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

A d v i s o r y C om m i tte e on Re ac to r Safe guards (ACRS ) Re gulator y Rulemaking , Policies and Practices:

Part 53 Subcommittee 10 CFR Part 53 Licensing and Regula on of A dvanced Nuclear Reactors October 18-19, 2022

Agenda - October 18th 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 2

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

Part 53 Licensing Subpart B - Safety Requirements Frameworks Subpart C - Design Requirements Framework A o Probabilistic Risk Assessment Subpart D - Siting (PRA)-led approach Subpart E - Construction/Manufacturing o Functional design criteria Subpart F - Operations Subpart G - Decommissioning Subpart H - Application Requirements Subpart I - License Maintenance Subpart J - Reporting Subpart K - Quality Assurance Subpart A - General Provisions Framework B Subpart N - Siting o Traditional use of risk insights Subpart O - Construction/Manufacturing o Principal design criteria Subpart P - Operations o Includes an AERI approach Subpart Q - Decommissioning Subpart R - Application Requirements Subpart S - License Maintenance Subpart T - Reporting Subpart U - Quality Assurance 4

Federal Register Notice (FRN)

Enclosure 1A Preamble ML22272A036 Enclosure 1B Section by Section, ML22272A038 Availability of Guidance Enclosure 1C Framework A ML22272A039 Enclosure 1D Framework B ML22272A040 Rule Package (ML22272A034) Guidance Documents DG-1413 Licensing Events ML22272A042 DG-1414 AERI Methodology ML22272A045 DRO-ISG-2023-01 Operator Licensing ML22272A047 Program Review ISG DRO-ISG-2023-02 Staffing Plan Review ISG ML22272A049 Augmenting NUREG-1791 DRO-ISG-2023-03 Scalable Human Factors ML22272A051 Engineering Review ISG

• Purpose

• Provide optional frameworks for the issuance, amendment, renewal, and Sections termination of licenses, permits, 53.000 certifications, and approvals for commercial nuclear plants and

• Frameworks 53.010

• 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

• Common Definitions

• Commercial Nuclear Plant Subpart A -

• Manufactured reactor General

• Manufactured reactor module Provisions

• Safety function (Definitions)

• 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

• 53.200 Safety objectives.

Subpart B -

• 53.210 Safety criteria for design basis Technology-

• 53.220 accidents.

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

Safety

• 53.230 Safety functions.

• 53.240 Licensing basis events.

Requirements

• 53.250 Defense-in-depth.

• 53.260 Normal operations.

• 53.270 Protection of plant workers.

§ 53.400 Design features for licensing basis events.

§ 53.410 Functional design criteria for design basis accidents.

Subpart C - § 53.415

§ 53.420 Protection against external hazards.

Functional design criteria for licensing basis Design and events other than design basis accidents.

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

Requirements § 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.

§ 53.500 General siting.

§ 53.510 External hazards.

Subpart D -

Siting § 53.520 Site characteristics.

Requirements § 53.530 Population-related considerations

§ 53.540 Siting interfaces.

• Scope and purpose.

• Reporting of defects and Subparts E & O noncompliance.

Construction

• Construction and Manufacturing

• Manufacturing

• Fuel loading for manufactured Requirements 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 Subparts E & O criticality should conditions result in the maximum reactivity being attained for the fissile material Fuel loading for

• Commission finding that a manufactured reactor manufactured module in required configuration is not a utilization reactor modules 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

§ 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.

Subpart F -

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

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

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

Facility safety

• Scope and purpose.

• Financial assurance for decommissioning.

• Cost estimates for decommissioning .

• Annual adjustments to cost estimates for decommissioning.

• Methods for providing financial assurance for Subpart G & Q decommissioning.

Decommissioning

• Limitations on the use of decommissioning trust funds.

Requirements

• 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.

§ 53.1100 - 53.1121 General/common requirements.

§ 53.1124 Relationship between sections.

§ 53.1130 Limited work authorizations.

§ 53.1140 Early site permits.

Subpart H - § 53.1200 Standard design approvals.

Licenses, § 53.1230 Standard design certifications.

Certifications, § 53.1270 Manufacturing licenses

§ 53.1300 Construction permits.

and Approvals § 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.

• 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)

Subparts I & S

• Other licensing basis information

• Evaluating changes to facility as described in final safety Maintaining and analysis reports (SAR).

Revising Licensing

• Program-related documents

• Transfer of licenses or permits.

Basis Information

• Termination of license.

• Information requests.

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

• Backfitting.

• Renewal.

• General information.

• Unfettered access for inspections.

• Maintenance of records, making of reports.

• Immediate notification requirements for operating commercial nuclear plants.

Subparts J & T

• Licensee event report system.

• Facility information and verification.

Reporting and

• Reporting of defects and noncompliance.

Other

• Financial requirements.

• Financial qualifications.

Administrative

• Annual financial reports.

Requirements

• Licensees change of status; financial qualifications.

• Creditor regulations.

• Financial protection.

• Insurance required to stabilize and decontaminate plant following an accident.

• Financial protection requirements.

10 CFR Part 50,

• General Provisions Appendix B Criteria

• Organization I

• Quality Assurance Program II

• Design Control III

• Procurement Document Control IV Subparts K & U

• Instructions, Procedures and Drawings V

• Document Control VI Quality Assurance

• Control of Purchased Material, Equipment and Services VII Criteria for

• Identification and Control of Materials, Parts and Components VIII Control of Special Processes IX Commercial

• Inspection X

• Test Control XI Nuclear Plants

• Control of Measuring and Test Equipment XII

• Handling, Storage and Shipping XIII

• Inspection, Test and Operating Status XIV

• Nonconforming Materials, Parts or Components XV

• Corrective Action XVI

• Quality Assurance Records XVII

• Audits XVIII

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

§ 53.3505 Scope.

Subpart N - § 53.3510 Definitions.

Siting

§ 53.3515 Factors to be considered when evaluating sites.

§ 53.3520 Non-seismic siting criteria.

§ 53.3525 Geologic and seismic siting criteria.

§ 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 Subpart P - General staffing, training, personnel qualifications, and human factors requirements.

Requirements § 53.4300 Programs for Operation 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.

§ 53.4700 - 53.4721 General/common requirements.

§ 53.4724 Relationship between sections.

§ 53.4730 General technical requirements.

§ 53.4731 Risk-informed classification of SSCs.

Subpart R - § 53.4733 Seismic design alternatives.

Licenses, § 53.4740 Limited work authorizations.

Certifications, § 53.4750 Early site permits.

§ 53.4800 Standard design approvals.

and 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

Existing Paradigm

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

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

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

2) Promote regulatory stability, predictability, and clarity, Findings 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 Comments generally fall into following groups:

Consideration

• Rule should not include a cumulative risk measure

• Rule should include alternative risk measures of Feedback on o Surrogates for the QHOs Including 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

§ 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 Subpart R - Part 52 requirements Licenses,

• Requirements in subparagraph (vi) modified to ensure rule Certifications, is technology-inclusive

• Fuel or core damage or potential for large radiological and Approvals 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 />)

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

• Objectives

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

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

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

(i) Analysis and Evaluation

• From § 52.79(a) with modifications to support technology-inclusiveness Subpart R - and Framework B event classifications.

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

Licenses, (ii) Design Basis Accidents Certifications,

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

and Approvals

• 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

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

(iv) Additional Licensing Basis Events

• Technology-inclusive requirements for relevant additional LBEs and Subpart R - analysis requirements for these events; similar to international Licenses, defense-in-depth (DID) requirements

• Changes clarify scope of initiators and event sequences that must be Certifications, considered and design requirements for SSCs used to mitigate additional LBEs and Approvals (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

§ 53.4730(a)(36) Containment requirements.

• Requirements split to acknowledge differences between non-LWR and LWR approaches to containment Subpart R -

• For non-LWRs, § 53.4730(a)(36)(i) addresses:

Licenses, o Set of barriers used to meet requirements for AOOs, DBAs, and siting criteria (functional containment)

Certifications, o Safety classification (i.e., safety-related) and qualification of SSCs making up functional containment barriers and Approvals 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)

Other General Technical Requirements

• § 53.4730(a)(2) Facility description.

Subpart R - * § 53.4730(a)(4) Design bases and principal design criteria.

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

Certifications, * § 53.4730(a)(14) Earthquake engineering criteria.

and Approvals * § 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

Introductions

• 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 40

The Graded PRA Working Group Membership Project Manager

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

• Katie Wagner, NRR/DANU

• Hanh Phan, NRR/DANU

• Sunil Weerakkody, NRR/DRA Technical Lead

• Marty Stutzke, NRR/DANU

• Robert Budnitz, consultant Working Group Members Management/Coordination

• Hosung Ahn*, previously on rotation from

• Candace de Messieres, NRR/DANU^

NRR/Division of Engineering and External Hazard

• Steve Lynch, NRR/DANU

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

• Nathan Sanfilippo*

• Anne-Marie Grady, NRR/DRA - Principal

• John Segala, NRR/DANU Co-Author of DG-1414

• Matt Humberstone, RES/DRA *Former WG member

• Ian Jung, NRR/DANU ^On rotation from current position 41

Recent Activities

• 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 42

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

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

Demonstrate that the AERI entry conditions 73 FR 60614, left column: the Commission has also issued policy statements on the use of PRA in regulatory activities (60 FR are met 42622; August 16, 1995), and severe accidents regarding future designs and existing plants (50 FR 32138; August 8, 1985). The use Develop a demonstrably conservative 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 risk estimate policy statements provide sufficient guidance to designers.

Search for severe accident vulnerabilities Policy Statement: Use of PRA Methods in Nuclear Regulatory Activities 60 FR 42622; August 16, 1995 use PRA or an 60 FR 42628, middle column: It is important to note that not all alternative Identify risk insights of the Commissions regulatory activities lend themselves to a risk risk-informed analysis approach that utilizes fault tree methods. In general, a approach as a fault tree method is best suited for power reactor events that design tool typically involve complex systemsthe Commission recognizes Evaluate DID adequacy that a single approach for incorporating risk analyses into the regulatory process is not appropriate. 44

AERI-Related Draft Proposed Rule Text The proposed AERI entry conditions are designed to limit use of the Current Draft Proposed Rule Text proposed AERI approach to commercial nuclear plants whose designs § 53.4730(a)(34) Description of risk evaluation. A description of the risk are relatively straightforward and do not involve overly complex evaluation developed for the commercial nuclear plant and its results.

systems and interactions and, accordingly, would not warrant The risk evaluation must be based on:

development of a PRA to provide quantitative risk insights. (i) A probabilistic risk assessment (PRA); or (ii) An alternative evaluation for risk insights (AERI), provided Draft Proposed Rule Text Presented to the ACRS that:

Regulatory Rulemaking, Policies and Practices:

(A) The analysis of a postulated bounding event demonstrates Part 53 Subcommittee that the consequence evaluated at a location 100 meters (328 feet)

June 23-24, 2022 away from the commercial nuclear plant does not exceed 10 mSv (1

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

second and subsequent years.

45

Changes to the AERI-Related Draft Proposed Rule Text

• 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 46

Proposed Uses of the AERI Entry Conditions

• 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) 47

Maintenance of Risk Evaluations

§ 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 48

Proposed AERI-Related FRN Questions

• 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.

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

• 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.

51

Analytic Expression Assumptions

• 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 52

Downwind Dose Reduction Coefficient 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- Elevated releases or plume rise will result in an increase in bouyant (i.e., ( , ) is monotonically concentration at short downwind distances as the plume disperses decreasing) 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 Highly unstable conditions can result in rapid vertical dispersion to the ground surface and is unconstrained by a top of the mixing layer due to insolation of ground surface mixing height The dose-distance reduction coefficient n is Although crosswind (transverse) dispersion is typically represented as assumed to be independent of distance r. a power law, vertical dispersion does not follow a power law relationship with distance 53

Downwind Dose Reduction Coefficient Elevated/Buoyant Plume 1

Normalized Relative Peak Dose 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0

0 2 4 6 8 10 12 14 Downwind Distance (km)

A B C D E F Normalized relative peak dose as a function of downwind distance and stability class Normalized to a constant core scaling factor and maximum peak dose 54

Downwind Dose Reduction Coefficient Effect of Protective Actions

• 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.

Mean value (across all weather trials) of peak total effective dose Mean value (across all weather trials) of peak total (rem) from early-phase exposure to the non-evacuating cohort effective dose (rem) from late-phase exposure Source: NRC 2022 55

Downwind Dose Reduction Coefficient Mixing Height 1000 900 800 700 Plume Sigma-z (m) 600 500 400 Isopleths of mean annual morning mixing height (m*10-2) as a function of downwind distance 300 Isopleth levels: 300 - 900 m 200 100 0

0 2000 4000 6000 8000 10000 12000 14000 Downwind Distance (m)

A B C D E F Isopleths of mean annual afternoon mixing height Plume sigma z (m) as a function of downwind distance (m*10-2) as a function of downwind distance Isopleth levels: 800 - 2600 m Sources: Case 2 Model Output Files; Holzworth 1972 56

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

Source: Reference 7 of NRC 1983 57

Single Plume Azimuthal Correction Factor

• 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:

Stability Class Ay By 100 m 10 mi 2 2 A 0.3658 0.9031 0.0934 0.0571 B 0.2751 0.9031 0.0702 0.0429

• An alternative would be to simply assume that C 0.2089 0.9031 0.0533 0.0326 the crosswind plume spread may be represented by D 0.1471 0.9031 0.0376 0.0230 a tophat with a width of one 22.5 sector, resulting E 0.1046 0.9031 0.0267 0.0163 in an azimuthal correction factor of 0.0625 (1/16) F 0.0722 0.9031 0.0184 0.0113 G 0.0481 0.9031 0.0123 0.0075 Figure Source: Jow et al. 1990 58

Maximum Dose vs Average Dose / Risk over an Annular Region 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 59

Approach

• 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. 60

Summary of Source Terms Source Term Characteristics

• 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.

PDELAY PLUDUR PLUDUR PLHITE PLHEAT RC Case Release Category Description NUMREL (hr) (50%) (hr) (100%) (hr) (m) (MW)

VF 5D Unscrubbed interfacing systems loss-of-coolant accident 86 3.2 4.5 68.8 11 19 with auxiliary building failure LCF 1B Late containment due to long-term quasi-static 179 48 32.1 120.0 0.36 5.9 overpressure, unscrubbed NOCF 2R1 Containment is not bypassed or failed, and radiological 199 13 89.9 154.5 32 0.0026 release to the environment occurs via design-basis containment leakage only.

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 Source: adapted from Tables 3.1-1 and A.1a in NRC 2022 61

Summary of Modeling Cases

• Modeling cases designed to test effect of key assumptions related to plume rise, wake effects, protective actions, plume segmentation, weather variability, and population density Dose Reduction Population Case Coefficient Effects Azimuthal Variation 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

• 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.

62

Case 0: Simple Model

• Simplest Case Results

• 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%

OVERALL EARLY CHRONC MACCS Analytic Peak dose (Sv) Peak dose (Sv) Peak dose (Sv) 10- mile MACCS 10-mile Percent Case at 100 m at 100 m at 100 m ILCFR P-G n fitted n ILCFR 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%

63

Case 1: Pasquill-Gifford Stability Results

• 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 OVERALL EARLY CHRONC Peak dose Peak dose Peak dose MACCS MACCS fitted Analytic Percent Case (Sv) at 100 m (Sv) at 100 m (Sv) at 100 m 10- mile ILCFR P-G n n 10-mile ILCFR 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%

64

Case 2: Plume Buoyancy Results

• 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 OVERALL EARLY CHRONC Peak dose (Sv) Peak dose (Sv) Peak dose (Sv) MACCS MACCS Analytic Percent Case at 100 m at 100 m at 100 m 10- mile ILCFR fitted n 10-mile ILCFR 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%

65

Case 3: Wake Effects Results

• 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 OVERALL EARLY CHRONC Peak dose (Sv) at Peak dose (Sv) at Peak dose (Sv) at MACCS MACCS Analytic Percent Case 100 m 100 m 100 m 10- mile ILCFR fitted n 10-mile ILCFR 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%

66

Case 4: Protective Actions Results

• 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 OVERALL EARLY CHRONC Peak dose (Sv) at Peak dose (Sv) at Peak dose (Sv) at MACCS MACCS Analytic Percent Case 100 m 100 m 100 m 10- mile ILCFR fitted n 10-mile ILCFR 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%

67

Case 5: Meteorological Sampling Results

• 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 OVERALL EARLY CHRONC Analytic Peak dose (Sv) Peak dose (Sv) Peak dose (Sv) MACCS MACCS 10-mile Percent Case at 100 m at 100 m at 100 m 10- mile ILCFR fitted n ILCFR* 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%

• Transverse dispersion assumed consistent with slightly unstable conditions 68

Case 6: Multiple Plumes Results

• 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 OVERALL EARLY CHRONC Peak dose (Sv) at Peak dose (Sv) at Peak dose (Sv) at MACCS MACCS Analytic Percent Case 100 m 100 m 100 m 10- mile ILCFR fitted n 10-mile ILCFR* 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 69

Case 7: Population Distribution Results

• 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.

OVERALL EARLY CHRONC Analytic Peak dose (Sv) Peak dose (Sv) Peak dose (Sv) MACCS MACCS 10-mile Percent Case at 100 m at 100 m at 100 m 10- mile ILCFR fitted n ILCFR* 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%

• Transverse dispersion assumed consistent with highly unstable conditions 70

Effect of Downwind Dose Reduction Coefficient on Individual Latent Fatality Risk within 10 miles 1.0E+00 10 Mile ILCF Conditional Risk Given D_100=25 1.0E-01 1.0E-02 1.0E-03 1.0E-04 rem 1.0E-05 1.0E-06 1.0E-07 1.0E-08

-1.5 -1.0 -0.5 0.0 0.5 1.0 1.5 2.0 2.5 3.0 Effective Dose-Distance Reduction Coefficient 71

Long-Term Time Dependence of Dose

• Accumulation of dose in years after the event occurs at different 1.00 Ratio of Dose at Given Year to First Year Dose rates for different source terms 0.90

• Therefore, there is likely no fixed ratio between early phase dose, 0.80 first year dose, and 50-year cumulative dose 0.70

• However, for the scaled source terms considered in this analysis, a 0.60 first-year dose of 2 rem appears to correspond to a lifetime dose* 0.50 less than 25 rem, probably due to radioactive decay and the effect 0.40 of weathering on groundshine and resuspension 0.30 First 50 year 50 year 0.20 Year Second Year Cumul Cumul 0.10 Case Early Phase CHRONC CHRONC CHRONC* TOTAL*

PAGs 1-5 2 0.5 Not specified Not specified -

VF/5D 0.70 2.0 1.0 9.0 9.7 0 10 20 30 40 50 LCF/1B 0.13 2.0 0.2 3.5 3.6 Years Since Release NOCF/2R1 0.11 2.0 0.4 5.2 5.3 VF/5D LCF/1B NOCF/2R1

• Cumul.: cumulative

• 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.

72

Summary

• 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.

73

Bibliography

• 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) 74

Confirmatory MACCS Calculations Supplemental Slides 75

Case 0: Simple Model Peak Dose vs Distance 1.E+00 Stability Class A: 1.E-01 OVERALL Peak Dose (Sv)

EARLY Extremely Unstable 1.E-02 CHRONC 1.E-03 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 Power (CHRONC) 0.1 1 10 Distance downwind (km) 1.E+00 Stability Class F: 1.E-01 OVERALL Peak Dose (Sv)

Strongly Stable 1.E-02 EARLY 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 Power (CHRONC) 0.1 1 10 Distance downwind (km) 76

Case 1: Pasquill-Gifford Stability Peak Dose vs Distance 1.E+00 Stability Class A: 1.E-01 OVERALL Peak Dose (Sv) 1.E-02 EARLY Extremely Unstable 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 0.1 1 10 Power (CHRONC)

Distance downwind (km) 1.E+00 Stability Class F: 1.E-01 OVERALL Peak Dose (Sv)

Strongly Stable 1.E-02 EARLY 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 1.E-06 Power (EARLY) 0.1 1 10 Power (CHRONC)

Distance downwind (km) 77

Case 2: Plume Buoyancy Peak Dose vs Distance Stability Class A: 1.E+01 1.E+00 OVERALL Peak Dose (Sv)

Extremely Unstable 1.E-01 1.E-02 EARLY 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 0.1 1 10 Power (CHRONC)

Distance downwind (km) 1.E+01 Stability Class F: 1.E+00 OVERALL Peak Dose (Sv) 1.E-01 Strongly Stable 1.E-02 EARLY 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 0.1 1 10 Power (CHRONC)

Distance downwind (km) 78

Case 3: Wake Effects Peak Dose vs Distance Stability Class A: 1.E+00 1.E-01 OVERALL Extremely Unstable Peak Dose (Sv) 1.E-02 EARLY 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 Power (CHRONC) 0.1 1 10 Distance downwind (km) 1.E+00 Stability Class F: 1.E-01 OVERALL Peak Dose (Sv)

Strongly Stable 1.E-02 EARLY 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 1.E-06 Power (EARLY) 0.1 1 10 Power (CHRONC)

Distance downwind (km) 79

Case 4: Protective Actions Peak Dose vs Distance 1.E+00 Stability Class A: 1.E-01 OVERALL Peak Dose (Sv)

Extremely Unstable 1.E-02 1.E-03 EARLY CHRONC 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 Power (CHRONC) 0.1 1 10 Distance downwind (km) 1.E+00 Stability Class F: 1.E-01 OVERALL Peak Dose (Sv)

Strongly Stable 1.E-02 EARLY 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 Power (CHRONC) 0.1 1 10 Distance downwind (km) 80

Case 5: Meteorological Sampling Peak Dose (Mean) vs Distance 1.E+01 1.E+00 OVERALL Peak Dose (Sv) 1.E-01 Without plume 1.E-02 EARLY CHRONC buoyancy 1.E-03 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 Power (CHRONC) 0.1 1 10 Distance downwind (km) 1.E+01 1.E+00 OVERALL With plume Peak Dose (Sv) 1.E-01 EARLY 1.E-02 buoyancy 1.E-03 CHRONC 1.E-04 Power (OVERALL) 1.E-05 Power (EARLY) 1.E-06 Power (CHRONC) 0.1 1 10 Distance downwind (km) 81

Case 6/7: Weather Sampling with Multiple Plumes Peak Dose (Mean) vs Distance 1.E+00 1.E+00 1.E+00 1.E-01 1.E-01 1.E-01 1.E-02 1.E-02 1.E-02 Peak Dose (Sv) Peak Dose (Sv) Peak Dose (Sv) 1.E-03 1.E-03 1.E-03 1.E-04 1.E-04 1.E-04 1.E-05 1.E-05 1.E-05 1.E-06 1.E-06 1.E-06 0.1 1 10 0.1 1 10 0.1 1 10 Distance downwind (km) Distance downwind (km) Distance downwind (km)

OVERALL EARLY CHRONC OVERALL EARLY CHRONC OVERALL EARLY CHRONC Power (OVERALL) Power (EARLY) Power (CHRONC) Power (OVERALL) Power (EARLY) Power (CHRONC) Power (OVERALL) Power (EARLY) Power (CHRONC)

VF/5D LCF/1A2 NOCF/2R1 82

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

Licensing Pathways and Licensing Events Regulation and Reactor Application Type Type Use of LMP Licensing Event Categories Risk Evaluation not applicable not required Part 50 (NEI 18-04, Rev. 1 and RG 1.233

• Design-basis events (DBEs) (§ 50.49): (Parts 50/52 lessons-CP, OL LWR currently only apply to o AOOs learned rulemaking)

Part 52 non-LWRs licensed under o DBAs (i.e., postulated accidents)

Parts 50 or 52) o External events PRA required DC, SDA, ML, COL o Natural phenomena not required Part 50

• Non-DBA (§ 50.2 alternate ac source)

(Parts 50/52 lessons-CP, OL

• Beyond-design-basis events (BDBE)

Non-LWR no learned rulemaking)

• Anticipated transient without scram Part 52

• Station black out PRA required DC, SDA, ML, COL Part 50 Licensing events are collectively referred to as licensing- PRA implied by CP, OL basis events (LBEs), which include the following use of LMP categories:

Non-LWR yes

• AOOs Part 52

• DBEs PRA required DC, SDA, ML, COL

• BDBEs

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

not applicable Part 53, Framework A LWR or

• AOOs (potential future update to PRA required CP, OL, DC, SDA, ML, COL non-LWR

• Unlikely event sequences NEI 18-04 and RG 1.233)

• Very unlikely event sequences

• DBAs

• AOOs Part 53, Framework B LWR or

• DBAs not applicable PRA or AERI required CP, OL, DC, SDA, ML, COL non-LWR

• Additional licensing-basis events 85

• Severe accidents

Overarching Principles 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.

86

1 Assemble Technology-Inclusive Identification Multi-disciplinary Team Collect Application-Specific Information of Licensing Events (Sheet 1 of 3) 2 3 4 5 Collect information on plant Identify Radiological Sources Establish Quality Identify Sources of Hazardous design, plant operating states, and Transport Barriers from Control Program Chemical Materials and site characteristics the Source to the Environment 6 Identify Plant-specific Safety Functions

• Systems needed to achieve Added guidance on establishing Updated text on safety safety functions a Quality Control program prior functions, consistent with Part

• Operator actions needed to engaging in the work 53 Framework A to achieve safety functions

• Success criteria Select Analysis Methods 7

8 Define Plant-specific End States for Event Sequences Select Initiating Event 10 Identification Methods 9 Select Analytical Methods

• Inductive methods Define Initiating Event for Event Sequences

• Deductive methods Grouping Strategy and X (e.g., Event Trees, Event

• Human-induced events Characteristics to Sequence Diagrams)

(Appendix provides Sheet 2 discussion and references) 87

Technology-Inclusive Identification of Licensing Events (Sheet 2 of 3)

Initiating Event Analysis 11 12 13 14 Account for Relevant Apply Initiating Event Apply Initiating Independent X Operating Experience and List of Identification Event Grouping Review and Quality for Insights from Earlier Initiating Events from Methods Strategy Control Relevant Analyses Sheet 1 Added references for listing of external hazards. The search for Initiating Events and Event Sequences is subject to Quality Control (not QA)

Event Sequence Selection 15 16 Apply Selected Analytical Methods 17

• Identify initiating event impact on safety functions Account for Relevant

• Identify the impact of front-line and support Operating Experience and Independent Review List of system dependencies on safety functions Y

for Insights from Earlier and Quality Control Event Sequences

• Identify the impact of operator actions on safety Relevant Analyses to functions Sheet 3

18 Is a PRA being developed to 19 Provide initiating Y support events and event Technology-Inclusive Identification yes (PRA) from the application? sequences to the PRA of Licensing Events (Sheet 3 of 3)

Sheet 2 no 20 Identify Required Categories Follow NEI 18-04, of Licensing Events for Rev. 1 as endorsed Licensing Framework in RG 1.233 Part 50 or 52 non-LWR applications based on LMP All other applications Clarified that the LMP guidance currently applies to non-LWRs 21 Define Licensing Event Grouping under parts 50 or 52.

Strategy and Characteristics Note: The staff intends to revise RG 1.233 to address licensing under Part

• Group by frequency 53 Framework A in the future.

o Qualitative o Quantitative

• Group by type o Plant response following the initiating The search for Licensing Events event (sequence of events, timing) is subject to QA o Similar challenge to safety functions o End state 24 23 Compare to Predefined Lists 25 (e.g., SRP Chapter 15, 22 Identify Limiting Cases for Apply Licensing Event previous CP, OL, DC, SDA, ML, Independent List of Each Group of Licensing Grouping Strategy or COL applications) and Review and QA Licensing Events Events identify differences from SRP (only for LWRs)

Quality Control Program

• 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.

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

90

DG-1414 (proposed new RG 1.255)

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

Alternative Evaluation for Risk Insights Methodology

• 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.

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.

92

Licensing Pathways - Risk Evaluation Perspective H I Parts 50 and 52 with LMP Perform Perform design basis Continue design transient and accident radiological and licensing Part 53 Framework A accident analyses consequences analyses activities C D E F Finish PRA Select LBEs Select DBAs Classify SSCs development G

A Comprehensive Evaluate DID and systematic initiator search and event B Select Notes:

sequence 1) Each step builds on all of the preceding steps (considers all information available at that point) licensing delineation framework 2) Feedback loops (e.g., the impact of design revisions) are not shown without preconceptions J K L M N or reliance on Select Perform Perform design basis yes Continue design Elect to Finish PRA predefined lists licensing transient and accident radiological and licensing develop PRA development events accident analyses consequences analyses activities Q AERI Parts 50 and 52 without LMP no Q1 Develop demonstrably Part 53 Framework B ONLY for Part 53 conservative risk estimate Framework B no using the bounding event Applicant decision Q2 Search all event O P sequences for severe Identify and yes Continue design AERI entry accident vulnerabilities DG-1413, Technology-Inclusive Identification of Licensing Events for analyze the and licensing condition met? Q3 Develop risk insights by Commercial Nuclear Plants bounding event activities reviewing all event DG-1414, Alternative Evaluation for Risk Insights (AERI) Methodology sequences Q4 Assess DID adequacy by Alternative Evaluation reviewing all event LMP guidance - NEI 18-04, Rev. 1, as endorsed in RG 1.233 93 for Risk Insights sequences 93

Elements of the AERI Methodology (1 of 4)

• 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 94

Elements of the AERI Methodology (2 of 4)

• 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.

95

Elements of the AERI Methodology (3 of 4)

• 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 96

Elements of the AERI Methodology (4 of 4)

• 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 97

Maintaining and Upgrading the AERI Risk Evaluation (1 of 2)

• 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 98

Maintaining and Upgrading the AERI Risk Evaluation (2 of 2)

• 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 99

Discussion 100

Final Discussion and Questions 101

Agenda - October 19th 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 102

Preliminary Requirements for Operations:

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

• 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 Agenda

• 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

• Theresa Buchanan, Senior Reactor Engineer (Examiner)

• Dr. David Desaulniers, Senior Technical Advisor for Human Factors and Human Overview of Performance Evaluation Primary Staff

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

Contributors

• Dr. Niav Hughes Green, Human Factors Psychologist (NRR & RES)

• 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

• 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 Updates to utilizing GLROs has been defined as self-reliant mitigation facilities Subpart F and

• Procedure program requirements have been consolidated P since the

• Staffing plan requirements for non-operations positions are now functional in nature 2nd Iteration

• 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

• 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 Status of use GLROs have been incorporated, in addition to those already in place for Framework A Engineering

• Framework B GLRO criteria vary depending on whether an AERI is used Irrespective of AERI, DID without human action is needed Expertise &

• For a non-AERI plant, the GLRO criteria are analogous to the equivalent criteria for Framework A, as adapted to the differing requirements of GLRO

• Framework B For an AERI plant, the GLRO criteria are met by meeting AERI criteria (plus requirements 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

• 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.

Response to

• Key points form the staff response included the following:

Interim Letter

• New framework for operator licensing under Part 53 is technology-Report from

• inclusive and creates significant flexibilities compared to Part 55 Accommodating such flexibilities while complying with statutory August 2022 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

• 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 Purpose 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

• 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 Goals program developments based on current best practices from research and expertise on the measurement and testing of KSAs 111

• 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 Section 1.0 planned to provide additional information in this area KSAs List

• Using this list as a starting point, a screening is Development 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

• 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.

Section 2.0

• What the format for the test will be Operator

• Developed detailed content specification

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

• How the KSAs are sampled for each examination developed

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

• Describe validation plan

• What evidence was collected to support validity of the test, that the test works and will work as intended Section 3.0

• Content validity, concurrent validity Examination

• Should require content validity at the least Validity 115

• Criterion-referenced

• Described how each test item is scored and how scores combined to get total Section 4.0 score Scoring

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

• Provided cut-off score 116

• 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 Section 5.0 Reliability of

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

• Companion to the test plan

• Provides more detail related to the specific types of tests

• Includes administrative aspects of test Section 6.0

• How to administer Test Manual

• Time to administer or time allowed to take the test

• Materials provided to test takers

• How to interpret test results 118

• This section is specifically for written and computer-based tests Section 7.0

• Provides additional characteristics Additional associated with psychometrics, test Characteristics instructions, objective scoring system, of High- and standardization Quality Test Materials 119

• This section references back to sections of NUREG-1021, Operator Licensing Examination Standards for Power Section 8.0 Reactors for items that are universally Other applicable, regardless of plant design Examination Program Considerations 120

• 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)

Section 9.0

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

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

• Examination procedures should be similar to those in NUREG-1021, as specific to the type of test administered Section 10.0

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

• Measures are in place to retain required records 122

• Documentation specifies what changes require NRC approval and which do not Section 11.0

• NRC approval Examination

• Exemption from regulation

• Change to technical specification Program

• Negative impact to examination Change security/integrity Management

• Negative impact on consistency Process 123

• Beyond the scope of the guidance

• The documentation would need to describe how this approach is Section 12.0 equivalent to the guidance provided Static in the ISG Computer-Based Testing 124

• Any requalification failures must be remediated and retested prior to returning to license duties Section 13.0

• For ROs and SROs Additional

• Periodicity not to exceed 24 months Guidance for

• For GLROs Requalification

• Periodicity defined by program Programs

• If >24 months, bases provided 125

• Actively perform the functions Section 14.0

• Maintain proficiency and familiarity Proficiency

• Re-establish proficiency if it cannot Programs for Specifically be maintained Licensed Operators and Senior Operators 126

• Appropriate criteria to waive requirements for an examination included in the program Section 15.0

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

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

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

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

Methods 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

• 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

Background:

50.54(m)

Current

• Developed with advanced reactors in mind

• Performance-based process for determining appropriate Practice 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

• 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 Part 53 applicants must include more detail supported by HFE analysis and assessments Approach to

• Operators may fill multiple roles (e.g., maintenance, Staffing 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

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

• A staffing plan must be developed to include the numbers, positions, and qualifications of operators Preliminary and senior operators or, if applicable, generally Part 53 licensed reactor operators across all modes of plant operations, as well as a description of how the Staffing numbers, positions, and responsibilities of personnel Requirement 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

• The staffing plan must include a description of how engineering expertise will be available to the on-shift Proposed crew during all plant conditions to assist in situations Part 53 not covered by procedures or training Requirement

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

Engineering

• Bachelors in in engineering or, Expertise

• Bachelors in engineering technology or a physical science or,

• PE license

[§ 53.730(f)(1)]

• 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

• 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 DRO-ISG-2023- maintained across all modes of plant operations 02: for review

• Approving staffing plans

• For plants that will have specifically licensed operators; of Part 53 could scale the review for plants with generally licensed staffing plans 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

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

• Education prerequisites DRO-ISG-2023-

• Training and qualification 02: for

• Responsibilities of the job reviewing

• Data needs if offsite

• Response time if on site engineering

• Expectations for one or multiple people filling the expertise 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

• 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,

Background:

Human Factors Engineering Program Review Current Practice 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

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

• HFE reviews to be application

Background:

specific (i.e., scaled) considering the Proposed Part characteristics of the facility design 53 Approach to and its operation HFE 138

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

• The plant design must reflect state-

Background:

of-the-art human factors principles Preliminary for safe and reliable performance in Part 53 HFE all locations that human activities Requirement are expected for performing or supporting the continued availability of plant safety or emergency response functions 139

• Objective is to guide reviewer through the process of:

• Developing an application specific review plan

Background:

• Identifying appropriate HFE review guidance Draft 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

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

• Concludes - with completion of application acceptance review Scaling Process:

Overview

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

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 /

Scaling Process: activities for review in conjunction with each target 5 Steps 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

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

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

• Applicability:

• SDAs, DCs, COLs and OLs

• Rationale for scaling reviews Scaling

• Regulatory basis / acceptance criteria Guidance: Main

• Guidance for each step of scaling process

• Objective Body - Key

• Process Features

• Reviewer Responsibilities

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

• Focus is on how to

• Recommended methods for each step of scaling process Scaling Guidance:

• Pointers to sources of additional Appendices - guidance Key Features 145

Characterization:

• What to include in the characterization -

essential elements Scaling

• How to organize and document the Guidance: characterization Appendix A

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

staffing, operator licensing, instrumentation and controls) 146

Targeting:

• General principles for target selection

• Descriptions of 38 prospective (example)

Scaling characteristics of advanced reactor Guidance: designs and operations

• Human performance implications Appendix B

• Availability of guidance to support reviews

Screening:

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

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

Grading:

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

• Considerations for use of documents that Guidance: lack prior NRC endorsement Appendix D

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

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 Scaling reasonable assurance determination Guidance:

• Guidance for documenting the review Appendix E 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 Concern addressed by requirements in both frameworks for novel designs with uncertainties due to requiring systematic assessments to identify events incompleteness in the knowledge base by performing supporting the design and licensing of commercial systematic searches for hazards, initiating events, and nuclear plants. Examples include §§ 53.240 and 53.450 accident scenarios with no preconceptions that could in Framework A and § 53.4730 in Framework B. In limit the creative process. addition, proposed guidance provided in DG-1413.

3. The rule should provide a pathway for licensing Existing pathway for prototype facilities maintained in prototype facilities, when uncertainties in the both frameworks. Provisions included in § 53.440 knowledge base and lack of operating experience (Subpart A, common), § 53.440 for Framework A, and § suggest that additional testing and monitoring are 53.4730 for Framework B. Existing guidance on needed. 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 Included some introduction-type sections to various intent of the rule and its associated design-specific subparts. However, most detailed explanation of the guidance should be developed as soon as possible and rule provided in Preamble.

enshrined in the rule itself.

Interim Letter Report; May 30, 2021

3. Regarding Subpart B:

a. To this point in the development, we find no value in the (a1) Revised Subpart B to eliminate reference to two two-tiered approach to safety tiers. However, safety objectives include: (1) ensuring no requirements. Alternative integral risk criteria to the QHOs immediate threat to public health and safety and (2) should be investigated. considering potential risks.

(a2) See previous discussion on QHOs.

b. Desired flexibility to address the broad range of technologies (c) Framework A continues to define a top-down methodology is provided 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

c. The rule should include a set of over-arching general general design criteria or other generally accepted standards.

principles in one place (Subpart B) that would apply to any reactor concept.

d. The rule should state that safety analyses must demonstrate (d) Framework A (§ 53.450(e)) requires establishing evaluation that for AOOs all safety related barriers to release are criteria for each AOO [anticipated event sequence].

maintained. 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 (e) Framework A (§ 53.450(f)) revised to require safe, stable that DBAs achieve and maintain a safe, stable, and subcritical end state for DBA and subcriticality following LBE required by § condition. 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. Rule language remains general (requiring use of PRA in

a. The requirement for risk-informed analysis is appropriate if Framework A) and flexibility afforded through key guidance the use of PRA is approached in a graded fashion commensurate such as RG 1.247 with the potential consequences and the simplicity of the b. Requirements to identify and assess DBAs provided design. in §§ 53.240 and 53.450(f) in Framework A

b. The requirements for selection and analysis of DBAs must be and § 53.4730(a)(5)(ii) in Framework B. Each clarified. maintains general alignment with Parts 50/52 in terms of

c. The rule eliminates single failure criteria but needs to define establishing design requirements for safety-related the process that replaces it. 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 Concern addressed by requirements in both frameworks Part 53 of October 21, 2020, still apply: for novel designs with requiring systematic assessments to identify events supporting uncertainties due to incompleteness in the knowledge base, the design and licensing of commercial nuclear plants. Examples systematic searches for hazards, initiating events, and accident include §§ 53.240 and 53.450 in Framework A and § 53.4730 scenarios should be required; and a licensing pathway including in Framework B. In addition, guidance provided by developing additional testing and monitoring akin to prototype testing DG-1413.

should be available.

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 Staff agrees with the ACRS, See subsequent iterations and letter to ensure the 10 CFR Part 53 approach yields equivalent discussions safety to current regulatory approaches.

3. The staff should approach the concept of not having a Shift See subsequent iterations and discussions 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.

4. The concept of non-licensed, certified operators should not See subsequent iterations and discussions 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.

5. Staff should develop guidance for judging the acceptability of See subsequent iterations and discussions limited scope simulators.

Interim Letter Report; August 2, 2022

• Discussed during previous session

1. There are limitations of the existing QHOs to fully capture the value and

• Additional questions/discussion ?

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.

Interim Letter Report;

• Preliminary proposed rule language includes a August 2, 2022 definition for safety function

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

2. Critical safety functions are

• Defining critical safety functions remains an foundational to the licensing process.

explicit requirement in Framework A (top-down As such, the requirements for approach); primary and secondary (additional) identifying critical safety functions safety functions made explicit should be common to both frameworks.

• Safety functions are addressed implicitly through the principal design criteria in Framework B, consistent with current bottom-up approach in existing framework

• Draft white paper on preapplication engagement Interim Letter Report; for advanced reactor applicants recommends August 2, 2022 early engagement in several topical areas:

• Principal design criteria

• Selection of LBEs

3. The staff should require, early in

• SSC classification the preapplication process, each

• Source term methodology applicant to identify numeric safety

• QA dose criteria, the critical safety functions, the safety design criteria,

• Probabilistic risk assessment and the underlying rationale for their

• Safety analysis methods selection and application in the

• Fuel qualification and testing design.

• Pre-application engagement is optional and at the discretion of the applicant

Interim Letter Report; August 2, 2022

4. The staff needs to ensure that the

• Fire protection provisions in Framework B have fire protection requirements in both been completely revised (aligned with frameworks are fully technology- Framework A) and are now technology-inclusive inclusive.

Interim Letter Report; August 2, 2022

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

5. The current approach with self-

• Each framework in the preliminary proposed rule contained requirements for each of language must be viewed independently (§ 53.010),

the two frameworks is very long. with some exceptions Furthermore, the rule has a significant amount of implementation

• Requirements in each framework largely replace detail that could be better located in existing requirements under Parts 50, 52, 55, and regulatory guidance. The optics of this 100; either framework is less than half of the approach run counter to a existing requirements 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.

Interim Letter Report; August 2, 2022

• 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

6. The proposed GLRO description of related regulatory requirements should provide for qualified operating personnel. However, the associated

• Significant amount of new guidance would need guidance for implementing to be developed to address recommended 10 CFR Part 55 can be amended to approach accommodate the objectives of the proposed rule without the additional

• Proposed approach should greatly reduce the voluminous text. need for exemptions while enhancing regulatory reliability and clarity

Interim Letter Report;

• Staff considers two tiers of SSC classification August 2, 2022 generally too limiting

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

7. The results of the PRA can be used

• At least one additional tier considered necessary to inform SSC classification by aligning for non-safety related SSCs warranting some type the risk assessment and deterministic of special treatment due to DID/risk safety analysis. This should result, in considerations most cases, in just two tiers for classification of SSCs:

• Framework A: Non-safety related with special Safety Related/Safety Significant and treatment Not-Safety Related/Low Safety

• Framework B: Important to safety Significant.

Interim Letter Report; August 2, 2022

• Staff agreed with the recommendation and are

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

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

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

Two tier safety criteria structure Eliminated two-tiered approach to safety criteria Enabled flexibility in using codes and standards; Unify QA requirements (allow broader set of QA requirements consolidated in rule and codes and standards) aligned with Appendix B to 10 CFR Part 50 Decoupled requirements for normal operation Normal operations from those for LBEs Add requirements for safe, stable end state Added requirement and clarified in Statements of conditions Consideration 166

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

NSRSS SSCs reduce sole reliance on safety-related Special treatment for non-safety related but SSCs; Requirements can be scaled to achieve desired safety significant SSCs capability/reliability/overall risk Staff views FSP as an operational benefit. Allows continued use of PRA for evaluating changes, Facility safety program (FSP) managing risks, and improving the relationship between the NRCs licensing and reactor oversight programs.

Staff agrees and has aligned with industry on future More guidance is needed to clarify regulations guidance needs 167

Industry Feedback on Framework B Feedback NRC Staff Perspectives Preamble discussion includes amplifying information to Objectives for chemical hazard requirements are address this feedback. Chemical hazards in question would include substances commingled with licensed material or those unclear 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 Staff revised several sections to ensure that the proposed rule beyond-design-basis events [MBDBE] is technology-inclusive, including MBDBE requirements requirements in § 50.155)

The requirement to have a PRA developed to support a CP PRA development at CP stage is not reasonable application is consistent with the 50/52 rulemaking and other Commission policies AERI entry conditions distinguish between plants with relatively straightforward designs and plants with relatively complicated Proposed entry conditions for AERI are too designs that warrant the development of a PRA in order to conservative understand their risk. The proposed AERI option is a departure from current Commission policy, which requires all new plants to have a PRA 168

Key Guidance Development Under Development Near-Term Part 53 Existing Future

• TICAP/ARCAP (NEI 21-07)

• DG-1413, Identification of

• LMP (RG 1.233)

• Analytical Margin

• Non-LWR PRA Standard Licensing Events

• Siting Criteria (RG 4.7)

• Chemical

• Non-LWR PRA Standard Applicability

• DG-1414, AERI Methodology

• Fuel Qualification Hazards ISG

• DRO-ISG-2023-01, Operator Framework (NUREG-

• Manufacturing

• High Temp Materials (ASME III-5) Licensing Program Review ISG 2246)

• Technical

• Reliability & Integrity Mgt (ASME XI-2)

• DRO-ISG-2023-02, Staffing Plan

• Developing Principal Specifications

• Molten Salt Reactor Fuel Qualification Review ISG Augmenting NUREG-Design Criteria for Non-1791

• FSP

• Seismic Design/Isolators LWR (RG 1.232)

• DRO-ISG-2023-03, Scalable

• Framework B

• Emergency Planning

• Change Evaluation (Southern Nuclear Human Content of Operating Company-led) Factors Engineering Review ISG Applications

• QA Alternatives (NEI-led)

• Part 26, FFD

• Facility Training Programs

• Part 26, Fatigue Management

• Materials Compatibility ISG

• Part 73, AA

• Part 73, Cyber Security

• Part 73 Security Programs

TICAP / ARCAP - Nexus Part 53

• Overall organization

• Use of QHOs FRN

• ALARA Section VII

• DID

• Earthquake Engineering Specific

• Construction and Manufacturing Requests for

• Use of references

• Manufacturing licenses Comments

• Staffing and GLROs

• OnShift engineering expertise

• Training program accreditation

• Use of simulation facilities

Part 53

• FSPs

• Integrity assessment programs FRN

• Decommissioning Section VII

• PRA information

• Changes to manufacturing licenses Specific

• Specific requirements for Technical Requests for Specifications

• AERI Comments

• 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

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

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

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

Backup Slides 179

• 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 Nuclear Energy reactors no later than December 2027 Innovation and * (9) REGULATORY FRAMEWORKThe term regulatory framework means the framework for reviewing requests for Modernization Act certifications, permits, approvals, and licenses for nuclear reactors.

(NEIMA)

• (14) TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK The term technology-inclusive regulatory framework means January 2019 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

• 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 Part 53 provided direction to the staff.

Rulemaking Plan

• 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

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 Part 53 requirements in 10 CFR Part 50 and 10 CFR Part 52, Rulemaking

4. Establish new requirements to address non-light-Objectives 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

Subparts H & R:

Leveraging and Combining Existing Licensing Processes Commercial Operations Fuel Load Operating License Site (OL) selected Combined License (COL)

Use OL or custom COL to develop a CP based on Site subsequent DC SDA or DC selected Standard Design Manufacturing Design Certification Approval (SDA) License (ML) (DC)

Part 50 Site selected Part 52 CP and COL may reference Early Site Permit (ESP) Part 53 Construction Permit (CP) 183