ML19351E380

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December 2019 - Advanced Reactors Stakeholders Meeting Presentation Slides
ML19351E380
Person / Time
Issue date: 12/12/2019
From: Vechioli L
NRC/NRR/DANU/UARL
To:
Vechioli L, NRR/DANU/UARL, 415-6035
References
Download: ML19351E380 (123)
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Public Meeting on Possible Regulatory Process Improvements for Advanced Reactor Designs December 12, 2019 Telephone Bridge: (888) 790 3549 Passcode: 9537029 1

  • Telephone Bridge (888) 790 3549 Passcode: 9537029
  • Opportunities for public comments and questions at designated times 2

Outline 9:00 - 9:10 am Opening Remarks 9:10 - 9:20 am Advanced Reactors Materials Workshop Dec 9-11 9:20 - 9:50 am Overview of NEI Micro-Reactor Paper 9:50 - 10:20 am Non-Light Water Reactor Review Strategy 10:20 - 10:50 am Advanced Reactor Preparations for Environmental Reviews 10:50 - 11:00 am Break 11:00 - 12:00 pm Physical Security Proposed Rule for Advanced Reactors 12:00 - 1:00 pm Lunch 1:00 - 2:15 pm Technology Inclusive Content of Applications Project (TICAP) 2:15 - 3:00 pm Canadian Nuclear Safety Commission (CNSC)

Licensing Approach 3:00 - 3:10 pm Advanced Reactor Export Working Group 3:10 - 3:30 pm Future Meetings planning and Open Discussion 3

Advanced Reactors - Materials and Component Integrity Workshop Dec 9-11, 2019

- Raj Iyengar, NRC 4

Overview of NEI Micro-Reactor paper

- Marc Nichol, NEI 5

Non-Light Water Reactor Review Strategy

- Jan Mazza, NRC 6

Advanced Reactor Preparations for Environmental Reviews

- Mallecia Sutton, NRC 7

Break Meeting/Webinar will begin shortly Telephone Bridge: (888) 790 3549 Passcode: 9537029 8

Physical Security Proposed Rule for Advanced Reactors

- Nanette Valliere/Joseph Rivers, NRC Potential Considerations for Physical Security for Advanced Reactors Rulemaking

-Kati Austgen, NEI 9

Lunch Meeting/Webinar will begin shortly Telephone Bridge: (888) 790 3549 Passcode: 9537029 10

Technology Inclusive Content of Applications Project (TICAP)

- Joseph Sebrosky, NRC/Thomas Hicks INL Technology Inclusive Content of Applications Project (TICAP)

- Amir Afzali, Southern Co./Steve Nesbit, LMNT Consulting 11

Canadian Nuclear Safety Commission Licensing Approach

- Chantal Morin, CNSC 12

Advanced Reactor Export Working Group

- Lauren Mayros, NRC 13

Open Discussion and Closing 14

2020 Tentative Schedule for Periodic Stakeholder Meetings February 6 March 26 May 7 June 18 August 6 September 24 November 5 December 17 15

Micro-Reactors Regulatory Topics Marc Nichol, Director New Reactor Deployment December 12, 2019

©2019 Nuclear Energy Institute

Overview of Micro-Reactors Typically 1 MWe to 10 MWe (2 MWth to 40 MWth)

Very small size

  • Building footprint as small as 1,000 sqft
  • Reactor w/ balance of plant is road shippable, minimal site work Non-traditional uses and locations
  • Remote villages (e.g., arctic and islands)
  • Mining operations
  • Defense and other mission critical installations
  • Micro-grids - in populated areas Rapidly maturing technology
  • First licensing applications expected in 2020

©2019 Nuclear Energy Institute 2

Micro-Reactor Safety Very small inventory of fission products, as compared to current large reactors Key Features*

  • Safety provided entirely by passive and inherent features
  • Fail-safe to shut down automatically
  • Accident and proliferation resistant fuel, enriched below 20% U-235
  • High fission product retention
  • Operator actions are not needed to assure safety of the reactor
  • Reactor can be located completely below ground
  • Operational simplicity with very few instruments and controls, and active SSCs
  • General description, all features may not be applicable to all designs ©2019 Nuclear Energy Institute 3

Micro-Reactor Regulatory Considerations Safety profile of micro-reactors fundamentally differ from other power reactors

  • Existing regulations and proposed rule changes are often not well suited
  • Near term - alternative approaches or exemptions, long-term may need rulemaking Micro-reactors are more similar to research and test reactors (RTR)
  • Similar in power level and potential consequences Some differences between micro-reactors and RTRs are expected
  • Micro-reactors operate at full power more frequently and for longer periods
  • Micro-reactors have balance of plant
  • Micro-reactors have more inherent safety features, such that some accident scenarios may not be relevant
  • Micro-reactors may not require human action for accident response
  • Micro-reactors do not perform test and experiments

©2019 Nuclear Energy Institute 4

Micro-Reactor Regulatory Issues Priority Issues Addressed in Broader Other Potential Issues Efforts

1. Review Scope, Duration,
  • Siting
  • QA Level of Effort
  • Environmental Reviews
2. Operations (auto/remote)
  • Fuel Qualification
  • Liability Insurance
3. Inspections
  • Decommissioning Funding
4. Physical Security
  • Annual Licensee Fees
5. Emergency
  • Generic License Preparedness
  • Transportation
6. Aircraft Impact

©2019 Nuclear Energy Institute 5

Review Scope, Duration, Level of Effort Recommend NRC establish target for review schedule

  • Less than 12 months from Acceptance to Final Safety Evaluation including environmental review Relevant recent NRC experience on level of effort
  • SHINE Construction Permit - 22,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />
  • Northwest Medical Construction Permit - 12,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />

©2019 Nuclear Energy Institute 6

Automatic / Remote Operations Automatic/remote operations may have very few controls and not require human action Applicant evaluation to determine whether there is need for operator actions Exemptions from Part 50.54 and Part 55 may be needed If no human action needed to protect public health and safety

  • No licensed operators required (numbers, presence, training, requal, simulator)

If some actions needed to protect public health and safety

  • Applicant propose licensed operator scheme (number, presence, training)
  • Ability to return within an acceptable time could be credited for presence Traditional control room may not be necessary If safety can be assured without operator action, and if unauthorized individual cannot compromise safety through manipulating controls:
  • No need for requirements relating to the control room, operator initiated shutdown or I&C to be in a restricted area

©2019 Nuclear Energy Institute 7

NRC Inspection and Oversight Licensee is responsible for safety and regulatory compliance NRC independently verifies licensees compliance through inspection

  • Performance based, focused on activities important to safety
  • Emphasis on observing activities over reviewing documents Micro-reactors have very few activities Recommended NRC inspection paradigm
  • Inspection frequency - once every one or two years
  • No need or requirement for resident inspectors
  • Construction inspection program emphasize vendor inspections

©2019 Nuclear Energy Institute 8

Emergency Preparedness EP rulemaking for SMRs and other nuclear technologies

  • Performance based, technology inclusive, consequence oriented
  • Do not explicitly refer to micro-reactors Draft rule 10 CFR 50.60
  • Appears flexible enough for micro-reactors DG-1530
  • Does not contemplate extremely low potential consequences of micro-reactors Need guidance on emergency plans for micro-reactors

©2019 Nuclear Energy Institute 9

Physical Security Security rulemaking for SMRs and other nuclear technologies

  • Not very performance based, technology inclusive, consequence oriented
  • Results in excessive burden on micro-reactors For micro-reactors where most severe scenarios do not result in undue risk to public
  • Should not be required to have additional design features or protective action to protect from DBT
  • Could determine DBT not applicable Requirements for micro-reactor security principally focused on theft and diversion
  • Access control, Intrusion detection, Communications with law enforcement Requirements related to protecting against DBT would not be necessary, e.g.,
  • Certain considerations for physical barriers
  • Force on force evaluation
  • May not require personnel on-site

©2019 Nuclear Energy Institute 10

Aircraft Impact Assessment Impact of large aircraft is beyond-design-basis-event

  • Requirements on design, construction, testing, ops and maintenance for design-basis events do not apply Aircraft impact on micro-reactors are highly unlikely
  • Not easily impacted: building is small in size, some may be below grade
  • Not high-value target: could not cause mass casualties Impact does not pose hazard to public: very little radionuclide inventory For micro-reactors where aircraft impact unlikely to pose substantial hazard to public
  • Meet intent of Part 50.150, but not specific criteria
  • Near-term - exemptions to 50.150 appropriate

©2019 Nuclear Energy Institute 11

Conclusions First micro-reactor application expected in 2020 Key Policy and Technical issues to address promptly

  • Review duration and level of effort
  • Automatic / remote operations
  • NRC Inspection
  • Physical Security
  • Aircraft Impact Additional work would be helpful
  • Details on design and performance of key safety features
  • Characterization of radionuclide inventories and source terms
  • Generic assessments (e.g., operator actions, likelihood of aircraft impact)

©2019 Nuclear Energy Institute 12

Non-Light Water Reactor Review Strategy -

Staff White Paper December 12, 2019

Introduction

Background

Purpose Contents Status

=

Background===

NRC recognized that review guidance was needed:

to support the near-term reviews of applications for non-LWR designs submitted prior to the development of the technology inclusive - risk informed performance based (TI-RIPB) regulatory framework in 2027, as required by the Nuclear Energy Innovation and Modernization Act (NEIMA).

Purpose The purpose of the Non-LWR Review Strategy is:

to provide acceptance criteria for non-LWR reviews to provide NRC staff an approach to reviewing the licensing basis information of a non-LWR application independent of the specific design or methodology used.

Contents Specific topics addressed in the review guide:

Section 2.0 - Non-LWR Vendor Approaches to Developing the Licensing Basis Section 3.0 - Acceptance Criteria and NRC Staff Review Approach Section 4.0 - References Section 5.0 - Acronyms Attachment 1 - Analysis of Applicability of NRC Regulations for Non-LWRs

Contents (Cont.)

Section 2.0 - Non-LWR Vendor Approaches to Developing the Licensing Basis:

describes the various approaches to developing the licensing basis for non-LWRs provides background and references for pre-application interactions discusses contents of applications discusses the development of the Safety Evaluation Report

Contents (Cont.)

Section 3.0 - Acceptance Criteria and NRC Staff Review Approach describes the scope and focus of the staffs technical review discusses, in general terms, the acceptance criteria that could be considered by NRC staff during the technical review of a non-LWR application provides guidance for the analysis and evaluation of the integrated system design provides expectations for probabilistic risk assessments for non-LWRs

Contents (Cont.) - Analysis of Applicability of NRC Regulations for Non-LWRs Table 1 - Areas of Expected Exemptions Table 2 - Part 50 Requirements Table 3 - Part 52 Subpart B Design Certification Requirements Table 4 - Other Applicable Regulations

Status Currently under review by OGC Once complete final review by DANU management Not formally soliciting public comments - but will consider any feedback that we receive

QUESTIONS?

Advanced Reactor Preparations for Environmental Reviews Mallecia Sutton Senior Project Manager Division of Advanced Reactors and Non-Power Production and Utilization Facilities 1

Status on Environmental Activities

  • Status update on:

Interim Staff Guidance for the environmental review of micro-reactors Guidance on addressing fuel cycle and fuel transportation impacts for non-LWRs Consideration of Generic Environmental Impact Statement for Advanced Reactors 2

Interim Staff Guidance

  • Assist the NRC staff in determining the scope and scaling the environmental reviews associated with micro-reactor applications
  • Internal Concurrence To be issued as a Draft for Comments

Fuel Cycle Impacts

  • Developing white paper on fuel cycle impacts for non-LWRs 10 CFR 51.51 only applies to LWRs and does not mention non-LWR license applicants Evaluate fuel cycle impacts to meet obligations under the National Environmental Policy Act (NEPA)
  • Developing white paper transportation impacts for non-LWRs 10 CFR 51.52 only applies to LWRs and does not mention non-LWR license applicants Evaluate transportation impacts to meet its obligations under NEPA 4

GEIS Exploratory Process

  • Exploratory Process Public meetings were held November 15th and 20th Extended comment period-January 24, 2020 Meeting summary to be issued within 30 days
  • Workshop January 8, 2019, 2-4 pm Commission Hearing Room
  • Report on Exploratory Process February 15, 2020 5

Rulemaking for PHYSICAL SECURITY FOR ADVANCED REACTORS Proposed Rule (NRC Docket ID: NRC-2017-0227)

December 12, 2019 1

Purpose

  • Discuss the path forward for the proposed rulemaking Physical Security for Advanced Reactors.
  • Solicit public feedback on the proposed rulemaking approach.

NRC will not provide formal comment responses to any oral remarks made at this meeting.

2

Background

  • SECY-18-0076, Options and Recommendation for Physical Security for Advanced Reactors, dated August 1, 2018 (ADAMS Accession No. ML18170A051)
  • The staff proposed 4 alternatives and recommended alternative 3:
1) No change / Status quo
2) Address possible requests for alternatives via guidance
3) Limited scope rulemaking
4) Broader based rulemaking
  • Staff Requirements Memorandum, SRM-SECY-18-0076, dated November 19, 2018 (ADAMS Accession No. ML18324A469)
  • The Commission approved a limited scope rulemaking (Alternative 3)
  • The Commission directed the staff to interact with stakeholders to identify specific requirements within existing regulations that would play a diminished role in providing physical security for advanced reactors 3

Regulatory Basis

  • Issued for public comment on July 16, 2019 Comment period closed on August 15, 2019 Six entities provided comments: 4 support, 2 oppose the rulemaking Comments on the regulatory basis will be addressed in the proposed rule
  • Permits future applicants and licensees to meet alternative requirements for a risk-informed, performance-based approach for designated portions of the physical security program.
  • Retains the current overall framework for security requirements while providing alternatives for advanced reactors to certain physical security regulations and guidance.
  • Most likely focus of the rulemaking is an alternative to the prescribed minimum number of armed responders currently defined in 10 CFR 73.55 (k) and prescriptive requirements in 10 CFR 73.55 for onsite secondary alarm stations.

Performance Criteria The staff identified the following three performance criteria that could be used to identify reactors that could make use of the alternative provided in this proposed rulemaking:

1) The radiological consequences from a hypothetical, unmitigated event involving the loss of engineered systems for decay heat removal and possible breaches in physical structures surrounding the reactor, spent fuel, and other inventories of radioactive materials result in offsite doses below the reference values defined in 10 CFR 50.34(a)(1)(ii)(D) and 52.79(a)(1)(vi); or
2) The plant features necessary to mitigate an event and maintain offsite doses below the reference values in 10 CFR 50.34(a)(1)(ii)(D) and 52.79(a)(1)(vi) cannot reasonably be compromised by the DBT for radiological sabotage; or
3) The plant features include inherent reactor characteristics combined with engineered safety and security features that allow for facility recovery and mitigation strategy implementation if a target set is compromised, destroyed, or rendered nonfunctional, such that offsite radiological consequences are maintained below the reference values defined in 10 CFR 50.34(a)(1)(ii)(D) and 52.79(a)(1)(vi).

5

Performance Criteria Is 1 Yes Reactor design met?

No Is 2 Yes Alternate Regulation Facility design met? Applicable No Is 3 Yes Mitigation strategies met?

No Alternate Regulation NOT Applicable 6

Proposed Approach

  • Alternative to the prescribed requirement in 10 CFR 73.55(k)(5)(ii) that the number of armed responders shall not be less than ten
  • Alternative to the prescriptive requirements for a secondary alarm station in:

7

Additional Potential Areas for Consideration

  • The staff is seeking input from the public on whether the performance criteria identified above should be applied to any additional prescriptive requirements, for example those found in:

8

Open Session /

Request Feedback 9

Next Steps

  • Proposed Rule and Draft Guidance Provide to the Commission in January 2021 Issue for public comment in 2021
  • Final Rule and Final Guidance Provide to the Commission in May 2022

Contact:

Dennis Andrukat, Rulemaking Project Manager Email: Dennis.Andrukat@nrc.gov 10

Potential Considerations for Physical Security for Advanced Reactors Rulemaking Kati Austgen NRC Public Meeting December 12, 2019

Draft Regulatory Basis for Rulemaking for Physical Security for Advanced Reactors Performance-based alternatives to prescribed physical security regulations

  • Minimum number of onsite armed responders

[10 CFR 73.55(k)(5)(ii)]

  • Onsite secondary alarm station [10 CFR 73.55(i)(4)(iii)]

Staff indicated a willingness to identify additional requirements that may be eliminated or modified to reduce the number of exemptions

©2018 Nuclear Energy Institute 2

Other Requirements for Consideration Permit security personnel to have additional duties and/or be located outside the Protected Area provided the NRC-approved security plan can be effectively implemented Consideration of requirements applicable to microreactors

  • Appendix E of NEI White Paper, Micro-Reactor Regulatory Issues, dated November 13, 2019 (ML19319C497)

©2018 Nuclear Energy Institute 3

Questions/Discussion

©2018 Nuclear Energy Institute 4

Technology-Inclusive Content of Applications Project (TI-CAP)

NRR/DANU - Advanced Reactor Policy Branch US Nuclear Regulatory Commission December 12, 2019

  • Purpose

- Build on progress from licensing modernization project to provide more guidance on content of applications

- Coordinate activities with utility-led TI-CAP

- Solicit stakeholder feedback on draft outline for advanced reactor applications based on a technology-inclusive, risk-informed, performance based process

- Solicit stakeholder feedback on portions of the application outside the scope of a final safety analysis report (FSAR);

with initial focus on technical specifications 2

Agenda

  • Background
  • Draft outline of final safety analysis report (FSAR)
  • Licensing documents outside the scope of an FSAR

- Technical specification development

  • Discussion
  • Next steps

- Process and timeline for providing comments 3

Background - Integrated Approach

  • Advanced Reactor Policy Statement

- Designs with some or all of these attributes are also likely to be more readily understood by the general public.

Indeed, the number and nature of the regulatory requirements may depend on the extent to which an individual advanced reactor design incorporates general attributes such as those listed previously.

  • SRM-SECY-10-0121

- New reactors with these enhanced margins and safety features should have greater operational flexibility than current reactors. This flexibility will provide for more efficient use of NRC resources and allow a fuller focus on issues of true safety significance.

4

Return to first principles 5

Fundamental Safety Functions

  • General Design Criteria (light water reactors)

- Protection and Reactivity Control Systems

- Fluid Systems

- Reactor Containment

- Fuel and Radioactivity Control

  • IAEA SSR 2/1, Safety of Nuclear Power Plants: Design

- Fulfilment of the following fundamental safety functions for a nuclear power plant shall be ensured for all plant states: (i) control of reactivity; (ii) removal of heat from the reactor and from the fuel store; and (iii) confinement of radioactive material, shielding against radiation and control of planned radioactive releases, as well as limitation of accidental radioactive releases.

- . fundamental safety functions such as controlling reactivity and reactor power, removing heat, and limiting the release of radioactive materials from a reactor facility.

  • Utility-led TI-CAP white paper on fundamental safety functions 6

Starting Point is Licensing Modernization Project (NEI 18-04) 7

SSC Classification and Level of Information in an Application

  • Expectation is that FSAR portion of an application would be more detailed for safety-related structures, systems, and components (SSCs) and less detailed for other SSCs
  • Level of information for non-safety related special treatment (NSRST) and non-safety related with no special treatment (NST) SSCs would be a function of its risk significance 8

Informing Content of Applications

  • NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Revision 1 (ADAMS Accession No. ML19241A336) considered in developing draft outline sections 1 through 14

- Staffs draft outline differs from traditional organization of information (e.g., RG 1.70, RG 1.206)

- Discussion on appropriate scope and level of detail might be helped by an early agreement on overall organization of information (format)

  • Draft outline addresses full scope of a combined license but it could be adapted for other applications including:

- Construction permit

- Operating license

- Design certification

- Standard design approval

- Manufacturing license 9

Informing Content of Applications

  • Staffs draft outline (with assistance from Idaho National Lab) found in ADAMS at Accession No. ML19325C089
  • Staffs draft outline Sections 1 through 14

- Final safety analysis report (FSAR) portion of application

  • Draft outline includes proposal for other portions of an application

- Technical specifications

  • Soliciting feedback on process to develop limited safety systems settings, and limiting conditions for operations or restructure technical specifications to support a more technology-inclusive approach 10

Informing Content of Applications Outline (FSAR) with major LMP areas highlighted Additional Portions of Application

1. General Information*
  • Technical Specifications
  • Technical Requirements Manual
2. Site Information
  • Quality Assurance Plan (design)
3. Licensing Basis Event (LBE) Analysis*
4. Description and Classification of SSCs*
  • Fuel qualification report
5. Design Basis Accidents Analysis (10 CFR 50.34)*
  • Exemptions
  • Quality Assurance Plan (construction and operations)
6. Integrated Plant Analysis*
7. Defense in Depth (DID)*
  • Physical Security Plan
8. Control of Routine Plant Radioactive Effluents and
  • SNM material control and accounting plan Solid Waste
  • New fuel shipping plan
9. Control of Occupational Dose
10. Human Factors Analysis*
  • Radiation Protection Program
11. Physical Security
  • Inservice inspection/Inservice testing (ISI/IST) Program
12. Overview of PRA*
  • Environmental Report
  • Site Redress Plan
13. Administrative Control Programs* (special treatment)
  • Exemptions, Departures, and Variances
14. Initial Startup Programs* (special treatment) 11

Summary

  • Discussion of staffs draft outline
  • Discussion of development of technical specifications
  • Next steps

- Major focus of discussions in upcoming stakeholder meetings and/or dedicated meetings; coordination with industry-led TI-CAP

- NRC will be interacting with Canadian Nuclear Safety Commission

- Staff will revise draft outline as appropriate and provide updated draft outline in March 2020 time frame

- Planned development of a regulatory guide 12

Fundamental Safety Functions (FSFs)

Technology Inclusive Content of Application Project (TICAP)

Steve Nesbit LMNT Consulting NRC Stakeholder Meeting December 12, 2019

2 FSF Presentation

  • Background
  • TICAP Goal
  • Desired Outcome of FSF Report
  • Definition of FSFs

- Controlling Reactivity

- Removing Heat from Reactor and Waste Stores

- Limiting Releases of Radioactive Materials

  • Summary 2

TICAP Overview

Purpose:

Develop a technology inclusive process for efficiently generating safety analysis report content at the level of detail that supports the safety case and simplifies the staff review processes

- Reliant on risk-informed, performance based Licensing Modernization Project (LMP) methodology Success Criteria: Develop a guidance document that can be submitted to the Nuclear Regulatory Commission (NRC) for endorsement by September 2021 3

TICAP Overview (cont.)

  • Product will be a process for developing and organizing the content of an application as opposed to a specific set of required information
  • Current content of application requirements reflect the safety case of large light water reactors

- Active engineered safety systems with multiple trains and many components

- Required operator actions

  • Expectation that advanced reactors will have a simpler and more straightforward safety case due to features such as inherent and passive safety features, slow system response times, etc.

- No need to describe design features or programs if it is demonstrated they are outside the safety case (e.g., emergency electrical power, human factors) 4

LMP Department of Energy-cost shared, Southern Company-led project Risk-Informed, Performance-Based NEI 18-04, Risk Informed Approach to Performance Based Guidance for Non-Light Water

  • Selection of Licensing Basis Events Reactor Licensing Basis
  • Classification of Structures, Systems, Development and Components (Sept 2018, Aug 2019)
  • Defense-in-Depth adequacy determination NRC Draft Guide 1353, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Approach to Non-Light-Water Reactors (April 2019) 5

TICAP Department of Energy-cost shared, Southern Company-led project Guidance for developing and organizing the content for key elements of the NRC license application Safety Analysis Report (SAR) Development team consisting of owner-

  • Applicable to all non-light water operators, advanced reactor (LWR) designs reactor developers and
  • Builds on foundation provided by consultants LMP
  • Focus on combined construction and operating license with no design certification or early site permit Ultimate product is an NRC-endorsable NEI guidance document 6

7 TICAP Organization DOE INL Southern Company Services Technical Team Review Teams Owner- Developers operators and (GE-Hitachi, Industry Kairos, Oklo, Senior Advisors NEI Advanced Consultants Westinghouse) (Former Reactor Regulatory Commissioners) Task Force 7

8 Key TICAP Products Formulation of Fundamental Safety Technology Inclusive Functions Definition Content of Application (to NRC 11/2019)

(to NRC 10/2020)

Regulation Mapping to FSFs (to NRC 5/2020)

NEI Guidance Denotes approximate date Document Annotated provided to NRC for review Outline (to NRC 10/2020)

LMP-Related Safety Case Tabletop (To NRC 6/2020) Exercises 2020 Nov Dec Jan Feb March April May June July Aug Sept Oct 8

Key TICAP Products (cont.)

Differences Between Licensing Paths (to NRC 1/2021)

NEI Guidance Document (draft to NRC 4/2021) NRC Review (final to NRC 7/2021)

Tabletop Exercises 2021 Nov Dec Jan Feb March April May June July Aug Sept Oct 9

Desired Outcomes of Definition of FSFs Report Alignment with the NRC

  • on definitions of FSFs

- categories

- breadth of coverage

  • that providing information to demonstrate FSFs are met by a design is sufficient for the Commission to conclude that there is reasonable assurance of adequate protection of public health and safety
  • that an application which demonstrates the FSFs are assured by the design does not need to request exemptions for

- regulations that do not apply to non-light water technologies, and

- regulations for which the intent has been satisfied by alternative means 10

Definition of FSFs

  • Functions that all technologies must be capable of performing in all plant states
  • Not specific to transient or accident mitigation only
  • Design-specific Required Safety Functions in LMP are derived from FSFs
  • FSFs chosen:

- Controlling Reactivity

- Removing Heat from Reactor and Waste Stores

- Limiting Releases of Radioactive Materials 11

Definition of FSFs (cont.)

  • Generally aligns with other FSFs used by other regulators
  • Technology inclusive
  • Applicable to technology dependent set of postulated transients or accidents initiated by internal events and external hazards

Controlling Reactivity

  • Proponents for controlling heat generation as an alternative FSF, per NEI 18-04
  • Inherent, passive or active means available to:

- Control the nuclear chain reaction consistent with intended plant operating conditions

- Terminate the nuclear chain reaction when transient or accident conditions dictate that the facility must be shut down

- Prevent inadvertent criticality in the reactor core, primary system, fuel handling system, or other areas where inadvertent criticality is an adverse condition that could result in unacceptable radiological consequences 13

Removing Heat from Reactor and Waste Stores

  • Inherent, passive, or active means available to:

- Remove heat generated from the nuclear chain reaction in the nuclear fuel, primary system, or fuel handling system during planned operating modes, or following postulated transients and accidents

- Remove the decay or residual heat from the reactor or primary system when the nuclear chain reaction is terminated and the plant is shut down so that unplanned releases of radioactive materials from the plant do not occur

- Remove the residual heat from radioactive material that is stored in the fuel handling or waste handling areas so that unplanned releases of radioactive materials from the plant do not occur 14

Limiting Releases of Radioactive Materials

  • Uses inherent, passive or active means available to prevent or mitigate releases of radioactive materials from the plant
  • Ultimate objective of protecting public health and safety
  • Achieves established public exposure limits 15

Mapping FSFs to Regulations

  • Activity underway to map underlying safety objectives of the regulations to one or more FSFs
  • An accountability step in overall TICAP efforts which will demonstrate that these FSFs provide a comprehensive list of functions that must be reliably performed to meet the intent of the existing regulations 16

Summary

  • FSFs for TICAP are:

- Controlling Reactivity

- Removing Heat from Reactor and Waste Stores

- Limiting Releases of Radioactive Materials

  • FSFs are technology inclusive and will be assessed using a technology-specific set of postulated transients and accidents.
  • FSFs will be mapped to demonstrate that LMP-developed safety bases will provide as comprehensive an evaluation of initiating events and external hazards as provided by existing regulations
  • Looking forward to NRC feedback 17

Canadian Licensing Approach USNRC Stakeholders Meeting on Possible Process Improvements for Advanced Reactor Designs Chantal Morin, Senior Project Officer New Major Facilities Licensing Division December 12, 2019 Canadian Nuclear Safety Commission Rockville, Maryland

Outline

  • Background
  • Regulatory Framework and Licensing Approach
  • License Application Guides
  • Pre-Licensing Engagments
  • Objectives of Design Requirements
  • International Collaboration
  • Conclusion 2

Canadian Nuclear Safety Commission OUR MANDATE REGULATE IMPLEMENT DISSEMINATE the use of nuclear energy and Canada's international objective scientific, technical materials to protect health, safety, commitments on the peaceful and regulatory information to and security and the environment use of nuclear energy the public OVER 70 YEARS OF REGULATORY EXPERIENCE 3

Section 26 of the Nuclear Safety and Control Act 26 Subject to the regulations, no person shall, except in accordance with a licence, (a) possess, transfer, import, export, use or abandon a nuclear substance, prescribed equipment or prescribed information; (b) mine, produce, refine, convert, enrich, process, reprocess, package, transport, manage, store or dispose of a nuclear substance; (c) produce or service prescribed equipment; (d) operate a dosimetry service for the purposes of this Act; (e) prepare a site for, construct, operate, modify, decommission or abandon a nuclear facility; or (f) construct, operate, decommission or abandon a nuclear-powered vehicle or bring a nuclear-powered vehicle into Canada.

REGULATION IS ASSOCIATED WITH THE CONDUCT OF ACTIVITIES 4

Scope of CNSC regulation - activities associated with:

Uranium mines Uranium fuel Nuclear Nuclear Industrial and and mills fabrication and power substance medical processing plants processing applications Nuclear research Transportation Nuclear Import and Waste and educational of nuclear security and export management activities substances safeguards controls facilities OVER THE FULL LIFECYCLE OF THESE FACILITIES / ACTIVITIES 5

REGULATORY FRAMEWORK AND LICENSING APPROACH 6

Regulatory Framework And Licensing Approach Five Stages in the Lifecycle of a Nuclear Facility Site preparation Construction Operation Decommissioning Release from CNSC

= = = regulatory control Licence to Licence to Licence to Licence to

prepare site construct operate decommission Licence to abandon COMBINED LICENCES ARE POSSIBLE 7

Regulatory Framework And Licensing Approach Regulatory Framework Act Provides detail and clarification of the licensing and other regulatory requirements of the Canadian Nuclear Safety Commission with respect to Regulations

  • The Nuclear Safety and Control Act
  • Regulations made under the Act Licences, - General Nuclear Safety and Control Regulations Licence Conditions - Class 1 Nuclear Facilities Regulations, etc.

Handbook and Certificates

  • Licence conditions REGDOCS and standards
  • Regulatory Documents (REGDOCs)

(legally envorceable when referenced in a Licence) 8

Regulatory Framework And Licensing Approach Regulations, REGDOCs and Industry Standards

  • Regulations provide further legislative authority under the NSCA
  • Regulatory documents (REGDOCs) explain how to meet the requirements in the Act and Regulations
  • The applicant propose the standards (proven practices) to be applied in the safety case
  • Industry or international standards may be referenced in CNSC regulatory documents and licences 9

Regulatory Framework And Licensing Approach Safety Case & the Licensing Basis Safety and Control Areas (SCAs)

  • SCAs are technical areas subject to CNSC Management System Human Performance Management assessment, review, verification and used to Operating Performance report on regulatory requirements and Safety Analysis performance across Physical Design all regulated facilities and activities Fitness for Service Radiation Protection
  • Regulatory framework documents exist for Conventional Health and Safety each Safety and Control Area (SCA)

Environmental Protection Emergency Management and Fire Protection

  • Licence Application Guides articulate scope Waste Management Security and depth considerations for each SCA Safeguards and Non-Proliferation Packaging and Transport
  • The combination of all constitutes the Other safety case and licensing basis for the Informing the public conduct of the licensed activity 10 Siting and EA

Safety and Control Areas and Regulatory Framework Functional Safety and Control Area Area CNSCs Regulatory Framework

  • REGDOC-2.1.1, Management System Management system
  • REGDOC-2.1.2 Safety Culture
  • REGDOC-2.2.2, Personnel Training, version 2 Human performance
  • REGDOC-2.2.3, Personnel Certification, Volume III: Certification of Persons Working at Nuclear Power Plants management
  • REGDOC-2.2.4: Fitness for Duty: Managing Worker Fatigue Management
  • REGDOC-2.2.4: Fitness for Duty: Volume II: Managing Alcohol and Drug Use
  • REGDOC 1.1.1, Licence to Prepare Site and Site Evaluation for New Reactor Facilities (in particular section 4)
  • REGDOC-1.1.2, Licence Application Guide: Licence to Construct a Nuclear Power Plant Operating performance
  • REGDOC 2.3.1, Conduct of Licensed Activities: Construction and Commissioning Programs
  • REGDOC-1.1.3: Licence Application Guide: Licence to operate a Nuclear Power Plant
  • REGDOC-2.3.2: Accident Management
  • CNSC REGDOC-2.4.1, Deterministic Safety Analysis Safety analysis
  • CNSC REGDOC-2.4.2, Probabilistic Safety Assessment (PSA) for Nuclear Power Plants
  • REGDOC-2.4.3, Nuclear Criticality Safety Facility and
  • REGDOC-2.5.1, General Design Considerations: Human Factors Physical design equipment
  • REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants
  • REGDOC-2.6.1, Reliability Programs for Nuclear Power Plants Fitness for service
  • REGDOC-2.6.2, Maintenance Programs for Nuclear Power Plants 11

Safety and Control Areas and Regulatory Framework

  • G-91, Ascertaining and Recording Radiation Doses to Individuals
  • G-129, Keeping Radiation Exposures and Doses "As Low as Reasonably Achievable (ALARA)

Radiation protection *Important to note is that REGDOC-2.7.1, Radiation Protection, and REGDOC-2.7.2, Dosimetry, Volume I: Ascertaining Occupational Dose, are currently under development by the CNSC. Once published, these REGDOCs will supersede the documents listed above.

Conventional health

  • REGDOC-2.8.1, Conventional Health and Safety and safety Core control Environmental protection
  • REGDOC-2.9.1, Environmental Protection: Environmental Principles, Assessments, and Protection Measures, Version 1.1 processes Emergency management
  • REGDOC-2.11, Framework for Radioactive Waste Management and Decommissioning in Canada Waste management
  • CNSC, REGDOC 2.11.1, Waste Management, Volume I: Management of Radioactive Waste, Ottawa, Canada, [DRAFT]
  • REGDOC-2.11.2, Decommissioning [DRAFT]
  • REGDOC-2.12.1, High-Security Facilities, Volume I: Nuclear Response Force. Document contains prescribed information.

Security

  • REGDOC-2.12.1, High-Security Facilities, Volume II: Criteria for Nuclear Security Systems and Devices.
  • REGDOC-2.12.2, Site Access Security Clearance Safeguards and
  • REGDOC-2.13.1, Safeguards and Nuclear Material Accountancy non-proliferation
  • REGDOC-2.13.2, Import and Export, Version 2 12

Regulatory Framework And Licensing Approach Basis for a Licensing Decision Section 24(4) of the Nuclear Safety and Control Act (NSCA)

No licence may be issued, renewed, amended or replaced unless, in the opinion of the Commission, the applicant:

(a) is qualified to carry on the activity that the licence will authorize the licensee to carry on; and (b) will, in carrying on that activity, make adequate provision for the protection of the environment, the health and safety of persons and the maintenance of national security and measures required to implement international obligations to which Canada has agreed THE REACTOR DESIGN IS ONLY PART OF THE CONVERSATION CONCERNING SAFE CONDUCT OF LICENSED ACTIVITIES 13

LICENCE APPLICATION GUIDES 14

Licence Application Guides Licence Application Guides (LAGs)

New Supplemental Guidance PROVIDES EXPECTATIONS ON CONTENT OF APPLICATIONS 15

Licence Application Guides Licence to Construct (REGDOC 1.1.2)

Application guide has a strong emphasis on:

  • Responsibility of applicant for all activities pertaining to design, procurement, manufacturing construction, commissioning
  • Appropriate knowledge, skills & abilities for all persons involved design, procurement, manufacturing construction & commissioning activities
  • Systematic control of procurement and manufacturing and of services purchased by the applicant, the applicants contractors, and their suppliers and sub-suppliers
  • High degree of design completion Approach is designed to:
  • Facilitate commissioning prior to fuel load
  • Facilitate fuel-in commissioning upon receipt of Operating Licence
  • Prepare for Operational Readiness (Operators certification, etc.)

16

Licence Application Guides Licence to Operate (REGDOC 1.1.3)

Application guide has a strong emphasis on:

  • Final safety analysis and design information
  • Final policies, programs and supporting procedures to ensure safe operation of the facility covering areas such as:

Operation and maintenance of the nuclear facility Control of the release of nuclear substances and hazardous materials into the environment Readiness of emergency preparedness measures, including assistance to deal with an abnormal off-site release Facility security appropriate number of qualified staff are in-place for facility operation Readiness of the Operating Organisation and supporting programs 17

Licence Application Guides Supplemental Guidance for SMR Proponents (REGDOC 1.1.5)

  • Provide guidance on the use of existing LAGs for SMRs and Advanced Reactor Technologies based on the Safety and Control Areas
  • Provide guidance on the application of the graded approach and use of alternative approaches for development of the Licensing Basis
  • Describe the two pre-licensing engagement processes by which a vendor or an applicant can engage with CNSC prior to licencing:

- Vendor Design Review

- Process for establishing an appropriate application assessment strategy 18

PERSPECTIVES ON PRE-LICENSING ENGAGEMENT PROCESSES FOR NEW TECHNOLOGY CONCEPTS 19

Perspectives on Pre-Licensing Engagement Processes for New Technology Concepts Pre-licensing Engagement and Licensing Process Relationship Environmental Assessment Optional Pre-licensing Engagement Licensing Process No regulatory decision making Publicly transparent regulatory decision making Determining Prepare application appropriate and submissions Site application Construction Operation Decommissioning preparation assessment strategy Vendor leverages VDR

  • Under licence to
  • Under licence
  • Under licence Under licence to results in discussions prepare site to construct to operate decommission with potential applicant
  • Licenceapplication
  • Licence application
  • Licence application Vendor guide (REGDOC-1.1.1) guide (RD/GD-369 - guide (REGDOC-1.1.3) design
  • Reactor vendor(s) under revision) review
  • REGDOC-3.5.4, Pre- REGDOC-1.1.5, Supplemental Information for Small Modular Reactor Proponents licensing Review of a Vendors Reactor Design 20

Perspectives on Pre-Licensing Engagement Processes for New Technology Concepts Pre-Licensing Engagement Triggers Pre-licensing engagement is productive when considering:

  • New organizational models for conducting a project
  • New types of activities being proposed with little or no past experience

- In a demonstration facility, what are the demonstration activities that will be performed?

  • New ways to conduct activities (e.g. construction approaches)
  • New technological approaches that require extensive interpretation of requirements 21

Perspectives on Pre-Licensing Engagement Processes for New Technology Concepts Vendor Design Review:

A tool for reactor vendors

  • The process should not be triggered unless the
  • An optional process to determine whether vendors conceptual design is essentially the vendor is ready for potential deployment complete and the basic engineering program in Canada. has begun (design requirements being established).
  • A proven and standardized process to evaluate, in principle, whether there are fundamental
  • Outcomes of the process helps the vendor barriers to licensing the vendors reactor design have discussions with potential future in Canada. licensees interested in their technology.

A PRE-LICENSING VDR IS NOT A LICENSING DISCUSSION -

IT IS A TECHNICAL CONVERSATION BETWEEN THE CNSC AND THE VENDOR 22

Perspectives on Pre-Licensing Engagement Processes for New Technology Concepts Licensee is accountable and responsible for the safe conduct of the activities being licensed Applicant specifies user Reactor Vendor + its supply chain requirements + must Applicant / Licensee

  • A vendor is part of the licensees have capability to assess
  • Licensing involves an applicant for a and accept what supply licence who is proposing to build and procurement process.

chain provides operate a vendors design.

  • They supply services and products which applicants generally expect
  • Expected to be intelligent customer to be adaptable to any project Vendor reconciles
  • Usually an owner/ operator of a plant site specific design configuration to meet the applicants user needs CNSC executes licensing OPTIONAL: CNSC can provide non-binding feedback (Vendor Design Review) and compliance for How well is vendor understanding and meeting new build project Canadian expectations in their design activities and outputs?

23

Perspectives on Pre-Licensing Engagement Processes for New Technology Concepts Licence Application Assessment Strategy Enables potential applicants to understand:

  • The licensing process and their obligations as a future licensee
  • The regulatory framework tools available to support the licensing process and how they are used in the establishment of a licensing basis
  • Licensing process considerations as a result of specific activities being considered for an application NO REGULATORY DECISION MAKING BY THE COMMISSION 24

DESIGN REQUIREMENTS OF REACTOR FACILITIES 25

Design Requirements of Reactor facilities Design of Reactor Facilities REGDOC-2.5.2 Design of Reactor Facilities

  • Sets out design requirements and guidance for new water-cooled NPPs
  • Aligned with international practice (IAEA SSR-2/1)
  • Includes safety goals and objectives as well as high-level safety concepts
  • Includes safety management of the design process
  • Both general and specific structures, systems and components (SSC) design requirements
  • Allows alternative approaches (Section 11)
  • Includes Appendix A on containment structural acceptance criteria 26

Design Requirements of Reactor facilities Requirements for use of Proven Engineering Practices REGDOC 2.5.2: Design of Reactor Facilities: Nuclear Power Plants Section 5.4: Proven Engineering Practices When a new SSC design, feature or engineering practice is introduced, adequate safety shall be demonstrated by a combination of supporting research and development programs and by examination of relevant experience from similar applications.

An adequate qualification program shall be established to verify that the new design meets all applicable safety requirements.

New designs shall be tested before being brought into service and shall be monitored while in service so as to verify that the expected behaviour is achieved.

BURDEN OF DEMONSTRATION IS ON THE LICENSEE 27

Design Requirements of Reactor facilities Design General Objectives and Safety Concepts (REGDOC 2-5-2) 4.1 General Nuclear Safety Objectives 4.2 Application of the technical safety objectives 4.2.1 Dose acceptance criteria 4.1.1 Radiation protection objective 4.2.2 Safety goals 4.1.2 Technical safety objectives 4.2.3 Safety analyses 4.1.3 Environmental protection objective 4.2.4 Accident mitigation and management 4.3 Safety Concepts 4.3.1 Defence in depth 4.3.2 Physical barriers 4.3.3 Operational limits and conditions 4.3.4 Interface of safety with security and safeguards 28

INTERNATIONAL COLLABORATION 29

International Collaboration Leveraging International Experience

  • Working with IAEA and NEA on sharing best practices
  • Bilaterals with other countries
  • MOC with USNRC to increase regulatory effectiveness through collaboration

- Exchange of lessons learned and leveraging reviews and research and development data to support regulatory reviews

- Compare regulatory practices, including areas of alignment, areas of potential convergence, and differences in criteria and practices to improve effectiveness

- Share research, training, and development activities to allow more cost effective in regulatory readiness 30

CONCLUSION 31

CONCLUSIONS

  • Licensing approach is flexible with high level objectives to be met
  • Licensees can suggest alternative approaches to meet requirements
  • Vendor Design Review is a pre-licensing engagement activity and not part of the licensing process
  • CNSC is collaborating internationally to improve effectiveness 32

THANK YOU!

Connect With Us Join the conversation nuclearsafety.gc.ca

Advanced Nuclear Reactors Justify Modernized Emergency Preparedness Requirements U.S. Nuclear Industry Council, ClearPath, and Third Way November 19, 2019 Findings A viable future U.S. advanced nuclear industry needs Emergency Planning Zones (EPZs) based on the specific safety characteristics of a reactor design. EPZs for advanced reactors should be appropriately based on the new generation of advanced reactor technologies. For example, advanced reactors should be regulated like industrial facilities that have similar levels of risk. The Tennessee Valley Authority (TVA) correctly recognized the contradiction between the existing Nuclear Regulatory Commission (NRC) EPZ regulations and the improved risk profiles of advanced reactors. TVA proposed an approach in their recent Early Site Permit (ESP) Application that is informed by enhanced design features and safety margins of light-water cooled small modular reactors (SMRs). This proposal establishes an important precedent for future advanced reactor emergency planning activities.

The U.S. Nuclear Industry Council (USNIC), ClearPath, and Third Way strongly support the NRC approval of the TVA plume exposure pathway EPZ sizing methodology for both the Clinch River ESP and for future advanced non-light water cooled advanced reactors. This methodology was initially presented in the TVA Clinch River ESP Application, which was reviewed and approved by the NRC staff. The TVA approach can result in an EPZ at the site boundary or at two miles, depending on the specific safety and design characteristics of the reactor selected within the envelope of the ESP.

Background

Historically, as a part of the licensing of a new nuclear reactor, the NRC has defined an EPZ surrounding the plant. The exact size and configuration of the EPZ can vary from plant to plant due to local emergency response needs and capabilities, the population surrounding the site, the topographic characteristics, access routes in the specific area, and the jurisdictional boundaries of the region.

However, the regulation stipulates a plume exposure pathway EPZ of about 10 miles in a radius around the plant.

1

In the U.S., each commercial nuclear power reactor has both onsite and offsite emergency plans to assure that adequate protective measures can be taken to protect the public in the event of a radiological emergency. Federal oversight of emergency preparedness for nuclear power plants is shared by the NRC and Federal Emergency Management Agency (FEMA). The NRC has statutory responsibility for the radiological health and safety of the public by overseeing onsite preparedness and the overall authority for both onsite and offsite emergency preparedness.

Before a plant is licensed to operate, the NRC must have "reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency." The NRC's decision of reasonable assurance is based on licensees meeting NRC regulations and guidance which demonstrate compliance with appropriate safety requirements. In addition, licensees and area response organizations must demonstrate that they can effectively implement emergency plans and procedures during periodic evaluated exercises. As part of the Reactor Oversight Process, the NRC reviews licensees' emergency planning procedures and training. These reviews include regular drills and exercises that assist licensees in identifying areas for improvement, such as in the interface of security operations and emergency preparedness. Each plant operator is required to exercise its emergency plan with offsite authorities at least once every two years to ensure state and local officials remain proficient in implementing their emergency plans. Those biennial exercises are inspected by the NRC and evaluated by FEMA.

Licensees also self-test their emergency plans regularly by conducting drills.i Discussion The licensing of reactors depends on their facilities meeting NRCs regulations for construction and operation. However, many regulations including those for EPZs, were developed and implemented based on the currently licensed fleet of large light water reactors (LWRs) which have characteristics that differ from advanced reactors (including light-water SMRs)1. Regulations should be appropriately tailored to meet the potential safety risk of the facility, and should recognize the significant value that passive and inherent safety systems can provide. An approach that requires advanced reactors to meet standards and practices consistent with their safety characteristics and risk, such as a reduced EPZ, should apply for all advanced reactors, consistent with the approach set forth in TVAs Clinch River ESP Application, and independent of location. The NRC is in the process of a rulemaking that would allow multiple advanced nuclear reactors (including SMRs and non-LWRs) to use a similar methodology to determine appropriate EPZs. Such an approach would provide the same level of public health and safety as is provided by the current operating fleet of large LWRs.

If the NRC determines that the appropriate EPZ should be located at the reactors site boundary, no formal off-site response plans would be required because the likelihood of accidents that have a significant public health and safety impact outside of the facility are extremely low. Any such hypothetical accident would have a risk commensurate with many other industrial hazards such that they would be covered by all hazard plans that are routinely developed and used by local and regional responders to respond to events such as natural disasters, industrial accidents, and transportation 1 Small modular reactors (SMRs) are generally considered to be 300 MWe or less. Advanced reactors typically refers to non-LWRs, i.e., using different fuel, coolant, and/or moderator types. Many non-LWRs currently under development are also SMRs. In the context of EPZ sizing, SMRs and non-LWRs share similar attributes of smaller source terms, increased accident progression times, and passive/inherent safety features.

2

accidents. Importantly, licensees of advanced nuclear facilities would still be required to have appropriate onsite response plans.ii TVAs Clinch River Proposal Leads the Way Right-sizing regulations for future advanced reactor designs (including light-water SMRs), which utilize passive and inherent safety features, is essential to affordably facilitating the next generation of nuclear technologies while maintaining the safety of people living and working near advanced reactors.

Approval of the TVA ESP will be the first formal regulatory action to recognize that future advanced reactor designs should have regulations that both credit their improved safety characteristics and are representative of a facilitys overall reduced risk. The TVA ESP Application uses a plume exposure pathway EPZ methodology that is a risk-informed, dose-based, and consequence-oriented approach that is appropriate for such a technology.iii The TVA ESP Application requested: (a) approval for plume exposure pathway EPZ sizing methodology, (b) exemptions for a site boundary EPZ or a 2-mile EPZ, and (c) approval for two major features emergency plans. The TVA ESP Application does not establish the final plume exposure pathway EPZ size for the Clinch River site. The final EPZ size for the Clinch River site would be determined in a future application based on a specific reactors safety and design characteristics.

Ultimately, the NRC staff supported the plume exposure pathway EPZ sizing methodology to be used in a future combined license or construction permit application.

The approach TVA used has broader applicability than just for a reactor built at the Clinch River site.

The TVA ESP Application described the characteristics of a nuclear plant using a composite of reactor and engineering parameters based on four U.S. light water SMR designs. Other advanced reactor designs, including non-LWRs, may fit into the same set of parameters if they have the appropriate accident source terms, risks, and other similar safety characteristics. Thus, other advanced reactors also may be able to use this risk-informed methodology and justify having either a site boundary EPZ or a less than 10-mile EPZ.

The NRC issued a Final Environmental Impact Statement for the TVA ESP Application in April 2019, iv and the Final Safety Evaluation Report in June 2019 v. The NRC staff has recommended, based upon the environmental and safety reviews, issuing an ESP for the Clinch River Nuclear site in Oak Ridge, TN.

The NRC held a mandatory public hearing on August 14, 2019. Based on a statement made at the August 14, 2019 NRC public hearing, the NRC Commissioners intend to decide on an appropriate methodology for determining EPZs in general after receiving any post-hearing input. FEMA submitted a post-hearing letter detailing their concerns with the scalable emergency planning approach on August 24, 2019. The NRC staff responded point by point to the FEMA post-hearing letter on September 5, 2019 substantiating that the basis for the NRC analysis, based in part on Environmental Protection Agency (EPA) policies, was appropriate for regulating future SMRs and non-LWRs.

In order to enable viable future U.S. advanced nuclear technologies, the NRC should approve the use of EPZs that are based on the specific safety characteristics of a reactor design. The TVA plume exposure 3

pathway methodology that is supported by the NRC staff is an appropriate methodology for the NRC to use not only for the Clinch River site, but also to apply to future advanced reactor designs.

About ClearPath ClearPath's mission is to develop and advance conservative policies that accelerate clean energy innovation. To advance that mission, we develop cutting-edge policy and collaborate with academics and industry. An entrepreneurial, young, strategic nonprofit, ClearPath (501(c)(3)) partners with in-house and external experts on nuclear, carbon capture, hydropower, natural gas, energy storage and energy innovation to advance our mission. For more information visit www.clearpath.org.

About USNIC The United States Nuclear Industry Council (USNIC) is the leading U.S. business consortium advocate for new nuclear and promotion of the American supply chain globally. Composed of over 80 companies USNIC represents the "Who's Who" of the nuclear supply chain community, including key utility movers, technology developers, construction engineers, manufacturers and service providers. USNIC encompasses eight working groups and select task forces including an Advanced Reactors Task Force. For more information visit www.usnic.org.

About ThirdWay Modern problems require fresh thinking. Our work is grounded in the mainstream American values of opportunity, freedom, and security.

But we identify as center-left, because we see that space in U.S. politics as offering the only real path for advancing those ideals in the century ahead.

Our agenda is ambitious, aspirational, and actionable. It is built on the bedrock belief that for political movements to succeed in our political system, they must relentlessly re-imagine their policies, strategies, and coalitions.

We are fighting for opportunity, so everyone has the chance to earn a good life; progress on social issues, so all have the freedom to live the lives they choose; and security, so we are protected from 21st century global threats. For more information visit www.thirdway.org.

Contact Information:

Nicholas McMurray Cyril Draffin Ryan Fitzpatrick Nuclear Program Director Senior Fellow, Advanced Nuclear Director of the Clean Energy Program ClearPath United States Nuclear Industry Council Third Way Phone: 216.407.2387 Phone: 301.233.4643 Phone: 202.384.1734 Email: mcmurray@clearpath.org Email: Cyril.Draffin@usnic.org Email: rfitzpatrick@thirdway.org i https://www.nrc.gov/reading-rm/doc-collections/fact-sheets/emerg-plan-prep-nuc-power.html#targetText=For%20planning%20purposes%2C%20the%20NRC,access%20routes%2C%20and%20jurisdictional%20boundaries ii https://www.nrc.gov/about-nrc/emerg-preparedness/respond-to-emerg/nuclear-facility-response.html iii https://www.nrc.gov/reactors/new-reactors/esp/clinch-river.html iv https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr2226/

v https://www.nrc.gov/reading-rm/doc-collections/news/2019/19-026.pdf 4

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