ML22301A107

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11/2/2022 - NRC Presentation to Support the ACRS Full Committee Meeting 10 CFR Part 53? Licensing and Regulation of ?Advanced Nuclear Reactors
ML22301A107
Person / Time
Issue date: 11/02/2022
From: Robert Beall
NRC/NMSS/DREFS/RRPB
To:
References
NRC-2019-0062, 10 CFR Part 53, RIN 3150-AK31
Download: ML22301A107 (30)
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Text

1 0 C F R Pa r t 5 3 L i c e n s i n g a n d R e g u l a o n o f A d v a n c e d N u c l e a r R e a c t o r s N o v e m b e r 2, 2 0 2 2 Advisory Committee on Reactor Safeguards (ACRS)

Agenda 2

8:35 am - 10:30 am Staff presentation on 10 CFR Part 53, Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants, Proposed Rulemaking Language Rulemaking Schedule Part 53 Licensing Frameworks Risk Insights/Quantitative Health Objectives (QHOs)

Fueled Modules Codes and Standards Alternative evaluation for risk insights (AERI)

Generally licensed reactor operators (GLROs), Human Factors, Engineering Expertise Guidance

Rulemaking Schedule 3

We are here

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

Rule Package (ML22272A034)

Sections 53.000 and 53.010

  • Purpose
  • Provide optional frameworks for the issuance, amendment, renewal, and termination of licenses, permits, certifications, and approvals for commercial nuclear plants
  • Frameworks
  • Framework A and Framework B are distinct
  • Applicants and licensees subject to the rules in this part must only use the subparts applicable to one framework 5

Subpart A -

General Provisions (Definitions)

  • Common Definitions
  • Commercial Nuclear Plant
  • Manufactured reactor
  • Manufactured reactor module
  • Safety function
  • Framework A Definitions
  • Construction, Licensing basis events (LBEs),

structure, system, and component (SSC) classifications

  • Framework B Definitions
  • Construction, Design basis, Functional containment, Safety-related SSCs, Severe nuclear accident 6

7 Framework A Subpart Title Topics Subpart B Technology-Inclusive Safety Requirements Risk Insights (QHOs)

Subpart C Design and Analysis Requirements Subpart D Siting Requirements Subpart E Construction and Manufacturing Requirements Fueled Modules Subpart F Requirements for Operation GLROs, Human Factors Subpart G Decommissioning Requirements Subpart H Licenses, Certifications and Approvals Subpart I Maintaining and Revising Licensing Basis Information Subpart J Reporting and Other Administrative Requirements Subpart K Quality Assurance Criteria for Commercial Nuclear Plants

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.

8

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.*

  • Existing Paradigm Does not specifically define adequate protection but compliance with NRC regulations and guidance may be presumed to assure adequate protection at a minimum Additional requirements as necessary or desirable to protect health or to minimize danger to life or property 9

Subparts E & O Fuel loading for manufactured reactor modules

§ 53.620(d) / § 53.4120(d) Fuel loading

  • A manufacturing license may include authorizing the loading of fuel into a manufactured reactor module
  • Specify required protections to prevent criticality o At least two independent mechanisms that can prevent criticality should conditions result in the maximum reactivity being attained for the fissile material
  • Commission finding that a manufactured reactor module in required configuration is not a utilization facility as defined in the Atomic Energy Act
  • Manufactured reactor module becomes a utilization facility in its final place of use after the Commission makes required findings on inspections, tests, analyses and acceptance criteria 10

11 Framework B Subpart Title Topics Subpart N Siting Subpart O Construction and Manufacturing Requirements Subpart P Requirements for Operation Codes and Standards Subpart Q Decommissioning Subpart R Licenses, Certifications and Approvals Codes and Standards AERI Subpart S Maintaining and Revising Licensing Basis Information Subpart T Reporting and Other Administrative Requirements Subpart U Quality Assurance

Codes and Standards (Clarification) 10 CFR 53.4730(a)(2)(ii)(A) would require applicants to provide a description and justification (for codes or standards not previously endorsed or accepted by the NRC) of the codes and standards to be used in the design Other Framework B requirements related to codes and standards are similar to those in the existing regulations 10 CFR 53.4360(a) would require boiling-water reactor (BWR) and pressurized-water reactor (PWR) licensees to meet requirements in 10 CFR 50.55a for inservice inspection and inservice testing programs 10 CFR 53.4730(a)(37)(ii) would require applicants for BWRs and PWRs to describe how they will comply with ASME Boiler and Pressure Vessel Code and ASME Operation and Maintenance Code requirements in 10 CFR 50.55a Conforming changes proposed for 10 CFR 50.55a would support use of existing requirements by applicants and licensees with BWRs or PWRs under Framework B 12

Subpart R -

AERI The AERI approach is consistent with Commission policy.

The AERI entry conditions in § 53.4730(a)(34)(ii) were revised after the ACRS Part 53 subcommittee meeting (October 18-19, 2022) to address stakeholder comments and reflect insights from the scoping MELCOR Accident Consequence Calculation System (MACCS) calculations.

Other provisions in Part 53 reference make use of the AERI entry conditions.

Two draft regulatory guides (DGs) developed:

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

DG-1414: Alternative Evaluation for Risk Insights Methodology (proposed new RG 1.255) 13

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

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

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

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

Why Revise the AERI Entry Conditions?

15

  • Some stakeholders have commented that the current proposed AERI entry conditions are overly conservative.
  • MACCS scoping calculations indicate that dose at 100 meters is an inadequate predictor of conditional risk. Depending on the assumptions (e.g., plume elevation or buoyancy), some conditional risks may be below the QHOs while others may be above the QHOs even though the current AERI entry condition is met.
  • Provide increased flexibility when determining if the AERI entry conditions are met.

Revised AERI Entry Conditions 16

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

i.

A probabilistic risk assessment (PRA); or ii.

An alternative evaluation for risk insights (AERI), provided that:

(A) The analysis of a postulated bounding event demonstrates that the consequence evaluated within the area between the commercial nuclear plants exclusion area boundary (EAB) and 16.1 kilometers (10 miles) from the EAB is less than 25 mSv (2.5 rem) TEDE in the first year; and (B) The identification of the postulated bounding event is informed by a systematic and comprehensive search for severe nuclear accident scenarios that considers:

(1) All radiological sources at the commercial nuclear plant; (2) Relevant internal and external hazards; (3) Combinations of plant equipment failures including common-cause failures, hazard-induced equipment failures, and equipment failures caused by severe nuclear accident phenomena; and (4) Credible human errors of commission and omission.

Rationale for the Revised AERI Entry Conditions 17

  • The change from dose at 100 meters to the peak dose within the 10-mile annulus addresses concerns about elevated releases and plume buoyancy.

EAB EAB + 10 miles Limit the peak dose within this annulus Buoyant Plume

Rationale for the Revised AERI Entry Conditions (Cont.)

18

  • The 2.5-rem criterion is consistent with MACCS scoping calculations:
  • A 25-rem lifetime (50-year) dose generally corresponds to a 10-mile population-weighted lifetime individual latent cancer fatality risk less than 2E-6 per event.
  • A first-year dose of 2.5 rem generally corresponds to a 50-year dose less than 25 rem, probably due to radioactive decay and the effect of weathering on groundshine and resuspension.
  • The 2.5-rem criterion is a small fraction (10%) of the traditional reference value (25 rem) used in Part 100 and § 50.34.
  • For example, see the Standard Review Plan (NUREG-0800), Section 15.0.3, Rev. 0: A small fraction is defined as less than 10% of the 10 CFR 50.34(a)(1) reference values, or 2.5 rem TEDE.
  • 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 must have GLROs in lieu of senior reactor operators and reactor operators Proposed Uses of the AERI Entry Conditions 19
  • All other applicable Framework B requirements must be met (AERI or PRA).
  • Applicants may elect to develop a PRA even if the AERI entry conditions are met.

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

  • Section A: Applies to light-water reactors (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) 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

o Reviews techniques for searching for initiating events and points the user to helpful references o Does not endorse or recommend any specific technique 20

  • 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.

DG-1414: Alternative Evaluation for Risk Insights Methodology (proposed new RG 1.255) 21 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.

Subparts F and P Staffing, HFE, Operator Licensing, and Training

  • During the 10/19/22 subcommittee meeting, the staff provided an update on the rule language, as well an overview of key guidance
  • Updates on the rule language status had included:

o Consolidating Frameworks A & B requirements using a common set of language under Subpart F o Extending provisions for GLROs to Framework B, to include facilities using an AERI approach o Retaining previous engineering expertise provisions (i.e., degreed individuals with plant familiarity)

  • Important points of ISG presentations included:

o Review guidance for tailored exam programs o Staffing review guidance for custom staffing plans o Guidance for conducting scalable human factors engineering (HFE) reviews 22

Follow-on Discussion of Operator Licensing Topics

  • Regarding Operator Licensing, the members asked that the staff discuss several areas further, including:

o Lack of approval preceding licensing of GLROs NRC approved program with inspections o How changes to operator tasks from plant mods translate into adjustments to exam program knowledge and abilities lists and change control process burdens Balances adaptability and program assurance o How the GLRO criteria interrelate with the AERI criteria and whether AERI is too restrictive The following slide provides an overview that builds on earlier AERI discussions 23

Follow-on Discussion of GLRO Criteria 24 Underlying Principle from Paper GLRO Criteria for Framework A GLRO Criteria for Framework B (PRA)

GLRO Criteria for Framework B (AERI)

Radiological consequence criteria met without human action Safety criteria (53.210 and 53.220 or 53.470) met without human actions for credited event mitigation Safety assessment (53.4730(a)(1)(vi))

demonstrates requirements met without credited human action Qualification for AERI (53.4730(a)(34)(ii))

must be demonstrated to be met Licensing basis events addressed without human action Analysis of LBEs and DBAs (53.450(e & f))

demonstrates criteria met without human actions for credited mitigation PRA (53.4730(a)(34))

demonstrates event sequences met without human actions for credited mitigation Safety functions not allocated to human action Safety functions (53.230) achieved without reliance on human actions for credited event mitigation FRA/FA (53.730(d))

demonstrates functions required for safety do not rely on credited human action Reliance on inherent or robust passive features Plant response to licensing basis events does not credibly rely on human actions to assure the performance of SSCs (e.g., SSCs function through inherent characteristics or have engineered protections against human failures)

Adequate DID without human action DID requirements (53.250) met without human actions for the purposes of credited DID Plant design must provide for layered DID without dependence upon any single barrier or reliance upon credited human action.

Follow-on Discussion of Staffing Topics

  • Regarding operational staffing, the members asked that the staff discuss several areas further, including:

o Potential for allowing plants with no operators There is no allowance for zero operator staffing o Engineering expertise degree requirement Complements/augments plant ops experience o Training requirements for engineering expertise role Systems approach to training required by § 53.830; topics covered by ISG o Availability of remote engineering expertise Not credited in event mitigation; supports crew o Requirements might allow remote operation Framework for staffing, HFE, operator licensing, and training is designed to adapt to future concept of operations; remote operations is a broader issue 25

Key Guidance Development

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

Existing

  • Analytical Margin
  • Chemical Hazards
  • Manufacturing
  • Technical Specifications
  • Facility Safety Program
  • Framework B Content of Applications Future Under Development Near-Term
  • ASME/ANS Non-LWR PRA Standard
  • Non-LWR PRA Standard Applicability ISG
  • High Temp Materials (ASME III-5)
  • Reliability & Integrity Mgt (ASME XI-2)
  • Molten Salt Reactor Fuel Qualification
  • Seismic Design / Isolators
  • Emergency Planning (50.160)
  • Change Evaluation (SNC-led)
  • QA Alternatives (NEI-led)
  • Facility Training Programs ISG
  • Materials Compatibility ISG
  • Treatment of Consequence Uncertainty Part 53
  • DG-1413, Identification of Licensing Events
  • DG-1414, AERI Methodology
  • DRO-ISG-2023-01, Operator Licensing Program Review ISG
  • DRO-ISG-2023-03, Scalable Human Factors Engineering Review ISG
  • Part 26, Fatigue Management
  • Part 73, Access Authorization
  • Part 73 Security Programs Part 53
  • DG-1413, Identification of Licensing Events
  • DG-1414, AERI Methodology
  • DRO-ISG-2023-01, Operator Licensing Program Review ISG
  • DRO-ISG-2023-03, Scalable Human Factors Engineering Review ISG
  • Part 26, Fatigue Management
  • Part 73, Access Authorization
  • Part 73 Security Programs 26

Discussion 27

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 28

ACRS Advisory Committee on Reactor Safeguards AERI Alternative evaluation for risk insights ANS American Nuclear Society ARCAP Advanced Reactor Content of Application Project ASME American Society of Mechanical Engineers BWR boiling-water reactor CFR Code of Federal Regulations DBA design-basis accident DG draft regulatory guidance DID defense-in-depth DRO Division of Reactor Oversight EAB exclusion area boundary Acronyms EDO Executive Director for Operations FA function allocation FR Federal Register FRA functional requirements analysis GLRO generally licensed reactor operator HFE human factors engineering ISG interim staff guidance LBE licensing basis events LMP Licensing Modernization Project LWR light-water reactor MACCS MELCOR accident consequence code system mSv millisievert 29

NEI Nuclear Energy Institute non-LWR non-light-water reactor NRC U.S. Nuclear Regulatory Commission NUREG U.S. Nuclear Regulatory Commission technical report designation PRA probabilistic risk assessment PWR pressurized-water reactor QA quality assurance Acronyms QHO quantitative health objective rem Roentgen equivalent man RG regulatory guide SNC Southern Nuclear Operating Company SSCs structures, systems, and components TEDE total effective dose equivalent TICAP Technology Inclusive Content of Application Project 30

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