ML22014A256

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

Slide 1

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 2

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

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) and Rulemaking Activities NRC 10:30 - 11:15 am Implementing Near-field Models in MACCS v4.1 for Better Near-field Dose Calculations NRC/SNL 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 4

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 5

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 (update from v1 to v2) v1 v2 Reference plant model for Sodium-Cooled Fast Reactor (update from v1 to v2) v1 v2 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 Code Assessment Report Volume 5 (Fuel Cycle Analysis) 1/7/22 Rulemaking 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 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 Slide 6

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/details#advSumISRA 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 Develop Environmental ISG for Micro Reactors - item complete and no longer being tracked - removed Slide 7

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/details#advSumISRA 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 New item: Revise Regulatory Guide (RG) 4.7 to implement SRM-SECY-20-0045 (SRM not issued yet)

Slide 8

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/details#advSumISRA 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 Emergency Preparedness Requirements for Small Modular Reactors and Other New Technologies.(NEIMA Section 103(a)(2)) - OEDO concurred and sent the package (SECY-22-0001) to the Commission on December 30, which is now with the Commission for their review and approval Slide 9

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

2 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

3 Current Status & Rulemaking Schedule November 2021

  • Proposed rule submitted to Commission on November 30, 2021.

May 2022 (estimated)

  • Proposed rule published for 75-day comment period (if approved by Commission)

May 2023 (estimated)

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

Slide 12

4 Proposed Rule Package

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

ML21222A044 Preliminary Draft Guide-4032 Package: Preliminary Draft Guide-4032 (RG 4.2), Preparation of Environmental Reports for Nuclear Power Stations ML21208A111 Preliminary Draft of Interim Staff Guidance COL-ISG-30: Draft Interim Staff Guidance COL-ISG-30: Advanced Reactor Applications - Environmental Considerations for Advanced Nuclear Applications that Reference the Generic Environmental Impact Statement (NUREG-2249)

ML21227A005 Slide 13

5 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.

Slide 14

6 QUESTIONS?

Slide 15

P R E S E N T E D B Y Sandia National Laboratories is a multimission 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 Implementing Nearfield Models in MACCS v4.1 for Better Nearfield Dose Calculations Dan Clayton MACCS Principal Investigator Sandia National Laboratories Advanced Reactor Stakeholder Meeting January 19, 2022 Slide 16

Agenda Motivation and Purpose

Background

Approach

  • Nearfield Code Comparisons
  • MACCS 4.1 Enhancements and Algorithms
  • Verification and Comparison Summary 2

Slide 17

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 3

Slide 18

=

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 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 4

C Q*

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2 1

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Slide 19

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 5

Slide 20

Nearfield Code List Four candidate codes were selected from the three main methods of atmospheric transport and dispersion (ATD) in the nearfield and evaluated

  • CFD models - OpenFOAM
  • Simplified wind-field models -

QUIC

  • Modified Gaussian models -

AERMOD and ARCON96 6

Based on these rankings, QUIC, AERMOD, and ARCON96 were selected for comparison with MACCS 4.0 (3.11.6)

Test cases developed varying

  • Weather conditions
  • Building configurations (HxWxL)
  • Power levels (heat content)

Slide 21

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 7

Slide 22

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 8

Slide 23

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 9

Slide 24

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 10 Reg. Guide 1.145 Ramsdell and Fosmire Slide 25

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

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

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

Model Comparisons (1/2) 14 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 29

Model Comparisons (2/2) 15 The three models converge with differences on the order of 5-10% at a distance of 35 km.

Slide 30

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 16 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 31

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) 17 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 32

Daniel Clayton MACCS Principal Investigator Sandia National Laboratories djclayt@sandia.gov Keith Compton Technical Monitor U.S. Nuclear Regulatory Commission Keith.Compton@nrc.gov 18 For questions or comments, please contact:

Slide 33

Backup slides Slide 34

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 20 Slide 35

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 21 Slide 36

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 22 Slide 37

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 23 Slide 38

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) 24 Slide 39

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 25 Slide 40

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 41

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.

2 Slide 42

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.

3 Slide 43

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 44

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.

5 Slide 45

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 46

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.

7 Slide 47

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.

8 Slide 48

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 49

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.

10 Slide 50

Questions and Answers Slide 51

Break Meeting will resume at 1pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 323 588 045#

Advanced Reactor Stakeholder Public Meeting Slide 52

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

Slide 53

2 NUREG/CR-7289 ORNL/TM-2021/2002

  • Oak Ridge National Laboratory (ORNL)

- F. Bostelmann

- G. Ilas

- C. Celik

- A.M. Holcomb

- W.A. Wieselquist Slide 54

Motivation/Background 3

Slide 55

Commercial Light Water Reactor Approach to Reactor Physics/Nuclear Design 4

Start with simplified geometry and detailed energy group structure, End with simplified group structure and 3D geometry Apply biases and uncertainties to calculated quantities of interest (QOIs):

Reactivity balance Shutdown margin Feedback coefficients Power distribution Data (e.g., ENDF/B-VII.1)

Cross Section Processing (e.g., AMPX, NJOY)

Output: 100s of energy groups 2-D Assembly (e.g., SCALE, CASMO)

Output: 2-4 Energy Groups, Cross Section and Discontinuity Factors 3-D Whole Core Simulator (e.g., PARCS, SIMULATE) 1-D Pin Cell (e.g., SCALE, CASMO)

Output: 20-100 energy groups Slide 56

Commercial Light Water Reactor Approach to Reactor Physics/Nuclear Design 5

Start with simplified geometry and detailed energy group structure, End with simplified group structure and 3D geometry Apply biases and uncertainties to calculated quantities of interest (QOIs):

Reactivity balance Shutdown margin Feedback coefficients Power distribution Data (e.g., ENDF/B-VII.1)

Cross Section Processing (e.g., AMPX, NJOY)

Output: 100s of energy groups 2-D Assembly (e.g., SCALE, CASMO)

Output: 2-4 Energy Groups, Cross Section and Discontinuity Factors 3-D Whole Core Simulator (e.g., PARCS, SIMULATE) 1-D Pin Cell (e.g., SCALE, CASMO)

Output: 20-100 energy groups Emphasized during safety review Slide 57

Impact of Data Uncertainty 6

  • 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 Slide 58

Data Uncertainty and Licensing 7

  • 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 Slide 59

Data Challenges for Advanced Reactor Licensing 8

  • 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?

Slide 60

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

Slide 61

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

11

  • 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)

Approach ADAMS Accession Nos.

ML20274A052 and ML21125A256 Slide 63

12 Sensitivity and Uncertainty Analysis Reactor technology Selected benchmarka Type High Temperature Gas Reactor HTR-10 Experiment Fluoride Salt Cooled High Temperature Reactor UC Berkeley Mark1 PB-FHR Computational benchmark Graphite-moderated Molten Salt Reactor MSRE Experiment Heat Pipe Microreactor (metal-fueled)

INL Megapower Design Ab Computational benchmark Sodium Cooled Fast Reactor (metal and oxide fueled)

EBR-II ABR-1000 Experiment 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.

Slide 64

13

  • 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)

Sensitivity and Uncertainty Analysis Slide 65

14 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 Slide 66

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

Sensitivity Analysis Results Key Nuclear Data Impacting Pebble Graphite Temperature Feedback Nuclide Reaction Sensitivity (reducing negative ) Nuclide Reaction Sensitivity (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 Slide 67

16 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%

Slide 68

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

Uncertainty Analysis Results Top Nuclear Data Contributors to Multiplication Factor Uncertainty Slide 69

18 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.

Conclusions

  • Results of study with respect to depletion/burnup are limited due to (1) unavailability of benchmarks and relevant data, and (2) capability not currently available to fully propagate uncertainty in depletion analyses.

Slide 70

19 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.

Conclusions Slide 71

1 SCALE and MELCOR development and application for non-LWRs Advanced Reactor Stakeholder Meeting January 19, 2022 Slide 72

NRC strategy for severe accident analysis Slide 73

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 3

Slide 74

4

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 SCALE MELCOR Non-LWR Demonstration Project - approach Slide 75

5 Slide 76

Molten Salt Reactor Experiment (MSRE) 6 Slide 77

Advanced Burner Test Reactor (ABTR) 7 Slide 78

SCALE analysis approach for MSR 3 models run in an iterative fashion to predict nuclide inventory, decay heat, and reactivity feedback coefficients at selected point in the operating cycle Time snapshot

  • predicts core neutron flux at a point in the operating cycle Simplified core + loop + offgas
  • predicts primary-system-average nuclide inventory over time 1D loop model
  • predicts nuclide inventory in each section of the loop Core+

Loop Offgas System Simplified Core+Loop+Offgas Xe, He, Kr, H Time snapshot 1D loop Slide 79

9

  • 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 Time snapshot model SCALE 3D full core MSRE model graphite fuel fuel fuel fuel Axial flux distribution Radial flux distribution Cross section of unit cell Slide 80

1D loop model

  • relative to the loop transit time

(~25 s for MSRE)

Short-lived nuclide (I-137, t1/2=24.5s) as a function of location in the loop Core (1)

Loop Short-lived nuclide as a function of time at the bottom of the core (zone 1) 2 4

3 5

6 7

8 9

  • Predicts nuclide inventory in each section of the loop
  • As fuel salt travels the loop Long-lived*

nuclides will slowly accumulate/be removed*

(same as solid fuel)

Short-lived* nuclides will oscillate about an equilibrium Slide 81

11 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.

SCALE analysis approach for SFR SCALE ABTR model ABTR model with individual assembly definitions and corresponding power map Slide 82

12 MELCOR Modeling Scope SCALE (ORNL)

Thermal hydraulics Fuel thermal-mechanical response Core degradation Ex-vessel damage progression Fission product release and transport Reactivity Effects Slide 83

13 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)

MELCOR Non-LWR Modeling Slide 84

14 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 TRISO Radionuclide Release Modeling Failing Intact TRISO Released to the matrix Transition from Intact-to-failed Failed TRISO Contamination Release from failed TRISO (Modified Booth)

Intact TRISO Failed TRISO recoil Released to the matrix Transfer to failed TRISO Distribution calculated from diffusion model Release from TRISO failure Diffusion Diffusion from intact TRISO Recoil fission source recoil Diffusion Diffusion Slide 85

15 MELCOR Generalized Radionuclide Transport and Retention (GRTR) Model Radionuclides grouped into forms found in the Molten Salt Reactor Experiment 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 Model Scope Slide 86

16 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 MELCOR Generalized Radionuclide Transport and Retention (GRTR) - States and State Transitions Note: MELCOR considers soluble, bulk colloid, interfacial colloid, and vapors as distinct chemical states

Slide 87

17 Fission product thermochemistry modeling sample demonstration

  • Exercise machinery
  • Focuses on Cs and CsF release from salt pool
  • Thermochimica Gibbs Energy Minimizer
  • Utilizing vapor phase data for CsF*

Demonstration calculation for LOCA 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 /> Model exhibits Cs and CsF vaporization to gas space at elevated salt temperatures Cesium Vaporization from Molten Salt - FHR Example
  • With modifications by Ontario Tech.

Cs Transport Pathway Fuel Pebbles Molten Salt Overlying Gas Atmosphere 600 700 800 900 1000 1100 1200 1300 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 0

6 12 18 24 Temperature (deg-C)

Fraction of initial invenory (-)

Time (hr)

Cesium Behavior Total released from pebbles Total in the liquid Vaporized from the liquid Core Fluid Temperature Slide 88

18 Some accidents may involve reactivity feedbacks For non-LWRs, MELCOR uses a point kinetics models Feedback models

  • User-specified external input
  • Doppler
  • Fuel and moderator density
  • Flow reactivity feedback effects integrated into the equation set FHR example calculation using MELCOR point kinetics model Point Kinetics Modeling 0

50 100 150 200 250 300

-2.0

-1.5

-1.0

-0.5 0.0 0.5 1.0 1

10 100 1000 Power (MW)

Reactivity ($)

Time (sec)

Core Reactivities Fuel Temperature Molten Salt Inner Reflector Outer Reflector Moderator Xenon Total Reactivity Power Slide 89

19 Extended static point kinetic equations to capture motion of delayed precursors through the reactor system Point Kinetics Modeling (MSR) 0 50 100 150 200 250 0

10 20 30 40 50 60 70 Compensating Control System Reactivity [pcm]

Time [s]

Guo Code MSRE Data MELCOR Validated against MSRE zero-power flow experiments Slide 90

NRC Non-LWR Vision and Strategy, Volume 5 Radionuclide Characterization, Criticality, Shielding, and Transport in the Nuclear Fuel Cycle 20 HTGR 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 Slide 91

Break Meeting will resume in 10 minutes Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 323 588 045#

Advanced Reactor Stakeholder Public Meeting Slide 92

NRC Activities on Advanced Manufacturing Technologies (AMTs)

Matthew Hiser NRC Office of Nuclear Regulatory Research January 19, 2022 Periodic Advanced Reactor Stakeholder Meeting Slide 93

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 94

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 3

https://www.osti.gov/pages/servlets/purl/1437906 Schematic of LPBF process

- Smaller Class 1, 2 and 3 components, fuel hardware, small internals Slide 95

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 4

https://www.neimagazine.com/news/newswestinghouse-produces-3d-printed-component-for-us-nuclear-plant-7911951 Slide 96

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 5

https://www.ornl.gov/news/additively-manufactured-components-ornl-headed-tva-nuclear-reactor?utm_source=miragenews&utm_medium=miragenews&utm_campaign=news Slide 97

Directed Energy Deposition

  • Process:

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

- Fundamentally welding using robotics/

computer controls

  • Potential Applications 6

- Similar to LPBF, although larger components possible due to faster production and greater build chamber volumes Schematic of DED process https://www.osti.gov/pages/servlets/purl/1437906 Slide 98

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 99

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 100

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 9

  • Potential Applications

- 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 Schematic of cold spray process*

Slide 101

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 102

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 103

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.

  • Revision 1 was published in June 2020 (ML19333B980) 12 Slide 104

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 105

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:

Task 2 - Regulatory Preparedness Activities Subtask 2A: Implementation using the 10 CFR 50.59 Process 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 107

  • A Technical Letter Report (TLR) is produced for each of the initial five AMTs
  • Provides technical basis information and gap analysis
  • Written by NRC contractor (to date, DOE labs)
  • A technical assessment (TA) is produced for each TLR by NRC staff which provides the NRC staff perspective on key aspects of the AMT for safety and component performance
  • A draft guidelines document (DGD), informed by the TA and TLR, will be generated by the NRC staff for each AMT.
  • The AMT-specific DGDs accompany and align with the generic Advanced Manufacturing Technologies Review Guidelines NRC AMT Guidelines Development Technical Letter Report LPBF ML20351A292 Technical Letter Report L-DED ML20233A693 Technical Letter Report Cold Spray ML21263A105 AMT-Specific (Initial 5 AMTs)

Generic Technical (Subtask 1A)

Final Guidance for Initial AMTs Regulatory Guidelines (draft for FRN public comment)

Technical Letter Report PM-HIP Technical Letter Report EBW NRC Staff-developed Contractor-developed Legend Technical Assessment LPBF ML20351A292 Draft Guidelines Document LPBF ML21074A040 Draft Guidelines Document L-DED Technical Assessment L-DED ML20233A693 Draft Guidelines Document Cold Spray Technical Assessment Cold Spray ML21263A105 Draft Guidelines Document PM-HIP Technical Assessment PM-HIP Technical Assessment EBW Draft Guidelines Document EBW Expected to be developed later after DOE-EPRI demo project Subtask 2C Draft AMT Review Guidelines ML21074A037 16 Slide 108

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 109

Status of Deliverables - Task 1 Subtask Actions/Deliverables Status 1A AMT processes under consideration Additive Manufacturing (AM) - Laser Powder Bed Fusion Complete - ML20351A292 AM - Directed Energy Deposition (DED)

Complete - ML20233A693 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 1C Modeling and Simulation of Microstructure ANL M&S gap analysis to predict microstructure Complete - ML20269A301 ANL M&S gap analysis to predict material performance Complete - ML20350B550 18 Slide 110

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

ML20240A077 Develop a document that describes the generic technical information to be addressed in AMT submittals.

Public meeting held on September 16, 2021 to discuss:

ML21074A037 - Draft AMT Review Guidelines ML21074A040 - Draft Guidelines Document for AM -

LPBF Public meeting to discuss draft document 3A/3B External/ Internal Interactions Continued communication with NRC staff, vendors, licensees and EPRI for future AMTs Ongoing as needed 3C Knowledge Management Plan Develop Knowledge Management Plan Complete - internal 3D Workshop Hold Public Workshop Complete - summary: ML20357B071 RIL: Part 1 Part 2 3E Material Information course Training course and course materials First 6 seminars complete - internal 19 Slide 111

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 112

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.

Slide 113

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