safety analysis report SRM Staff Requirements Memorandum

ISI Inservice inspection SSCs Structures, systems, and components

IST Inservice testing UFSAR Updated final safety analysis report

39 Background Slides

40 First Principles

See: SECY 0096, Functional Containment Performance Criteria for Non-Light-Water-Reactors, and INL/EXT 58717, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities 41 Integrated Approach

Siting near densely populated Functional areas EP for SMRs Licensing Containment and ONTs

Modernization Project

Insurance and Liability

Environmental Reviews

Consequence Based Security

42 Part 53 Rulemaking

The Part 53 Rulemaking Process*

The process depicted in this schematic is unique to the Part 53 rulemaking and varies in some ways compared to a similar A Typical Rulemaking Process schematic available on the NRCs public website.

43

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,

44 NRC Staff Plan to Develop Part 53

Subpart B Subpart C Subpart DSubpart E Subpart F Subpart G Project Life Cycle

Design and Siting Construction Operation Retirement Requirements Definition Analysis Facility Safety

Safety Objectives External

Safety Criteria System Hazards Construction/ Program

Safety Functions & Component Manufacturing Surveillance Design Site Characteristics Ensuring Maintenance Analysis Capabilities/ Configuration Requirements Environmental Reliabilities Considerations Control Safety Change Control Staffing &

Categorization

& Special Environmental Human Factors Treatment Considerations Programs Security, EP

Other Plant/Site (Design, Construction, Configuration Control)

Subpart A Clarify

General ProvisionsAnalyses (Prevention, Mitigation, Compare to Criteria) Controls Subpart J and

Admin & Reporting Plant Documents (Systems, Procedures, etc.) Distinctions Other 10 CFR Between Parts LB Documents (Applications, SAR, TS, etc.) Subparts H & I

45

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