ML20127H907

May 7, 2020 Advanced Reactor Stakeholder Public Meeting - Combined Presentation

ML20127H907
Person / Time
Issue date:	05/07/2020
From:	Jordan Hoellman NRC/NRR/DANU/UARP
To:
Hoellman J, NRR/DANU/UARP, 301-415-5481
References
Download: ML20127H907 (97)

Text

Advanced Reactor Stakeholder Public Meeting May 7, 2020 Telephone Bridgeline: (888) 390-0788 Passcode: 4560771#

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Time Agenda Speaker 10:00 - 10:10 am Opening Remarks NRC 10:10 - 10:25 am Promoting Preapplication Participation B. Beasley, NRC Introduction to Annual Fee Regulations for Non-Light-Water 10:25 - 10:40 am A. Cubbage, NRC Reactors Update on Status of NRC draft Interim Staff Guidance (ISG) 10:40 - 10:50 am for Micro Reactors and Advanced Nuclear Reactor Generic M. Sutton, NRC Environmental Impact Statement (GEIS)

NEI Feedback Regarding PNNL Reports on Approach to 10:50 - 11:00 am Determine the Environmental Data for Table S-3 of 10 CFR K. Austgen, NEI 51.51 and Table S-4 of 10 CFR 51.52 for Non-LWRs Advanced Reactor Fuel Qualification Guidance - Outline and 11:00 - 12:00 pm T. Drzeweicki, NRC Evaluation Criteria 12:00 - 12:30 pm Break All 12:30 - 1:00 pm Overview of Proposed Rulemaking on Spent Fuel Reprocessing Y. Faraz, NRC 1:00 - 1:30 pm Discussion of Category II Fuel Cycle Facility Security T. Harris, NRC Discussion of Review and Potential Endorsement of ASME 1:30 - 1:45 pm Section XI Division 2 (a.k.a., Reliability Integrity Management T. Lupold, NRC or RIM)

Advanced Reactor Inspection and Oversight Contract M. Khan and J.

1:45 - 2:00 pm Development Sebrosky, NRC 2:00 - 2:15 pm Closing Remarks and Future 2 of 97 Meeting Planning NRC/All

Promoting Preapplication Participation Ben Beasley, Chief Advanced Reactor Licensing Branch 3 of 97

• Pre-application interaction:

- White paper, audit No SER

- Topical report, Preliminary Safety Write SER Information Document

• Value

- Reliable regulatory findings early

- More efficient permit or license review

- More visibility for public on key topics 4 of 97 2

Current Generic Schedules Time to issue final safety evaluation LWR Non-LWR DC 42 months 36 months COL referencing DC 30 months 30 months Custom COL 42 months 36 months CP 36 months 36 months OL 42 months 36 months 5 of 97 3

What and Why?

• Add definition

• Specify key activities

• Promote use

• Offer clear strategies

• Caveats

• No substantive design changes

• Timely RAI responses 6 of 97 4

Key Interactions - Topical Reports

• Principle design criteria

• Classification of SSCs

• Fuel qualification

• Source term development

• QA Program

• Safeguards Information Plan

• Accident analysis method 7 of 97 5

Key Interactions - Papers and Meetings

• Overview of environmental (NEPA)

• Novel design features or preparations approaches

• Meet with other agencies on

• Consensus codes and standards endangered species and cultural

• Engineering computer codes resources affected

• Readiness assessment

• PRA

• Regulatory exemptions

• Policy issues 8 of 97 6

Strategies What would be meaningful?

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References

• Policy Statement on the Regulation of Advanced Reactors (73 FR 60612; October 14, 2008)

• A Regulatory Review Roadmap for Non-Light Water Reactors, ML17312B567

• Generic schedules are found at https://www.nrc.gov/about-nrc/generic-schedules.html 10 of 97

Annual Fees for Non-LWRs May 7, 2020 Amy Cubbage, NRR Jo Jacobs, OCFO 11 of 97

NRC Fee Requirements

• Hourly Fees under 10 CFR Part 170

- Fee for services

- Billed as hours expended times NRC professional hourly rate

- Billed to pre-applicants, applicants, and licensees

• Annual Fees under 10 CFR Part 171

- Collect approximately 90% of budget authority by end of fiscal year (FY)

- Excludes certain activities such as advanced reactor regulatory infrastructure, nuclear waste fund, etc.

- Recover through annual fees

• Research costs

• Rulemaking costs

• Other agency costs not recovered under IOAA

- Billed to licensees only 12 of 97

Annual Fees for SMRs

• Variable annual fee structure was established for light-water SMRs in June 2016 (81 FR 45963).

• The SMR fee structure has three parts:

- a minimum fee for bundled units less than 250 MWt

- a variable fee for bundled units between 250 and 2,000 MWt

- a maximum fee equivalent to the flat annual fee charged to current operating fleet reactors for bundles units between 2,000 and 4,500 MWt 13 of 97

Considerations for Non-LWRs

• Status Quo?

• Variable annual fee similar to current SMR fee rule?

• Evaluate new annual fee policy for all non-LWRs?

• Evaluate micro reactors separately?

• Other considerations?

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Next Steps

• Discuss in more detail in future stakeholder meetings

- Industry proposals?

- Address in FY2022 fee rule?

- Conduct separate rulemaking?

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Advanced Reactor Preparations for Environmental Reviews Mallecia Sutton Senior Project Manager Division of Advanced Reactors and Non-Power Production and Utilization Facilities 16 of 97

Status on Environmental Activities

• Status update on:

Interim Staff Guidance (ISG) for the environmental review of micro-reactors Generic Impact Statement (GEIS) for Advanced Reactors 17 of 97 2

Interim Staff Guidance

• ISG issued:

Public Comment on February 21, 2020 Comment period ends May 11, 2020 Comment on regulations.gov at https://www.regulations.gov/docket?D=NRC-2020-0051 18 of 97 3

GEIS for Advanced Reactors

• GEIS Federal Register Notice Issued-April 30, 2020 Scoping Meeting-5/28/2020 Scoping Period ends-6/30/2020

• Scoping Meeting May 28,2020 1-4pm Webinar Meeting Notice ADAMS Accession Number: ML20122A049

• Additional Information on the GEIS advanced reactor web page https://www.regulations.gov/document?D=NRC-2020-0101-0001 19 of 97 4

Questions 20 of 97 5

NEI Feedback on draft non-LWR Fuel Cycle Environmental PNNL reports May 7, 2020 21 of 97

©2020 Nuclear Energy Institute

Fuel Cycle Environmental Topics NRC should not assume that a different fuel form results in a substantially different environmental impact

• Staff should conduct an evaluation to demonstrate non-LWR fuels are adequately characterized by analysis of LWR fuel Should be addressed generically

• Update, or parallel, Tables S-3 and S-4

• If needed, GEIS on mining, milling, and enrichment Provide clarity on appropriate level of detail requested of applicants 22 of 97

©2020 Nuclear Energy Institute 2

Fuel Qualification (FQ) for Advanced Reactors Advanced Reactor Stakeholder Meeting May 7, 2020 23 of 97 1

Outline Background/Motivation Activity affecting FQ guidance FQ report/FQ assessment framework Next steps/stakeholder input 24 of 97 2

NEIMA SEC. 103. ADVANCED NUCLEAR REACTOR PROGRAM (c) REPORT TO INCREASE THE USE OF RISK-INFORMED AND PERFORMANCE-BASED EVALUATION TECHNIQUES AND REGULATORY GUIDACNE 25 of 97 3

Regulatory Aspects of Nuclear Fuel Qualification Regulations No requirements specific to nuclear fuel qualification Requirements on fuel qualification are provided by top level requirements attributed to the facility 10 CFR 50.43(e)

GDC/ARDC 2, Design bases for protection against natural phenomena GDC/ARDC 10, Reactor design 26 of 97 4

Regulatory Aspects of Nuclear Fuel Qualification

• Guidance

- NUREG-0800, Standard Review Plan

• Section 4.2, Fuel System Design

- Identifies acceptance criteria derived from know fuel failure/degradation mechanisms for light water reactor fuel

- ATF-ISG-2020-01

• Significant changes to fuel design must be assessed for potentially new failure/degradation mechanisms 27 of 97 5

Outline Background/Motivation Activity affecting FQ guidance FQ report/FQ assessment framework Next steps/stakeholder input 28 of 97 6

FQ Activity NRC reviewed the Electric Power Research Institute (EPRI) TRISO fuel qualification report (ACRS subcommittee meeting on May 6, 2020 - yesterday)

NRC reviewed and approved the quality assurance program for legacy metallic fuel data, ANL/NE-16/17 (i.e.,

EBR-II data)

MSR fuel qualification work with Oak Ridge National Lab Accelerated fuel qualification reports (General Atomics and TerraPower)

NEA - Working Group on the Safety of Advanced Reactors (WGSAR)

Fuel Qualification Report (Draft)

Work going on in parallel to with NRC effort to address NEIMA requirement 29 of 97 7

Outline Background/Motivation Activity affecting FQ guidance FQ report/FQ assessment framework Next steps/stakeholder input 30 of 97 8

FQ Framework - Literature JNM 2007 Paper Accelerated FQ Paper (Unpublished) 31 of 97 9

FQ Framework - Scope Broad interpretation of fuel qualification (many aspects of nuclear safety are impacted by the fuel)

Neutronic performance Thermal-fluid performance (e.g., margin to critical heat flux limits)

Seismic behavior Fuel transportation and storage Need to restrict the scope of the report The scope of this report focuses on the identification and evaluation of safety relevant phenomena for fuel performance including the understanding of fuel life limiting failure and degradation mechanisms which occur as a result of irradiation during reactor operation.

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FQ Framework - Other Considerations Definition of fuel qualification (from JNM 2007)

The objective of nuclear fuel qualification is the demonstration that a fuel product fabricated in accordance with a specification behaves as assumed or described in the applicable licensing safety case, and with the reliability necessary for economic operation of the reactor plant Clarify safety case The role of nuclear fuel in the safety case can vary significantly between different reactor designs (e.g. TRISO fuel contains fission product barriers within the fuel itself) 33 of 97 11

FQ Framework Development of a generic assessment framework for fuel qualification:

Top-down approach used to decompose the top level goal of fuel is qualified into lower level supporting goals Lower level supporting goals are further decomposed until clear objective goals are identified that can be satisfied with direct evidence 34 of 97 12

FQ Assessment Framework:

Intro and Nomenclature

• This is a top down approach that attempts identify specific goals that can be directly supported by evidence

• A high level or abstract goal is given by an empty rectangle

• A concrete goal that is broken down no further is given by a shaded rectangle

• A goal that leads to the use of a separate framework. The framework will be identified directly under the goal.

• Clarifying notes are provided in rounded rectangles 35 of 97 13

FQ Assessment Framework: Goal Goal: Fuel is qualified for use

= High confidence exists that the fuel fabricated in accordance its specification will perform as described in the applicable licensing safety case Goal: Fuel is qualified for use A fuel manufacturing specification controls the key Safety criteria can be fabrication parameters that satisfied with high significantly affect fuel confidence [G2]

performance [G1]

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G2: Safety Criteria Safety criteria can be satisfied with high confidence [G2]

Margin to design limits can be Margin to radionuclide Ability to achieve and demonstrated for normal and release limits under accident maintain safe shutdown can off normal conditions with conditions can be be assured [G2.3]

high confidence [G2.1] demonstrated with high confidence [G2.2]

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G2.1: Design Limits for Normal and Off-Normal Operation Margin to design limits can be demonstrated for normal operations and off-normal conditions with high confidence [G2.1]

An evaluation model is available to assess fuel performance The fuel performance envelope against design limits to protect is defined [G2.1.1] against fuel failure and degradation (i.e., life-limiting) mechanisms [G2.1.2]

Note: The fuel performance envelope specifies the environmental conditions and radiation exposure that the fuel is expected to encounter. The envelope is typically specified by fuel designers and provides constraints on the design of the reactor and associated systems.

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G2.2: Radionuclide Release Limits Margin to radionuclide release limits under accident conditions can be demonstrated with high confidence [G2.2]

Criteria for barrier degradation Radionuclide retention and failure under accident Radionuclide retention and requirements of the fuel conditions (e.g., PCMI and high release behavior of the fuel under accident conditions is enthalpy cladding failure, matrix under accident specified [G2.2.1] temperature induced reactions conditions is modeled and phase transformations) is conservatively [G2.2.3]

supported by quality experimental data [G2.2.2]

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G2.2.2: Criteria for Barrier Degradation Criteria for barrier degradation and failure under accident conditions (e.g., PCMI and high enthalpy cladding failure, temperature induced reactions and phase transformations) is supported by quality experimental data [G2.2.2]

Criteria are shown to provide conservative Experimental data is prediction of barrier appropriate degradation and failure [G2.2.2(b)]

[G2.2.2(a)]

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G2.2.3: Conservative Modeling Radionuclide retention and release behavior of the fuel matrix under accident conditions is modeled conservatively [G2.2.3]

Radionuclide transport model is shown to provide conservative prediction of Experimental data is radionuclide retention and appropriate release behavior of fuel [G2.2.3(b)]

matrix

[G2.2.3(a)]

Note: Testing at environmental conditions consistent with accident conditions is expected (e.g., elevated fuel temperatures) 41 of 97 19

G2.3: Safe Shutdown Ability to achieve and maintain safe shutdown can be assured

[G2.3]

Maintaining coolable geometry Control element insertion can under accident conditions can be demonstrated with high be demonstrated with high confidence [G2.3.2]

confidence [G2.3.1]

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G2.3.1: Maintaining Coolable Geometry Maintaining coolable geometry under accident conditions can be demonstrated with high confidence [G2.3.1]

Criteria for maintaining coolable Criteria to ensure coolable Criteria are shown to provide geometry under accident geometry are identified (e.g. fuel conservative prediction of conditions (e.g., fuel dispersal) melt, fuel fragmentation, fuel coolable geometry loss are supported by quality ballooning) [G2.3.1(a)] mechanisms [G2.3.2(b)]

experimental data [G2.3.3(c)]

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G2.3.2: Control Element Insertion Control element insertion can be demonstrated with high confidence [G2.3.2]

Criteria are provided that An evaluation model is ensure that the control available to assess geometry element insertion path is not changes as a result of normal obstructed during conditions of operation and accident normal operation or accident conditions [G2.3.2(b)]

conditions [G2.3.2(a)]

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Evaluation Model (EM)

Assessment Framework Goal: The evaluation model is acceptable The evaluation model contains The evaluation model has the appropriate modeling been adequately assessed capabilities against experimental data

[EM G1] [EM G2]

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EM G1: EM Capabilities The evaluation model contains the appropriate modeling capabilities

[EM G1]

The evaluation model is The evaluation model is The evaluation model is capable of modeling the capable of modeling the capable of modeling material of the material of the fuel and the geometry of the necessary physics for associated environment fuel [EM G1.1] fuel performance

[EM G1.2]

[EM G1.3]

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EM G2: EM Assessment The evaluation model has been adequately assessed against experimental data

[EM G2]

The evaluation model has demonstrated the ability to Experimental data used for predict fuel failure and assessment is appropriate degradation mechanisms over

[EM G2.1]

the test envelope

[EM G2.2]

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EM G2.2: Demonstrated Ability over Test Envelope The evaluation model has demonstrated the ability to predict fuel failure and degradation mechanisms over the test envelope

[EM G2.2]

Evaluation model error Evaluation model The evaluation model is quantified through error is determined Sparse data regions is restricted to use assessment against throughout the fuel are justified within its test experimental data performance envelope [EM G2.2.3] envelope

[EM G2.2.1] [EM G2.2.2] [EM G2.2.4]

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Experimental Data (ED)

Assessment Framework Goal: Experimental data used for assessment is appropriate Validation data is Data has been collected Experimental data Test specimens are independent of data over a test envelope have been representative of used to develop/train that covers the fuel accurately measured prototypical fuel the evaluation model performance envelope

[ED G3] [ED G4]

[ED G1] [ED G2]

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ED G3: Data Measurement Experimental data have been accurately measured

[ED G3]

Experimental data accounts for The test facility has an Experimental data is sources of experimental appropriate quality collected using established uncertainty, including assurance program measurements techniques instrumentation uncertainty

[ED G3.1] [ED G3.2]

[ED G3.3]

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ED G4: Test Specimens Test specimens are representative of prototypical fuel

[ED G4]

Test specimens are fabricated Distortions are justified and consistent with the prototypical accounted for in the fuel manufacturing specification experimental data

[ED G4.1] [ED G4.2]

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Outline Background/Motivation Activity affecting FQ guidance FQ report/FQ assessment framework Next steps/stakeholder input 52 of 97 30

Current Status

• An Assessment Framework has been developed

- Based on current review guidance, literature, WGSAR member input, and additional NRC input

- Multiple iterations with NRC working group

- Beginning to share with advanced reactor stakeholders 53 of 97 31

Next Steps/Stakeholder Input

• Need further writing to provide:

- Supporting/clarifying language

- Standards for evidence with clarifying examples

• Completed draft expected August 2020

• To be placed on stakeholder meeting agenda one month after draft release (planning on September stakeholder meeting) 54 of 97 32

Break Meeting/Webinar will begin shortly Telephone Bridgeline: (888) 390-0788 Passcode: 4560771#

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Status of Spent Fuel Reprocessing Rulemaking Periodic Advanced Reactor Stakeholders Meeting May 7, 2020 56 of 97 1

Historical Perspective

• 2006 Global Nuclear Energy Partnership (GNEP) launched which included

- establishing domestic reprocessing and burner reactor capability

- take-back of spent fuel from foreign countries

• 2007/2008 Congress reduced GNEP funding

• 2009 domestic aspect of GNEP ended, but DOE moved ahead with reprocessing R&D and industry remained interested in domestic reprocessing 57 of 97 2

Historical Perspective (cont.)

• 2008 - 2013 NRC received letters from four companies supporting update of reprocessing regulatory framework

• 2009 NRC conducted a gap analysis of the reprocessing regulation (reprocessing regs to be in Part 70)

- 23 gaps identified (14 high priority, 5 medium priority)

• 2011 Draft Regulatory Basis

- 19 high and medium priority gaps addressed

- Part 7x recommended for reprocessing 58 of 97 3

Historical Perspective (cont.)

• 2013 (Aug) staff provided the Commission its resource estimate for completing Part 7x rulemaking activities

• 2013 (Nov) Commission directed staff to complete the regulatory basis for Gap 5 only

• 2016 (Oct) NRC suspended work on Part 7x rulemaking

- budgetary constraints

- apparent lack of industry interest in constructing and operating a commercial spent fuel reprocessing facility in the United States 59 of 97 4

Recent Public Engagement

• NRC held a public meeting on March 4, 2020, to seek stakeholder input on rulemaking

• Posed two discussion questions to participants:

- Should the NRC discontinue Part 7x rulemaking?

- What is the intention of industry with regard to the construction, licensing and operation of spent fuel reprocessing facilities?

• Several organizations and many public citizens opposed reprocessing on safety and environmental grounds

• Some industry representatives voiced support for continuing the rulemaking indicating that having a better framework of regulations would encourage companies to engage in reprocessing of spent fuel

• NuScale commented that potential customers in foreign nations have expressed interest in a U.S. fuel take-back option that involves reprocessing 60 of 97 5

Current Status

• The Commission is expecting a final technical basis on the Gap 5 resolution and proposed path forward on Part 7x rulemaking in early 2021

• Pending NEIs letter to the NRC anticipated around the end of May 2020 regarding its position on reprocessing, the staff intends to inform the Commission in a COMSECY of its recommendation regarding reprocessing rulemaking 61 of 97 6

Background Slides 62 of 97 7

References

• SECY-06-0066, Regulatory and Resource Implications of a Department of Energy Spent Nuclear Fuel Recycling Program, dated March 22, 2006 (ADAMS Accession No. ML060370037).

• SECY-09-0082, Update on Reprocessing Regulatory Framework -

Summary of Gap Analysis, dated May 28, 2009 (ADAMS Accession No. ML091520243).

• SECY-11-0163, Reprocessing Rulemaking: Draft Regulatory Basis and Path Forward, dated November 18, 2011 (ADAMS Accession No. ML113202350).

• SRM-SECY-11-0163, Reprocessing Rulemaking: Draft Regulatory Basis and Path Forward, dated August 30, 2012 (ADAMS Accession No. ML122430189).

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References (cont.)

• SECY-13-0093, Reprocessing Regulatory Framework - Status and Next Steps, dated August 30, 2013 (ADAMS Accession No. ML13178A243).

• SRM-SECY-13-0093, Reprocessing Regulatory Framework - Status and Next Steps, dated November 4, 2013 (ADAMS Accession No. ML13308A403).

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Gap Summaries

• Gap 1 - Licensing under Part 50 regulations could require many exemptions since these have always focused, for the most part, on reactors

• Gap 2 - The current Part 72 regulations do not provide for interim, commercial independent storage of solidified HLW from reprocessing facilities

• Gap 3 - The NRC lacks regulations defining waste incidental to reprocessing, since not all waste such as HLW tank residues, chopped and leached fuel hulls, irradiated fuel hardware, and reprocessing facility equipment is HLW. This would result in regulatory uncertainty for an applicant with regard to differentiating HLW from incidental wastes produced at its facility.

• Gap 4 - 10 CFR 74.51, currently excludes irradiated fuel reprocessing facilities from Category 1 MC&A requirements

• Gap 5 - Part 50 does not require risk assessment for reprocessing nor is there any associated guidance for conducting risk assessments for reprocessing such as an enhanced ISA or a PRA 65 of 97 10

Gap Summaries (cont.)

• Gap 6 - The current regulations do not define terms such as reprocessing, recycling or vitrification

• Gap 7 - Part 55 does not require operator licensing for reprocessing facilities

• Gap 8 - The security categorization schemes in Part 73 and 74 may place an undue burden on licensees for portions of their reprocessing facilities

• Gap 9 - Part 50 does not contain any General Design Criteria for reprocessing facilities. The NRC staff identified 78 potential GDCs for reprocessing in its draft regulatory basis document.

• Gap 10 - Part 50 does not allow one-step licensing for reprocessing facilities

• Gap 11 - Part 50 does not contain criteria for identifying technical specifications for reprocessing facilities as it does for reactors 66 of 97 11

Gap Summaries (cont.)

• Gap 12 - Price Anderson protection and indemnity fees and amounts for reprocessing facilities are currently not included in Part 140

• Gap 13 - The scope of Part 170 does not include reprocessing outside of Part 50

• Gap 14 - Part 171 does not address annual fees for a reprocessing facility

• Gap 15 - Potential long-term storage of HLW at a reprocessing facility will need to be addressed

• Gap 16 - The tables in 10 CFR 61.55, Waste Classification, do not include all reprocessing-related radionuclides. As a result, some waste streams may be considered Class A but may not be generally acceptable for near surface disposal.

• Gap 17 - There are no existing regulations for a diversion path analysis requirement for a reprocessing facility under Part 74 67 of 97 12

Gap Summaries (cont.)

• Gap 18 - 10 CFR 74 does not appropriately address material accounting timeliness and goal quantities for a reprocessing facility

• Gap 19 - Part 70 does not adequately address effluent controls and monitoring for reprocessing facilities

• Gap 20 - Existing regulations do not address security risks for certain fissile material other than uranium and plutonium

• Gap 21 - Tables S-3 and S-4 of Part 51 do not address a closed fuel cycle involving reprocessing

• Gap 22 - Part 70 does not adequately address 1-step vs 2-step licensing

• Gap 23 - Part 110 Appendix I Illustrative List of Reprocessing Plant Components under NRC Export Licensing Authority, does not include equipment related to pyroprocessing or vitrification 68 of 97 13

Category II Fuel Cycle Facility Security Tim Harris, Senior Program Manager Materials Security Branch Division of Physical and Cyber Security Policy Office of Nuclear Security and Incident Response 69 of 97

Topics

• Current NRC Approach

• Pre-application Discussions

• Supplemental Security Measures 70 of 97 2

Current Approach

• Use a risk-informed analysis on a case-by-case basis

• Use site-specific license conditions

• Ensure that requirements are fairly and reasonably applied

• Continue to interface with the interagency community 71 of 97 3

Pre-application Discussions

• Applicant describes

- Facility setting

- Facility processes

- Types of materials (physical/chemical forms, enrichment, quantity)

- Facility Layout

- Material flow (transportation, storage, use) 72 of 97 4

Regulatory Discussions

• Applicable Regulatory Requirements

• Available Guidance

• Available Reference material

• Information protection 73 of 97 5

10 CFR 73.67(d) - Fixed Site

• Use the material only within a controlled access area

• Store the material within a vault-type room

• Monitor controlled access area with an intrusion alarm

• Conduct screening of individuals with unescorted access

• Develop and maintain a controlled badging and lock system

• Establish a security organization of at least one watchman per shift able to assess and respond

• Provide a communication capability between the security organization and appropriate response force

• Search on a random basis vehicles and packages leaving the controlled access areas

• Establish and maintain written response procedures 74 of 97 6

Supplemental Measures

• Discussions will be iterative and interactive

• Potential supplemental security measures will be site-specific

• Security could be zoned/partitioned

• Security can be achieved in multiple ways that balance the need to detect, assess, and delay potential adversaries and effectively respond to potential threats

• Applied fairly and reasonably 75 of 97 7

Questions 76 of 97 8

Additional Slides 77 of 97 9

Existing Requirements - In Transit

• Provide advance notification

• Check the integrity of the container and locks or seals prior to shipment and upon receipt of the shipment

• Notify the shipper of receipt of the material

• Arrange for telephone or radio communications between the transport and the licensee

• Minimize the time that the material is in transit

• Conduct screening of all licensee employees involved in the transportation

• Establish and maintain written response procedures

• Initiate immediately a trace investigation of any shipment that is determined to be lost 78 of 97 10

Potential Supplemental Measures - Fixed Sites

• Better defined access controls (background checks)

• Random entry searches

• Greater control over material during use

• Alarm station

• Maintenance program 79 of 97 11

Potential Supplemental Measures - Fixed Sites

• For site with larger quantities, the following may also apply

- Protected area

- Armed guards

- Expanded intrusion and detection 80 of 97 12

Potential Supplemental Measures - In Transit

• Transfers occur in controlled access area

• Increased key control

• Transport in closed and locked conveyance

• Increased searches

• Increased custody verification 81 of 97 13

In-Service Inspection Programs for Advanced Reactors

• The American Society of Mechanical Engineers, Boiler & Pressure Vessel Code,Section XI, Division 2 has developed a probabilistic risk based approach for establishing inspection and monitoring activities for advanced reactors.

- ASME Code,Section XI, Division 2

• Reliability and Integrity Management (RIM)

• ASME has requested NRC review and endorse the code in 10 CFR 50.55a 82 of 97 1

RIM Process Overview

• Step 1 Determine Scope of SSCs for RIM Program

• Step 2 Evaluate SSC Damage Mechanisms

• Step 3 Determine Plant and SSC Level Reliability and Capability Requirements

• Step 4 Identify and Evaluate RIM Strategies to Achieve Reliability Targets

• Step 5 Evaluate Uncertainties in Reliability Performance

• Step 6 Implement RIM Program

• Step 7 Monitor SSC Reliability Performance and Update RIM Program 83 of 97 2

Interest in RIM Use

• The NRC is considering the ASME request.

NRC management wants to understand the interest from potential vendors prior to expending resources and initiating a review.

• If there is interest in using RIM in a future application submittal, please contact one of the individuals below, preferably by May 22, 2020.

- Tim Lupold: timothy.Lupold@nrc.gov

- Bruce Lin: Bruce.Lin@nrc.gov 84 of 97 3

Development of a New Inspection and Oversight Framework Document to Support Construction and Operation of Advanced Reactors NRR/DANU/UARP - Maryam Khan, Joe Sebrosky (PM)

May 7, 2020 85 of 97

• Purpose

- Brief Stakeholders on NRC staffs intention to place a contract in Fiscal Year 2021 to develop a new inspection and oversight framework document to support construction and operation of advanced reactors

• Outcome

- Stakeholders have an understanding of the reason for a new framework and near-term path forward (issue contract) 86 of 97 2

Agenda

• Background

• Advanced Reactors within the Scope of the Work

• Examples of Issues to be Considered Under Contract

• Time Fame

• Questions

• Wrapup 87 of 97 3

Background - Implementation Action Plans Strategy 1 Strategy 2 Strategy 3 Strategy 4 Strategy 5 Strategy 6 Knowledge, Skills Computer Codes Flexible Review Consensus Codes Policy and Key Communication and Capability & Review Tools Processes and Standards Technical Issues Identification & ASME BPVC Consequence ONRL Molten Salt Regulatory NRC DOE Assessment of Section III Based Security Reactor Training Roadmap Workshops Available Codes Division 5 (SECY-18-0076)

ASME/ANS Non- EP for SMRs Periodic Code Non-LWR Design LWR HTGR Training and ONTs Stakeholder Development/V&V Criteria PRA Standard (SECY-18-0103) Meetings Fast Reactor Environmental ANS Standards Functional NRC DOE DOD Training Review Micro 20.1, 20.2 Containment Micro Reactor Reactor ISG 30.2, 54.1 (SECY-18-0096) MOU Licensing Siting near NRC/DOE GAIN, Competency Modernization densely populated VTR & NEICA Modeling Project (LMP) areas MOUs Technology Memorandum of Knowledge Environmental Inclusive Content Cooperation Management GEIS Applications with CNSC Project (TICAP)

Insurance and New Rulemaking Liability WGSAR 10 CFR Part 53 Micro Reactor Inspection and Policy issues Oversight 88 of 97 4

Background - Licensing Modernization Project

• Described in SECY-19-0117

- Technology inclusive, risk informed performance based methodology

- Major elements of the approach are:

• identifying licensing basis events (LBEs);

• classifying structures, systems, and components (SSCs);

• and assessing the adequacy of defense in depth (DID).

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Background - Advanced Reactor Designs Liquid-Metal-Cooled High-Temperature Molten Salt Reactors Micro-Fast Reactors (LMFR) Gas-Cooled Reactors (MSR) Reactors (HTGR)

GEH PRISM (VTR) X-energy Kairos Westinghouse TerraPower Framatome Liquid Salt Cooled Others ARC StarCore Transportable Sodium Cooled TRISO Fuel Terrestrial Oklo Westinghouse TerraPower Others General Atomics Columbia Basin Elysium Stationary Hydromine Thorcon Lead Cooled Muons Flibe Alpha Tech Liquid Salt Fueled 90 of 97 6

Background - Results of Advanced Reactor LMP Table Top Exercises

• Several Table Top Exercises Performed

- Limited number of safety-related SSCs identified

- Some Safety-related SSCs that have been identified do not have a nexus to SSCs found in large light water reactors

- Non safety related SSCs with special treatment an outcome of the LMP approach 91 of 97 7

Background

• Challenges with Current Inspection and Oversight Programs (examples)

- Micoreactor

• Could have limited technical specifications and inspections tests analysis and acceptance criteria (ITAAC)

• Oversight could be more like a Research and Test Reactor than a large light water reactor

- Molten salt fueled reactors

• There is not a traditional containment

• Fission product inventory inside reactor and outside reactor in waste tanks

• Complex waste tanks to ensure decay heat is removed and criticality is 8 prevented 92 of 97

Advanced Reactors within the Scope of the Work

• Non light water reactors

• Small modular reactors (i.e., less than 300 MWe)

• Fusion Reactors 93 of 97 9

Examples of Issues to be Considered Under Contract

• Covers both construction inspection and oversight and operating plant inspection and oversight

• Prioritize development of microreactor guidance first

• Use of risk insights and concepts from LMP process

• Consideration of advanced reactor construction techniques including reactors being assembled in a factory and shipped to the site

• Flexible such that it can be used under Part 50 or Part 52 process

• Includes development of risk-informed performance indicators

• Includes consideration of virtual inspections.

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Timeframe

• Spring 2020 - solicit request for proposals from commercial contractors interested in the work

• Summer 2020 - finalize statement of work, subject to the availability of funds

• October 1, 2020 - work begins

• Envision stakeholder interactions starting in Calendar Year 2021

• December 2021 - final version of framework document provided to NRC

• Inspection procedures and manual chapters (as appropriate) to be developed longer term based on concepts identified in framework document 95 of 97 11

Questions 96 of 97 12

Future Meeting Planning and Open Discussion 2020 Tentative Schedule for Periodic Stakeholder Meetings June 18 August 6 September 24 November 5 97 of 97