ML22291A004

10/18-19/2022 - NRC Slides to Support the Advisory Committee on Reactor Safeguards Regulatory Rulemaking, Policies and Practices: Part 53 Subcommittee

ML22291A004
Person / Time
Issue date:	10/18/2022
From:	Robert Beall NRC/NMSS/DREFS/RRPB
To:
Beall, Robert
References
10 CFR Part 53, NRC-2019-0062, RIN 3150-AK31
Download: ML22291A004 (183)

Text

1 0 C F R Pa r t 5 3 L i c e n s i n g a n d R e g u l a o n o f A d v a n c e d N u c l e a r R e a c t o r s O c t o b e r 1 8 - 1 9, 2 0 2 2 Advisory Committee on Reactor Safeguards (ACRS) Regulatory Rulemaking, Policies and Practices:

Part 53 Subcommittee

Agenda - October 18th 2

8:35 am - 10:00 am Staff Introduction and Overview of Frameworks A and B 10:00 am - 11:45 am Draft Proposed Language for Quantitative Health Objectives (QHOs)/Safety Analysis 11:45 am - 12:45pm Lunch 12:45 pm - 5:00 pm Draft Proposed Language for Alternative Evaluation for Risk Insights (AERI) Methodology and Guidance Documents

Rulemaking Schedule Oct/Nov 2022 ACRS Interactions on Rulemaking Package for Proposed Rule

Part 53 Licensing Frameworks Framework A o Probabilistic Risk Assessment (PRA)-led approach o Functional design criteria Framework B o Traditional use of risk insights o Principal design criteria o Includes an AERI approach Subpart A - General Provisions 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 4

Rule Package (ML22272A034)

Federal Register Notice (FRN) A Preamble ML22272A036 B Section by Section, Availability of Guidance ML22272A038 C Framework A ML22272A039 D Framework B ML22272A040 Guidance Documents DG-1413 Licensing Events ML22272A042 DG-1414 AERI Methodology ML22272A045 DRO-ISG-2023-01 Operator Licensing Program Review ISG ML22272A047 DRO-ISG-2023-02 Staffing Plan Review ISG Augmenting NUREG-1791 ML22272A049 DRO-ISG-2023-03 Scalable Human Factors Engineering Review ISG ML22272A051

Sections 53.000 and 53.010

• Purpose

• Provide optional frameworks for the issuance, amendment, renewal, and termination of licenses, permits, certifications, and approvals for commercial nuclear plants

• Frameworks

• Framework A and Framework B are distinct

• Applicants and licensees subject to the rules in this part must only use the subparts applicable to one framework

Subpart A -

General Provisions (Definitions)

• Common Definitions

• Commercial Nuclear Plant

• Manufactured reactor

• Manufactured reactor module

• Safety function

• Framework A Definitions

• Construction, Licensing basis events (LBEs)

• Framework B Definitions

• Construction, Design basis, Functional containment, Safety-related structures, systems, and components (SSCs), Severe nuclear accident

Subpart A -

Safety Function Definition

• Safety function means a purpose served by a design feature, human action, or programmatic control to prevent or mitigate unplanned events and thereby demonstrate compliance with requirements in part 53 for limiting risks to public health and safety. Safety functions can be performed by any combination of the elements listed above and can be specified at the plant level or at the level of a particular barrier or system. The approach to identifying and addressing safety functions in Frameworks A and B are as follows:

(1) Within Framework A, the primary safety function is stated to be limiting the release of radioactive materials. Additional safety functions supporting the retention of radioactive materials, such as controlling reactivity, heat generation, heat removal, and chemical interactions, are determined for each reactor design by analyzing a spectrum of unplanned events.

(2) Within Framework B, multiple plant-level safety functions are assumed to apply to all reactor designs based on established requirements and historical practices. These fundamental safety functions include the control of reactivity, removal of heat, and limiting the release of radioactive materials. The protection of a specific barrier or system that contributes to meeting plant-level safety criteria may also be referred to as a safety function.

Framework A 9

Subpart B -

Technology-Inclusive Safety Requirements

• 53.200 Safety objectives.

• 53.210 Safety criteria for design basis accidents.

• 53.220 Safety criteria for licensing basis events other than design basis accidents. (including QHOs)

• 53.230 Safety functions.

• 53.240 Licensing basis events.

• 53.250 Defense-in-depth.

• 53.260 Normal operations.

• 53.270 Protection of plant workers.

Subpart C -

Design and Analysis 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 treatment.

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

§ 53.480 Earthquake engineering.

Subpart D -

Siting Requirements

§ 53.500 General siting.

§ 53.510 External hazards.

§ 53.520 Site characteristics.

§ 53.530 Population-related considerations

§ 53.540 Siting interfaces.

Subparts E & O Construction and Manufacturing Requirements Scope and purpose.

Reporting of defects and noncompliance.

Construction Manufacturing

• Fuel loading for manufactured reactor modules

Subparts E & O Fuel loading for manufactured reactor modules

§ 53.620(d)/53.4120(d) Fuel loading

• A manufacturing license may include authorizing the loading of fuel into a manufactured reactor module

• Specify required protections to prevent criticality o At least two independent mechanisms that can prevent criticality should conditions result in the maximum reactivity being attained for the fissile material

• Commission finding that a manufactured reactor module in required configuration is not a utilization facility as defined in the Atomic Energy Act

• Manufactured reactor module becomes a utilization facility in its final place of use after the Commission makes required findings on inspections, tests, analyses and acceptance criteria

Subpart F -

Requirements for Operation

§ 53.700 Operational objectives.

§ 53.710 Maintaining capabilities and availability of structures, systems, and components.

§ 53.715 Maintenance, repair, and inspection programs.

§ 53.720 Response to seismic events.

§ 53.725 General staffing, training, personnel qualifications, and human factors requirements.

§ 53.845 Programs Radiation Protection Emergency preparedness Security Quality Assurance (QA)

Integrity Assessment Fire protection Inservice inspection (ISI) and inservice testing (IST)

Facility safety

Subpart G & Q Decommissioning Requirements

• Scope and purpose.

• Financial assurance for decommissioning.

• Cost estimates for decommissioning.

• Annual adjustments to cost estimates for decommissioning.

• Methods for providing financial assurance for decommissioning.

• Limitations on the use of decommissioning trust funds.

• NRC oversight.

• Reporting and recordkeeping requirements.

• Termination of license.

• Program requirements during decommissioning

• Release of part of a commercial nuclear plant or site for unrestricted use.

Subpart H -

Licenses, Certifications, and Approvals

§ 53.1100 - 53.1121 General/common requirements.

§ 53.1124 Relationship between sections.

§ 53.1130 Limited work authorizations.

§ 53.1140 Early site permits.

§ 53.1200 Standard design approvals.

§ 53.1230 Standard design certifications.

§ 53.1270 Manufacturing licenses

§ 53.1300 Construction permits.

§ 53.1360 Operating licenses.

§ 53.1410 Combined licenses.

§ 53.1470 Standardization of commercial nuclear power plant designs: licenses to construct and operate nuclear power reactors of identical design at multiple sites.

Subparts I & S Maintaining and Revising Licensing Basis Information Licensing basis information.

Specific terms and conditions of licenses Changes to licensing basis information requiring prior NRC approval.

License amendments.

Specific provisions (e.g., changes to standard designs)

Other licensing basis information Evaluating changes to facility as described in final safety analysis reports (SAR).

Program-related documents Transfer of licenses or permits.

Termination of license.

Information requests.

Revocation, suspension, modification of licenses, permits, and approvals for cause.

Backfitting.

Renewal.

Subparts J & T Reporting and Other Administrative Requirements General information.

Unfettered access for inspections.

Maintenance of records, making of reports.

Immediate notification requirements for operating commercial nuclear plants.

Licensee event report system.

Facility information and verification.

Reporting of defects and noncompliance.

Financial requirements.

Financial qualifications.

Annual financial reports.

Licensees change of status; financial qualifications.

Creditor regulations.

Financial protection.

Insurance required to stabilize and decontaminate plant following an accident.

Financial protection requirements.

Subparts K & U Quality Assurance Criteria for Commercial Nuclear Plants

• General Provisions

• Organization

• Quality Assurance Program

• Design Control

• Procurement Document Control

• Instructions, Procedures and Drawings

• Document Control

• Control of Purchased Material, Equipment and Services

• Identification and Control of Materials, Parts and Components

• Control of Special Processes

• Inspection

• Test Control

• Control of Measuring and Test Equipment

• Handling, Storage and Shipping

• Inspection, Test and Operating Status

• Nonconforming Materials, Parts or Components

• Corrective Action

• Quality Assurance Records

• Audits 10 CFR Part 50, Appendix B Criteria I

II III IV V

VI VII VIII IX X

XI XII XIII XIV XV XVI XVII XVIII

Framework B

Subpart N -

Siting New subpart that facilitates risk-informed, performance-based approaches to siting and seismic design

§ 53.3505 Scope.

§ 53.3510 Definitions.

§ 53.3515 Factors to be considered when evaluating sites.

§ 53.3520 Non-seismic siting criteria.

§ 53.3525 Geologic and seismic siting criteria.

Subpart P -

Requirements for Operation

§ 53.4200 Operational objectives.

§ 53.4210 Maintaining capabilities and availability of structures, systems, and components.

§ 53.4213 Technical specifications.

§ 53.4215 Response to seismic events.

§ 53.4220 General staffing, training, personnel qualifications, and human factors requirements.

§ 53.4300 Programs Radiation Protection Emergency Preparedness Security QA Integrity Assessment Fire Protection ISI and IST Environmental qualification of electric equipment Procedures and guidelines Primary containment leakage testing

§ 53.4420 Mitigation of beyond-design-basis events.

Subpart R -

Licenses, Certifications, and Approvals

§ 53.4700 - 53.4721 General/common requirements.

§ 53.4724 Relationship between sections.

§ 53.4730 General technical requirements.

§ 53.4731 Risk-informed classification of SSCs.

§ 53.4733 Seismic design alternatives.

§ 53.4740 Limited work authorizations.

§ 53.4750 Early site permits.

§ 53.4800 Standard design approvals.

§ 53.4830 Standard design certifications.

§ 53.4870 Manufacturing licenses

§ 53.4900 Construction permits.

§ 53.4960 Operating licenses.

§ 53.5010 Combined licenses.

§ 53.5070 Standardization of commercial nuclear power plant designs: licenses to construct and operate nuclear power reactors of identical design at multiple sites.

Draft Proposed Language for QHOs / Safety Analysis

Framework A Integrated Approach to Ensure Comparable Findings Existing Paradigm

• Does not specifically define adequate protection but compliance with NRC regulations and guidance may be presumed to assure adequate protection at a minimum

• Additional requirements as necessary or desirable to protect health or to minimize danger to life or property Part 53 (SECY-20-0032)

1) Continue to provide reasonable assurance of adequate protection of public health and safety and the common defense and security,

2) Promote regulatory stability, predictability, and clarity,

3) Reduce requests for exemptions from the current requirements in 10 CFR Part 50 and 10 CFR Part 52,

4) Establish new requirements to address non-light-water reactor (LWR) technologies,

5) Recognize technological advancements in reactor design, and

6) Credit the response of advanced nuclear reactors to postulated accidents, including slower transient response times and relatively small and slow release of fission products.

Framework A Integrated Approach to Ensure Comparable Findings

Framework A Ensuring Comparable Level of Safety Additional discussion in Preamble on how an integrated assessment like that in Regulatory Guide (RG) 1.174 can be used to support the comparisons to existing requirements and related regulatory findings.

Framework A QHOs as one of several performance standards for LBEs Additional discussion in Preamble on how QHOs are considered as one of several performance measures within Framework A. Including the QHOs as one of several performance measures does not equate to the QHOs defining adequate protection of public health and safety.

Framework A Consideration of Feedback on Including QHOs Comments generally fall into following groups:

Rule should not include a cumulative risk measure Rule should include alternative risk measures o

Surrogates for the QHOs Develop new safety goals

• It is appropriate to include a risk-related performance standard in Framework A as part of an integrated decisionmaking process, especially given the importance of risk assessments and consideration of risk-insights within the licensing process

• In SRM-SECY-10-0121, the Commission reaffirmed that existing safety goals, safety performance expectations, subsidiary risk goals and associated risk guidance are sufficient for new plants

• Surrogate measures tend to be technology-or design-specific. However, the Preamble reinforces that technology-or design-specific surrogates for the QHOs may be developed and proposed for use in supporting licensing under Framework A

• Major efforts such as developing new safety goals not included in rulemaking plan and not feasible considering project constraints

Subpart R -

Licenses, Certifications, and Approvals

§ 53.4730(a)(1) Site safety analysis.

Proposed rule language derived from current requirements in § 52.79(a)(1); (i) through (v) are essentially identical to Part 52 requirements Requirements in subparagraph (vi) modified to ensure rule is technology-inclusive Fuel or core damage or potential for large radiological releases from sources other than the reactor system replaces fission product release from the core into the containment Fission product release analyses can be performed using a mechanistic source term or bounding assessment Applicant may elect to comply with more restrictive dose consequence criteria (e.g., 1 rem [roentgen equivalent man]

TEDE [total effective dose equivalent] over 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />)

Subpart R -

Licenses, Certifications, and Approvals

§ 53.4730(a)(5) Initiating events and accident analysis.

• Objectives Provide an equivalent level of safety by developing technology-inclusive analogs to applicable Part 50 and 52 requirements for initiating events and accident analyses

• Provide an approach that better aligned with international regulatory paradigms, as appropriate and consistent with Commission policy

• Leveraged previously developed language from the Part 5X effort

• Preliminary proposed rule language maintains top-level acceptance criteria from Part 50 and 52

Subpart R -

Licenses, Certifications, and Approvals

§ 53.4730(a)(5) Initiating events and accident analysis.

(i) Analysis and Evaluation From § 52.79(a) with modifications to support technology-inclusiveness and Framework B event classifications.

Recent changes to acknowledge multi-unit facilities (e.g., SMRs)

(ii) Design Basis Accidents Technology-inclusive requirements for DBA analyses and SSC classification drawing from §§ 50.34(a)(4) and 50.46.

Includes deterministic classification approach for safety-related SSCs (iii) Normal Operation and Anticipated Operational Occurrences (AOOs)

Consistent with existing requirements including Part 20 acceptance criteria Changes clarify applicability of requirements to normal operations

Subpart R -

Licenses, Certifications, and Approvals

§ 53.4730(a)(5) Initiating events and accident analysis.

(iv) Additional Licensing Basis Events Technology-inclusive requirements for relevant additional LBEs and analysis requirements for these events; similar to international defense-in-depth (DID) requirements Changes clarify scope of initiators and event sequences that must be considered and design requirements for SSCs used to mitigate additional LBEs (v) Severe Accidents Derived from § 52.79(a)(38), with modifications to support technology-inclusiveness Definition of severe nuclear accident moved to § 53.028 (vi) Chemical hazard requirements address substances commingled with licensed material or those produced by a reaction with licensed material

Subpart R -

Licenses, Certifications, and Approvals

§ 53.4730(a)(36)

Containment requirements.

Requirements split to acknowledge differences between non-LWR and LWR approaches to containment For non-LWRs, § 53.4730(a)(36)(i) addresses:

o Set of barriers used to meet requirements for AOOs, DBAs, and siting criteria (functional containment) o Safety classification (i.e., safety-related) and qualification of SSCs making up functional containment barriers o

Functional containment now defined in § 53.028 For LWRs, § 53.4730(a)(36)(ii) addresses the need for a leak-tight primary containment that:

o Meets the requirements of Part 50 Appendix J (also addressed in Subpart P) o Addresses any technically relevant requirements from LWR operating experience (containment isolation systems, penetrations, venting/purging)

Subpart R -

Licenses, Certifications, and Approvals Other General Technical Requirements

§ 53.4730(a)(2) Facility description.

§ 53.4730(a)(4) Design bases and principal design criteria.

§ 53.4730(a)(11) Dose to members of the public.

§ 53.4730(a)(14) Earthquake engineering criteria.

§ 53.4730(a)(34) Description of risk evaluation.

§ 53.4730(a)(37) Water-cooled reactor requirements.

Changes to other paragraphs under § 53.4730 largely organization since last iteration was issued

10 CFR Part 53, Framework B Alternative Evaluation for Risk

Insights, DG-1413, and DG-1414

Introduction Katie Wagner Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission

Agenda

• Introductions & Recent Activities

• Proposed AERI Entry Conditions

• Evaluation of Dose-Based AERI Entry Criteria Using MELCOR Accident Consequence Code System (MACCS)

• DG-1413 (proposed new RG 1.254), "Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants"

• DG-1414 (proposed new RG 1.255), "Alternative Evaluation for Risk Insights (AERI) Methodology" 39

• Marty Stutzke - Technical Lead of the Graded PRA Working Group (WG), Senior Level Advisor for Probabilistic Risk Assessment, Division of Advanced Reactors and Non-power Production and Utilization Facilities (DANU), Office of Nuclear Reactor Regulation (NRR)

• Keith Compton - Lead for MACCS calculations related to the AERI entry conditions, Senior Reactor Scientist, Division of Systems Analysis, Office of Nuclear Regulatory Research (RES)

• Mihaela Biro - Principal Author of DG-1413 (proposed new RG 1.254), "Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants," Senior Reliability and Risk Analyst, Division of Risk Assessment (DRA), NRR

• Anne-Marie Grady - Principal Co-author of DG-1414 (proposed new RG 1.255),

"Alternative Evaluation for Risk Insights (AERI) Methodology," Reliability and Risk Analyst, DRA, NRR

• Katie Wagner - Project Manager of the Graded PRA WG, Project Manager, DANU, NRR

Introductions

40

The Graded PRA Working Group Membership Project Manager

• Katie Wagner, NRR/DANU Technical Lead

• Marty Stutzke, NRR/DANU Working Group Members

• Hosung Ahn*, previously on rotation from NRR/Division of Engineering and External Hazard

• Mihaela Biro, NRR/DRA - Principal Author of DG-1413

• Anne-Marie Grady, NRR/DRA - Principal Co-Author of DG-1414

• Matt Humberstone, RES/DRA

• Ian Jung, NRR/DANU

• Alissa Neuhausen, NRR/DRA*^ - Principal Co-Author of DG-1414

• Hanh Phan, NRR/DANU

• Sunil Weerakkody, NRR/DRA

• Robert Budnitz, consultant Management/Coordination

• Candace de Messieres, NRR/DANU^

• Steve Lynch, NRR/DANU

• Nathan Sanfilippo*

• John Segala, NRR/DANU

• Former WG member

^On rotation from current position 41

• Latest ACRS Interactions and Communications o ACRS Subcommittee Meeting - June 23-24, 2022 (ML22172A091) o ACRS Full Committee Meeting - July 6, 2022 (ML22186A166) o ACRS Letter - August 2, 2022 (ML22196A292)

• Path forward discussion in late-June 2022 o DG-1413 & DG-1414

Make revisions in response to ACRS and stakeholder feedback

Monitor changes to preliminary proposed rule text o DG-1414

Develop guidance for AERI maintenance and upgrades Recent Activities 42

AERI-Related Draft Proposed Rule Text and FRN Sections 43 Marty Stutzke Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission

Regulatory Basis for the AERI Approach 44 Policy Statement on the Regulation of Advanced Reactors 73 FR 60612; October 14, 2008 73 FR 60616, left column: The Commission also expects that advanced reactor designs will comply with the Commissions safety goal policy statement (51 FR 28044; August 4, 1986, as corrected and republished at 51 FR 30028; August 21, 1986),

73 FR 60614, left column: the Commission has also issued policy statements on the use of PRA in regulatory activities (60 FR 42622; August 16, 1995), and severe accidents regarding future designs and existing plants (50 FR 32138; August 8, 1985). The use of PRA as a design tool is implied by the policy statement on the use of PRA and the NRC believes that the current regulations and policy statements provide sufficient guidance to designers.

Policy Statement: Use of PRA Methods in Nuclear Regulatory Activities 60 FR 42622; August 16, 1995 60 FR 42628, middle column: It is important to note that not all of the Commissions regulatory activities lend themselves to a risk analysis approach that utilizes fault tree methods. In general, a fault tree method is best suited for power reactor events that typically involve complex systemsthe Commission recognizes that a single approach for incorporating risk analyses into the regulatory process is not appropriate.

AERI Elements Evaluate DID adequacy Identify risk insights Search for severe accident vulnerabilities Develop a demonstrably conservative risk estimate Demonstrate that the AERI entry conditions are met Identify and characterize the postulated bounding event use PRA or an alternative risk-informed approach as a design tool

AERI-Related Draft Proposed Rule Text 45 Current Draft Proposed Rule Text

§ 53.4730(a)(34) Description of risk evaluation. A description of the risk evaluation developed for the commercial nuclear plant and its results.

The risk evaluation must be based on:

(i) A probabilistic risk assessment (PRA); or (ii) An alternative evaluation for risk insights (AERI), provided that:

(A) The analysis of a postulated bounding event demonstrates that the consequence evaluated at a location 100 meters (328 feet) away from the commercial nuclear plant does not exceed 10 mSv (1 rem) total effective dose equivalent (TEDE) over the first four days following a release, an additional 20 mSv (2 rem) TEDE in the first year, and 5 mSv (0.5 rem) TEDE per year in the second and subsequent years; and (B) The qualification in § § 53.4730(a)(34)(ii)(A) is demonstrated to be met without reliance on active safety features or passive safety features except for those passive safety features that do not require any equipment actuation or operator action to perform their required safety functions, that are expected to survive accident conditions, and that cannot be made unavailable or otherwise defeated by credible human errors of commission and omission.

The proposed AERI entry conditions are designed to limit use of the proposed AERI approach to commercial nuclear plants whose designs are relatively straightforward and do not involve overly complex systems and interactions and, accordingly, would not warrant development of a PRA to provide quantitative risk insights.

Draft Proposed Rule Text Presented to the ACRS Regulatory Rulemaking, Policies and Practices:

Part 53 Subcommittee June 23-24, 2022

§ 53.4730(a)(34) Description of risk evaluation. A description of the risk evaluation developed for the commercial nuclear plant and its results. The risk evaluation must be based on:

(i)

A PRA, or (ii)

An AERI, provided that the dose from a postulated bounding event to an individual located 100 meters (328 feet) away from the commercial nuclear plant does not exceed 1 rem total effective dose equivalent (TEDE) over the first four days following a release, an additional 2 rem TEDE in the first year, and 0.5 rem TEDE per year in the second and subsequent years.

• 53.4730(a)(34)(ii)(A) o The consequence criteria in the AERI entry condition were originally inspired by the U.S. Environmental Protection Agency (EPA)

Protective Action Guidelines (PAGs); however:

The EPA PAGs are used in response to an actual event; in contrast, the AERI entry conditions refer to a postulated bounding event that is used to help establish the licensing basis.

The Commission has never stated that the EPA PAGs are limits. In addition, the PAGs state: protective action guide doses represent trigger points for taking protective actions. They are not dose limits that cannot be exceeded.

Stakeholders may misconstrue the previous draft proposed AERI entry conditions to mean that it is an acceptable limit for an emergency dose to the public under accident conditions.

o Changes to the draft proposed rule text were made during extensive discussions with the Office of Nuclear Security and Incident Response.

o Conforming changes were made to the FRN preamble and to DG-1414.

• 53.4730(a)(34)(ii)(B) o Changes made in concert with changes to Part 53, Framework A, Subpart F concerning operator licensing.

o Current draft proposed rule text is consistent with:

Draft staff white paper, Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors, March 2021, ML21069A003

Section 2.7 of DOE-HDBK-1224-2018, DOE Handbook: Hazard and Accident Analysis Handbook (Interim Use), August 2018 Changes to the AERI-Related Draft Proposed Rule Text 46

• Would be used to determine:

o Which applicants could develop an AERI in lieu of a PRA to demonstrate compliance with the proposed risk evaluation requirement in 53.4730(a)(34) o When the requirements to address the mitigation of beyond-design-basis events in 53.4420 must be met o When the requirements to address combustible gas control in 53.4730(a)(7) must be met

• In addition, the proposed AERI entry conditions would be used in combination with other conditions to determine when a commercial nuclear plant is a self-reliant mitigation facility, as provided in 53.800(a)(2) o A self-reliant mitigation facility may have generally licensed reactor operators (GLROs) in lieu of senior reactor operators (SROs) and reactor operators (ROs)

Proposed Uses of the AERI Entry Conditions 47

Maintenance of Risk Evaluations 48

§ 53.6052 Maintenance of risk evaluations.

Applicants or licensees required to submit a risk evaluation under § 53.4730(a)(34) must meet the following requirements:

(a) No later than the scheduled date for initial loading of fuel, each holder of an operating or combined license for a commercial nuclear plant under Framework B of this part must develop a risk evaluation.

(b) Each licensee required to develop a risk evaluation under paragraph (a) of this section must maintain the risk evaluation to reflect the as-built, as-operated facility. The risk evaluation must be maintained at least every five years until the permanent cessation of operations under § 53.4670. If a PRA is performed under § 53.4730(a)(34)(i), the licensee must upgrade the PRA to cover initiating events and modes of operation contained in consensus standards on PRA that are endorsed by the NRC. The upgrade must be completed within five years of NRC endorsement of the standard.

(c) Each licensee required to develop a risk evaluation based on a PRA must, no later than the date on which the licensee submits an application for a renewed license, upgrade the PRA required by paragraph (a) of this section to cover all modes and all initiating events.

(d) Each licensee who developed an alternative evaluation for risk insights under § 53.4730(a)(34)(ii) must, no later than the date on which the licensee submits an application for a renewed license, confirm that the alternative evaluation for risk insights reflects the as-built, as-operated facility.

Definitions from the non-LWR PRA standard (ASME/ANS Ra-S-1.4-2022)

• PRA maintenance: a change in the PRA that does not meet the definition of PRA upgrade.

o Peer review not required by the standard

• PRA upgrade: a change in the PRA that results in the applicability of one or more supporting requirements or Capability Categories (e.g., the addition of a new hazard model) that were not previously assessed in a peer review of the PRA, an implementation of a PRA method in a different context, or the incorporation of a method not previously used.

o Peer review required by the standard

• The NRC is seeking comment on whether the NRC should retain this AERI approach under Framework B. If so, what changes, if any, would be recommended to the proposed criteria and approach in proposed Framework B? Please provide the considerations and rationale for your answer.

• Could the AERI criteria as written or potentially as revised and the related analyses of bounding events be used to support other regulatory decisions in Framework B (e.g., physical security, cyber security, AA (access authorization), FFD (fitness for duty) and emergency preparedness)? If so, which design areas and programs could logically use the AERI criteria and related analyses and how could requirements in those areas be scaled or graded based on the proposed 53.4730(a)(34)(ii) or a similar concept?

• The NRC is seeking comment on the criteria and how they are used in both justifying an alternative to PRAs and in allowing the use of GLROs, as well as possible alternatives to the proposed criteria. Please provide your considerations and rationale for your recommendation.

Proposed AERI-Related FRN Questions 49

Evaluation of Dose-Based AERI Entry Criteria Using MACCS Keith L. Compton Division of Systems Analysis Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission

Objectives 51

• Evaluate the relationship between dose computed at 100 m and the population-weighted individual latent cancer fatality risk (ILCFR) averaged over 10 miles using MACCS o Develop a closed-form analytic approximation to this relationship o Identify assumptions needed to develop the closed-form approximation o Test the impact of these assumptions using suitable calculations with MACCS

• The analyses and results in this presentation provide a status report on work-in-progress. They do not represent the staffs final analyses or conclusions.

Analytic Expression Assumptions 52

• Individual doses from ingestion pathways are not explicitly considered

• The maximum individual dose max at a distance r is assumed to be related to the maximum individual dose max,0 at the distance r0 as follows:

• All material is released in a single plume (i.e., there are no wind shifts during release)

• The population density N is assumed to be constant and independent of distance r

• The latent cancer proportionality constant is assumed to be constant and independent of dose

Downwind Dose Reduction Coefficient 53 The maximum individual dose max at a distance r is assumed to be related to the maximum individual dose max,0 at the distance r0 as follows:

Subsidiary Assumption Rationale The release is from ground level and non-bouyant (i.e., (,) is monotonically decreasing)

Elevated releases or plume rise will result in an increase in concentration at short downwind distances as the plume disperses overhead before contacting ground Protective actions to limit dose are not taken Protective actions may constrain dose at short downwind distances The plume is completely reflected at the ground surface and is unconstrained by a mixing height Highly unstable conditions can result in rapid vertical dispersion to the top of the mixing layer due to insolation of ground surface The dose-distance reduction coefficient n is assumed to be independent of distance r.

Although crosswind (transverse) dispersion is typically represented as a power law, vertical dispersion does not follow a power law relationship with distance

Downwind Dose Reduction Coefficient Elevated/Buoyant Plume 54 Normalized relative peak dose as a function of downwind distance and stability class Normalized to a constant core scaling factor and maximum peak dose 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0 2

4 6

8 10 12 14 Normalized Relative Peak Dose Downwind Distance (km)

A B

C D

E F

Downwind Dose Reduction Coefficient Effect of Protective Actions 55 Mean value (across all weather trials) of peak total effective dose (rem) from early-phase exposure to the non-evacuating cohort Mean value (across all weather trials) of peak total effective dose (rem) from late-phase exposure

• The flatness of the ICF-BURN (red) curve out to 20 miles, and the latent cancer fatality (LCF) (magenta) curve out to 15 miles, is due to early-phase hotspot relocation within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> coupled with a relatively prolonged release

• Doses incurred during the late phase are low near the site, but do not appreciably decline with distance from the site for the most severe scenarios.

Source: NRC 2022

Downwind Dose Reduction Coefficient Mixing Height 56 0

100 200 300 400 500 600 700 800 900 1000 0

2000 4000 6000 8000 10000 12000 14000 Plume Sigma-z (m)

Downwind Distance (m)

A B

C D

E F

Plume sigma z (m) as a function of downwind distance Sources: Case 2 Model Output Files; Holzworth 1972 Isopleths of mean annual morning mixing height (m*10-2) as a function of downwind distance Isopleth levels: 300 - 900 m Isopleths of mean annual afternoon mixing height (m*10-2) as a function of downwind distance Isopleth levels: 800 - 2600 m

Downwind Dose Reduction Coefficient Power Law Coefficient with Distance 57 Lateral diffusion without meander and building wake effects (y) vs. downwind distance from source for Pasquill's turbulence types (atmospheric stability)

Vertical diffusion without meander and building wake effects (z) vs. downwind distance from source for Pasquill's turbulence types (atmospheric stability)

Source: Reference 7 of NRC 1983

Single Plume Azimuthal Correction Factor 58

• A single plume azimuthal correction factor (r) is defined as the ratio between peak individual dose max from a single plume at a distance r and the individual dose averaged across the circumference of a circle of radius r.

• Assuming (Tadmor and Gur, 1969) that the crosswind dispersion factor may be represented as a power function given by,

the azimuthal correction factor may be represented as:

2

2

Stability Class Ay By

100 m

10 mi A

0.3658 0.9031 0.0934 0.0571 B

0.2751 0.9031 0.0702 0.0429 C

0.2089 0.9031 0.0533 0.0326 D

0.1471 0.9031 0.0376 0.0230 E

0.1046 0.9031 0.0267 0.0163 F

0.0722 0.9031 0.0184 0.0113 G

0.0481 0.9031 0.0123 0.0075 An alternative would be to simply assume that the crosswind plume spread may be represented by a tophat with a width of one 22.5 sector, resulting in an azimuthal correction factor of 0.0625 (1/16)

Figure Source: Jow et al. 1990

Maximum Dose vs Average Dose / Risk over an Annular Region 59 For nBy+1, the average individual dose in the annular region between r0 and x may be expressed as:

The average individual cancer risk in the annular region between r0 and x may be expressed as :

Where:

is the power law linear coefficient for transverse dispersion is the power law exponent for transverse dispersion

,is the peak centerline dose at the inner annular radius (e.g., 100 m) is the inner annular radius (e.g., 100 m) is the outer annular radius (e.g., 16,090 m (10 mi))

n is the downwind dose reduction coefficient

Approach 60

• Develop a set of MACCS modeling cases to quantitatively examine impact of assumptions

• Use source terms from NRC Level 3 PRA reactor at-power internal events and internal floods Level 2 analyses to represent a range of source term compositions

• Apply scaling factors to source terms to yield a 25 rem (0.25 Sv) lifetime* dose at 100 m

• Use combinations of constant weather conditions, constant population density, and meteorological and site files from SOARCA (state-of-the-art reactor consequence analyses) analyses to examine impact of variability in weather condition and population density

• Lifetime dose, in this analysis, is assumed to be the dose resulting from a 96-hour (4 day) early phase exposure and a 50-year late phase exposure.

Summary of Source Terms Source Term Characteristics 61 RC Case Release Category Description NUMREL PDELAY (hr)

PLUDUR (50%) (hr)

PLUDUR (100%) (hr)

PLHITE (m)

PLHEAT (MW)

VF 5D Unscrubbed interfacing systems loss-of-coolant accident with auxiliary building failure 86 3.2 4.5 68.8 11 19 LCF 1B Late containment due to long-term quasi-static overpressure, unscrubbed 179 48 32.1 120.0 0.36 5.9 NOCF 2R1 Containment is not bypassed or failed, and radiological release to the environment occurs via design-basis containment leakage only.

199 13 89.9 154.5 32 0.0026 Source: adapted from Tables 3.1-1 and A.1a in NRC 2022 RC Case Xe Cs Ba I

Te Ru Mo Ce La VF 5D 8.6E-01 1.3E-01 2.1E-03 1.4E-01 1.3E-01 2.6E-03 3.3E-02 9.3E-05 2.7E-06 LCF 1B 9.1E-01 9.9E-03 3.0E-04 1.2E-02 1.1E-02 6.6E-06 4.0E-02 1.4E-06 5.8E-07 NOCF 2R1 1.0E-02 7.4E-05 2.4E-06 8.5E-05 7.9E-05 3.7E-06 2.0E-04 2.3E-08 2.0E-08 All source terms are inventory-scaled to yield 25 rem overall (EARLY+CHRONC) dose at 100 m Base case plume is based on intersystem loss-of-coolant accident (VF/5D) source term Scaled source terms may vary in relative radionuclide composition and release duration Single segment plume are created by summing/averaging properties for individual plume segments.

Multi-plume releases capture the time dependence of the release.

Summary of Modeling Cases 62 Case Dose Reduction Coefficient Effects Azimuthal Variation Population Density 0A-F* Single Stabilities - A-F Power Law Stability Single Plume - VF Constant 1A-F* Single Stabilities - A-F Pasquill-Gifford Stability Single Plume - VF Constant 2A-F* Single Stabilities - A-F Plume Rise Single Plume - VF Constant 3A-F* Single Stabilities - A-F Wake Effects Single Plume - VF Constant 4A-F* Single Stabilities - A-F Protective Actions Single Plume - VF Constant 5A-B Met Sampling - PB None/Plume Rise Single Plume - VF Constant 6A-C Met Sampling - PB None Multiplume - VF/LCF/NOCF Constant 7A-C Met Sampling - PB None Multiplume - VF/LCF/NOCF PB Modeling cases designed to test effect of key assumptions related to plume rise, wake effects, protective actions, plume segmentation, weather variability, and population density

• Each stability class (A-F) represent a separate subcase for these cases. For example, Case 2A represents Case 2 with stability class A, Case 3F represents Case 3 with stability class F, etc.

Case 0: Simple Model Results 63

• Simplest Case

• Power law representation for Y and z with constant parameters

• Constant weather conditions - specified stabilities, 2.5 m/s, no rain, mixing layer depth 10 km

• Constant deposition velocity (0.003 m/s)

• Single plume - scaled VF source term, ground level release with no plume buoyancy (plume heat of 0 MW)

• Uniform population density with no protective actions

• Single cancer risk coefficient based on total effective dose

• Fitted n derived from power law regression of MACCS results (see supplemental slides)

• Lifetime dose of 25 rem yields 10-mile ILCFR from 3.6e-8 to 3.4e-7

• All cases produce MACCS ILFCR <2e-6

• Difference between MACCS and analytic calculation ranges from 3.6% to 470%

Case OVERALL Peak dose (Sv) at 100 m EARLY Peak dose (Sv) at 100 m CHRONC Peak dose (Sv) at 100 m MACCS 10- mile ILCFR P-G n MACCS fitted n Analytic 10-mile ILCFR Percent Difference 0A 0.25 0.02 0.23 3.6E-08 3.0 2.4 2.0E-07 470%

0B 0.25 0.02 0.23 5.5E-08 2.5 2.4 1.4E-07 160%

0C 0.25 0.02 0.23 2.9E-07 1.8 1.8 4.2E-07 47%

0D 0.25 0.02 0.23 3.4E-07 1.6 1.6 4.6E-07 32%

0E 0.25 0.02 0.23 2.9E-07 1.5 1.6 3.6E-07 21%

0F 0.25 0.02 0.23 2.0E-07 1.5 1.7 2.0E-07 3.6%

Case 1: Pasquill-Gifford Stability Results 64

• Differences from Case 0:

• 1000-m deep boundary layer

• Eimutis and Konicek representation for Y and z with spatially variable parameters for z

• Particle-size-dependent deposition velocity

• Organ-specific cancer risk coefficients

• Lifetime dose of 25 rem yields 10-mile ILCFR from 1.4e-7 to 3.3e-7

• Difference between MACCS and analytic calculation ranges from 40% to 264%

• Analytic calculation is conservative relative to MACCS calculation

• All cases produce MACCS ILFCR <2e-6 Case OVERALL Peak dose (Sv) at 100 m EARLY Peak dose (Sv) at 100 m CHRONC Peak dose (Sv) at 100 m MACCS 10- mile ILCFR P-G n MACCS fitted n

Analytic 10-mile ILCFR Percent Difference 1A 2.5E-01 2.3E-02 2.3E-01 3.3E-07 3.0 1.6 1.2E-06 260%

2B 2.5E-01 2.3E-02 2.3E-01 2.5E-07 2.5 1.8 5.0E-07 100%

2C 2.5E-01 2.3E-02 2.3E-01 2.2E-07 1.8 1.8 3.7E-07 68%

2D 2.5E-01 2.4E-02 2.3E-01 2.4E-07 1.6 1.7 4.0E-07 65%

2E 2.5E-01 2.5E-02 2.3E-01 2.0E-07 1.5 1.7 3.1E-07 54%

2F 2.5E-01 2.7E-02 2.2E-01 1.4E-07 1.5 1.7 1.9E-07 40%

Case 2: Plume Buoyancy Results 65 Case OVERALL Peak dose (Sv) at 100 m EARLY Peak dose (Sv) at 100 m CHRONC Peak dose (Sv) at 100 m MACCS 10- mile ILCFR MACCS fitted n Analytic 10-mile ILCFR Percent Difference 2A 2.5E-01 2.6E-02 2.3E-01 2.5E-05 0.9 1.6E-05

-38%

2B 2.5E-01 1.0E-01 1.5E-01 8.0E-04 0.1 4.3E-04

-47%

2C 2.5E-01 2.5E-01 4.9E-04 2.3E-03

-0.6 8.0E-03 244%

2D 2.5E-01 2.5E-01 9.8E-13 1.9E-03

-0.8 1.3E-02 566%

2E 2.5E-01 2.5E-01 0.0E+00 5.4E-03

-1.0 2.3E-02 331%

2F 2.5E-01 2.5E-01 0.0E+00 6.1E-03

-1.0 2.2E-02 256%

Difference from Case 1: Ground-level release with plume buoyancy based on 19 MW plume heat Lifetime dose of 25 rem yields 10-mile ILCFR from 2.5e-5 to 6.1 e-3 Difference between MACCS and analytic calculation ranges from 38% to 566%

Analytic calculation can be either conservative or non-conservative relative to MACCS calculation All cases produce MACCS ILFCR > 2e-6

Case 3: Wake Effects Results 66 Case OVERALL Peak dose (Sv) at 100 m EARLY Peak dose (Sv) at 100 m CHRONC Peak dose (Sv) at 100 m MACCS 10- mile ILCFR MACCS fitted n Analytic 10-mile ILCFR Percent Difference 3A 2.5E-01 2.3E-02 2.3E-01 5.5E-07 1.5 1.7E-06 210%

3B 2.5E-01 2.3E-02 2.3E-01 5.3E-07 1.7 8.1E-07 53%

3C 2.5E-01 2.3E-02 2.3E-01 6.3E-07 1.6 7.3E-07 16%

3D 2.5E-01 2.3E-02 2.3E-01 1.2E-06 1.3 1.3E-06 13%

3E 2.5E-01 2.3E-02 2.3E-01 1.5E-06 1.2 1.5E-06 2.7%

3F 2.5E-01 2.3E-02 2.3E-01 1.9E-06 1.1 1.7E-06

-11%

Difference from Case 1: Eimutis and Konicek representation for Y and z coupled with Ramsdell-Fosmire model for plume meander and wake effects Lifetime dose of 25 rem yields 10-mile ILCFR from 5.5e-7 to 1.9e-6 Difference between MACCS and analytic calculation ranges from 3% to 210%

Analytic calculation generally conservative relative to MACCS calculation All cases produce MACCS ILFCR <2e-6

Case 4: Protective Actions Results 67 Case OVERALL Peak dose (Sv) at 100 m EARLY Peak dose (Sv) at 100 m CHRONC Peak dose (Sv) at 100 m MACCS 10- mile ILCFR MACCS fitted n Analytic 10-mile ILCFR Percent Difference 4A 2.5E-01 1.9E-01 6.1E-02 4.1E-06 1.2 6.0E-06 47%

4B 2.5E-01 1.9E-01 6.1E-02 3.1E-06 1.4 2.2E-06

-29%

4C 2.5E-01 1.9E-01 6.0E-02 2.6E-06 1.3 1.8E-06

-31%

4D 2.5E-01 1.9E-01 6.0E-02 2.8E-06 1.2 2.3E-06

-17%

4E 2.5E-01 1.9E-01 5.9E-02 2.2E-06 1.2 1.7E-06

-23%

4F 2.5E-01 1.9E-01 5.7E-02 1.5E-06 1.2 9.1E-07

-38%

• Difference from Case 1: Early phase relocation at 1-5 rem and late phase interdiction/decontamination at 2 rem in first year and 500 mrem in second year

• Lifetime dose of 25 rem yields 10-mile ILCFR from 1.5e-6 to 4.1e-6

• Difference between MACCS and analytic calculation ranges from 17% to 47%

• Analytic calculation is generally non-conservative relative to MACCS calculation

• Most cases produce MACCS ILFCR >2e-6

Case 5: Meteorological Sampling Results 68 Case OVERALL Peak dose (Sv) at 100 m EARLY Peak dose (Sv) at 100 m CHRONC Peak dose (Sv) at 100 m MACCS 10- mile ILCFR MACCS fitted n Analytic 10-mile ILCFR*

Percent Difference 5A 2.5E-01 2.7E-02 2.2E-01 1.3E-07 1.8 4.5E-07 238%

5B 2.5E-01 2.3E-02 2.3E-01 1.4E-06 1.1 4.5E-06 222%

Difference from Case 1: Weather sampled from SOARCA Peach Bottom meteorological file without (5A) and with (5B) plume buoyancy Lifetime dose of 25 rem yields 10-mile ILCFR from 1.3e-7 to 1.4e-6 Difference between MACCS and analytic calculation ranges from 220% to 240%

Analytic calculation is conservative relative to MACCS calculation

• Transverse dispersion assumed consistent with slightly unstable conditions

Case 6: Multiple Plumes Results 69

• Difference from Case 1:

• Weather sampled from SOARCA Peach Bottom meteorological file

• Multiple plume segments - scaled VF (6A) / LCF (6B) / NOCF (6C) source terms

• Lifetime dose of 25 rem yields 10-mile ILCFR from1.3e-7 to 2.9e-7 for different source terms

• Difference between MACCS and analytic calculation ranges from 45% to 460% for different source terms

• Analytic calculation is conservative relative to MACCS calculation

• MACCS ILCFR is comparable to Case 5 (single plume) for source term 5D Case OVERALL Peak dose (Sv) at 100 m EARLY Peak dose (Sv) at 100 m CHRONC Peak dose (Sv) at 100 m MACCS 10- mile ILCFR MACCS fitted n Analytic 10-mile ILCFR*

Percent Difference 6A 2.5E-01 2.2E-02 2.3E-01 1.3E-07 1.8 7.0E-07 460%

6B 2.5E-01 1.8E-02 2.3E-01 2.9E-07 2.0 4.3E-07 45%

6C 2.5E-01 6.7E-03 2.4E-01 1.3E-07 2.0 4.0E-07 210%

• Transverse dispersion assumed consistent with highly unstable conditions

Case 7: Population Distribution Results 70 Case OVERALL Peak dose (Sv) at 100 m EARLY Peak dose (Sv) at 100 m CHRONC Peak dose (Sv) at 100 m MACCS 10- mile ILCFR MACCS fitted n Analytic 10-mile ILCFR*

Percent Difference 7A 2.5E-01 2.2E-02 2.3E-01 7.3E-08 1.8 7.0E-07 866%

7B 2.5E-01 1.8E-02 2.3E-01 1.5E-07 2.0 4.3E-07 180%

7C 2.5E-01 6.7E-03 2.4E-01 6.3E-08 2.0 4.0E-07 537%

• Difference from Case 1:

• Weather sampled from SOARCA Peach Bottom meteorological file

• Multiple plume segments - scaled VF (7A) / LCF (7B) / NOCF (7C) source terms

• Population distribution based on Peach Bottom site file

• Lifetime dose of 25 rem yields 10-mile ILCFR from 6.3e-8 to 1.5e-7 for different source terms

• Difference between MACCS and analytic calculation ranges from 180% to 866%

• Realistic population distribution resulted on lower ILCFR relative to Case 6, particularly for pulse type releases such as VF/5D.

• Transverse dispersion assumed consistent with highly unstable conditions

Effect of Downwind Dose Reduction Coefficient on Individual Latent Fatality Risk within 10 miles 71 1.0E-08 1.0E-07 1.0E-06 1.0E-05 1.0E-04 1.0E-03 1.0E-02 1.0E-01 1.0E+00

-1.5

-1.0

-0.5 0.0 0.5 1.0 1.5 2.0 2.5 3.0 10 Mile ILCF Conditional Risk Given D_100=25 rem Effective Dose-Distance Reduction Coefficient

Long-Term Time Dependence of Dose 72 Accumulation of dose in years after the event occurs at different rates for different source terms Therefore, there is likely no fixed ratio between early phase dose, first year dose, and 50-year cumulative dose However, for the scaled source terms considered in this analysis, a first-year dose of 2 rem appears to correspond to a lifetime dose*

less than 25 rem, probably due to radioactive decay and the effect of weathering on groundshine and resuspension Case Early Phase First Year CHRONC Second Year CHRONC 50 year Cumul CHRONC*

50 year Cumul TOTAL*

PAGs 1-5 2

0.5 Not specified Not specified VF/5D 0.70 2.0 1.0 9.0 9.7 LCF/1B 0.13 2.0 0.2 3.5 3.6 NOCF/2R1 0.11 2.0 0.4 5.2 5.3

• Cumul.: cumulative 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00 0

10 20 30 40 50 Ratio of Dose at Given Year to First Year Dose Years Since Release VF/5D LCF/1B NOCF/2R1

• Lifetime dose, in this analysis, is assumed to be the dose resulting from a 96-hour (4 day) early phase exposure and a 50-year late phase exposure.

Summary 73

• Analytic derivation of relationship between 100 m lifetime dose and 10-mile population-weighted ILCFR developed and used to identify assumptions for examination with MACCS.

• A 25-rem lifetime dose at 100 meters generally corresponds to a 10-mile population-weighted lifetime ILCFR less than 2e-6, unless buoyant releases or protective actions are credited for computing dose at 100 m.

• The relationship is sensitive to the value used for the downwind dose reduction coefficient.

• There is likely no fixed ratio between early phase dose, first year dose, and 50-year cumulative dose.

• For the scaled source terms considered in this analysis, a first-year dose of 2 rem appears to correspond to a 50-year dose less than 25 rem, probably due to radioactive decay and the effect of weathering on groundshine and resuspension.

Bibliography 74

• U.S. Nuclear Regulatory Commission 2022, U.S. NRC Level 3 Probabilistic Risk Assessment (PRA) Project, Volume 3d: Reactor, At-Power, Level 3 PRA for Internal Events and Floods, Draft for Comment (ML22067A215)

• Holzworth, G.C, 1972. Mixing Heights, Wind Speeds, and Potential for Urban Air Pollution Throughout the Contiguous United States (AP-101), Research Triangle Park, NC: Office of Air Programs, U.S. Environmental Protection Agency, January 1972.

• U.S. Nuclear Regulatory Commission, 1983. Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Rev. 1 (RG 1.145), Washington DC: U.S. Nuclear Regulatory Commission, November 1982, Reissued February 1983 (ML003740205)

• Jow, H-N, J.L. Sprung, J.A. Rollstin, L.T. Ritchie, and D.I Chanin, 1990. MELCOR Accident Consequence Code System (MACCS): Volume 2, Model Description (NUREG/CR-4691 / SAND86-1562), Albuquerque, NM: Sandia National Laboratories, February 1990. (ML063560409)

Confirmatory MACCS Calculations Supplemental Slides 75

Case 0: Simple Model Peak Dose vs Distance 76 Stability Class A:

Extremely Unstable 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC) 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

Stability Class F:

Strongly Stable

Case 1: Pasquill-Gifford Stability Peak Dose vs Distance 77 Stability Class A:

Extremely Unstable Stability Class F:

Strongly Stable 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC) 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

Case 2: Plume Buoyancy Peak Dose vs Distance 78 Stability Class A:

Extremely Unstable Stability Class F:

Strongly Stable 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC) 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

Case 3: Wake Effects Peak Dose vs Distance 79 Stability Class A:

Extremely Unstable Stability Class F:

Strongly Stable 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC) 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

Case 4: Protective Actions Peak Dose vs Distance 80 Stability Class A:

Extremely Unstable Stability Class F:

Strongly Stable 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC) 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

Case 5: Meteorological Sampling Peak Dose (Mean) vs Distance 81 Without plume buoyancy With plume buoyancy 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC) 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

Case 6/7: Weather Sampling with Multiple Plumes Peak Dose (Mean) vs Distance 82 VF/5D 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

LCF/1A2 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

NOCF/2R1 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 0.1 1

10 Peak Dose (Sv)

Distance downwind (km)

OVERALL EARLY CHRONC Power (OVERALL)

Power (EARLY)

Power (CHRONC)

DG-1413 (proposed new RG 1.254)

Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants Mihaela Biro Division of Risk Assessment Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission

Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants (DG-1413)

• Section A: Applies to LWRs and non-LWRs licensed under Parts 50, 52, and 53 (Frameworks A and B)

• Section B (Discussion):

o Identifies licensing events for each licensing framework o Provides historical perspectives (early licensing, development of the standard review plan

[SRP])

o Addresses ACRS recommendations to start with a blank sheet of paper (10/7/2019, 10/21/2020, 5/30/2021, and 10/26/2021)

• Section C (Staff Guidance) provides an integrated approach for:

o Conducting a systematic and comprehensive search for initiating events o Delineating a systematic and comprehensive sets of event sequences o Grouping the lists of initiating events and event sequences into licensing events

• Appendix A (Comprehensive Search for Initiating Events):

o Reviews techniques for searching for initiating events and points the user to helpful references o Does not endorse or recommend any specific technique 84

85 Licensing Pathways and Licensing Events Regulation and Application Type Reactor Type Use of LMP Licensing Event Categories Risk Evaluation Part 50 CP, OL LWR not applicable (NEI 18-04, Rev. 1 and RG 1.233 currently only apply to non-LWRs licensed under Parts 50 or 52)

• Design-basis events (DBEs) (§ 50.49):

o AOOs o DBAs (i.e., postulated accidents) o External events o Natural phenomena

• Non-DBA (§ 50.2 alternate ac source)

• Beyond-design-basis events (BDBE)

• Anticipated transient without scram

• Station black out not required (Parts 50/52 lessons-learned rulemaking)

Part 52 DC, SDA, ML, COL PRA required Part 50 CP, OL Non-LWR no not required (Parts 50/52 lessons-learned rulemaking)

Part 52 DC, SDA, ML, COL PRA required Part 50 CP, OL Non-LWR yes Licensing events are collectively referred to as licensing-basis events (LBEs), which include the following categories:

• AOOs

• DBEs

• BDBEs

• DBAs PRA implied by use of LMP Part 52 DC, SDA, ML, COL PRA required Part 53, Framework A CP, OL, DC, SDA, ML, COL LWR or non-LWR not applicable (potential future update to NEI 18-04 and RG 1.233)

Licensing events are collectively referred to as LBEs, which include the following categories:

• AOOs

• Unlikely event sequences

• Very unlikely event sequences

• DBAs PRA required Part 53, Framework B CP, OL, DC, SDA, ML, COL LWR or non-LWR not applicable

• AOOs

• DBAs

• Additional licensing-basis events

• Severe accidents PRA or AERI required

Identify application-specific factors (licensing framework, plant-specific design features, and site characteristics).

Conduct a systematic and comprehensive search for initiating events.

Use a systematic process to delineate a comprehensive set of event sequences.

Group initiating events and event sequences into designated licensing event categories according to the selected licensing framework.

Provide assurance that the set of licensing events is sufficient.

Overarching Principles 86

Establish Quality Control Program Select Initiating Event Identification Methods

• Inductive methods

• Deductive methods

• Human-induced events (Appendix provides discussion and references)

Define Initiating Event Grouping Strategy and Characteristics Collect information on plant design, plant operating states, and site characteristics 1

3 6

9 8

Identify Plant-specific Safety Functions

• Systems needed to achieve safety functions

• Operator actions needed to achieve safety functions

• Success criteria X

Assemble Multi-disciplinary Team 2

Collect Application-Specific Information Select Analysis Methods Select Analytical Methods for Event Sequences (e.g., Event Trees, Event Sequence Diagrams) 7 Define Plant-specific End States for Event Sequences 10 Identify Radiological Sources and Transport Barriers from the Source to the Environment 5

Identify Sources of Hazardous Chemical Materials 4

Technology-Inclusive Identification of Licensing Events (Sheet 1 of 3) to Sheet 2 87 Updated text on safety functions, consistent with Part 53 Framework A Added guidance on establishing a Quality Control program prior to engaging in the work

Apply Selected Analytical Methods

• Identify initiating event impact on safety functions

• Identify the impact of front-line and support system dependencies on safety functions

• Identify the impact of operator actions on safety functions Independent Review and Quality Control List of Event Sequences 17 16 Account for Relevant Operating Experience and for Insights from Earlier Relevant Analyses 15 Independent Review and Quality Control Account for Relevant Operating Experience and for Insights from Earlier Relevant Analyses List of Initiating Events 13 14 Initiating Event Analysis Event Sequence Selection Apply Initiating Event Identification Methods Apply Initiating Event Grouping Strategy 12 11 X

Y Technology-Inclusive Identification of Licensing Events (Sheet 2 of 3) from Sheet 1 to Sheet 3 Added references for listing of external hazards.

The search for Initiating Events and Event Sequences is subject to Quality Control (not QA)

Is a PRA being developed to support the application?

Define Licensing Event Grouping Strategy and Characteristics

• Group by frequency o Qualitative o Quantitative

• Group by type o Plant response following the initiating event (sequence of events, timing) o Similar challenge to safety functions o End state Apply Licensing Event Grouping Strategy Compare to Predefined Lists (e.g., SRP Chapter 15, previous CP, OL, DC, SDA, ML, or COL applications) and identify differences from SRP (only for LWRs)

Identify Limiting Cases for Each Group of Licensing Events Independent Review and QA List of Licensing Events 21 22 23 24 25 Follow NEI 18-04, Rev. 1 as endorsed in RG 1.233 20 18 Identify Required Categories of Licensing Events for Licensing Framework Provide initiating events and event sequences to the PRA All other applications Part 50 or 52 non-LWR applications based on LMP no yes (PRA) 19 Y

Technology-Inclusive Identification of Licensing Events (Sheet 3 of 3) from Sheet 2 Clarified that the LMP guidance currently applies to non-LWRs under parts 50 or 52.

Note: The staff intends to revise RG 1.233 to address licensing under Part 53 Framework A in the future.

The search for Licensing Events is subject to QA

• A Quality Control Program should be established prior to engaging in the work; includes personnel, procedures, documentation.

• The initiating event and event sequence analyses are not subject to QA requirements (PRA is not part of the design-basis information).

• Existing programs may be leveraged:

o If a PRA is developed, PRA Configuration Control can be used for analysis documentation.

o If a PRA is developed, PRA peer review can be used for independent review.

Quality Control Program 90 The licensing event selection informs the design basis and licensing basis; therefore, it is subject to QA requirements.

DG-1414 (proposed new RG 1.255)

Alternative Evaluation for Risk Insights Methodology 91 Anne-Marie Grady Division of Risk Assessment Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission

• This RG provides the NRC staffs guidance on the use of an AERI methodology to inform the content of applications and licensing basis for LWRs and non-LWRs.

• 10 CFR 53.4730(a)(34)(ii) establishes AERI as an alternative to a PRA for a risk evaluation if the entry conditions A and B for an AERI are met.

• The title of this DG-1414 is now AERI Methodology, to distinguish it from Part 53 Frameworks A and B. This new title does not signal any change in approach.

Alternative Evaluation for Risk Insights Methodology 92 Applicants who meet the AERI entry conditions may elect to develop an AERI in lieu of a PRA.

However, PRA confers additional benefits such as:

• A means to optimize the design, and

• The ability to take advantage of various risk-informed initiatives, for example risk-informed completion times, risk-informed categorization of SSCs.

93 93 Perform transient and accident analyses Perform design basis accident radiological consequences analyses Identify and analyze the bounding event Finish PRA development Select LBEs Select DBAs Classify SSCs Continue design and licensing activities Evaluate DID Comprehensive and systematic initiator search and event sequence delineation without preconceptions or reliance on predefined lists Select licensing events Select licensing framework Perform transient and accident analyses Perform design basis accident radiological consequences analyses Elect to develop PRA Finish PRA development Continue design and licensing activities Continue design and licensing activities A

Parts 50 and 52 with LMP Part 53 Framework A Parts 50 and 52 without LMP Part 53 Framework B B

C D

E F

G H

I J

K L

M N

O yes no Applicant decision DG-1413, Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants DG-1414, Alternative Evaluation for Risk Insights (AERI) Methodology LMP guidance - NEI 18-04, Rev. 1, as endorsed in RG 1.233 AERI entry condition met?

P yes no Q

Licensing Pathways - Risk Evaluation Perspective Alternative Evaluation for Risk Insights Notes:

1)

Each step builds on all of the preceding steps (considers all information available at that point) 2)

Feedback loops (e.g., the impact of design revisions) are not shown AERI Q1 Develop demonstrably conservative risk estimate using the bounding event Q2 Search all event sequences for severe accident vulnerabilities Q3 Develop risk insights by reviewing all event sequences Q4 Assess DID adequacy by reviewing all event sequences ONLY for Part 53 Framework B 93

• DG-1414 applies only to LWRs and non-LWRs licensed under Part 53, Framework B

• Identification and characterization of the postulated bounding event(s):

o Selection of licensing events is covered in DG-1413 o

Consider both core and non-core radiological sources o

Perform consequence analysis for selected licensing event(s) o Multiple bounding events could be considered for events with approximately similar likelihoods of occurrence and similar overall radiological impacts, but with different radiological release characteristics

• Estimate dose consequence for the postulated bounding event to confirm that the reactor design meets the AERI entry conditions:

o Condition A - Consequences evaluated at 100m (328 feet) from plant do not exceed:

10 mSv (1 rem) TEDE over the first four days following a release,

An additional 20 mSv (2 rem) TEDE in the first year, and

5 mSv (0.5 rem) TEDE in second and subsequent years o

Condition B - Condition A must be met without reliance on active safety features or passive safety features, except passive safety features that:

Do not require equipment actuation or operator action to perform their required safety functions,

Are expected to survive accident conditions, and

Cannot be made unavailable or otherwise defeated by credible human errors of commission and omission o

One acceptable approach to developing a dose consequence estimate is to provide the postulated bounding event source term to MACCS or a comparable analytical model Elements of the AERI Methodology (1 of 4) 94

• Determination of a demonstrably conservative risk estimate for the postulated bounding event to demonstrate that the QHOs are met:

o Utilize consequence estimate.

o Assume a frequency of 1/yr to represent the sum of the event sequence frequencies (based on LWR statistics equal to the sum of initiating event frequencies).

o Compare to QHOs.

o Applicant may use a different frequency, with justification, which NRC staff will review on a case-by case basis.

o One acceptable approach to developing a dose consequence estimate is to provide the postulated bounding event source term to MACCS, or a comparable analytical model.

o The applicant should identify the software codes used for the consequence analyses and provide information on how the development and maintenance of these software codes meets quality standards commensurate with the application.

Elements of the AERI Methodology (2 of 4) 95

• Search for severe accident vulnerabilities:

o Severe accidents are those events that progress beyond the DBAs, in which substantial damage is done to the reactor core or to any other structure, vessel, or retention system containing a significant inventory of radiological material, whether or not there are serious offsite consequences o

Severe accident vulnerabilities are aspects of a design which represent an overreliance on a single design feature, either for accident prevention or mitigation, that could lead to a severe accident o

Encompasses the entire set of licensing events and any additional severe accidents o

Search for cliff-edge effects o

Consider external hazards

• Address how identifying severe accident vulnerabilities could enable the design to prevent or mitigate severe accidents

• Justify why a severe accident vulnerability is acceptable for the design Elements of the AERI Methodology (3 of 4) 96

• Identification of risk insights:

o The objective of the search for risk insights is to understand issues that are important to plant operation and safety such as:

important hazards and initiators

important event sequences and their associated SSC failures and human error

system interactions

vulnerable plant areas

likely outcomes

sensitivities

areas of uncertainty o Search encompasses the entire set of licensing events o Provides an understanding of the hierarchy of event sequences ranked by frequency

• Assessment of DID adequacy:

o Encompasses the entire set of previously identified licensing events o Facility design should include a reasonable balance among the layers of defense, to ensure that failure of a single barrier does not result in a severe accident Elements of the AERI Methodology (4 of 4) 97

• Assure that the AERI risk evaluation continues to be valid, useful, and an adequate basis for regulatory decision-making throughout the plant operating lifetime.

o The initial risk evaluation must be performed by the scheduled fuel load date o The risk evaluation should be maintained/upgraded every five years

• Regularly assess that the postulated bounding event selection remains current o If not, identify new postulated bounding event to be used in the upgraded risk evaluation

• As-built, as-operated facility o Ascertain if any important aspects of the facilitys design or operational scheme have changed since the prior risk evaluation, and if so, maintain/upgrade the risk evaluation

• New safety issue(s) o Ascertain if any new safety issues have arisen since the prior risk evaluation, and if so, maintain/upgrade the risk evaluation

• New data, information, or analyses o Ascertain if any relevant new data, information or analyses have arisen since the prior risk evaluation, and if so, maintain/upgrade the risk evaluation Maintaining and Upgrading the AERI Risk Evaluation (1 of 2) 98

• QHO comparison o If the AERI risk evaluation requires upgrading, the QHO comparison should be revisited and modified, if appropriate

• Vulnerability search o If the AERI risk evaluation requires upgrading, the severe accident vulnerability search should be revisited and modified, if appropriate

• Search for Risk Insights o If the AERI risk evaluation requires upgrading, the search for risk insights should be revisited and modified, if appropriate

• DID o If the AERI risk evaluation requires upgrading, the DID evaluation should be revisited and modified, if appropriate Maintaining and Upgrading the AERI Risk Evaluation (2 of 2) 99

Discussion 100

Final Discussion and Questions 101

Agenda - October 19th 102 8:35 am - 1:00 pm Requirements for Operations: Draft Proposed Language for Staffing, Role of STA, and Guidance 1:00 pm - 2:00 pm Lunch 2:00 pm - 5:00 pm Draft Proposed Language Addressing Other ACRS Comments and Major Industry Comments

Preliminary Requirements for Operations:

Rule Language Updates, Staffing Topics, and Overview of Key Guidance 103

Agenda

• Introduction

• Updates to Subparts F and P since the 2nd Iteration Consolidation of requirements under Subpart F Current status of engineering expertise requirements Current status of GLRO requirements Response to recent ACRS letter

• Overview of ISG for Operator Licensing Program Reviews

• Overview of ISG for Staffing Plan Reviews

• Overview of ISG for Scalable Human Factors Engineering (HFE) Reviews

• Questions 104

Overview of Primary Staff Contributors (NRR & RES)

• Theresa Buchanan, Senior Reactor Engineer (Examiner)

• Dr. David Desaulniers, Senior Technical Advisor for Human Factors and Human Performance Evaluation

• Dr. Brian Green, Senior Human Factors Engineer (Team Lead)

• Dr. Niav Hughes Green, Human Factors Psychologist

• Dr. Stephanie Morrow, Human Factors Psychologist

• Lauren Nist, Branch Chief, Operator Licensing and Human Factors Branch

• Maurin Scheetz, Reactor Engineer (Examiner)

• Jesse Seymour, Senior Reactor Engineer (Examiner) 105

Updates to Subpart F and P since the 2nd Iteration Requirements for HFE, staffing, operator licensing, and training have all been consolidated under Subpart F, with Subpart P now just containing a single pointer located at 53.4220 (i.e., Framework A and B now use a common set of requirements in these areas)

The class of reactors meeting the technical requirements for utilizing GLROs has been defined as self-reliant mitigation facilities Procedure program requirements have been consolidated Staffing plan requirements for non-operations positions are now functional in nature Examination programs are required to provide for validity and reliability in testing Remedial training is mandated for operators failing requalification examinations Commission approval is no longer required for simulation facilities 106

Status of Engineering Expertise &

GLRO requirements Engineering expertise remains a required element of staffing plans for all facilities under both Frameworks A and B, including for those facilities staffed by GLROs Criteria for potentially allowing facilities under Framework B to use GLROs have been incorporated, in addition to those already in place for Framework A Framework B GLRO criteria vary depending on whether an AERI is used Irrespective of AERI, DID without human action is needed For a non-AERI plant, the GLRO criteria are analogous to the equivalent criteria for Framework A, as adapted to the differing requirements of Framework B For an AERI plant, the GLRO criteria are met by meeting AERI criteria (plus DID)

These various sets of criteria have a common goal of identifying when operators are not expected to significantly influence safety outcomes based on the design GLRO criteria now are specific to limiting analysis to credited human actions 107

Response to Interim Letter Report from August 2022

• ACRS letter included a recommendation that the associated guidance for implementing 10 CFR Part 55 can be amended to accommodate the objectives of the proposed rule without the additional voluminous text.

• Key points form the staff response included the following:

New framework for operator licensing under Part 53 is technology-inclusive and creates significant flexibilities compared to Part 55 Accommodating such flexibilities while complying with statutory requirements necessitates requirements for GLROs being codified in regulations Absent Part 53s alternative, applicants would be required to adhere to Part 55 While revised or new guidance could be developed, applicants would be required to seek exemptions and justify pursuing alternative approaches, requiring NRC staff reviews on an application-specific basis; proposed Part 53 will remove the need for exemptions and enhance regulatory reliability and clarity 108

Overview of ISG for Operator Licensing Program Reviews DRO-ISG-2023-01 Operator Licensing Programs Draft Interim Staff Guidance

Purpose

• To assist staff reviews of applications under 10 CFR Part 53 related to the operator licensing examination program

• To provide guidance for review of tailored initial and requalification examination programs

• For specifically licensed operators (SROs and ROs)

• For generally licensed operators (GLROs)

• To address proficiency for SROs and ROs

• To assist staff reviews of exemptions from 10 CFR Part 55 for non-LWR, power reactor examination programs 110

Goals

• Enable facility applicants/licensees to identify knowledge, skills, and abilities (KSAs) necessary for safe operation as the basis for the examination standards

• Establish reliable guidelines for exam program developments based on current best practices from research and expertise on the measurement and testing of KSAs 111

Section 1.0 KSAs List Development

• Systems approach to training-based processes are used to identify a training KSA list

• This list is not solely limited to tasks related to safe plant operation

• DRO-ISG-2023-04, Facility Training Programs, is planned to provide additional information in this area

• Using this list as a starting point, a screening is performed to identify those tasks important to safe plant operation and/or related to the foundational theory of plant operations to develop the KSA list for the exam program

• Depending on the original list, may have needed to add or remove items to get the necessary KSAs for testing 112

113

Section 2.0 Operator Licensing Test Development

• Developed Test Plan

• How the testable KSAs will be measured

• For example, what KSAs will be tested using a written test, or a walkthrough format, etc.

• What the format for the test will be

• Developed detailed content specification

• What specific KSAs the exam type (written, oral, scenario, job performance measure, etc.) covers

• How the KSAs are sampled for each examination developed

• How the test items are reviewed for clarity, quality, and other psychometric issues 114

Section 3.0 Examination Validity

• Describe validation plan

• What evidence was collected to support validity of the test, that the test works and will work as intended

• Content validity, concurrent validity

• Should require content validity at the least 115

Section 4.0 Scoring Specifications

• Criterion-referenced

• Described how each test item is scored and how scores combined to get total score

• If based on scorer observation, described steps to eliminate any bias in judgments

• Provided cut-off score 116

Section 5.0 Reliability of the Test

• If individual repeats the test, the result would be similar to the original result

• Documentation that the tests will have stability of test performance over time

• Documentation of findings that are adequate to justify use of the test for operator licensing 117

Section 6.0 Test Manual

• Companion to the test plan

• Provides more detail related to the specific types of tests

• Includes administrative aspects of test

• How to administer

• Time to administer or time allowed to take the test

• Materials provided to test takers

• How to interpret test results 118

Section 7.0 Additional Characteristics of High-Quality Test Materials

• This section is specifically for written and computer-based tests

• Provides additional characteristics associated with psychometrics, test instructions, objective scoring system, and standardization 119

Section 8.0 Other Examination Program Considerations

• This section references back to sections of NUREG-1021, Operator Licensing Examination Standards for Power Reactors for items that are universally applicable, regardless of plant design 120

Section 9.0 Simulation Facilities

• Documentation on how the simulation facility provides a level of fidelity sufficient to assess KSAs as required by 10 CFR Part 53.780(e) or 53.815(e)

• Simulation facilities should have same cognitive requirements as the real environment

• For simulation-based assessment, documentation provided on how that examination is valid 121

Section 10.0 Administering Operating Tests

• Examination procedures should be similar to those in NUREG-1021, as specific to the type of test administered

• Measures are in place to ensure examiners behave in accordance with codes of conduct to ensure examination integrity

• Measures are in place to retain required records 122

Section 11.0 Examination Program Change Management Process

• Documentation specifies what changes require NRC approval and which do not

• NRC approval

• Exemption from regulation

• Change to technical specification

• Negative impact to examination security/integrity

• Negative impact on consistency 123

Section 12.0 Static Computer-Based Testing

• Beyond the scope of the guidance

• The documentation would need to describe how this approach is equivalent to the guidance provided in the ISG 124

Section 13.0 Additional Guidance for Requalification Programs

• Any requalification failures must be remediated and retested prior to returning to license duties

• For ROs and SROs

• Periodicity not to exceed 24 months

• For GLROs

• Periodicity defined by program

• If >24 months, bases provided 125

Section 14.0 Proficiency Programs for Specifically Licensed Operators and Senior Operators

• Actively perform the functions

• Maintain proficiency and familiarity

• Re-establish proficiency if it cannot be maintained 126

Section 15.0 Waivers for GLROs

• Appropriate criteria to waive requirements for an examination included in the program

• If similar to 10 CFR 55.47, no further NRC review

• Else, a basis is provided that describes how the criteria ensures individuals are able to safely and competently operate the facility 127

Appendix A Currently Approved Examination Methods

• Methods currently approved in NUREG-1021 can be used without needing further basis from the facility or additional NRC review

• Example: use of a 4-part multiple choice written examination with 80%

cut score 128

Overview of ISG for Staffing Plan Reviews DRO-ISG-2023-02 Interim Staff Guidance Augmenting NUREG-1791, Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR 50.54(m), for Licensing Plants under Part 53

Background:

Current Practice

• Current 10 CFR 50/52 staffing requirement (i.e.,

50.54(m)) is prescriptive

• NRC reviews exemptions to this requirement using NUREG-1791, Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR 50.54(m)

Developed with advanced reactors in mind Performance-based process for determining appropriate number of licensed control room operators 11 steps including a staffing plan validation

• Staff used NUREG-1791 to evaluate novel control room staffing models for NuScale SMR design and concept of operations

• Cannot use NUREG-1791 as written for Part 53 staffing plan reviews because it relies on exemptions to Part 50 requirements 130

Part 53 Approach to Staffing

• Applicant proposes minimum staffing level by submitting a staffing plan with application

• Consider differences in staffing level when operators have/do not have a safety role (i.e., for specific or generally licensed operators) - if specific licenses then applicants must include more detail supported by HFE analysis and assessments

• Operators may fill multiple roles (e.g., maintenance, radiation protection, etc.) so must include these responsibilities in staffing plan submittal

• The staff will review and approve the staffing plan.

Changes to approved staffing plans are subject to administrative controls 131

Preliminary Part 53 Staffing Requirement

• Addressed under the preliminary requirements of § 53.730(f):

• A staffing plan must be developed to include the numbers, positions, and qualifications of operators and senior operators or, if applicable, generally licensed reactor operators across all modes of plant operations, as well as a description of how the numbers, positions, and responsibilities of personnel contained within those plans will adequately support all necessary functions within areas such as plant operations, equipment surveillance and maintenance, radiological protection, chemistry control, fire brigades, engineering, security, and emergency response.

132

Proposed Part 53 Requirement for On-Shift Engineering Expertise

[§ 53.730(f)(1)]

• The staffing plan must include a description of how engineering expertise will be available to the on-shift crew during all plant conditions to assist in situations not covered by procedures or training

• A person available to support the crew at all times. This person is familiar with the operation of the facility and has a technical degree:

Bachelors in in engineering or, Bachelors in engineering technology or a physical science or, PE license

• Basis: Commission policy for, Education for Senior Reactor Operators and Shift Supervisors at Nuclear Power Plants, (published in the Federal Register (54 FR 33639) on August 15, 1989) 133

DRO-ISG-2023-02: for review of Part 53 staffing plans

• Objective is to guide reviewer through the process of:

Evaluating staffing plans and support analyses submitted under § 53.730(f)

Determining whether the proposed minimum staffing level provides assurance that plant safety functions can be maintained across all modes of plant operations Approving staffing plans

• For plants that will have specifically licensed operators; could scale the review for plants with generally licensed operators

• Use in conjunction with NUREG-1791

• 11 steps that rely on other Human Factors elements

• Includes review guidance for engineering expertise requirement

• Developed as an Interim Staff Guide (ISG)

Following experience with using the ISG the staff plans to update NUREG-1791 134

DRO-ISG-2023-02: for reviewing engineering expertise

• Guidance on what staff will look at for satisfying engineering expertise requirement to include:

• Education prerequisites

• Training and qualification

• Responsibilities of the job

• Data needs if offsite

• Response time if on site

• Expectations for one or multiple people filling the job

• Communication needs

• Cybersecurity expectations

• Include job in validation activities 135

Overview of ISG for Scalable Human Factors Engineering Reviews DRO-ISG-2023-03 Development of Scalable Human Factors Engineering Review Plans

Background:

Current Practice

• Current 10 CFR 50 HFE requirement (i.e.,

50.34(f)(2)(iii)) is focused on the main control room

• NRCs HFE reviews for large light-water reactors have been conducted using NUREG-0711, Human Factors Engineering Program Review Model

• Systems engineering based approach

• 12 program elements and 300+ criteria

• Lessons-learnt from recent Part 52 reviews indicated a need for a new approach to regulation and review of HFE for advanced reactor technologies 137

Background:

Proposed Part 53 Approach to HFE

• HFE to be required where necessary to support important human actions

• HFE reviews to be application specific (i.e., scaled) considering the characteristics of the facility design and its operation 138

Background:

Preliminary Part 53 HFE Requirement

• Addressed by the preliminary requirement of § 53.730(a)

• The plant design must reflect state-of-the-art human factors principles for safe and reliable performance in all locations that human activities are expected for performing or supporting the continued availability of plant safety or emergency response functions 139

Background:

Draft Guidance

• Objective is to guide reviewer through the process of:

• Developing an application specific review plan

• Identifying appropriate HFE review guidance

• To be used in place of NUREG-0800, Chapter 18, Human Factors Engineering

• Developed as an ISG

• Following experience with using the ISG the staff plans to make the guidance a NUREG 140

Scaling Process:

Overview

• Begins - during pre-application engagements (if conducted)

• Concludes - with completion of application acceptance review

• Conducted - in 5 steps leading to the staff assembling the review plan 141

Scaling Process:

5 Steps 1.

Characterization - establishing a documented understanding of the design and its operation from an HFE perspective 2.

Targeting - identifying aspects of the design and operation for HFE review 3.

Screening - selecting HFE program elements /

activities for review in conjunction with each target 4.

Grading - selecting specific standards and guidance documents to be applied to the review 5.

Assembling the review plan - integrating results of prior steps to produce a plan that supports an efficient, risk-informed, reasonable assurance determination

Scaling Guidance:

Overview

• Main body (22 pages) - provides essential guidance for developing the review plan

• Appendices (88 pages) - provide supporting guidance for implementing each step of the process 143

Scaling Guidance: Main Body - Key Features

• Applicability:

• SDAs, DCs, COLs and OLs

• Rationale for scaling reviews

• Regulatory basis / acceptance criteria

• Guidance for each step of scaling process

• Objective

• Process

• Reviewer Responsibilities

• Focus is on what to do / accomplish when scaling reviews 144

Scaling Guidance:

Appendices -

Key Features

• Focus is on how to

• Recommended methods for each step of scaling process

• Pointers to sources of additional guidance 145

Scaling Guidance:

Appendix A Characterization:

• What to include in the characterization -

essential elements

• How to organize and document the characterization

• Use of the characterization to aid coordination with related reviews (e.g.,

staffing, operator licensing, instrumentation and controls) 146

Scaling Guidance:

Appendix B Targeting:

• General principles for target selection

• Descriptions of 38 prospective (example) characteristics of advanced reactor designs and operations

• Human performance implications

• Availability of guidance to support reviews

Scaling Guidance:

Appendix C Screening:

• General strategies and specific considerations for selecting which HFE activities to review or screen out

• Implications / challenges of advanced reactor design characteristics for certain HFE activities or their review 148

Scaling Guidance:

Appendix D Grading:

• Guidance for selection of standards and guidance documents to support the review

• Considerations for use of documents that lack prior NRC endorsement

• Reference table of HFE standards and guidance documents in both nuclear and non-nuclear domains 149

Scaling Guidance:

Appendix E Assembling the Review Plan:

• Strategies for integrating the results of Steps A-D to develop a plan that is efficient yet sufficient to support a reasonable assurance determination

• Guidance for documenting the review plan and gaining management approval 150

Discussion 151

Draft Proposed Language Addressing Other ACRS Comments and Major Industry Comments 152

Interim Letter Report; October 21, 2020

1. The staffs proposed approach for developing the Title 10 of the Code of Federal Regulations (10 CFR) Part 53 rule is viable.

2. The staff should ensure that applicants compensate for novel designs with uncertainties due to incompleteness in the knowledge base by performing systematic searches for hazards, initiating events, and accident scenarios with no preconceptions that could limit the creative process.

Concern addressed by requirements in both frameworks requiring systematic assessments to identify events supporting the design and licensing of commercial nuclear plants. Examples include §§ 53.240 and 53.450 in Framework A and § 53.4730 in Framework B. In addition, proposed guidance provided in DG-1413.

3. The rule should provide a pathway for licensing prototype facilities, when uncertainties in the knowledge base and lack of operating experience suggest that additional testing and monitoring are needed.

Existing pathway for prototype facilities maintained in both frameworks. Provisions included in § 53.440 (Subpart A, common), § 53.440 for Framework A, and § 53.4730 for Framework B. Existing guidance on prototype plants is applicable to Part 53.

Interim Letter Report; May 30, 2021

1. The overall structure of Subparts A through I provides a logical framework for the rule. It is complete with respect to topics that must be covered and addresses the lifetime of a power reactor. It will be helpful to all applicants and to the NRC staff.

2. A coherent and detailed explanation of the integrated intent of the rule and its associated design-specific guidance should be developed as soon as possible and enshrined in the rule itself.

Included some introduction-type sections to various subparts. However, most detailed explanation of the rule provided in Preamble.

Interim Letter Report; May 30, 2021

3. Regarding Subpart B:

a. To this point in the development, we find no value in the two-tiered approach to safety requirements. Alternative integral risk criteria to the QHOs should be investigated.

(a1) Revised Subpart B to eliminate reference to two tiers. However, safety objectives include: (1) ensuring no immediate threat to public health and safety and (2) considering potential risks.

(a2) See previous discussion on QHOs.

b. Desired flexibility to address the broad range of technologies is provided (c) Framework A continues to define a top-down methodology based on criteria, safety functions, and related requirements for SSCs, personnel, and programmatic controls. Framework B requires development of principal design criteria based on LWR general design criteria or other generally accepted standards.

c. The rule should include a set of over-arching general principles in one place (Subpart B) that would apply to any reactor concept.

d. The rule should state that safety analyses must demonstrate that for AOOs all safety related barriers to release are maintained.

(d) Framework A (§ 53.450(e)) requires establishing evaluation criteria for each AOO [anticipated event sequence].

Framework B ((§ 53.4730(a)(5)(iii)) limits offsite dose for AOOs and requires demonstration that events do not escalate to DBA.

e. The rule should state that safety analyses must demonstrate that DBAs achieve and maintain a safe, stable, and subcritical condition.

(e) Framework A (§ 53.450(f)) revised to require safe, stable end state for DBA and subcriticality following LBE required by § 53.440(g). Framework B (§ 53.4730(a)(5)(ii)) requires acceptance criteria for SR SSCs to demonstrate they adequately mitigate the consequences of DBAs. Additional requirements provided though principal design criteria.

Interim Letter Report; May 30, 2021

4. Subpart C, Design and Analysis Requirements, is generally in good shape.

a. The requirement for risk-informed analysis is appropriate if the use of PRA is approached in a graded fashion commensurate with the potential consequences and the simplicity of the design.

b. The requirements for selection and analysis of DBAs must be clarified.

c. The rule eliminates single failure criteria but needs to define the process that replaces it.

a.

Rule language remains general (requiring use of PRA in Framework A) and flexibility afforded through key guidance such as RG 1.247 b.

Requirements to identify and assess DBAs provided in

§§ 53.240 and 53.450(f) in Framework A and

§ 53.4730(a)(5)(ii) in Framework B. Each maintains general alignment with Parts 50/52 in terms of establishing design requirements for safety-related SSCs. Additional information available in guidance documents (e.g., RG 1.233 for Framework A) c.

Use of probabilistic (reliability) criteria instead of single failure criteria for Framework A discussed in Preamble (see also SECY-03-0047)

5. The two recommendations in our first letter report on 10 CFR Part 53 of October 21, 2020, still apply: for novel designs with uncertainties due to incompleteness in the knowledge base, systematic searches for hazards, initiating events, and accident scenarios should be required; and a licensing pathway including additional testing and monitoring akin to prototype testing should be available.

Concern addressed by requirements in both frameworks requiring systematic assessments to identify events supporting the design and licensing of commercial nuclear plants. Examples include §§ 53.240 and 53.450 in Framework A and

§ 53.4730 in Framework B. In addition, guidance provided by developing DG-1413.

Existing pathway for prototype facilities maintained in both frameworks.

Interim Letter Report; February 17, 2022

1. The staff is methodically working through the delicate balance of flexibility and predictability in regulations for operator staffing.

2. The staff should consider the suggestions identified in this letter to ensure the 10 CFR Part 53 approach yields equivalent safety to current regulatory approaches.

Staff agrees with the ACRS, See subsequent iterations and discussions

3. The staff should approach the concept of not having a Shift Technical Advisor (STA) by having the applicant justify why the STA is not needed rather than a blanket elimination of this position. This is particularly important for the expected wide application of first-of-a-kind technologies that may be licensed under this rule.

See subsequent iterations and discussions

4. The concept of non-licensed, certified operators should not be pursued. Staff should focus on adapting the existing approach to the NRC operator licensing process to produce training, qualification, and licensing requirements based on the degree of safety reliance attributed to operator actions for the specific plant design. This should take advantage of inherent and passive safety features of the nuclear power plant.

See subsequent iterations and discussions

5. Staff should develop guidance for judging the acceptability of limited scope simulators.

See subsequent iterations and discussions

• Discussed during previous session

• Additional questions/discussion ?

Interim Letter Report; August 2, 2022

1. There are limitations of the existing QHOs to fully capture the value and risk of nuclear technologies and the large uncertainties associated with evaluating individual and societal risk.

This could inhibit flexibility and opportunities for more innovative approaches as the regulator and applicants learn from new nuclear technologies and associated missions.

• Preliminary proposed rule language includes a definition for safety function

• Definition has generic elements, but is bifurcated to acknowledge fundamental differences between the frameworks

• Defining critical safety functions remains an explicit requirement in Framework A (top-down approach); primary and secondary (additional) safety functions made explicit

• Safety functions are addressed implicitly through the principal design criteria in Framework B, consistent with current bottom-up approach in existing framework Interim Letter Report; August 2, 2022

2. Critical safety functions are foundational to the licensing process.

As such, the requirements for identifying critical safety functions should be common to both frameworks.

• Draft white paper on preapplication engagement for advanced reactor applicants recommends early engagement in several topical areas:

• Principal design criteria

• Selection of LBEs

• SSC classification

• Source term methodology

• QA

• Probabilistic risk assessment

• Safety analysis methods

• Fuel qualification and testing

• Pre-application engagement is optional and at the discretion of the applicant Interim Letter Report; August 2, 2022

3. The staff should require, early in the preapplication process, each applicant to identify numeric safety dose criteria, the critical safety functions, the safety design criteria, and the underlying rationale for their selection and application in the design.

• Fire protection provisions in Framework B have been completely revised (aligned with Framework A) and are now technology-inclusive Interim Letter Report; August 2, 2022

4. The staff needs to ensure that the fire protection requirements in both frameworks are fully technology-inclusive.

• NRC staff agrees that streamlined and efficient regulatory frameworks are desirable and that guidance used where practicable to reduce the size of the rule

• Each framework in the preliminary proposed rule language must be viewed independently (§ 53.010),

with some exceptions

• Requirements in each framework largely replace existing requirements under Parts 50, 52, 55, and 100; either framework is less than half of the existing requirements Interim Letter Report; August 2, 2022

5. The current approach with self-contained requirements for each of the two frameworks is very long.

Furthermore, the rule has a significant amount of implementation detail that could be better located in regulatory guidance. The optics of this approach run counter to a streamlined more efficient licensing process, which is an expectation for many stakeholders. As a result, the rule may be too cumbersome to implement and may not be used.

• Draft requirements in Part 53 are technology-inclusive and significantly more flexible than those in Part 55

• Development of a new category of license operators and facility class requires codification of related regulatory requirements

• Significant amount of new guidance would need to be developed to address recommended approach

• Proposed approach should greatly reduce the need for exemptions while enhancing regulatory reliability and clarity Interim Letter Report; August 2, 2022

6. The proposed GLRO description should provide for qualified operating personnel. However, the associated guidance for implementing 10 CFR Part 55 can be amended to accommodate the objectives of the proposed rule without the additional voluminous text.

• Staff considers two tiers of SSC classification generally too limiting

• Both frameworks generally address safety-related SSCs in a manner consistent with current requirements

• At least one additional tier considered necessary for non-safety related SSCs warranting some type of special treatment due to DID/risk considerations

• Framework A: Non-safety related with special treatment

• Framework B: Important to safety Interim Letter Report; August 2, 2022

7. The results of the PRA can be used to inform SSC classification by aligning the risk assessment and deterministic safety analysis. This should result, in most cases, in just two tiers for classification of SSCs:

Safety Related/Safety Significant and Not-Safety Related/Low Safety Significant.

• Staff agreed with the recommendation and are currently evaluating the most appropriate format for documenting the technical basis for AERI entry criteria, including MACCS validation Interim Letter Report; August 2, 2022

8. The simple novel analysis that provides the technical basis for the entry criteria to be able to use the AERI should be documented either in an appendix to the DG-1414 or in another appropriate document (e.g., NUREG).

Major Industry Feedback 166 Feedback NRC Staff Perspectives Duplicative/overlapping programs Added flexibility for licensees to organize and combine programs as appropriate to avoid duplication.

Manufacturing license expansion Expanded activities to include fabrication of entire reactor including fuel loading.

Two tier safety criteria structure Eliminated two-tiered approach to safety criteria Unify QA requirements (allow broader set of codes and standards)

Enabled flexibility in using codes and standards; QA requirements consolidated in rule and aligned with Appendix B to 10 CFR Part 50 Normal operations Decoupled requirements for normal operation from those for LBEs Add requirements for safe, stable end state conditions Added requirement and clarified in Statements of Consideration

Major Industry Feedback 167 Feedback NRC Staff Perspectives Not require or rely on just LMP or International Atomic Energy Association approach; Part 53 can be methodology neutral Created two distinct frameworks within Part 53 to provide clarity and predictability for applicants using either approach; developed DG-1413 and AERI approach Questioned as low as reasonably achievable (ALARA) in regulations Staff has added Part 20 references to Part 53. Clarified to recognize that a combination of design features and programmatic controls may fulfill ALARA requirements, as appropriate.

Special treatment for non-safety related but safety significant SSCs NSRSS SSCs reduce sole reliance on safety-related SSCs; Requirements can be scaled to achieve desired capability/reliability/overall risk Facility safety program (FSP)

Staff views FSP as an operational benefit. Allows continued use of PRA for evaluating changes, managing risks, and improving the relationship between the NRCs licensing and reactor oversight programs.

More guidance is needed to clarify regulations Staff agrees and has aligned with industry on future guidance needs

Industry Feedback on Framework B 168 Feedback NRC Staff Perspectives Objectives for chemical hazard requirements are unclear Preamble discussion includes amplifying information to address this feedback. Chemical hazards in question would include substances commingled with licensed material or those produced by a reaction with licensed material, consistent with similar requirements in Part 70 Rule language is not technology-inclusive in some areas (e.g., references to mitigation of beyond-design-basis events [MBDBE]

requirements in § 50.155)

Staff revised several sections to ensure that the proposed rule is technology-inclusive, including MBDBE requirements PRA development at CP stage is not reasonable The requirement to have a PRA developed to support a CP application is consistent with the 50/52 rulemaking and other Commission policies Proposed entry conditions for AERI are too conservative AERI entry conditions distinguish between plants with relatively straightforward designs and plants with relatively complicated designs that warrant the development of a PRA in order to understand their risk. The proposed AERI option is a departure from current Commission policy, which requires all new plants to have a PRA

Key Guidance Development

• LMP (RG 1.233)

• Siting Criteria (RG 4.7)

• Fuel Qualification Framework (NUREG-2246)

• Developing Principal Design Criteria for Non-LWR (RG 1.232)

Existing

• Analytical Margin

• Chemical Hazards

• Manufacturing

• Technical Specifications

• FSP

• Framework B Content of Applications Future Under Development Near-Term

• TICAP/ARCAP (NEI 21-07)

• Non-LWR PRA Standard

• Non-LWR PRA Standard Applicability ISG

• High Temp Materials (ASME III-5)

• Reliability & Integrity Mgt (ASME XI-2)

• Molten Salt Reactor Fuel Qualification

• Seismic Design/Isolators

• Emergency Planning

• Change Evaluation (Southern Nuclear Operating Company-led)

• QA Alternatives (NEI-led)

• Facility Training Programs

• Materials Compatibility ISG Part 53

• DG-1413, Identification of Licensing Events

• DG-1414, AERI Methodology

• DRO-ISG-2023-01, Operator Licensing Program Review ISG

• DRO-ISG-2023-02, Staffing Plan Review ISG Augmenting NUREG-1791

• DRO-ISG-2023-03, Scalable Human Factors Engineering Review ISG

• Part 26, FFD

• Part 26, Fatigue Management

• Part 73, AA

• Part 73, Cyber Security

• Part 73 Security Programs Part 53

• DG-1413, Identification of Licensing Events

• DG-1414, AERI Methodology

• DRO-ISG-2023-01, Operator Licensing Program Review ISG

• DRO-ISG-2023-02, Staffing Plan Review ISG Augmenting NUREG-1791

• DRO-ISG-2023-03, Scalable Human Factors Engineering Review ISG

• Part 26, FFD

• Part 26, Fatigue Management

• Part 73, AA

• Part 73, Cyber Security

• Part 73 Security Programs

TICAP / ARCAP - Nexus

FRN Section VII Specific Requests for Comments Part 53

• Overall organization

• Use of QHOs

• ALARA

• DID

• Earthquake Engineering

• Construction and Manufacturing

• Use of references

• Manufacturing licenses

• Staffing and GLROs

• OnShift engineering expertise

• Training program accreditation

• Use of simulation facilities

FRN Section VII Specific Requests for Comments Part 53

• FSPs

• Integrity assessment programs

• Decommissioning

• PRA information

• Changes to manufacturing licenses

• Specific requirements for Technical Specifications

• AERI

• Reporting

• Financial qualifications

Discussion 173

Final Discussion and Questions 174

Additional Information Additional information on the 10 CFR Part 53 rulemaking is available at https://www.nrc.gov/reactors/new-reactors/advanced/rulemaking-and-guidance/part-53.html For information on how to submit comments go to https://www.regulations.gov and search for Docket ID NRC-2019-0062 For further information, contact Robert Beall, Office of Nuclear Material Safety and Safeguards, telephone: 301-415-3874; email:

Robert.Beall@nrc.gov

AA Access authorization ACRS Advisory Committee on Reactor Safeguards AERI Alternative evaluation for risk insights ALARA As low as reasonably achievable AOO Anticipated operational occurrence ARCAP Advanced Reactor Content of Application Project ASME American Society of Mechanical Engineers BDBE Beyond-design-basis event CFR Code of Federal Regulations COL Combined license CP Construction permit DANU Division of Advanced Reactors and Non-Power Production and Utilization Facilities Acronyms DBA Design-basis accident DBE Design-basis event DC Design certification DG Draft regulatory guidance DID Defense-in-depth DRA Division of Risk Assessment DRO Division of Reactor Oversight EPA U.S. Environmental Protection Agency ESP Early site permit FFD Fitness for duty FR Federal Register FRN Federal Register Notice

FSP Facility safety program GLRO Generally licensed reactor operator HFE Human factors engineering ILCFR Individual latent cancer fatality risk INL Idaho National Labs ISG Interim staff guidance ISI Inservice inspection IST Inservice testing KSAs Knowledge, skills, and abilities LBE Licensing basis events LCF Latent cancer fatality LMP Licensing modernization project Acronyms LWR Light-water reactor MACCS MELCOR accident consequence code system MBDBE Mitigation of beyond-design-basis events ML Manufacturing license NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSRSS Non-safety related but safety significant NUMREL Number of released plume segments NUREG U.S. Nuclear Regulatory Commission technical report designation OL Operating license PAG Protective action guideline

PRA Probabilistic risk assessment QA Quality assurance QHO Quantitative health objectives REM Roentgen equivalent man RES Office of Nuclear Regulatory Research RG Regulatory guide RO Reactor operator SAR Safety analysis report SDA Standard design approval SECY Office of the Secretary SMR Small modular reactor Acronyms SNM Special nuclear material SOARCA State-of-the-art reactor consequence analyses SRM Staff requirements memorandum SRO Senior reactor operator SRP Standard review plan SSCs Structures, systems, and components STA Shift technical advisor Sv Sievert TEDE Total effective dose equivalent TICAP Technology Inclusive Content of Application Project WG Working group

Backup Slides 179

Nuclear Energy Innovation and Modernization Act (NEIMA)

January 2019 NEIMA Section 103(4) 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

• (9) REGULATORY FRAMEWORKThe term regulatory framework means the framework for reviewing requests for certifications, permits, approvals, and licenses for nuclear reactors.

• (14) TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK The term technology-inclusive regulatory framework means a regulatory framework developed using methods of evaluation that are flexible and practicable for application to a variety of reactor technologies, including, where appropriate, the use of risk-informed and performance-based techniques and other tools and methods.

180

Part 53 Rulemaking Plan SECY-20-0032, Rulemaking Plan on Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors, dated April 13, 2020 (ADAMS ML19340A056).

In SRM-SECY-20-0032, dated October 2, 2020 (ADAMS ML20276A293), the Commission provided direction to the staff.

On November 2, 2020, staff submitted a Commission memorandum responding to the SRM direction to provide a schedule with milestones and resources to complete the final rule by October 2024 (ADAMS ML20288A251).

On November 23, 2021, the Commission approved the NRC staffs schedule extension request 181

Part 53 Rulemaking Objectives

1. Continue to provide reasonable assurance of adequate protection of public health and safety and the common defense and security,

2. Promote regulatory stability, predictability, and

clarity,

3. Reduce requests for exemptions from the current requirements in 10 CFR Part 50 and 10 CFR Part 52,

4. Establish new requirements to address non-light-water reactor technologies,

5. Recognize technological advancements in reactor design, and

6. Credit the response of advanced nuclear reactors to postulated accidents, including slower transient response times and relatively small and slow release of fission products.

182

Site selected Part 50 Part 52 Part 53 Subparts H & R:

Leveraging and Combining Existing Licensing Processes Operating License (OL)

CP based on SDA or DC Construction Permit (CP)

Commercial Operations Site selected Site selected Fuel Load Combined License (COL)

Manufacturing License (ML)

Standard Design Approval (SDA)

Use OL or custom COL to develop a subsequent DC Design Certification (DC)

CP and COL may reference Early Site Permit (ESP) 183