ML21312A055

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Nov. 10 2021 Periodic Advanced Reactor Stakeholder Meeting Slides
ML21312A055
Person / Time
Issue date: 11/10/2021
From: Prosanta Chowdhury
NRC/NRR/DANU/UARP
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Chowdhury P
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Download: ML21312A055 (120)
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Advanced Reactor Stakeholder Public Meeting November 10, 2021 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 272 279 12#

Slide 1

Time Agenda Speaker 10:00 - 10:15 am Opening Remarks NRC 10:15 - 11:15 am The Technology-Inclusive, Risk-Informed Maximum Accident (TI-RI-MA) Approach

- An Alternative to Probabilistic Risk Assessment NRC 11:15 - 11:45 am An Overview of NRCs Regulatory Requirements and Guidance on Counterfeit, Fraudulent, and Suspect Items NRC 11:45 am - 12:15 pm Advanced Reactor Content of Application Revised Chapter 11, "Organization and Human-System Considerations and Chapter 12, Post-Construction Inspection, Testing and Analysis Program NRC 12:15 - 1:00 pm Lunch Break All 1:00 - 1:45 pm Accelerated Fuel Qualification (AFQ) White Paper General Atomics 1:45 - 2:45 pm Update on the Development of a Flexible Operator and Staffing Licensing Framework for Advanced Reactors NRC 2:45 - 2:55 pm Break All 2:55 - 3:25 pm FAQ for Physical Security Cat. II Fuel Cycle Facilities NRC 3:25 - 3:55 pm Fuel Qualification for Molten Salt Reactors NRC/ORNL 3:55 - 4:00 pm Future Meeting Planning and Concluding Remarks NRC Slide 2

Advanced Reactor Integrated Schedule of Activities The updated Advanced Reactor Integrated Schedule is publicly available on NRC Advanced Reactors website at:

https://www.nrc.gov/reactors/new-reactors/advanced/details#advSumISRA Slide 3

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/details#advSumISRA Knowledge, Skills, and Capability Computer Codes and Review Tools Concurrence (Division/Interoffice)

Guidance Federal Register Publication Commission Review Period**

Consensus Codes and Standards Public Comment Period ACRS SC/FC (Scheduled or Planned)

Policy and Key Technical Issues Draft Issuance of Deliverable External Stakeholder Interactions Communication Final Issuance of Deliverable

Public Meeting (Scheduled or Planned)

Present Day Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec x

x x

x x

x x

x x

x Reference plant model for Heat Pipe-Cooled Micro Reactor x

Reference plant model for Sodium-Cooled Fast Reactor x

Reference plant model for Molten-Salt-Cooled Pebble Bed Reactor x

Reference plant model for Monolith-type Micro-Reactor Reference plant model for Gas-Cooled Pebble Bed Reactor x

FAST code assessment for metallic fuel x

FAST code assessment for TRISO fuel x

x Non-LWR MELCOR (Source Term) Demonstration Project x

Reference SCALE/MELCOR plant model for Heat Pipe-Cooled Micro Reactor x

Reference SCALE/MELCOR plant model for High-Temperature Gas-Cooled Reactor x

Reference SCALE/MELCOR plant model for Molten Salt Cooled Pebble Bed Reactor x

Reference SCALE/MELCOR plant model for Sodium-Cooled Fast Reactor Reference SCALE/MELCOR plant model for Molten Salt Fueled Reactor MACCS radionuclide screening analysis MACCS near-field atmospheric transport and dispersion model assessment x

MACCS radionuclide properties on atmospheric transport and dosimetry MACCS near-field atmospheric transport and dispersion model improvement x

Phase 1 - Atmospheric Code Consolidation x

x 2022 Complete Regulatory Activity NEIMA Development of non-Light Water Reactor (LWR) Training for Advanced Reactors (Adv. Rxs) (NEIMA Section 103(a)(5))

FAST Reactor Technology High Temperature Gas-cooled Reactor (HTGR) Technology Molten Salt Reactor (MSR) Technology Code Assessment Reports Volume 1 (Systems Analysis)

Code Assessment Reports Volume 2 (Fuel Perf. Anaylsis)

Code Assessment Reports Volume 3 (Source Term Analysis)

Competency Modeling to ensure adequate workforce skillset Identification and Assessment of Available Codes Advanced Reactor Program - Summary of Integrated Schedule and Regulatory Activities*

Legend Strategy 1 Strategy 2 Strategy 3 Strategy 4 Strategy 5 Strategy 6 EDO Concurrence Period Version 11/4/21 Rulemaking 2021 Strategy 1

2 Development of Non-LWR Computer Models and Analytical Tools Guidance Research plan and accomplishments in Materials, Chemistry, and Component Integrity for Adv. Rxs.

Code Assessment Report Volume 4 (Licensing and Siting Dose Assessments)

Commission Papers Research on risk-informed and performance-based (RIPB) seismic design approaches and adopting seismic isolation technologies Code Assessment Report Volume 5 (Fuel Cycle Analysis)

Slide 4

The Technology-Inclusive, Risk-Informed Maximum Accident (TI-RI-MA) Approach - An Alternative to Probabilistic Risk Assessment Advanced Reactor Stakeholder Meeting Marty Stutzke Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission November 10, 2021 Slide 5

Outline

Background:

Quick history and definitions related to risk-informed regulation

  • Uses of PRA in initial licensing
  • Three potential licensing pathways
  • Guidance development
  • Path forward Slide 6

Early (Post-TMI) Recommendations: Use Quantitative Risk Assessment

  • ACRS letter1 (May 16, 1979)

- The ACRS believes that it is time to place the discussion of risk, nuclear and non-nuclear, on as quantitative basis as possible.

  • Kemeny Report1 (October 30, 1979) Recommendation #4:

- The [Presidential] Commission recommends that continuing in-depth studies should be initiated on the probabilities and consequences (on-site and off-site) of nuclear power plant accidents, including the consequences of meltdown.

Recommendation #8:

- The best way to improve the existing design review process is by relying in a major way upon quantitative risk analyses, and by emphasizing those accident sequences that contribute significantly to risk.

1Available from the Idaho National Laboratory Knowledge Management Library for the Three Mile Island Unit 2 Accident of 1979 at https://tmi2kml.inl.gov/HTML/Page1.html Slide 7

Definitions (1 of 2)

Risk triplet (Kaplan and Garrick2, SRM-SECY-98-1443):

- What can go wrong?

- How likely is it?

- What are the consequences?

Risk assessment (SRM-SECY-98-144): A systematic method for addressing the risk triplet as it relates to the performance of a particular system (which may include a human component) to understand likely outcomes, sensitivities, areas of importance, system interactions and areas of uncertainty.

Risk insights (SRM-SECY-98-144): The results and findings that come from risk assessments.

Risk-informed approach (SRM-SECY-98-144): A philosophy whereby risk insights are considered together with other factors to establish requirements that better focus licensee and regulatory attention on design and operational issues commensurate with their importance to public health and safety.

2Kaplan, S. and Garrick, B. J., On the Quantitative Definition of Risk, Risk Analysis, Vol. 1, Issue 1, March 1981.

3NRC, Staff Requirements - SECY-98-144 - White Paper on Risk-Informed and Performance-Based Regulation, February 24, 1998, ML003752593.

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Slide 8

Definitions (2 of 2)

Probabilistic risk assessment

- NRC online glossary4: A systematic method for assessing three questions that the NRC uses to define "risk." These questions consider (1) what can go wrong, (2) how likely it is, and (3) what its consequences might be. These questions allow the NRC to understand likely outcomes, sensitivities, areas of importance, system interactions, and areas of uncertainty, which the staff can use to identify risk-significant scenarios. The NRC uses PRA to determine a numeric estimate of risk to provide insights into the strengths and weaknesses of the design and operation of a nuclear power plant.

- RG 1.2005: An approach is considered to be a PRA when it (1) provides a quantitative assessment of the identified risk in terms of scenarios that result in undesired consequences (e.g., core damage or a large early release) and their frequencies and (2) comprises specific technical elements in performing the quantification.

- Draft RG 1.2476: A risk assessment approach is considered to be a PRA when it (1) provides a quantitative assessment of the identified risk in terms of scenarios that result in undesired consequences (e.g., releases of radioactive material, radiological consequences) and their frequencies and (2) comprises specific PRA elements for quantifying risk.

4https://www.nrc.gov/reading-rm/basic-ref/glossary/probabilistic-risk-assessment-pra.html 5NRC, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, RG 1.200, Rev. 3, December 2020, ML20238B871.

6NRC, Acceptability of Probabilistic risk Assessment Results for Advanced Non-Light Water Reactor Risk-Informed Activities, draft trial use RG 1.247, September 3, 2021, ML21246A216.

Slide 9

Uses of Risk Assessment in Initial Licensing

Purpose:

Thoroughly understand how and why risk assessment is used to support initial licensing.

- Required uses of PRA

- Expected uses of PRA

  • Review of information sources:

- Regulations

- Rulemakings

- Regulatory guides

- Commission policy statements

- Commission staff requirement memoranda

- Standard review plans

- IAEA SSR-2/1 Slide 10

Role of the PRA in Initial Licensing Traditional role

- Consistent with previous DC and COL applications

- Includes, but not limited to:

  • Searching for severe accident vulnerabilities (severe accident policy statement)
  • TMI requirement § 50.34(f)(1)(i), which under Part 52 requires LWR applicants to Perform a plant/site specific probabilistic risk assessment, the aim of which is to seek such improvements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant.
  • Demonstrating that the QHOs are met (safety goal policy statement)
  • Using PRA in the design process (PRA policy statement)

- Previously referred to as PRA in a supporting role Enhanced role

- Any use of PRA beyond its traditional role

- Includes, but not limited to:

  • Certain proposed required uses of PRA in preliminary 10 CFR Part 53 rule text (e.g., identifying licensing basis events; classifying systems, structures, and components; evaluating defense-in-depth)
  • Voluntary risk-informed applications (e.g., risk-managed technical specifications, risk-informed fire protection)

- Previously referred to as PRA in a leading role Slide 11

Requirements/Uses Part 50 Part 52 Part 53 Preliminary Rule Text Submit description of PRA and its results Currently none.

Proposed in Part 50/52 lessons learned rulemaking (NRC-2019-0196; RIN 3150-AI66)

All applicants All applicants Develop, maintain, and upgrade PRA COL holders COL and OL holders Required uses of PRA Meet the TMI requirements in

§ 50.34(f)(1)(i) - Seek improvements in core and containment heat removal systems reliability Use PRA to:

  • Search for severe accident vulnerabilities
  • Demonstrate that safety goals are met
  • Use PRA to evaluate changes to the facility described in FSAR (§ 53.1322)
  • Use PRA or generally accepted risk-informed approaches for systematically evaluating engineered systems to:

Identify LBEs Evaluate DID Classify SSCs Support the FSP Commission expectations (e.g., policy statements and SRMs)

Voluntary uses of PRA Voluntary risk-informed applications to establish or change the licensing basis Leveraged uses by the staff

  • Focus the staff review
  • Inform the development of ITAACS, COL action items, D-RAP, etc.
  • Support oversight and inspections Uses of the PRA Traditional role of PRA Enhanced role of PRA
  • Search for severe accident vulnerabilities
  • Demonstrate that safety goals are met
  • Use PRA in design Slide 12

Observations A risk-informed approach may be based on risk insights developed from:

- A PRA (i.e., quantitative), or

- A qualitative risk assessment PRA not used to support NPUF licensing

- Not addressed in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors

- Discussion with NRR/DANU staff Integrated safety analysis (ISAs) required by 10 CFR Part 70 for certain licensees The current preliminary rule text for Part 53:

- Codifies the traditional role of the PRA

- Adds requirements that use PRA in an enhanced role Three potential licensing pathways

  • Enhanced PRA approach
  • Traditional PRA approach
  • Technology-inclusive, risk-informed maximum accident (TI-RI-MA) approach Slide 13

Initial Thoughts: Three Pathways Applicant elects to develop PRA Comprehensive search for initiators Select design-basis accidents (DBAs)

Perform DBA consequence analyses Identify the MCA or MHA Conservative risk estimate QHOs Applicant elects enhanced uses of PRA Develop conservative risk estimate TI-RI-MA Approach Traditional PRA Approach Enhanced PRA Approach yes no yes no no Comprehensive definition of event sequences A

G B

C D

E F

H J

I L

K yes technology-inclusive, risk-informed maximum accident approach quantitative risk insights qualitative risk insights Slide 14

Development of How-To Guidance for the TI-RI-MA Approach (1 of 3)

The staff intends to develop guidance for:

- Box A: Comprehensive search for initiating events

- Box B: Comprehensive definition of event sequences

- Box D: Design-basis accident selection

- Box E: Design-basis accident consequence analysis

- Box F: Maximum accident (MCA or MHA) identification

- Box G: Conservative risk estimation Leverage existing guidance and studies such as, but not limited to:

- NUREG-1513, Integrated Safety Analysis Guidance Document

- NUREG/CR-2300, PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants

- Occupational Safety and Health Administration regulations (29 CFR 1910.119),

standards, handbooks, and guidance

- EPRI TR 3002011801, Program on Technology Innovation: Early Integration of Safety Assessment into Advanced Reactor Design - Preliminary Body of Knowledge and Methodology Guidance and preliminary rule text to be developed in parallel Slide 15

Development of How-To Guidance for the TI-RI-MA Approach (2 of 3)

  • Initial thoughts:

- Start with a blank sheet of paper

- Use a combination of inductive and deductive methods

- How much searching is enough? How do you know when you are finished?

- Focus on how plant design actually works vs. how plant design is supposed to work

- Consolidate/group similar items

- Be (very) careful when screening Slide 16

Step 1 Initiating Event Identification Categorization (AOO, DBA, BDBA, DBE, BDBE)

Scenario Analysis (Mitigation Scenarios vs. Prevention Scenarios; MCA or MHA)

Step 1B Deductive Methods Step 1A Inductive Methods Step 1C Human-Induced Events Step 2 Verification Operating Experience Analysis Step 3 Candidate Initiating Events Step 4 Initiating Event Grouping Quality Control Design/

Configuration Control Well-Established Methods MLD FTA Well-Established Methods HAZOP FMEA Well-Established Method HRA Initiating Event and Scenario Search Slide 17

Development of How-To Guidance for the TI-RI-MA Approach (3 of 3)

  • Multi-disciplinary team effort
  • Independent review
  • Documentation

- Tell the story

- Capture assumptions and decisions The Advisory Committee on Reactor Safeguards, July 9, 2008.

Source: USNRC. Cropped from original image.

Slide 18

Path Forward NRC staff will continue to develop guidance NRC staff will start to develop preliminary rule text Slide 19

November 10, 2021 Deanna Zhang NRR/DRO/IQVB An Overview of NRCs Regulatory Requirements and Guidance on Counterfeit, Fraudulent, and Suspect Items (CFSIs)

Slide 20

Objective

  • Present an overview of the NRCs regulatory requirements and guidance for addressing CFSIs and identify available external CFSI training.
  • Engage advanced reactor vendors to raise awareness on CFSI and discuss means to prevent or mitigate CFSI in the supply chain Slide 21

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Background===

Concerns related to CFSI affecting NRC regulated entities prompted several NRC initiatives to support addressing CFSI concerns.

Recent CFSI events both domestically and overseas resulted in:

- Issuance of guidance to heighten awareness of the existing NRC regulations and how they apply to CFSI

- Issuance of information notices on certain CFSI events

- Creation of CFSI Technical Review Group (TRG) to evaluate events to determine whether they involve CFSI and their applicability to NRC regulated facilities NRC advocates a proactive approach to detect and prevent the intrusion of CFSI into SSCs intended for use as a basic component.

Slide 22

CFSI is encompassed in NRC regulations for quality assurance and defect reporting Appendix B to 10 CFR Part 50 areas:

1) Design control
2) Procurement document control
3) Control of purchased materials, equipment, and services
4) Identification and control of material, parts, and components
5) Disposition of nonconforming materials, parts, or components
6) Corrective action and program effectiveness reviews 10 CFR Part 21 and 10 CFR 50.55(e)
1) Evaluation of deviations and failures to comply to identify defects and failures to comply associated with substantial safety hazard
2) Notification to NRC when there is information indicating a failure to comply or a defect Slide 23

Key Guidance Documents for CFSI and Recent Information Notices on CFSI events Information Notice (IN) 2018-11 Supplement 1: Kobe Steel Quality Assurance Record of Falsification IN 2013-15 Willful Misconduct/Record Falsification and Nuclear Safety Culture IN 2013-02 Issues Potentially Affecting Nuclear Facility Fire Safety IN 2008-04 Counterfeit Parts Supplied to Nuclear Power Plants NRC Bulletin 1988-010 Nonconforming Molded-Case Circuit Breakers Regulatory Issue Summary (RIS)-15-08, Oversight of Counterfeit, Fraudulent, and Suspect Items in Nuclear Industry heightens awareness of the existing NRC regulations and how they apply to CFSI within the scope of NRCs regulatory jurisdiction.

(ADAMS Accession No. ML15008A191)

Generic Letter 89-02, Actions to Improve the Detection of Counterfeit and Fraudulently Marketed Products shares information regarding elements of programs that appear to be effective in providing the capability to detect counterfeit or fraudulently marked products.

https://www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1989/gl89002.html Slide 24

Other Sources of Guidance International Guidance IAEA Publication on Managing Counterfeit and Fraudulent Items in the Nuclear Industry NEA MDEP CP-VICWG-04 Common Position on Counterfeit, Fraudulent, and Suspect Items Procedure Nuclear Industry Guidance EPRI Technical Report 3002002276 Plant Support Engineering: Counterfeit and Fraudulent Items - Mitigating the Increasing Risk NEI 14-09, Revision 1, Guidelines for Implementation of 10 CFR Part 21 Reporting of Defects and Noncompliance, as endorsed in NRC RG 1.234 Resources from other Federal Agencies Department of Homeland Security National Intellectual Property Rights Coordination Center https://www.iprcenter.gov/

Department of Energy (DOE) Operating Experience Committee https://www.energy.gov/ehss/doe-corporate-operating-experience-program Slide 25

CFSI Training Offerings

  • NRC has issued a list of CFSI-related training offerings as part of IN 2012-22, Counterfeit, Fraudulent, Suspect Item (CFSI) Training Offerings, which has been updated in 2019 (ADAMS Accession No. ML19017A117).
  • NRC continues to engage stakeholders to enhance awareness on CFSI and disseminate information on CFSI-related events in the nuclear industry.

Slide 26

Prevention and Mitigation of CFSI in Advanced Reactor Supply Chain: Key Takeaways

  • Incorporate processes to verify that products are authentic using receipt inspection, procurement controls, vendor authentication tools
  • Maintain traceability of products within the supply chain and reduce risk of counterfeit products by procuring from authorized resellers mitigate
  • Increase awareness of CFSI through training and coordinating with industry and government organizations involved in preventing and mitigating CFSI Slide 27

Questions Slide 28

Acronyms CFR: Code of Federal Regulations CFSI: counterfeit, fraudulent, and suspect item DHS: Department of Homeland Security DOE: Department of Energy EPRI: Electric Power Research Institute IAEA: International Atomic Energy Agency IN: Information Notice IPR: Intellectual Property Rights (Coordination Center)

MDEP: Multi-national Design Evaluation Program NEA: Nuclear Energy Agency NEI: Nuclear Energy Institute RIS: Regulatory Information Summary RG: Regulatory Guide VICWG: Vendor Inspection Co-operation Working Group Slide 29

Resources https://www.nrc.gov/about-nrc/cfsi.html https://www.iprcenter.gov/

https://www.epri.com/research/products/300 2002276 https://www.iaea.org/publications/11182/man aging-counterfeit-and-fraudulent-items-in-the-nuclear-industry https://www.energy.gov/ehss/doe-corporate-operating-experience-program Slide 30

Stakeholders Meeting Advanced Reactor Content of Application Project (ARCAP)

Chapter 11 Organization and Human-System Consideration Interim Staff Guidance (Draft)

November 2021 Slide 31

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Background===

The Advanced Reactor Content of Application Project (ARCAP) is developing guidance to support the review of non-light-water reactors (non-LWRs), modular LWRs and stationary micro-reactors. This project encompasses industry-led Technology-Inclusive Content of Application Project (TICAP).

SAR structure consists of 12 main chapters. TICAP is applicable to portions of first 8 SAR chapters. ARCAP addresses SAR Chapters 9, 10, 11, and 12.

This guidance for SAR Chapter 11, Organization and Human-System Considerations, under development as a draft Interim Staff Guidance (ISG) document, currently a draft white paper (ML21309A020).

This latest version of the guidance expands on the earlier version to provide additional proposed guidance for human factors engineering, operator licensing, operator training and staffing.

Slide 32

ARCAP and Technology Inclusive Content of Application Project (TICAP) - Nexus Outline Safety Analysis Report (SAR) -

Based on TICAP Guidance

1. General Plant Information, Site Description, and Overview of the Safety Case
2. Methodologies and Analyses
3. Licensing Basis Event (LBE) Analysis
4. Integrated Evaluations
5. Safety Functions, Design Criteria, and SSC Safety Classification
6. Safety Related SSC Criteria and Capabilities
7. Non-safety related with special treatment SSC Criteria and Capabilities
8. Plant Programs Additional Portions of Application
  • Technical Specifications
  • Technical Requirements Manual
  • Quality Assurance Plan (design)
  • Quality Assurance Plan (construction and operations)
  • Physical Security Plan
  • SNM material control and accounting plan
  • Radiation Protection Program
  • Inservice inspection/Inservice testing (ISI/IST) Program
  • Environmental Report
  • Site Redress Plan
  • Exemptions, Departures, and Variances
  • Facility Safety Program (under consideration for Part 53 applications)

Audit/inspection of Applicant Records Calculations Analyses P&IDs System Descriptions Design Drawings Design Specs Procurement Specs Probabilistic Risk Assessment Safety Analysis Report (SAR) structure based on clean sheet approach Additional SAR Content -Outside the Scope of TICAP

9. Control of Routine Plant Radioactive Effluents, Plant Contamination, and Solid Waste
10. Control of Occupational Doses
11. Organization and Human-System Considerations
12. Post-construction Inspection, Testing and Analysis Programs
  • Additional contents of application outside of SAR are still under discussion. The above list is draft and for illustration purposes only.

Slide 33

Updated Guidance Areas Included applicability section This ISG will be applicable to non-LWRs, stationary micro reactors and small modular LWRs submitting applications for a construction permit (CP) or operating license (OL) under Part 50 or for a combined license (COL) under Part 52.

Human Factors Engineering (HFE)

NRC staff identified need to provide guidance in this area to supplement licensing modernization project (LMP) and pending associated TICAP guidance LMP provides insights but provides limited guidance on how to develop a HFE program ARCAP Chapter 11 ISG covers HFE information that would support NRC findings Slide 34

Updated Guidance Areas Human Factors Engineering (continued)

References guidance found in:

NUREG-0711, Human Factors Engineering Program Review Model," and Scalable HFE approach being considered for Part 53 (Final Report Development of HFE Review Guidance for Advanced Reactors, (ML21287A088)

Slide 35

Updated Guidance Areas Licensed Operator Training Staff interested in stakeholder feedback on how best to capture full scope of proposed guidance Staff notes that proposed ARCAP guidance includes information that is not necessary to support an operating license or combined license issuance Staffs proposed guidance provides a holistic approach for operator licensing Based on insights from recent LWR applications Portions of guidance could be split out to other non-application guidance (e.g., guidance that supports operator licensing and inspection processes)

Slide 36

Updated Guidance Areas Licensed Operator Training (continued)

Proposed guidance includes areas such as:

Description of how chosen examination methods, structures, and passing scores, support examination validity, reliability, and fairness Procedure development used to ensure examination material:

1) writers guide requirements are met that addresses technical specification and FSAR requirements, 2) technical review to ensure procedure is correct for proper operation of the plant Description and qualification of simulator used to administer initial operator licensing examinations Timeline for operator licensing examinations Slide 37

Updated Guidance Areas Licensed Operator Training (continued)

Proposed guidance includes areas such as (continued):

Use of simulator for operation training experience and examinations during construction Operator license issuance prior to fuel load Operator Staffing Additional proposed guidance provided Option of providing technical basis for control room staffing in conjunction with control room configuration that would support capturing requirements in design certification rulemaking Provide technical basis that could support a future exemption from §§ 50.54(m) and/or 50.54(k) requirements Slide 38

QUESTIONS?

Slide 39

Stakeholders Meeting Advanced Reactor Content of Application Project (ARCAP)

Chapter 12 Post-construction Inspection, Testing, and Analysis Program Interim Staff Guidance (Draft)

November 2021 Slide 40

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Background===

The Advanced Reactor Content of Application Project (ARCAP) has been developing guidance to support the review of non-light-water reactors (non-LWRs), modular LWRs and stationary micro-reactors.

This project encompasses industry-led Technology-Inclusive Content of Application Project (TICAP).

SAR structure consists of 12 main chapters. TICAP is applicable to portions of first 8 SAR chapters. ARCAP addresses SAR Chapters 9, 10, 11, and 12.

This guidance for SAR Chapter 12, 'Post-Construction Inspection and Analysis Program,' (PITAP) under development as a draft Interim Staff Guidance (ISG) document, currently a draft white paper (ML21294A266).

This latest version of the guidance expands on the earlier version to address inspections and analysis verification, including inspections, tests, analyses, and acceptance criteria (ITAAC).

Slide 41

ARCAP and Technology Inclusive Content of Application Project (TICAP) - Nexus Outline Safety Analysis Report (SAR) -

Based on TICAP Guidance

1. General Plant Information, Site Description, and Overview of the Safety Case
2. Methodologies and Analyses
3. Licensing Basis Event (LBE) Analysis
4. Integrated Evaluations
5. Safety Functions, Design Criteria, and SSC Safety Classification
6. Safety Related SSC Criteria and Capabilities
7. Non-safety related with special treatment SSC Criteria and Capabilities
8. Plant Programs Additional Portions of Application
  • Technical Specifications
  • Technical Requirements Manual
  • Quality Assurance Plan (design)
  • Quality Assurance Plan (construction and operations)
  • Physical Security Plan
  • SNM material control and accounting plan
  • Radiation Protection Program
  • Inservice inspection/Inservice testing (ISI/IST) Program
  • Environmental Report
  • Site Redress Plan
  • Exemptions, Departures, and Variances
  • Facility Safety Program (under consideration for Part 53 applications)

Audit/inspection of Applicant Records Calculations Analyses P&IDs System Descriptions Design Drawings Design Specs Procurement Specs Probabilistic Risk Assessment Safety Analysis Report (SAR) structure based on clean sheet approach Additional SAR Content -Outside the Scope of TICAP

9. Control of Routine Plant Radioactive Effluents, Plant Contamination, and Solid Waste
10. Control of Occupational Doses
11. Organization and Human-System Considerations
12. Post-construction Inspection, Testing and Analysis Programs
  • Additional contents of application outside of SAR are still under discussion. The above list is draft and for illustration purposes only.

Slide 42

Background (cont.)

This ISG will be applicable to non-LWRs, stationary micro reactors and small modular LWRs submitting applications for a construction permit (CP) or operating license (OL) under 10 CFR Part 50 or for a design certification (DC), a combined license (COL), or a manufacturing license (ML) under 10 CFR Part 52.

This ISG differentiates between 10 CFR Part 52 applicants that must include ITAAC and 10 CFR Part 50 applications that are not required to include ITAAC.

This ISG will be updated to apply to applications under 10 CFR Part 53, when 10 CFR Part 53 is issued.

Slide 43

Requirements Post-construction inspection, testing, and analysis are required, in part, by regulations for applicants to provide a description of their quality assurance programs, as required by 10 CFR Part 50, Appendix B.

These quality assurance requirements are also included in 10 CFR 50.34(a)(7) for CP applicants and in 10 CFR 50.34(b)(6) for OL applicants. In addition, similar COL requirements associated with quality assurance are contained in 10 CFR 52.79(a)(25). Some advanced reactor applicants for whom the requirements of 10 CFR 50.43 apply will find similar requirements contained in 10 CFR 50.43(e)(1).

Requirements to describe preoperational testing and initial operations in OL and COL applications are contained in 50.34(b)(6)(iii) and 52.79(a)(28), respectively.

Slide 44

Requirements (continued)

The following regulations require that applications contain the proposed ITAAC that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, the facility has been constructed and will be operated in conformity with the license, the design certification for a facility that incorporates one, the provisions of the Act, and the Commission's rules and regulations:

o For DC applications: 10 CFR 52.47(b)(1) o For COL applications: 10 CFR 52.80 o For ML applications: 10 CFR 52.158(a)

Slide 45

Requirements (continued)

Requirements regarding Commission findings:

The following regulations require, in part, that the Commission make a finding that the facility has been constructed and will be operated in accordance with the [design certification or license], the provisions of the Act, and the rules and regulations of the Commission.

o For DCs: 10 CFR 52.54, Issuance of standard design certification o For OLs: 10 CFR 50.57, Issuance of operating license o For COLs: 10 CFR 52.97, Issuance of combined licenses Slide 46

Referenced Guidance The ISG references the guidance in NUREG-0800, Standard Review Plan, (SRP) Section 14.3, Inspections, Tests, Analysis, and Acceptance Criteria, with the caveat that the guidance in SRP 14.3 is applicable to LWRs but may contain insights that are useful for non-LWR application reviews. SRP 14.3 guidance in Appendix C, Detailed Review Guidance, may only be applicable if the features described are considered within the scope of Safety Related or safety-significant systems covered by this ISG.

Guidance regarding PITAP and ITAAC for Emergency Planning and Physical Security Hardware are not addressed in this ISG; but the LWR guidance for these topics in SRP Sections 14.3.10 and 12, respectively, is generally technology-neutral and may be adapted for use with non-LWRs.

Although guidance provided in Regulatory Guide (RG) 1.68, Initial Test Programs For Water-Cooled Nuclear Power Plants, is specific to LWRs, applicants may use this RG to gain insights that could inform the development of initial test programs for advanced reactors.

Slide 47

Scope The ISG consists of guidance related to o post-construction inspection, preoperational testing (i.e., tests conducted following construction and construction-related testing, but prior to initial fuel load), analysis verification, and o initial startup testing (i.e., tests conducted during and after initial fuel load, up to and including initial power ascension).

This ISG is intended to provide guidance to the NRC staff regarding application content that would support making the finding that the applicant has met the referenced regulations.

Slide 48

Scope (continued)

Slide 49

Objectives

  • The primary objective of the PITAP is to demonstrate, to the extent possible, that the Safety-Related (SR) and safety-significant structures, systems and components (SSCs) have been constructed and will be operated in accordance with the design and as described in the safety analysis report.

Slide 50

Objectives (cont.)

  • Additional objectives of the PITAP include:

Providing reasonable assurance that the facility exhibits the performance and associated safety margins that are described in the design; Satisfying any license conditions associated with the PITAP; Obtaining as-built data to validate the analytical assumptions, limits, and/or models; Familiarizing the plants operating and technical staff with operation of the facility; and Verifying the adequacy of the plant operating and emergency procedures.

Slide 51

Guidance Topics Phase 1 - Preoperational Inspection, Testing, and Analysis Verification Inspections - The PITAP (or referenced elements of the quality assurance program) should include a post-construction (preoperational) inspection program that addresses verification of items such as:

o Basic configuration and key design features for SR and safety-significant SSCs. This activity includes inspection of the functional arrangement of the as-built SR and safety-significant SSCs described in the safety analysis report.

o Electrical separation for SR and safety-significant SSCs where required.

o Materials of construction for SR and safety-significant SSCs per approved design codes and standards (e.g., ASME Code Section III,Section VIII, etc.)

Slide 52

Guidance Topics Phase 1 - Preoperational Inspection, Testing, and Analysis Verification (continued)

Testing - The PITAP should include a post-construction (preoperational) testing program for SR and safety-significant SSCs that addresses items such as:

o Reactivity control functions o

Heat removal functions o

Containment of radioactive material o

Testing required by consensus design codes and standards applied in the design (e.g., ASME, IEEE) o Flow induced system vibration and thermal expansion tests o

Electrical system performance for normal and emergency power o

Equipment identified as necessary for defense-in-depth o

Instrumentation and control (I&C) systems relied upon in the safety analysis to perform SR or safety-significant functions Slide 53

Guidance Topics Phase 1 - Preoperational Inspection, Testing, and Analysis Verification (continued)

Analysis - The PITAP (or referenced quality assurance program element) should include a description of what important analysis of SR and safety-significant SSCs should be verified including areas such as:

o Thermal and hydraulic analysis important to the performance of required safety functions o

Seismic analysis o

Verification of equipment required to be qualified for a harsh environment o

Critical assumptions from transient and accident analysis including barrier performance and effluent release calculations o

For I&C systems, analytical limits associated with each key variable, the ranges (normal, abnormal, and accident conditions), and the rates of change for these variables Slide 54

Guidance Topics (cont.)

Phase 2 - Initial Startup Testing Testing - The PITAP should include a post-construction (initial startup) testing program for SR and safety-significant SSCs that addresses items such as:

o Initial fuel loading and reactor physics tests o Low power testing o Power ascension testing o Performance of residual heat removal systems o Performance of liquid and gaseous waste systems o Performance of first-of-a-kind, inherent or passive safety features o Flow induced vibration and thermal expansion within design limits Slide 55

Guidance Topics (cont.)

General Guidelines Guidance for ensuring that the description of the PITAP in the application addresses programmatic items related to the development and conduct of the PITAP, such as:

o The PITAP objectives, including the objectives of each phase of the program.

o The scope of each phase of the PITAP.

o The organization and responsibilities for conduct and control of the inspection and testing program.

o A general schedule and sequence for conducting the inspections and tests, including established hold points.

o The extent to which the test program will use plant operating, emergency and surveillance procedures and technical specifications.

Slide 56

Guidance Topics (cont.)

Guidelines for Testing Guidance for ensuring that the application includes a general description for each test, or group of similar tests (i.e., test abstract), to be conducted.

The focus of the test descriptions should be on providing the bases for the tests and test conditions selected, instrumentation to be used, and a description of how the tests will confirm the performance of the SSCs.

The PITAP development should also take into consideration PITAP experience at other similar facilities and include measures to avoid problems they have had.

Slide 57

Guidance Topics (cont.)

General Responsibilities Guidance for ensuring that the application describes the responsibilities and guidelines for conduct of the PITAP, such as:

o Defining the qualifications of the personnel managing, conducting, and reviewing the inspection, test, and analysis verification program results.

o Providing training as necessary to ensure that personnel are ready to perform their functions.

o Developing the testing objectives, schedule, sequence, prerequisites, procedures safety precautions and acceptance criteria.

o Managing, controlling, and approving key aspects (e.g.,

prerequisites, procedures) of the test program.

o Establishing a plant review committee to review, evaluate, and disposition the inspection, test, and analysis verification results.

Slide 58

QUESTIONS?

Slide 59

Break Meeting will resume at 1pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 272 279 12#

Advanced Reactor Stakeholder Public Meeting Slide 60

Accelerated Fuel Qualification (AFQ) White Paper Overview Presented at the NRC Advanced Reactors Stakeholder Meeting November 10, 2021 Presented by Ron S. Faibish, PhD General Atomics Electromagnetic Systems (GA-EMS)

The AFQ White Paper Task Force (in alphabetic order):

Framatome General Atomics Idaho National Laboratory Los Alamos National Laboratory Oak Ridge National Laboratory Westinghouse Electric Company Slide 61

The AFQ Working Group

  • The AFQ Working Group is a grassroots, industry-driven group with participants from industry, national labs, academia, DOE and NRC
  • The AFQ Working Group (WG) is charged to develop a methodology that is generalized and is applicable to all fuel types
  • Three key elements to the WG charge:
  • 1) adoption of physics-informed advanced nuclear fuel performance modeling and simulation (M&S) tools
  • 2) use of targeted experiments to validate the modeling, under conditions that are as prototypic as possible
  • Main tool of implementation: information exchange workshops Slide 62

Motivation

  • Overarching motivation: Remove fuel qualification from the critical path of advanced reactors licensing to support more timely deployment of new reactors with new fuel systems
  • Formalize the AFQ methodology with details on enabling experimental and modeling tools and their use
  • Promote the Adoption of the AFQ methodology by industry, and recognition of the methodology by the Nuclear Regulatory Commission (NRC), would facilitate more efficient and timely qualification of new fuel systems.

Slide 63

Introduction

  • The AFQ methodology brings together advanced, physics-informed nuclear fuel performance M&S with targeted experiments to significantly reduce the time and cost to qualify new fuels
  • Use of targeted, separate-effects and accelerated experiments enables higher quality, fewer, more effective integral irradiation experiments to validate and acquire the data needed to support the safety case
  • Enabling AFQ: Three main phases discussed in the paper
  • Phase 1: Data Compilation and Physics-Based Modeling
  • Phase 2: Model Validation
  • Phase 3: Essential and Limited Integral Testing The goal of AFQ is to reduce the time to qualify new fuels from 20 years to as few as 5 years Slide 64

Key Elements of AFQ

  • High-fidelity, physics-based modeling and simulation (M&S) tools that adequately describe the fuel performance
  • Out-of-pile and in-pile targeted experiments that efficiently span the range of relevant parameters to provide data that may either be used to construct semi-empirical models or to efficiently validate physics and mechanistic models
  • Execution of coordinated experimental testing and M&S activities, in parallel.
  • Incorporation of specialized and accelerated testing methods to obtain relevant data more quickly, such as the Fission Accelerated Steady State Test (FAST) or HFIR MiniFuel irradiation methods
  • The implementation of AFQ must be tailored to specific reactor type, fuel form, and associated safety case Adoption of the AFQ methodology by industry, and its recognition by the NRC would facilitate efficient and timely qualification of new fuel systems Slide 65

AFQ-Enabling Modeling & Simulation Tools M&S are key in the AFQ methodology. Mechanistic models of fuel performance based on a multiscale methodology can be utilized to:

  • Accurately interpolate between sparse experimental data on irradiated fuels
  • Provide a detailed analysis of experimental results to reveal and understand governing phenomena
  • Design future experiments to strategically target key unknowns or regimes.
  • Potentially identify optimized fuel compositions
  • Be a key aid to informing fabrication studies or manufacturing activities (although this is not the subject of this white paper)

Data produced via M&S do NOT replace integral tests and other experimental data -

they are augmenting them Slide 66

Multi-Scale Modeling & Simulation Least Linear Scale (LLS) tools - Density functional theory (DFT), molecular dynamics (MD), dislocation dynamics (DD),

cluster dynamics, phase-field simulations, crystal plasticity methods, and more, are available to be applied to nuclear fuel problems Bison - Engineering-scale fuel performance code built on MOOSE with inherent ability to utilize modern HPC platforms, can assess multiple fuel types and complex geometries - uses both empirical materials models and multi-scale models Data Science - scale bridging by development of reduced order models (ROMs) from mechanistic models and uncertainty quantification and sensitivity analysis - the glue that ties experiments, mechanistic and empirical models together Example: Fission gas release in doped UO2 M.W.D Cooper, et al., Fission gas diffusion and release for Cr2O3-doped UO2:

From the atomic to the engineering scale, Journal of Nuclear Materials 545, 152590 (2021).

Mechanistic multi-scale modeling, validated by separate effects data, can provide additional data to that obtained by traditional integral tests Slide 67

Accelerated Irradiation Capabilities Modeling &

Simulation Post Irradiation Examinations Advanced Experimental Tools

  • Old Paradigm: Serial and Limiting
  • Numerous prototypic integral test campaigns
  • Develop empirical fits with limited operational regime
  • Applicability limited to data regime
  • Additional test campaigns required to broad applicability
  • New Paradigm: Ties accelerated irradiation testing, advanced M&S, and high-throughput characterization at microstructural scale together
  • Reduced integral data needs and accelerated irradiations
  • Bridge phenomenological and length scale gaps with advanced PIE
  • Develop mechanistic models bounding the application AFQ Slide 68

Advanced Experimental Tools Accelerated Experimental Capabilities

  • Separate Effects: MiniFuel
  • Simplified design, analysis, and PIE
  • Multiple samples and conditions
  • Option to accelerate sample burnup
  • Integral Irradiations: FAST (Fission Accelerated Steady-state Testing)
  • Burnup acceleration
  • Reduce irradiation times
  • Maintains semi-prototypic fuel conditions
  • Inform/validate advanced M&S
  • Understand and assess non-prototypicalities Advanced PIE Capabilities
  • Types of Data
  • microstructure and form
  • chemistry and composition
  • thermo-mechanical properties
  • three-dimensional reconstructions
  • High-throughput capabilities generate comprehensive data sets spanning length scales
  • Identifies key structure-processing-properties that impact performance
  • Identifies materials strengths and weaknesses
  • Used to inform the development of physics-based material models
  • Enable increasingly predictive models Slide 69

The Overall Reactor System and AFQ

  • For any reactor system, we need to qualify fuel such that we demonstrate[e] 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
  • A licensing methodology - including one which utilizes AFQ - needs to:
  • Define an envelope of normal operating conditions and accident scenarios
  • Define failure modes in the above conditions
  • Define requirements to ensure that the failures will not happen, with a high degree of certainty Slide 70
  • As stated on the prior slide, a reactor systems fuel and structural materials must be able to withstand the fuel performance envelope of a given design
  • AFQ can be used to accelerate feasibility assessments of fuel and reactor materials (i.e. screen materials, expedite experimental testing, etc.)
  • AFQ models can be used to explore sensitivities in key parameters and to understand the influence of materials selection criteria on ultimate reactor system performance
  • Novel experiments, advanced modeling, and on-line instrumentation are designed to expedite materials model development for use in codes and methods
  • In new reactor designs, AFQ modeling and advanced experiments can aid in the identification and characterization of novel failure modes The Overall Reactor System and AFQ Slide 71

Importance of the Phenomena Identification and Ranking Table (PIRT)

  • The PIRT is a systematic way of organizing information to help guide research or development of regulatory requirements
  • Once the system constraints (reactor system and operating parameters) during normal operation and postulated accident scenarios and the fuel design concepts are identified, potential failure mechanisms are considered with respect to achieving basic design functions
  • The phenomena leading to the failure are then investigated and ranked
  • PIRTs identify gaps in the understanding of fuel failure and damage mechanisms to focus integral irradiation testing and continued development of M&S Slide 72

The PIRT Process

  • A collaborative effort between fuel developers and national labs is utilized in the iterative process of ranking the phenomena to prioritize the use of engineering-scale M&S tools in combination with separate-effects testing
  • The complexity of the new fuel design, or change from the currently licensed, in combination with the reactor design, determine the degree of the PIRT that needs to be completed
  • A well-informed, focused and collaborative effort prioritizing R&D on lower-knowledge phenomena will contribute to a reduction in overall fuel qualification development time Slide 73

Fuel fabrication Fuel performance models AND Targeted testing:

Physics validation Engineering models using FEM code AND Targeted testing:

Separate effects Out-of-pile measurements In-pile measurements Incorporation in conventional fuel performance codes AND Integral testing Accelerated Fuel Qualification (AFQ) Methodology Essence of AFQ: To Build on Existing Efforts and Incorporate New Tools For Efficient Demonstration and Adoption of New Fuels Fuel characterization measurements Multi-scale modeling and simulation Material property measurements Empirical Methodology (Traditional)

  • Figure from the AFQ white paper Slide 74

AFQ Methodology Has Three Phases and Facilitates Judicious Use of Resources

  • Phase 1: Design phase
  • Design of fuel system takes advantage of physics-based understanding
  • Assessment identifies gaps and early PIRT analysis
  • Phase 2: Detailed analysis and evaluation, involving targeted experiments and dedicated simulations to model constituent behavior
  • Use of M&S in designing experiments and enhancing analysis leads to efficient acquisition of necessary data
  • Phase 3: Integral irradiation tests
  • Design of complex integral experiments is better informed by physics-based understanding from separate effects tests
  • Use of AFQ methodology enables quality data and bounding of uncertainties Slide 75

Employing Modern Tools of AFQ Produces the Most Time Savings During Phase 1 and 2 and Contributes to Higher Quality Data in Phase 3

  • Figure from Terrani, et al., JNM 539 (2020) 152267 Slide 76

Phase 2: Focus on Obtaining Data Through Targeted Experiments and Appropriate Use of Modeling and Simulation

  • Figure from AFQ white paper, and Bolin et al., GA report 30533R00003 submitted to DE-NE0008331 Slide 77

Summary

  • The AFQ methodology offers a path to qualify new nuclear fuels in a timely and cost-effective way by leveraging the most advanced M&S and experimental tools that are available today
  • The AFQ methodology is a suggested guide to the qualification of new nuclear fuels
  • The AFQ methodology must be tailored for the specific reactor type, fuel form, and safety case
  • Best-practice updates of AFQ implementation can only be enabled by shared experience and data from specific user applications Slide 78

Path Forward

  • Conduct technical workshops to share knowledge on the development, validation and implementation of AFQ tools
  • Establish DOE funding opportunities to industry, labs and academia for developing and validating AFQ tools for specific case studies
  • Continue engagement with the NRC for acceptance and adoption of the AFQ methodology as a valid, optional toolkit for accelerating advanced fuel qualification
  • The AFQ Working Group welcomes the mention of AFQ in the NRCs Draft NUREG-2246, "Fuel Qualification for Advanced Reactors. This is a great first step in including a high-level summary of the AFQ methodology and stating that the AFQ process appear to be consistent with the considerations in the experimental data assessment framework Slide 79

Thank You!

Ron S. Faibish Ron.faibish@ga.com Slide 80

Update on the Development of Part 53 Key Regulatory Guidance for Tailored Operator Licensing and Flexible Staffing Juan Uribe, NRR/DANU Jesse Seymour, NRR/DRO Maurin Scheetz, NRR/DRO November 10, 2021 Slide 81

Agenda

  • Background
  • Part 53 Tailored Operator Licensing Guidance Development
  • Part 53 Flexible Staffing Guidance Development
  • Questions Slide 82

Risk-Informed Thinking White Paper: Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors Released March 25, 2021 (ML21069A003)

Key messages:

  • If advanced reactor designs present very low radiological risk, then current regulatory framework of large LWRs may be unnecessary for reasonable assurance of safety.
  • A new regulatory framework for advanced reactors (10 CFR 53) should be capable of addressing novel operational concepts for a wide variety of advanced reactor technologies.
  • A risk-informed, performance-based, and technology-inclusive regulatory framework for advanced reactors must appropriately consider the role of humans and human-system integration.

Slide 83

White Paper: Key Components Operations Guidance Scalable HFE Reviews Scalable Staffing Tailored Operator Licensing Criteria for NL Operators CONOPS HFE in ARCAP Legend Ongoing Activities Project under development Topics addressed via rule language Main document - Operations Roadmap

  • Rule language
  • Supporting Guidance Scalable HFE Tailored Operator Licensing Scalable Staffing Slide 84

Innovative Part 53 Rule Language Subpart F - Covers the areas of staffing, training, personnel qualifications, and human factors.

Preliminary Subpart F rule language publicly available via ADAMS at ML21267A006.

Discussed at 10/26/21 Advanced Reactor stakeholder meeting.

Introduction with ACRS late September (9/23/2021) and full discussion late November*.

  • tentative schedule Slide 85

Scalable Human Factors Engineering Discussed at 09/29/2021 Advanced Reactor Stakeholder meeting.

NRC/BNL - Purpose is to develop a method for scaling the scope and depth of HFE reviews for advanced reactors.

Method enables staff to readily adjust the focus and level of HFE review based upon risk/safety insights and the unique characteristics of the facility design/operation.

Staff plans to issue a white paper with guidance on scalable HFE reviews by December 2021.

Released October 12, 2021 (ML21287A088)

Slide 86

Part 53 Operator Licensing

  • Two main categories of operators
1. Licensed Operators
  • Includes RO and SRO license levels
  • Initial licensed operator training and examination programs require Commission approval
  • Examinations approved by the Commission; may be administered by facility staff
  • Operator licenses issued by the Commission
  • Requalification training program requires Commission approval and includes periodic requalification examinations
  • Simulators used for reactivity manipulations and licensing exams require Commission approval Slide 87

Part 53 Operator Licensing

  • Two categories of operators (continued)
2. Certified Operators (non-licensed)
  • Initial training programs and examination programs require Commission approval
  • Examinations developed and implemented by facility
  • Facility administers certifications and issues certificates
  • Program subject to ongoing inspection by Commission
  • Continuing training program requires Commission approval and includes periodic requalification examinations
  • Simulators used in certified operator programs do not require Commission approval Slide 88

Operator Licensing Guidance

  • Part 53 requires that initial training programs, examination programs, and requalification programs be developed using SAT

- Focus on process/method

- Less prescriptive than Part 55

  • Guidance is under development on how to review and approve these programs

- Specialized experience under contract to ensure guidance reflects sound competency assessment testing practices Slide 89

Operator Licensing Guidance

  • Facilities propose programmatic elements

- What knowledge warrants testing

- How knowledge will be sampled

- Examination methods used (written, JPM, etc)

- What constitutes passing performance

  • Comparable to basis of current NUREG-1021, but technology-inclusive and tailored to each facility
  • NRC reviews and approves facility proposed examination process Slide 90

Operator Licensing Guidance

  • Currently under development
  • Multiple opportunities for public engagement
  • Staff anticipate that specific opportunities for stakeholder engagement via workshops and comment period on draft guidance will be provided as part of ongoing DOE contract
  • Intent is to issue draft guidance to accompany proposed rule language in 2022 Slide 91

Part 53 Staffing Requirements

  • Part 53 approach to licensed operator staffing allows facilities to propose staffing plans appropriate for their concept of operation

- No equivalent to 10CFR50.54 (m) control room staffing levels

- No requirements for the location of operators

  • Applicants need to support staffing plans with relevant HFE-based analyses and assessments to demonstrate that safety functions will be maintained
  • Approved staffing plan informs the facility's minimum staffing requirements Slide 92

Flexible Staffing Reviews

  • Scope of the staffing review depends on the type of staffing required by Part 53:

- Plants with licensed operators: need to demonstrate how safety functions are maintained by proposed staffing (HFE analyses/assessments)

- Plants with certified operators: need to detail how proposed staffing supports certain duties (i.e., admin, monitoring, EP, etc.)

  • Certified operators are not credited for event mitigation; this is central to their relative level of requirements

- Both sets of requirements are performance-based Slide 93

Flexible Staffing Guidance NUREG-1791 provides a structured review method for evaluating 10 CFR 50.54(m) exemption requests

  • Staff is drafting guidance to review staffing plans submitted under Part 53 for plants with licensed operators

- The guidance under development augments the review guidance in NUREG-1791 to facilitate staffing plan reviews for Part 53 applicants/licensees (no exemption necessary)

  • Additional guidance may be developed for review of other staffing plans; these may be incorporated into other ARCAP interim staff guidance documents Slide 94

Questions?

Slide 95

Acronyms Used ARCAP - Advanced Reactor Content of Application Project DANU - Division of Advanced Reactors and Non-Power Production and Utilization Facilities DOE - Department of Energy DRO - Division of Reactor Oversight EP - Emergency Preparedness HFE - Human Factors Engineering ISG - Interim Staff Guidance JPM - Job Performance Measure NRR - Office of Nuclear Reactor Regulation RO - Reactor Operator SAT - Systems Approach to Training SRO - Senior Reactor Operator Slide 96

Break Meeting will resume in 10 minutes Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 272 279 12#

Advanced Reactor Stakeholder Public Meeting Slide 97

Physical Security Q&As for Category II Fuel Cycle Facilities November 10, 2021 Jason Piotter Office of Nuclear Material Safety and Safeguards (NMSS)

Division of Fuel Management (DFM)

Slide 98

=

Background===

Physical Protection Requirements for Cat II Fuel Facilities Q&A Content Key Messages Overview Slide 99

Current or recent fuel facility licensing actions have been related to preliminary updates to support higher enrichment Extent of the timing and scope of category II quantities (e.g.

HALEU) fuel licensing actions is developing with some preliminary pre application interactions taking place An important component of those upcoming licensing interactions is the physical security of a Category II quantity of special nuclear material (SNM) of moderate strategic significance

Background

Slide 100

Governed by 10 CFR 73.67(d) which presents requirements for each applicant who possesses, stores, or uses SNM of moderate strategic significance Existing guidance for the physical security plan is Regulatory Guide 5.59 Standard Format and Content for a Licensee Physical Security Plan for the Protection of Special Nuclear Material of Moderate or Low Strategic Significance.

Physical Protection Requirement for Category II Fuel Facilities Slide 101

Q&A Content

1. What are the types of SNM that the NRC regulates?
2. What types of facilities possess Category II quantities of SNM? How are these facilities related to advanced reactor fuels?
3. Can a licensee possess some amount of high-enriched uranium (20% U-235 or greater) and still be considered a Category II facility?
4. What are the existing physical protection requirements and current regulatory approach for Category II quantities of SNM?
5. How will the NRC staff provide consistent and transparent Part 73 reviews?

Slide 102

Q&A Content (cont)

6. What information should an applicant be prepared to discuss to ensure productive discussions with NRC during early interactions (e.g., pre-application meetings)?
7. Has the NRC staff identified lessons learned during the recent SHINE and Centrus license reviews?
8. Is SECY-11-0184, Security Regulatory Framework for Certifying, Approving, and Licensing Small Modular Reactors," applicable for physical security requirements for fuel cycle facilities with Category II quantities of SNM?
9. Does the Physical Security for Advanced Reactors rulemaking include requirements for Category II SNM at facilities other than power reactors?
10. Will the Part 53 rulemaking include new requirements for Category II quantities of SNM?

Slide 103

Supplemental security measures for the protection of Category II quantities of SNM may be required to address the current threat environment and the changing understanding of the risks associated with facilities possessing Category II quantities of SNM.

Staff uses a risk-informed analysis on a case-by-case basis to develop appropriate site-specific supplemental security measures, if needed, that would be implemented through license conditions to ensure the security of Category II quantities of SNM.

Key Messages Slide 104

Supplemental security requirements could include measures to provide greater security or control over material in use and storage and vital equipment To ensure a timely and efficient review, applicants planning to possess Category II quantities of SNM should engage with NRC staff early in the licensing process. The early establishment of an information security program allows for more detailed information to be shared expeditiously.

Key Messages (cont.)

Slide 105

Jason Piotter U. S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards Division of Fuel Management jason.piotter@nrc.gov 301-415-7739 Slide 106

ORNL is managed by UT-Battelle, LLC for the US Department of Energy US MSR Fuel Salt Qualification Process NRC - Advanced Reactors Stakeholders Meeting David Holcomb, George Flanagan, and Mike Poore November 10th, 2021 Slide 107

22 November 10, 2021 Advanced Reactors Stakeholders Meeting Fuel Qualification is an Element in Achieving Sufficient Understanding of Fuel Behavior Fuel qualification is a process which provides high confidence that physical and chemical behavior of fuel is sufficiently understood so that it can be adequately modeled for both normal and accident conditions, reflecting the role of the fuel design in the overall safety of the facility. Uncertainties are defined so that calculated fission product releases include the appropriate margins to ensure conservative calculation of radiological dose consequences.

NRC Presentation on Possible Regulatory Process Improvements for Advanced Reactor Designs, August 3rd, 2017 (ML17220A315)

Slide 108

33 November 10, 2021 Advanced Reactors Stakeholders Meeting Liquid Fuel Has Substantial, Fundamental Differences From Solid Fuel Liquid Fuel Chemically damageable -

may be reparable during use Composition may be adjustable during use Properties depend on composition and state Container breach could release nearly all radionuclides Solid Fuel Mechanically damageable Composition set prior to use Properties depend on fabrication process

  • Liquid salt fuel

- Serves as nuclear fuel and primary heat transfer media

- Must meet requirements for both purposes Slide 109

44 November 10, 2021 Advanced Reactors Stakeholders Meeting Common Salt Properties and Plant Functions Enable a General Liquid Fuel Salt Evaluation Method

  • Specific accident sequences are design dependent
  • Basic operational and fundamental safety functions are common to any nuclear power plant
  • Halide salt characteristics are common to any MSR

- High boiling points (low pressure)

- Low Gibbs free energy (low chemical potential energy)

- Natural circulation heat transfer properties

  • Fuel salt interacts with its container layers via common chemical and physical mechanisms - for example via

- Thermal energy transfer, chemical reactions, and mechanical processes Slide 110

55 November 10, 2021 Advanced Reactors Stakeholders Meeting Key Issue is What Constitutes Fuel Salt?

  • Fuel salt does not come in discrete elements (rods or assemblies) and moves independently of its container during normal operations

- Cladding and fuel assembly structures are qualified as part of solid fuel

  • Fuel salt includes all of the material containing fissionable elements or radionuclides that remain in hydraulic communication, but does not include the surrounding systems, structures, or components

- Salt vapors and aerosols remain part of the fuel salt system until they become adequately trapped

- Container corrosion products become part of the fuel salt

  • Fresh and used fuel salt in on-site storage are within scope Slide 111

66 November 10, 2021 Advanced Reactors Stakeholders Meeting Functional Containment is Important to How MSRs Provide Adequate Radionuclide Retention

  • Barrier performance must be degraded to release radionuclides into the environment

- Performance degradation can occur through failure or bypass

  • Fuel salt properties that stress barriers cause them to be more likely to release radionuclides - for example

- Increased temperature increases radionuclide vapor pressure in cover gas and well as decreasing strength of container

  • Different performance requirements for materials normally in contact with salt versus those that only need to withstand accidents Slide 112

77 November 10, 2021 Advanced Reactors Stakeholders Meeting Fuel Salt Boundary Breach Accident Progression Part of Performance Based and Deterministic Fuel Qualification

  • Multiple locations in the Code of Federal Regulations require evaluation of a postulated fission product release from core into containment
  • Fuel salt or cover gas cannot directly stress exterior containment layers without first breaching an inner containment layer
  • High radiation and high temperatures immediately outside fuel salt boundary substantially circumscribes characteristics of materials adjacent to fuel salt container
  • Focus is on fuel salt properties that must be known to adequately model accident progression and interaction characteristics with materials within containment Slide 113

88 November 10, 2021 Advanced Reactors Stakeholders Meeting Fuel Related Advanced Reactor Requirements Are Similar for Liquid and Solid Fuel

  • Example

- 10 CFR 50.43(e)(1)(i) requires that the performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof

- Fuel salt thermophysical and thermochemical properties provide the information necessary to model its role in enabling plant safety features to perform safety functions

- Fuel salt properties vary with both composition and temperature

- Fuel salt properties need to be determined across the range of temperatures and compositions that span potential operational and accident conditions

- Quality of the fuel salt property data needs to be sufficient to enable modeling the role of the fuel salt in achieving the plant FSFs Slide 114

99 November 10, 2021 Advanced Reactors Stakeholders Meeting Liquid Salt Fuel Assessment Framework Follows Template Developed for Solid Fueled Advanced Reactors

  • Top-down approach used to decompose top level goal of fuel is qualified to lower level supporting goals

- Qualifying fuel develops high confidence that the fuel will adequately perform its role in enabling the facility to achieve its safety objectives

  • Lower level supporting goals are further decomposed until clear objective goals are identified that can be satisfied with direct evidence Slide 115

10 10 November 10, 2021 Advanced Reactors Stakeholders Meeting Qualification is Based Upon Understanding the Chemical and Physical Properties of Representative Fuel Samples

  • Liquid state significantly changes the physical behavior of fuel

- Liquids do not accumulate internal stresses

  • No history dependent properties

- Flow homogenizes fluid properties

  • No position dependent properties
  • No size dependent properties
  • Chemical and physical properties are set by elemental composition and temperature

- Independent of isotopic content Small minimally-radioactive liquid fuel salt samples provide representative physical and chemical properties Slide 116

11 11 November 10, 2021 Advanced Reactors Stakeholders Meeting Liquid Fuel Salt Property Database Relates Composition to Physical and Chemical Properties to Aid Developers

  • Database development underway sponsored by DOE-NE

- Database development guided by modeling and simulation

- Requires appropriate quality assurance for both new and existing data

  • Safety evaluations / accident models performed with bounding values to establish acceptable performance range Slide 117

12 12 November 10, 2021 Advanced Reactors Stakeholders Meeting Fuel Salt Supports the Plant SSCs in Achieving the FSFs and Regulatory Requirements

  • Qualification focuses on identification and understanding of fuel salt property degradation mechanisms that occur as a result of irradiation during reactor operation

- Property repair (composition adjustment) may be incorporated into normal operation

  • During normal operations and AOOs fuel salt properties must result in sufficient margin from damage to safety-related SSCs
  • Under accident conditions the fuel salt properties must not result in sufficient damage to safety-related SSCs to prevent them from achieving their function Slide 118

13 13 November 10, 2021 Advanced Reactors Stakeholders Meeting Fuel Qualification Draft NUREG/CR is Available for Review and Comment

  • Suggestions for improvements to the approach can be provided at any time
  • Comments and suggestions can be provided to the NRC or ORNL contacts

- Chris Van Wert, Christopher.VanWert@nrc.gov

- Richard Rivera, Richard.Rivera@nrc.gov

- David Holcomb, HolcombDE@ornl.gov Slide 119

Future Meeting Planning

  • The next periodic stakeholder meeting is scheduled for January 19, 2022.
  • If you have suggested topics, please reach out to Prosanta.Chowdhury@nrc.gov.

Slide 120

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