ML22014A256

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January 19 2022 Advanced Reactor Stakeholder Meeting Slides
ML22014A256
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Issue date: 01/14/2022
From: Prosanta Chowdhury
NRC/NRR/DANU/UARP
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Download: ML22014A256 (113)
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Slide 1 Advanced Reactor Stakeholder Public Meeting January 19, 2022 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 323 588 045#

Slide 2 New names of GovDelivery categories: from NRC-DOE non-LWR workshops to Advanced Reactor Stakeholder Meetings; from Advanced Reactor Guidance Initiative to Advanced Reactor Rulemaking and Guidance Development https://service.govdelivery.com/accounts/USNRC/subscriber/new

Slide 3 https://service.govdelivery.com/accounts/USNRC/subscriber/topics

Slide 4 Time Agenda Speaker 10:00 - 10:20 am Opening Remarks / Adv. Rx Integrated Schedule NRC 10:20 - 10:30 am Status Overview of the Adv. Rx Generic Environmental Impact Statement (GEIS) NRC and Rulemaking Activities 10:30 - 11:15 am Implementing Near-field Models in MACCS v4.1 for Better Near-field Dose NRC/SNL Calculations 11:15 am - 12:00 pm Light Water Reactor Construction Permit Interim Staff Guidance NRC 12:00 - 1:00 pm Lunch Break All 1:00 - 1:45 pm Nuclear Data Assessment for Advanced Reactors NRC/ORNL 1:45 - 2:30 pm SCALE/MELCOR Development and Applications for non-LWRs NRC/SNL & ORNL 2:30 - 2:40 pm Break All 2:40 - 3:20 pm Advanced Manufacturing Technologies NRC 3:20 - 3:30 pm Future Meeting Planning and Concluding Remarks NRC

Slide 5 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 6 Advanced Reactor Integrated Schedule of Activities Advanced Reactor Program - Summary of Integrated Schedule and Regulatory Activities*

Strategy 1 Knowledge, Skills, and Capability Legend Strategy 2 Computer Codes and Review Tools Concurrence (Division/Interoffice) EDO Concurrence Period Strategy 3 Guidance Federal Register Publication Commission Review Period**

Strategy 4 Consensus Codes and Standards Public Comment Period ACRS SC/FC (Scheduled or Planned)

Strategy 5 Policy and Key Technical Issues Draft Issuance of Deliverable External Stakeholder Interactions Strategy 6 Communication Final Issuance of Deliverable Public Meeting (Scheduled or Planned) Version Present Day 1/7/22 2021 2022 Commission Strategy Regulatory Activity Papers Guidance Rulemaking NEIMA Complete Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Development of non-Light W ater Reactor (LW R) Training for x

Advanced Reactors (Adv. Rxs) (NEIMA Section 103(a)(5))

FAST Reactor Technology x x 1

High Temperature Gas-cooled Reactor (HTGR) Technology x x Molten Salt Reactor (MSR) Technology x x Competency Modeling to ensure adequate workforce skillset x Identification and Assessment of Available Codes x Development of Non-LW R Computer Models and Analytical Tools Code Assessment Reports Volume 1 (Systems Analysis) x Reference plant model for Heat Pipe-Cooled Micro v1 v2 Reactor (update from v1 to v2)

Reference plant model for Sodium-Cooled Fast Reactor v1 v2 (update from v1 to v2)

Reference plant model for Molten-Salt-Cooled Pebble x

Bed Reactor Reference plant model for Monolith-type Micro-Reactor Reference plant model for Gas-Cooled Pebble Bed Reactor Code Assessment Reports Volume 2 (Fuel Perf. Anaylsis) x FAST code assessment for metallic fuel x FAST code assessment for TRISO fuel x Code Assessment Reports Volume 3 (Source Term Analysis) x Non-LW R MELCOR (Source Term) Demonstration x

Project Reference SCALE/MELCOR plant model for Heat x

Pipe-Cooled Micro Reactor Reference SCALE/MELCOR plant model for High-x 2 Temperature Gas-Cooled Reactor Reference SCALE/MELCOR plant model for Molten x

Salt Cooled Pebble Bed Reactor 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 x

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

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

Phase 1 - Atmospheric Code Consolidation Code Assessment Report Volume 5 (Fuel Cycle Analysis) x Research plan and accomplishments in Materials, Chemistry, and x

Component Integrity for Adv. Rxs.

Research on risk-informed and performance-based (RIPB) seismic design approaches and adopting seismic isolation technologies https://www.nrc.gov/reactors/new-reactors/advanced/details#advSumISRA

Slide 7 Advanced Reactor Integrated Schedule of Activities UPDATES:

Strategy 2, Computer Codes and Review Tools:

  • Reference plant model for Heat Pipe-Cooled Micro Reactor - task complete
  • Reference plant model for Sodium-Cooled Fast Reactor (update from version 1 to 2) - v1 complete; v2 completion Sept. 2022
  • Reference plant model for Monolith-type Micro-Reactor - completion Jul. 2022
  • Reference plant model for Gas-Cooled Pebble Bed Reactor - completion Dec. 2022
  • MACCS near-field atmospheric transport and dispersion model assessment - Marked complete
  • MACCS radionuclide properties on atmospheric transport and dosimetry - Final issuance of deliverable now Sept.

2022 from June 2022 Strategy 3, Guidance:

  • Develop Advanced Reactor Technology Inclusive Content of Application Project (TICAP) Regulatory Guidance -

Added a TICAP public meeting in January 2022

  • Develop Advanced Reactor Inspection and Oversight Framework Document - Draft issuance of deliverable moved to February 2022 from December 2021

Slide 8 Advanced Reactor Integrated Schedule of Activities UPDATES (contd.):

Strategy 3, Guidance (contd.):

  • Develop MC&A guidance for Cat II facilities (NUREG-2159) - Draft of NUREG at end of Sept. 2021; 60-day comment period, extended to Dec. 3 per NEI request. Issue final by March 2022 (shifted by five months)

Strategy 4, Consensus Codes and Standards:

  • Develop Regulatory Guide for endorsement of the ASME Section XI, Division 2 Standard (Reliability and Integrity Management) - Draft Guide issued 9/30/21; public comment period closed 11/15/21 - staff working to resolve comments; plan to issue Final RG ~June 2022 Strategy 5, Policy and Key Technical Issues:
  • Report regarding review of the insurance and liability for advanced reactors (Price-Anderson Act) - completed 12/21/21 (due date 12/31/21)
  • Develop SECY Paper regarding Population-Related Siting Considerations for Advanced Reactors - marked complete with issuance of SECY-20-0045

Slide 9 Advanced Reactor Integrated Schedule of Activities UPDATES (contd.):

Rulemaking:

  • Part 53 Plan - Risk-Informed, Technology Inclusive Regulatory Framework for Advanced Reactors (NEIMA Section 103(a)(4)) - Extension request approved. This version reflects new schedule including interactions with ACRS -

concurrence in Sept - Dec 2022; ACRS meetings in Feb, Apr, Jun, Aug-Oct

  • Physical Security for Advanced Reactors - Extension request approved. Changes reflect new schedule
  • Develop draft Generic Environmental Impact Statement for Advanced Reactors. Final GEIS.*(Has been voted to rulemaking by Comm.) - Draft issuance of deliverable May 2022

Slide 10 Advanced Reactor Generic Environmental Impact Statement and Rulemaking Status Laura Willingham, Environmental Project Manager Environmental Center of Expertise, U.S. NRC

Slide 11 Rulemaking Process

  • The Proposed Rule Package is publicly available while it is with the Commission for review.

No public comments taken during the Commission review Commission will vote on publishing the proposed rule package If Commission votes to approve publication of the proposed rule package Proposed rule to be issued in the Federal Register with a 75-day public comment period.

Public meetings will be held during the comment period

Slide 12 Current Status & Rulemaking Schedule

  • Proposed rule submitted to Commission on November 2021 November 30, 2021.
  • Proposed rule published for 75-day comment May 2022 (estimated) period (if approved by Commission)

May 2023

  • Final rule submitted to Commission (estimated)
  • Final rule publication (if approved by Jan 2024 (estimated)

Commission) 3

Slide 13 Proposed Rule Package

Proposed Rule Package: SECY-21-0098: Proposed Rule: Advanced ML21222A044 Nuclear Reactor Generic Environmental Impact Statement (RIB3150-AK55; NRC-2020-0101)

Preliminary Draft Guide-4032 Package: Preliminary Draft Guide-4032 (RG ML21208A111 4.2), Preparation of Environmental Reports for Nuclear Power Stations Preliminary Draft of Interim Staff Guidance COL-ISG-30: Draft Interim Staff ML21227A005 Guidance COL-ISG-30: Advanced Reactor Applications - Environmental Considerations for Advanced Nuclear Applications that Reference the Generic Environmental Impact Statement (NUREG-2249) 4

Slide 14 Proposed Rule Package (con't)

The following documents can be found at Regulations.gov SECY paper Draft Advanced Reactor GEIS Draft Guide-4032 Draft Regulatory Analysis Draft COL-ISG-30 The Docket ID on Regulations.gov for the ANR GEIS is NRC-2020-0101.

Hit "Subscribe" to get notifications when new content is added.

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Slide 15 QUESTIONS?

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Slide 16 Implementing Nearfield Models in MACCS v4.1 for Better Nearfield Dose Calculations PRESENTED BY Dan Clayton MACCS Principal Investigator Sandia National Laboratories Advanced Reactor Stakeholder Meeting Sandia National Laboratories is a multimission January 19, 2022 laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energys National Nuclear Security Administration under contract DE-NA0003525.

SAND2022-0282 PE

Slide 17 2

Agenda Motivation and Purpose

Background

Approach

  • Nearfield Code Comparisons
  • MACCS 4.1 Enhancements and Algorithms
  • Verification and Comparison Summary

Slide 18 3

Motivation and Purpose Motivation: Resolve the technical issues with the adequacy of MACCS in the nearfield (i.e., at distances less than 500 m) that are identified in a non-Light Water Reactor (LWR) vision and strategy report that discusses computer code readiness for non-LWR applications developed by the Nuclear Regulatory Commission (NRC)

The purpose of this presentation is threefold:

  • Summarize the technical issues associated with the use of MACCS in the nearfield and approach used to resolve them
  • Alert stakeholders that improved nearfield modeling capabilities have been added to MACCS 4.1
  • Familiarize stakeholders with the improved nearfield capabilities available in MACCS 4.1

Slide 19 4

Background

MACCS 4.0 uses the general gaussian plume equation with reflective boundaries and includes models for plume meander and building wake effects based on building dimensions

  • -y 2 1 + H -z- 2 Q

C = ---------------------- exp ---------

1 2nh

- H-z- 2

- 2nh 2 y z u 2 2y exp -

2 z + exp -

2 z n = -

Previous (4.0 and earlier) versions of MACCS include only a simple model for building wake effects. The MACCS Users Guide suggests that this simple building wake model should not be used at distances closer than 500 m. This statement raised the question of whether MACCS can reliably be used to assess nearfield doses, i.e., at distances less than 500 m

Slide 20 5

Approach Identify candidate codes considered adequate for use in nearfield modeling Benchmark MACCS 4.0 nearfield results against results from candidate codes Identify model input recommendations or code updates for improved nearfield modeling Implement the code updates in MACCS 4.1 Verify that the MACCS 4.1 code updates adequately reflect the results from the candidate codes Exercise new capabilities in MACCS 4.1

Slide 21 6

Nearfield Code List Four candidate codes were Based on these rankings, QUIC, selected from the three main AERMOD, and ARCON96 were methods of atmospheric selected for comparison with MACCS transport and dispersion (ATD) 4.0 (3.11.6) in the nearfield and evaluated

  • CFD models - OpenFOAM Test cases developed varying
  • Simplified wind-field models -
  • Weather conditions QUIC
  • Building configurations (HxWxL)
  • Modified Gaussian models -

AERMOD and ARCON96

  • Power levels (heat content)

Slide 22 7

MACCS 4.0 Nearfield Comparison Results At 50 m, order from highest to lowest is ARCON96, AERMOD, QUIC, MACCS Order changes with distance Need to modify MACCS input to bound results of other codes

Slide 23 8

MACCS 4.0 Nearfield Comparison Results with Updated Inputs MACCS input modified to reflect a ground-level (1), non-buoyant (2) release (grey) bounds AERMOD and QUIC up to 1 km and ARCON96 from 200 m up to 1 km MACCS input modified to reflect a ground-level (1), non-buoyant (2),

point-source (3) release (light blue) bounds all three up to 1 km

Slide 24 9

MACCS 4.1 Enhancements Add two new capabilities in MACCS 4.1 to facilitate simulating or bounding nearfield calculations performed with other codes:

  • Implemented Ramsdell and Fosmire wake and meander algorithms used in ARCON96
  • Updated existing meander model to fully implement wake and meander model equations from US NRC Regulatory Guide 1.145 as implemented in PAVAN Maintain existing MACCS capabilities to bound results with AERMOD and QUIC

Slide 25 10 New MACCS 4.1 Algorithms Ramsdell and Fosmire meander model used in ARCON96 US NRC Regulatory Guide 1.145 meander model as implemented in PAVAN Ramsdell and Fosmire Reg. Guide 1.145

Slide 26 11 Verification-Ramsdell and Fosmire meander model Generate results comparable to those from ARCON96 with MACCS when using the Ramsdell and Fosmire meander model

Slide 27 12 Verification-US NRC Reg Guide 1.145 meander model as implemented in PAVAN Generate results comparable to those from PAVAN with MACCS when using the full US NRC Regulatory Guide 1.145 meander model

Slide 28 13 Verification-US NRC Reg Guide 1.145 meander model as implemented in MACCS 4.0 Maintain capability to bound AERMOD and QUIC results using recommended MACCS parameter choices

Slide 29 14 Model Comparisons (1/2)

When using the full US NRC Regulatory Guide 1.145 meander model, the /Q values for the test cases are higher than for the other two models The /Q values for the test cases with MACCS Ramsdell and Fosmire plume meander model are lower than the other two models except at distances of less than 200-300 m

Slide 30 15 Model Comparisons (2/2)

The three models converge with differences on the order of 5-10% at a distance of 35 km.

Slide 31 16 Summary Assessment of MACCS 4.0 ARCON96, AERMOD, and QUIC selected for comparison with MACCS 4.0 based on initial evaluation Based on the comparison, MACCS 4.0 can be used in a conservative manner at distances significantly shorter than 500 m downwind from a containment or reactor building However, the MACCS user needs to select the MACCS input parameters appropriately to generate results that are adequately conservative for a specific application Additional information available from final technical report (Clayton D.J and N.E. Bixler, Assessment of the MACCS Code Applicability for Nearfield Consequence Analysis SAND2020-2609, Sandia National Laboratories, Albuquerque, NM, February 2020, ADAMS Accession Number ML20059M032)

Slide 32 17 Summary of New MACCS 4.1 Capabilities Additional nearfield meander models are included with MACCS 4.1

  • Generate results comparable to those from ARCON96 with MACCS when using the Ramsdell and Fosmire meander model
  • Generate results comparable to those from PAVAN with MACCS when using the full US NRC Regulatory Guide 1.145 meander model
  • Maintain capability to bound AERMOD and QUIC results using recommended MACCS parameter choices Comparing the plume meander model results shows
  • When using the full US NRC Regulatory Guide 1.145 meander model, the /Q values for the test cases are higher than for the other two models
  • The /Q values for the test cases with MACCS Ramsdell and Fosmire plume meander model are lower than the other two models except at distances of less than 200-300 m
  • Beyond 1 km, the three models converge with differences on the order of 5-10% at a distance of 35 km.

MACCS 4.1 also available as Linux version (see https://maccs.sandia.gov for more information)

Additional information available from final technical report (Clayton D.J, Implementation of Additional Models into the MACCS Code for Nearfield Consequence Analysis SAND2021-6924, Sandia National Laboratories, Albuquerque, NM, June 2021)

Slide 33 18 For questions or comments, please contact:

Daniel Clayton MACCS Principal Investigator Sandia National Laboratories djclayt@sandia.gov Keith Compton Technical Monitor U.S. Nuclear Regulatory Commission Keith.Compton@nrc.gov

Slide 34 Backup slides

Slide 35 20 MACCS 4.0 Results Building and elevation effects greatly diminished at 800 m downwind Building significantly increases dispersion at short distances Dilution for stable conditions generally higher than the corresponding dilution for neutrally-stable conditions Buoyant plumes that escape building wake produce significantly lower dilution values due to fast plume rise compared with dispersion

Slide 36 21 ARCON96 Results Minimal change due to inclusion of building or elevated release within 1 km Dilution for stable conditions generally higher than the corresponding dilution for neutrally-stable conditions No plume rise model implemented; buoyant cases were not modeled

Slide 37 22 AERMOD Results Building and elevation effects greatly diminished at 500 m downwind Building significantly increases dispersion at short distances Dilution for stable conditions generally higher than the corresponding dilution for neutrally-stable conditions Minor differences due to buoyancy

Slide 38 23 QUIC Results (1/2)

Building and elevation effects greatly diminished at 1 km downwind Building significantly increases dispersion at short distances Dilution for stable conditions generally higher than the corresponding dilution for neutrally-stable conditions No straightforward way to implement buoyancy; buoyant cases were not modeled

Slide 39 24 QUIC Results (2/2)

Horizontal and vertical slices for a 4 m/s, neutrally-stable weather condition with a non-buoyant, elevated release from a 20 m x 100 m x 20 m building (Case 01)

Slide 40 25 Potential Modifications to MACCS Input

1. Specify a ground-level release, instead of a release at the height of the building
  • ARCON96 model showed little dependence on elevation of release
  • Wake-induced building downwash observed in QUIC output
  • Regulatory Guide 1.145 discusses releases less than 2.5 times building height should be modeled as ground-level releases
2. Specify no buoyancy (plume trapped in building wake)
  • AERMOD model showed little dependence on buoyancy
3. If additional conservatism needed or desired, model as a point source
  • ARCON96 model showed little dependence on building size
  • DOE approach used for collocated workers
  • If point source too bounding, use an intermediate building wake size

Slide 41 Draft Interim Staff Guidance for the Safety Review of Light-Water Power Reactor Construction Permit Applications Carolyn Lauron New Reactor Licensing Branch (NRLB)

Division of New and Renewed Licenses (DNRL)

Office of Nuclear Reactor Regulation (NRR)

Slide 42 What is the purpose of todays presentation?

To facilitate stakeholder understanding of the information contained in the construction permit interim staff guidance recently noticed in the Federal Register for comment. (86 FR 71101)

This presentation should aid in the development and submission of stakeholder written comments consistent with the instructions in the Federal Register notice.

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Slide 43 Why was the interim staff guidance developed?

  • NRC anticipates the submission of construction permit applications.
  • NRC last reviewed and issued a light-water power-reactor construction permit in the 1970s.
  • Recently, NRC reviewed and issued licenses using the one-step process in 10 CFR Part 52.
  • There are ongoing NRC activities to realign the requirements in 10 CFR Parts 50 and 52, and to develop guidance for non-light-water reactor designs.

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Slide 44 Availability of Draft ISG DNRL-ISG-2022-XX On December 14, 2021, the NRC published a notice in the Federal Register requesting comments on the draft interim staff guidance by January 28, 2022. (86 FR 71101)

The draft interim staff guidance may be found in the NRCs Agencywide Documents Access and Management System at this link: ML21165A157 4

Slide 45 Scope of Draft ISG DNRL-ISG-2022-XX The scope of the interim staff guidance is the safety review of light-water power-reactor construction permit applications.

The interim staff guidance supplements the existing review guidance for light-water power-reactor applications found in NUREG-0800.

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Slide 46 Parts of Draft ISG DNRL-ISG-2022-XX

  • Main Body of Document

- Purpose, Background, Rationale, Applicability

- Guidance

- Implementation

- Backfitting and Issue Finality Discussion, Congressional Review Act

- Final Resolution

- References

  • Appendix 6

Slide 47 Guidance in Draft ISG DNRL-ISG-2022-XX Guidance Subsections

  • Requirements for a Power Reactor Construction Permit Application
  • Light-Water-Reactor Safety Review Guidance
  • Special Topics

- Relationship between the Construction Permit and Operating License reviews

- Purposes and benefits of preapplication activities

- Lessons learned from recently issued construction permits

- Approach for reviewing concurrent license applications and applications incorporating prior NRC approvals

- Potential effect of ongoing regulatory activities on construction permit reviews and

- Licensing requirements for byproduct, source, or special nuclear material.

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Slide 48 Appendix to Draft ISG DNRL-ISG-2022-XX

- Reiterates the context, expected engagement, and review approach

- Clarifies guidance for selected safety-related topics

  • Not intended to include all topics expected and reviewed in a construction permit application.

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Slide 49 Clarifications in Appendix to Draft ISG DNRL-ISG-2022-XX Select topics discussed:

- Siting

- Radiological Consequence Analyses

- Transient and Accident Analyses

- Structures, Systems, and Components

- Protective Coatings Systems

- Instrumentation and Control

- Electrical System Design and

- Radioactive Waste Management 9

Slide 50 Submitting Comments on DNRL-ISG-2022-XX Link to Federal Register notice: 86 FR 71101 Two ways to submit comments:

1. Federal Rulemaking Website: Go to https://www.regulations.gov/

and search for Docket ID NRC-2021-0162.

- Address questions about Docket IDs in Regulations.gov to Stacy Schumann; telephone: 301-415-0624; email:

Stacy.Schumann@nrc.gov

- For technical questions, contact Carolyn Lauron, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-2736, email: Carolyn.Lauron@nrc.gov

2. Mail comments to: Office of Administration, Mail Stop: TWFN A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Program Management, Announcements and Editing Staff.

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Slide 51 Questions and Answers

Slide 52 Advanced Reactor Stakeholder Public Meeting Break Meeting will resume at 1pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 323 588 045#

Slide 53 NUREG/CR-7289, Nuclear Data Assessment for Advanced Reactors Advanced Reactor Stakeholder Meeting January 19, 2022 1

Slide 54 NUREG/CR-7289 ORNL/TM-2021/2002

  • Oak Ridge National Laboratory (ORNL)

- F. Bostelmann

- G. Ilas

- C. Celik

- A.M. Holcomb

- W.A. Wieselquist 2

Slide 55 Motivation/Background 3

Slide 56 Commercial Light Water Reactor Approach to Reactor Physics/Nuclear Design Data Start with simplified (e.g., ENDF/B-VII.1) geometry and detailed Cross Section Processing energy group structure, (e.g., AMPX, NJOY) End with simplified group Output: 100s of energy groups structure and 3D geometry 1-D Pin Cell Apply biases and (e.g., SCALE, CASMO)

Output: 20-100 energy groups uncertainties to calculated quantities of interest 2-D Assembly (e.g., SCALE, CASMO)

(QOIs):

Output: 2-4 Energy Groups, Reactivity balance Cross Section and Discontinuity Factors Shutdown margin Feedback coefficients 3-D Whole Core Simulator (e.g., PARCS, SIMULATE) Power distribution 4

Slide 57 Commercial Light Water Reactor Approach to Reactor Physics/Nuclear Design Data Start with simplified (e.g., ENDF/B-VII.1) geometry and detailed Cross Section Processing energy group structure, (e.g., AMPX, NJOY) End with simplified group Output: 100s of energy groups structure and 3D geometry 1-D Pin Cell Apply biases and (e.g., SCALE, CASMO)

Output: 20-100 energy groups uncertainties to calculated quantities of interest 2-D Assembly (e.g., SCALE, CASMO)

(QOIs):

Output: 2-4 Energy Groups, Reactivity balance Cross Section and Discontinuity Factors Shutdown margin Feedback coefficients 3-D Whole Core Simulator (e.g., PARCS, SIMULATE) Power distribution Emphasized during safety review 5

Slide 58 Impact of Data Uncertainty

  • QOIs verified via (1) startup physics testing, and (2) surveillance requirements
  • Advanced Reactor examples*:

- Changes in graphite data from ENDF/B-VII.0 to B-VII.1 (capture cross section) had a 1% k/k impact

- No data for FLiBe/FLiNak thermal scattering, possible 2% k/k impact for thermal spectrum

  • Uncertainties in nuclear data/physics modeling has the potential to adversely impact reactor operation
  • Based on 2018 work performed at ORNL and available literature in 2019 6

Slide 59 Data Uncertainty and Licensing

  • NRC review of nuclear design expected to emphasize uncertainty management

- Appropriate application/justification of design margin into QOIs

- Uncertainty update methodologies

- Commitment to measurements/surveillances to verify design margin

- Commitment to required actions in the event that measurements/surveillances fail to meet acceptance criteria 7

Slide 60 Data Challenges for Advanced Reactor Licensing

  • Confidence in current nuclear data needs to be confirmed for non-LWRs:

- Unique materials and neutron energy spectra

- Nontraditional fuel forms

- Limited integral validation data

  • Nuclear data expertise:

- Gaps in current nuclear data libraries?

- Impact of gaps/uncertainties on QOIs?

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Slide 61 Overview of NUREG/CR-7289, Nuclear Data Assessment for Advanced Reactors 9

Slide 62 Technologies Considered Molten High Chloride Fast Temperature Spectrum Gas Reactor Reactor Fluoride Salt-Cooled High Heat Pipe Temperature Microreactor Reactor Graphite Sodium-Moderated Cooled Fast Molten Salt Reactor Reactor 10

Slide 63 Approach

  • 4 Phases:

- Phase 1 and 2: Identify and assess key data impacting reactivity in non-LWRs based on literature review

- Phase 3: Identify relevant benchmarks

- Phase 4: Assess the impact of nuclear data uncertainty through propagation to key QOIs

  • Sensitivity and uncertainty analysis (performed using SCALE 6.3) ADAMS Accession Nos.

ML20274A052 and ML21125A256 11

Slide 64 Sensitivity and Uncertainty Analysis Reactor technology Selected benchmarka Type High Temperature Gas Reactor HTR-10 Experiment Fluoride Salt Cooled High UC Berkeley Mark1 PB-FHR Computational Temperature Reactor benchmark Graphite-moderated Molten Salt MSRE Experiment Reactor Heat Pipe Microreactor (metal- INL Megapower Design Ab Computational fueled) benchmark EBR-II Experiment Sodium Cooled Fast Reactor (metal and oxide fueled) ABR-1000 Computational benchmark a Although Fast Spectrum Molten Salt Reactors were identified as a relevant reactor concept, a concept with details sufficient for modeling could not be found in the open literature.

b The original design contains oxide fuel. However, for this project, metal fuel was assumed.

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Slide 65 Sensitivity and Uncertainty Analysis

  • Analyses were performed using ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0
  • Sensitivity coefficients:

o ,, =  ; ( is the QOI, and ,

is the data) o NUREG/CR-7289 reports sensitivity coefficients using ENDF/B-VII.1 (results using ENDF/B-VII.0 and ENDF/B-VIII.0 obtained values that are very close to ENDF/B-VII.1) 13

Slide 66 Results and Key Nuclear Data (Subset of results from HTR-10 benchmark)

Nominal Results Nominal Reactivity Impacts for QOIs QOIs ENDF/B-VII.0 ENDF/B-VII.1 ENDF/B-VIII.0 .

Fuel temperature -243 +/- 22 -241 +/- 25 -222 +/- 25 3 +/- 33 19 +/- 36 Pebble gr. density 1182 +/- 23 1175 +/- 23 1201 +/- 27 -8 +/- 32 26 +/- 35 Pebble gr. impurities -602 +/- 23 -623 +/- 23 -588 +/- 25 -21 +/- 32 35 +/- 34 Pebble gr. temperature -1948 +/- 23 -1960 +/- 22 -1701 +/- 25 -11 +/- 32 259 +/- 33 Structural gr. density 546 +/- 25 504 +/- 22 543 +/- 24 -43 +/- 33 40 +/- 32 Structural gr. impurities -3947 +/- 26 -3877 +/- 25 -3807 +/- 25 70 +/- 36 70 +/- 35 Structural gr. temperature 780 +/- 24 783 +/- 22 798 +/- 24 4 +/- 33 14 +/- 33 14

Slide 67 Results and Key Nuclear Data (Subset of results from HTR-10 benchmark)

Sensitivity Analysis Results Key Nuclear Data Impacting Pebble Graphite Temperature Feedback Sensitivity Sensitivity Nuclide Reaction Nuclide Reaction (reducing negative ) (increasing negative )

u-235 fission 1.196e+00 +/- 6.070e-03 b-10 n, -9.273e-02 +/- 1.440e-03 u-235 9.976e-01 +/- 6.552e-04 u-238 n, -3.655e-02 +/- 1.764e-03 s-28 elastic 9.796e-03 +/- 6.801e-03 n-14 n,p -5.147e-03 +/- 1.908e-04 c elastic 9.083e-03 +/- 9.656e-03 u-235 elastic -3.560e-03 +/- 3.272e-03 u-238 elastic 8.487e-03 +/- 9.148e-03 si-28 n, -4.577e-04 +/- 2.769e-05 o-16 elastic 6.737e-03 +/- 8.590e-03 graphite n, -8.149e-04 +/- 2.176e-04 u-235 n, 6.585e-03 +/- 1.145e-03 si-28 n,n -3.930e-04 +/- 4.912e-04 n-14 elastic 6.281e-03 +/- 6.051e-03 n-14 n, -2.084e-04 +/- 7.821e-06 graphite n,n 4.702e-03 +/- 2.311e-03 ar-40 elastic -1.988e-04 +/- 1.457e-04 u-238 nu-fission 2.402e-03 +/- 6.552e-04 n-14 n, -4.236e-05 +/- 1.867e-06 15

Slide 68 Results and Key Nuclear Data (Subset of results from HTR-10 benchmark)

Uncertainty Analysis Results Uncertainty in QOIs due to nuclear data QOIs ENDF/B-VII.0 ENDF/B-VII.1 ENDF/B-VIII.0 .

keff 0.607% 0.668% 0.690% 10.1% 3.3%

Fuel temperature 1.124% 1.192% 1.030% 6.1% -13.6%

Pebble gr. density 0.667% 0.848% 0.618% 27.1% -27.1%

Pebble gr. impurities 0.639% 0.749% 1.126% 17.2% 50.3%

Pebble gr. temperature 0.694% 0.753% 0.972% 8.4% 29.1%

Structural gr. density 0.873% 0.952% 0.820% 9.1% -13.9%

Structural gr. impurities 0.921% 1.109% 0.990% 20.3% -10.7%

Structural gr. temperature 0.998% 1.135% 0.920% 13.7% -18.9%

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Slide 69 Results and Key Nuclear Data (Subset of results from HTR-10 benchmark)

Uncertainty Analysis Results Top Nuclear Data Contributors to Multiplication Factor Uncertainty 17

Slide 70 Conclusions

  • Major data gaps from the libraries:

- Thermal scattering kernel for molten salts

- Uncertainty for thermal scattering (e.g., graphite)

- Angular scattering uncertainty for fast spectrum reactors

  • In general, the most important reactions were shown to be:

- Neutron multiplicity, fission and radiative capture cross sections of fissile isotopes (e.g., U-235)

- Radiative capture cross sections of fertile isotopes (e.g., U-238)

  • Other significant contributors:

- Capture cross sections of fission products*

- Capture cross sections of neutron absorbing material (e.g., Gd or B)

- Scattering reactions with the coolant and structural materials for fast spectrum systems

  • For Molten Salt Reactors, in particular, additional neutron capture reactions such as (n,p) and (n,t) for salt components (e.g., Li and Cl) are significant contributors to the reactivity balance.
  • Results of study with respect to depletion/burnup are limited due to (1) unavailability of benchmarks and relevant data, and (2) capability not 18 currently available to fully propagate uncertainty in depletion analyses.

Slide 71 Conclusions

  • Calculated uncertainty in reactivity balance due to nuclear data is generally greater than what is used in LWR nuclear design.
  • Large uncertainties that are not considered relevant in LWRs studies were found to be significant for several advanced reactor systems:

- All fast spectrum systems impacted by larger uncertainties in U-238 inelastic scattering and U-235 radiative capture at higher energies

- A large uncertainty in the Li-7 capture cross section causes larger uncertainty in all QOIs for systems that use lithium as part of a salt coolant.

  • No performance differences observed between the different libraries (i.e.,

ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0)

- One exception being ENDF/B-VII.1 and ENDF/B-III.0 perform better for high temperature gas reactors because of the adjusted carbon capture cross section.

  • NUREG/CR-7278 provides useful insight regarding nuclear design margins to accommodate gaps and uncertainty in the nuclear data.

19

Slide 72 1

SCALE and MELCOR development and application for non-LWRs Advanced Reactor Stakeholder Meeting January 19, 2022

Slide 73 NRC strategy for severe accident analysis

Slide 74 3 SCALE MELCOR Non-LWR Demonstration Project - objectives Understand severe accident behavior and provide insights for regulatory guidance Facilitate dialogue on staffs approach for source term Demonstrate use of SCALE and MELCOR

  • Identify accident characteristics and uncertainties affecting source term
  • Develop publicly available input models for representative designs

Slide 75 SCALE MELCOR Non-LWR Demonstration Project - approach

1. Use SCALE to estimate core decay heat, radionuclide inventory, reactivity coefficients
2. Build MELCOR full-plant input model
3. Select accident scenarios
4. Perform MELCOR simulations for the selected scenarios and debug
  • Base case
  • Sensitivity cases
5. Public workshops to discuss the modeling and sample results 4

Slide 76 5

Slide 77 6 Molten Salt Reactor Experiment (MSRE)

Slide 78 7 Advanced Burner Test Reactor (ABTR)

Slide 79 SCALE analysis approach for MSR 3 models run in an iterative fashion to predict nuclide inventory, decay heat, Time snapshot Simplified and reactivity feedback coefficients at Core+Loop+Offgas 1D loop selected point in the operating cycle Time snapshot

  • predicts core neutron flux at a point Core+

Loop in the operating cycle Simplified core + loop + offgas

  • predicts primary-system-average Xe, He, Kr, H nuclide inventory over time 1D loop model Offgas
  • predicts nuclide inventory in each System section of the loop

Slide 80 Time snapshot model

  • Predicts 3D flux profiles via axial/radial discretization
  • Currently using 30 axial levels, 7 radial rings
  • Investigating sensitivity of reactivity feedback to various modeling parameters Radial flux distribution fuel fuel graphite fuel fuel SCALE 3D full Cross section of core MSRE model unit cell Axial flux distribution 9

Slide 81 1D loop model

  • Predicts nuclide inventory in each Short-lived 2

3 4

section of the loop 5

nuclide (I-137, 6

t1/2=24.5s) as a 7 function of 8

location in the Core (1)

  • As fuel salt loop Loop 9 travels the loop

- Long-lived*

nuclides will slowly accumulate/be removed*

Short-lived (same as solid fuel) nuclide as a function of

- Short-lived* nuclides will time at the oscillate about an equilibrium bottom of the core (zone 1)

  • relative to the loop transit time

(~25 s for MSRE)

Slide 82 SCALE analysis approach for SFR

  • Development of fully heterogeneous full-core model for continuous-energy Monte Carlo calculation
  • Power-profile calculation via axial and radial discretization of fuel region
  • Full-core depletion calculation to obtain core inventory at end of cycle
  • Reactivity effect calculations via direct perturbations: coolant density, fuel temperature, fuel axial expansion, radial core expansion, etc.

ABTR model with individual assembly definitions and SCALE ABTR model corresponding power map 11

Slide 83 MELCOR Modeling Scope Thermal hydraulics SCALE (ORNL)

Fuel Reactivity thermal-Effects mechanical response Fission product Core release and degradation transport Ex-vessel damage progression 12

Slide 84 MELCOR Non-LWR Modeling Hydrodynamic modeling

  • Generalized working fluid treatment
  • Conduction heat transfer within working fluids (under development)
  • Generalized convection and flow models to capture flow through new core geometries (e.g., pebble beds)

Core models

  • TRISO pebble and compact core components
  • Heat pipe reactor core component
  • Graphite oxidation
  • Intercell and intracell conduction
  • Fast reactor core degradation (under development)

Fission product release

  • Generalized release modeling for metallic fuels
  • Radionuclide transport and release from TRISO particles, pebbles and compacts
  • Generalized Radionuclide Transport and Retention (GRTR) model (under development)

Simplified neutronic modeling

  • Solid fuel core point kinetics
  • Fluid point kinetics (liquid-fueled molten salt reactors) 13

Slide 85 TRISO Radionuclide Release Modeling Recent failures - particles failing within latest time-step (burst release, diffusion release in time-step)

Previous failures - particles failing on a previous time-step (time history of diffusion release)

Contamination and recoil Diffusion from intact TRISO Transfer to Released failed to the TRISO matrix Distribution calculated from Release from TRISO failure diffusion model Diffusion Released to the matrix Diffusion recoil Failing Intact TRISO Intact TRISO Transition from Intact-to-failed Release from recoil Failed failed TRISO TRISO Contamination (Modified Failed Recoil fission source Booth)

TRISO Diffusion 14

Slide 86 MELCOR Generalized Radionuclide Transport and Retention (GRTR) Model Model Scope Uses 5 radionuclide physico-chemical forms in liquid pool Soluble fission products Insoluble fission products suspended in working fluid Insoluble fission products deposited on structures Insoluble fission products at liquid-gas interface Fission product gases Generalized Gibbs Energy Minimization approach Fission product solubility Fission product vapor pressure Model generically applies to range of non-LWR working fluids Molten salt systems Liquid metal systems Radionuclides grouped into forms found in the Molten Salt Reactor Experiment 15

Slide 87 MELCOR Generalized Radionuclide Transport and Retention (GRTR) - States and State Transitions Radionuclides characterized in terms of Isotopic state

  • Fission product decay Distribution of fission products in reactor system
  • Hydrodynamic flows moving fission products within system Physico-chemical form and ability of fission products to be transported out of the liquid
  • Deposition on structures from the liquid
  • Vaporization into gas atmospheres from the liquid
  • Attachment to gas bubbles
  • Aerosolization of fission products into atmosphere above the liquid via bursting of bubbles Note: MELCOR considers soluble, bulk colloid, interfacial colloid, and vapors as distinct chemical states 16

Slide 88 Cesium Vaporization from Molten Salt - FHR Example Fission product thermochemistry modeling sample demonstration Cesium Behavior

  • Exercise machinery 1.E-01 1300
  • Focuses on Cs and CsF release from salt pool Total released from pebbles Total in the liquid
  • Thermochimica Gibbs Energy Minimizer 1.E-02 Vaporized from the liquid 1200
  • Utilizing vapor phase data for CsF* Core Fluid Temperature Fraction of initial invenory (-)

1.E-03 1100 Temperature (deg-C)

Demonstration calculation for LOCA 1.E-04 1000 sequence

  • No core uncovery through 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.E-05 900 Model exhibits Cs and CsF vaporization to 1.E-06 800 gas space at elevated salt temperatures 1.E-07 700 1.E-08 600 0 6 12 18 24 Cs Transport Pathway Time (hr)

Overlying Fuel Pebbles Molten Salt Gas Atmosphere

  • With modifications by Ontario Tech. 17

Slide 89 Point Kinetics Modeling Some accidents may involve reactivity Core Reactivities feedbacks 1.0 300 For non-LWRs, MELCOR uses a point kinetics models 0.5 250 Feedback models 0.0 200

  • User-specified external input Reactivity ($) Power (MW)
  • Doppler -0.5 Fuel Temperature 150
  • Fuel and moderator density Molten Salt Inner Reflector
  • Flow reactivity feedback effects integrated -1.0 Outer Reflector 100 Moderator into the equation set Xenon

-1.5 Total Reactivity 50 Power FHR example calculation using MELCOR point kinetics model -2.0 1 10 100 1000 0

Time (sec) 18

Slide 90 Point Kinetics Modeling (MSR)

Extended static point kinetic equations to capture motion of delayed precursors through the reactor system 250 Validated against MSRE zero-Compensating Control System Reactivity [pcm]

power flow experiments 200 150 Guo Code MSRE Data 100 MELCOR 50 0

0 10 20 30 40 50 60 70 Time [s]

19

Slide 91 NRC Non-LWR Vision and Strategy, Volume 5 20 Radionuclide Characterization, Criticality, Shielding, and Transport in the Nuclear Fuel Cycle Project goal: Demonstration of capabilities to simulate accident scenarios during the fuel cycle with MELCOR and SCALE for HTGR, SFR, MSR, HPR, FHR Current effort is the development of the project plan:

Determine boundary conditions for each stage of the fuel cycle Identify potential hazards and accident scenarios for each stage of the fuel cycle From these, select accident scenarios for SCALE/MELCOR to simulate Challenges encountered:

Some stages of the fuel cycle are not yet developed Many documents are proprietary (e.g., safety analysis reports)

Current status:

HTGR fuel cycle developed and discussed between ORNL/SNL/NRC MSR and SFR fuel cycle discussions scheduled for end of January/early February HTGR fuel cycle

Slide 92 Advanced Reactor Stakeholder Public Meeting Break Meeting will resume in 10 minutes Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 323 588 045#

Slide 93 NRC Activities on Advanced Manufacturing Technologies (AMTs)

Matthew Hiser NRC Office of Nuclear Regulatory Research January 19, 2022 Periodic Advanced Reactor Stakeholder Meeting

Slide 94 Advanced Manufacturing Technologies

  • Techniques and material processing methods that have not been:

- Traditionally used in the U.S. nuclear industry

- Formally standardized/codified by the nuclear industry

  • Key AMTs based on industry interest:

- Laser Powder Bed Fusion (LPBF)

- Directed Energy Deposition (DED)

- Electron Beam Welding (EBW)

- Powder Metallurgy - Hot Isostatic Pressing (PM-HIP)

- Cold Spray (CS) 2

Slide 95 Laser Powder Bed Fusion

  • Process:

- Uses laser to melt or fuse powder particles together within a bed of powder

- Generally most advantageous for more complex geometries

  • Potential LWR Applications Schematic of LPBF process

- Smaller Class 1, 2 and 3 components, fuel hardware, small internals 3

https://www.osti.gov/pages/servlets/purl/1437906

Slide 96 First US Application of Additive Manufacturing

  • Thimble Plugging Device

- Installed in March 2020 in Byron Unit 1

- 316L stainless steel -LPBF

- Very low safety significant component (Non ASME B&PV Code class)

- PWR environment with irradiation

- Installation done without prior NRC approval under 10 CFR 50.59 https://www.neimagazine.com/news/newswestinghouse-produces-3d-printed-component- 4 for-us-nuclear-plant-7911951

Slide 97 Second US Application of Additive Manufacturing

  • Channel Fastener

- Installed in April 2021 at Browns Ferry Unit 2

- 316L stainless steel - LPBF

- Non ASME B&PV Code Class

- BWR environment with irradiation

- Installation done without prior NRC approval under 10 CFR 50.59 https://www.ornl.gov/news/additively-manufactured-components-ornl-headed-tva-nuclear- 5 reactor?utm_source=miragenews&utm_medium=miragenews&utm_campaign=news

Slide 98 Directed Energy Deposition

  • Process:

- Wire or powder fed through nozzle into laser or electron beam

- Fundamentally welding using robotics/

computer controls

  • Potential Applications Schematic of DED process

- Similar to LPBF, although larger components possible due to faster production and greater build chamber volumes https://www.osti.gov/pages/servlets/purl/1437906 6

Slide 99 Powder Metallurgy - Hot Isostatic Pressing (PM-HIP)

  • Process:

- Metal powder is encapsulated in a form mirroring the desired part

- The encapsulated powder is exposed to high temperature and pressure, densifying the powder and producing a uniform microstructure

- After densification, the capsule is removed, yielding a near-net shape component where final machining and inspection can be performed

  • Potential Applications

- All sizes of Class 1, 2 and 3 components and reactor internals

- EPRI / DOE focused on use with electron beam welding to fabricate NuScale reactor vessel 7

Slide 100 Electron Beam Welding

  • Process:

- Fusion welding process that uses a beam of high-velocity electrons to join materials

- Single pass welding without filler metal

- Welding process can be completed much more quickly due to deep penetration

  • Potential Applications

- For welding medium and large components, such as NuScale upper head 8

Slide 101 Cold Spray

  • Process:

- Powder is sprayed at supersonic velocities onto a metal surface and forms a bond with the part

- This can be used to repair existing parts or as a mitigation process

  • Potential Applications Schematic of cold spray process*

- Mitigation or repair of potential chloride-induced stress corrosion cracking (CISCC) in spent fuel canisters

- Mitigation or repair of stress corrosion cracking (SCC) in reactor applications https://www.army.mil/article/148465/army_researchers_develop_cold_spray_system_transition_to_industry 9

Slide 102 Industry and Research Activities

  • Variety of stakeholders are working towards more widespread use in both existing and future nuclear applications

- Vendors and licensees/applicants

  • Identifying candidate applications
  • Developing technical basis for gaining regulatory acceptance

- Nuclear Energy Institute - Developed roadmap to understand industry needs/interests and assist with regulatory acceptance

- Electric Power Research Institute - Developing techniques for large components in small modular reactors, developed data package for 316L L-PBF ASME draft Code case

- US Department of Energy - Performing basic and applied research and technology development to support AMT implementation 10

Slide 103 Codes and Standards

  • Codes and Standards Organizations (eg ASTM, ASME) - addressing standardization gaps, Code Cases (PM-HIP, LPBF)

- ASME Special Working Group -

  • Developing guidelines for use of additive manufacturing (AM), Criteria for Pressure Retaining Metallic Components Using Additive Manufacturing. Was published as an ASME Pressure Technology Book
  • 316L L-PBF Data Package and Code Case under development

- ASME Task Group on AM for High Temperature Applications

  • Developing Code actions for incorporating AM materials/components in ASME Section III, Division 5 (high temperature reactors) for elevated temperature nuclear construction

- ASME PM-HIP Code Case approved for use by US NRC

  • Code Case N-834 allows use of ASTM A988/A988M Standard Specification for Hot Isostatically-Pressed Stainless Steel Flanges, Fittings, Valves, and Parts for High Temperature Service in Section III, Division 1 Class 1 components
  • October 2019 - RG 1.84, Revision 38 approved this Code Case as acceptable for use without conditions 11

Slide 104 NRC Action Plan

  • NRC activities related to AMTs have been organized and planned through the AMT action plan with the following objectives:

- Assess the safety significant differences between AMTs and traditional manufacturing processes, from a performance-based perspective.

- Prepare the NRC staff to address industry implementation of AMT-fabricated components through the 10 CFR 50.59 process.

- Identify and address AMT characteristics pertinent to safety, from a risk-informed and performance-based perspective, that are not managed or addressed by codes, standards, regulations, etc.

- Provide guidance and tools for review consistency, communication, and knowledge management for the efforts associated with AMT reviews.

- Provide transparency to stakeholders on the process for AMT approvals.

Slide 105 Action Plan - Rev. 1 Tasks

  • Task 1 - Technical Preparedness

- Technical information, knowledge and tools to prepare NRC staff to review AMT applications

  • Task 2 - Regulatory Preparedness

- Regulatory guidance and tools to prepare staff for efficient and effective review of AMT-fabricated components submitted to the NRC for review and approval

  • Task 3 - Communications and Knowledge Management

- Integration of information from external organizations into the NRC staff knowledge base for informed regulatory decision-making

- External interactions and knowledge sharing, i.e. AMT Workshop (held in Dec. 2020) 13

Slide 106 Task 1 Technical Preparedness Activities

  • Subtask 1A: AMT Processes under Consideration

- Perform a technical assessment of multiple selected AMTs of interest

- Gap assessment for each selected AMTs vs traditional manufacturing techniques

- Technical letter report and technical assessment for each AMT: LPBF - ML20351A292

  • Subtask 1B: NDE Gap Assessment

- Literature survey of the current state of the art of non-destructive examination (NDE) of components made using advanced manufactured technologies (AMTs) (ML20349A012).

  • Subtask 1C: Microstructural and Modeling

- Evaluate modeling and simulation tools used to predict the initial microstructure, material properties and component integrity of AMT components

- Identify existing gaps and challenges that are unique to AMT compared to conventional manufacturing processes:

Slide 107 Task 2 - Regulatory Preparedness Activities

- Provide guidance and support to regional inspectors regarding AMTs implemented under quality assurance and 50.59 programs. Complete: ML21155A043

  • Subtask 2B: Assessment of Regulatory Guidance

- Assess whether any regulatory guidance needs to be updated or created to clarify the process for reviewing submittals with AMT components. Complete: ML20233A693

  • Subtask 2C: AMT Guidelines Document

- Develop a report which describes the generic technical information to be addressed in AMT submissions.

Technology specific guidelines are also being developed.

- Public meeting held on September 16, 2021 to discuss Draft AMT Review Guidelines ML21074A037 and Draft Guidelines Document for AM -LPBF ML21074A040 15

Slide 108 NRC AMT Guidelines Development AMT-Specific (Initial 5 AMTs) Generic

  • A Technical Letter Report (TLR) is produced for each of the initial five AMTs Technical Regulatory Guidelines
  • Provides technical basis information and gap analysis (Subtask 1A)

Technical Technical (draft for FRN public comment)

Draft

  • Written by NRC contractor (to date, DOE labs) Letter Report Assessment LPBF Guidelines Document LPBF LPBF ML20351A292 ML20351A292 ML21074A040
  • A technical assessment (TA) is produced for each TLR by Technical Technical Draft Letter Report NRC staff which provides the NRC staff perspective on Assessment Guidelines L-DED L-DED Document ML20233A693 key aspects of the AMT for safety and component ML20233A693 L-DED Subtask 2C Final performance Technical Technical Draft Draft AMT Letter Report Assessment Guidelines Review Guidance Cold Spray Cold Spray Document ML21263A105 Guidelines for Initial ML21263A105 Cold Spray ML21074A037 AMTs
  • A draft guidelines document (DGD), informed by the TA Technical Technical Draft and TLR, will be generated by the NRC staff for each AMT. Letter Report Assessment Guidelines Document
  • The AMT-specific DGDs accompany and align with the PM-HIP PM-HIP PM-HIP Expected to generic Advanced Manufacturing Technologies Review Technical Technical Draft be developed Legend later after Guidelines Contractor-developed Letter Report Assessment Guidelines EBW EBW Document EBW DOE-EPRI NRC Staff-developed demo project 16

Slide 109 Communications and KM Activities

  • Subtask 3A: Internal Interactions

- Internal coordination with NRC staff in other areas (e.g., advanced reactors, dry storage, fuels)

  • Subtask 3B: External Interactions

- Engagement with codes and standards, industry, research, international

  • Subtask 3C: Knowledge Management

- Seminars, public meetings, training, knowledge capture tools

  • Subtask 3D: Public Workshop

- RIL 2021-03: Part 1 Part 2

  • Subtask 3E: AMT Materials Information Course

- Internal NRC staff training

- Six seminars to date on a variety of topics 17

Slide 110 Status of Deliverables - Task 1 Subtask Actions/Deliverables Status Additive Manufacturing (AM) - Laser Powder Bed Complete - ML20351A292 Fusion AM - Directed Energy Deposition (DED) Complete - ML20233A693 1A AMT processes under consideration Cold Spray Complete - ML21263A105 Powder Metallurgy (PM) - Hot Isostatic Pressing (HIP) Draft report under NRC review Electron Beam (EB) welding Draft report under NRC review 1B Inspection and NDE PNNL NDE gap analysis Complete - ML20349A012 ANL M&S gap analysis to predict microstructure Complete - ML20269A301 1C Modeling and Simulation of Microstructure Complete - ML20350B550 ANL M&S gap analysis to predict material performance 18

Slide 111 Status of Deliverables - Tasks 2 and 3 Subtask Actions / Deliverables Status Finalize document incorporating feedback from Regional staff 2A 50.59 process Complete - ML21200A222 regarding the 10 CFR 50.59 process 2B Assessment of regulatory guidance Path forward on guidance development or modification Complete - ML20233A693 Public meeting on guidance concept / framework Public meeting held on July 30, 2020 - summary:

ML20240A077 Develop a document that describes the generic technical Public meeting held on September 16, 2021 to 2C AMT Guidance Document information to be addressed in AMT submittals. discuss:

ML21074A037 - Draft AMT Review Guidelines Public meeting to discuss draft document ML21074A040 - Draft Guidelines Document for AM -

LPBF Continued communication with NRC staff, vendors, licensees and 3A/3B External/ Internal Interactions Ongoing as needed EPRI for future AMTs 3C Knowledge Management Plan Develop Knowledge Management Plan Complete - internal Complete - summary: ML20357B071 3D Workshop Hold Public Workshop RIL: Part 1 Part 2 3E Material Information course Training course and course materials First 6 seminars complete - internal 19

Slide 112 Path Forward

  • Complete remaining activities under Rev. 1 AMT Action Plan:

- EBW and PM-HIP technical report and assessment

- L-DED and Cold spray DGDs

  • Plan and initiate future work likely focused on:

- Additional AMTs

- In-process NDE and digital data for qualification

- AMT guidance development

- Knowledge management and staff training on AMTs 20

Slide 113 Future Meeting Planning

  • The next periodic stakeholder meeting is scheduled for March 16, 2022.
  • If you have suggested topics, please reach out to Prosanta.Chowdhury@nrc.gov.
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