ML20232D114

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August 20 2020 Advanced Reactor Stakeholder Public Meeting - Combined Presentation
ML20232D114
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Issue date: 08/20/2020
From: Jordan Hoellman
NRC/NRR/DANU/UARP
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Hoellman J,NRR/Danu/UARP,415-5181
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Advanced Reactor Stakeholder Public Meeting August 20, 2020 Telephone Bridgeline: (888) 810-4937 Passcode: 8854397#

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Time Agenda Speaker 10:00 - 10:10 am Opening Remarks NRC Presentation on ANL Report, The Assessment of Tritium 10:10 - 11:10 am NRC/ANL Detection and Control in Molten Salt Reactors Presentation on INL Report, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-11:10 - 12:10 pm NRC/INL Related Assessments for Advanced Nuclear Reactor Facilities NRC discussion of Advanced-Reactor Source Term - Pilot 12:10 - 12:40 pm NRC Studies 12:40 - 1:00 pm BREAK All Discussion of Considerations for Annual Fee Regulations for 1:00 - 1:30 pm NEI Microreactors 1:30 - 2:15 pm Discussion of Part 53 Rulemaking Plan and White Paper NRC 2:15 - 2:30 pm Industry Stakeholders Perspectives on Part 53 NEI/USNIC Discussion of Status of Spent Fuel Reprocessing 2:30 - 2:45 pm NRC Rulemaking Overview of ORNL Report on Preparing and Reviewing a 2:45 - 3:00 pm NRC Molten Salt Non-Power Reactor Application 3:00 - 3:15 pm Concluding Remarks and2 ofFuture 146 Meeting Planning NRC/All

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced.html 3 of 146

ASSESSMENT OF TRITIUM DETECTION AND CONTROL IN MOLTEN SALT REACTORS erhtjhtyhy David Grabaskas, Tingzhou Fei, James Jerden Argonne National Laboratory 4 of 146

OBJECTIVES n Assist NRC:

  • Expanding capacity and capabilities for licensing non-LWRs through knowledge base and skillset development n Technical Assessment of Tritium Behavior in MSRs:
  • Location and pathways of tritium generation
  • Tritium transport and retention phenomena
  • Applicable experience and existing data on tritium behavior and control
  • Available modeling and simulation tools n Regulatory Considerations:
  • Applicability of current regulations
  • Associated limits and constraints on tritium handling and release
  • Areas of consideration during NRC review of MSR licensing applications
  • Assessment of the adequacy of the current regulation and guidance 5 of 146 2

REPORT n ANL/NSE-20-15:

  • Available on ADAMS and NRC Advanced Reactor Webpage 6 of 146 3

BACKGROUND ON TRITIUM n 3H or T

  • Radioactive isotope of hydrogen with 12.3 year half-life
  • Naturally occurring due to cosmic ray interaction with the atmosphere
  • Additional environmental tritium from nuclear weapons tests and nuclear reactor effluents n Health Hazard
  • Low energy beta emitter (max ~18keV):

o Internal exposure is the only concern as beta has insufficient energy to penetrate dead skin layer

  • Differing chemical forms and biological impact:

o HT/T2 Gas: Exhaled quickly from the body o HTO Water: Mostly eliminated with biological half-life of water (10 days)

Courtesy of the Canadian Nuclear Safety Commission (CNSC) o OBT: Organically bound tritium, can act like carbon in body with longer biological half-life (40 days) 7 of 146 4

MOLTEN SALT REACTORS: NOMENCLATURE 8 of 146 5

MSR: SALT SELECTION n Considerations:

  • Neutronics, material compatibility, dissolution properties, stability, and thermophysical properties
  • Lithium- and beryllium-bearing salts are popular choices for thermal reactors due to moderating ability
  • Most past experience with FLiBe 9 of 146 6

MSR Tritium MSRE Phenomena Experience

  • Production
  • Design
  • Transport
  • Operation
  • Control
  • Tritium Modeling and Regulatory Simulation Considerations
  • Requirements
  • Quantitative Limits
  • Current Capabilities
  • Regulation
  • Assessment 10 of 146
  • Assessment

MSR Tritium MSRE Phenomena Experience

  • Production
  • Design
  • Transport
  • Operation
  • Control
  • Tritium Modeling and Regulatory Simulation Considerations
  • Requirements
  • Quantitative Limits
  • Current Capabilities
  • Regulation
  • Assessment 11 of 146
  • Assessment

TRITIUM PRODUCTION: WATER REACTORS n Ternary Fission

  • All fission nuclear reactors create tritium from ternary fission (fission with three products)
  • Approximately 1 in 10,000 fissions
  • Largely contained within the fuel in water reactors n Other Factors:
  • Boron neutron capture in control elements (BWRs) or coolant (PWRs)
  • Deuterium neutron capture in heavy water reactors (HWRs), such as CANDUs 12 of 146 9

TRITIUM PRODUCTION: MSR n MSRs

  • Two major factors in the production of tritium: Lithium and Beryllium 7Li 9Be

$ + # + & + # + $

+  % + #

6Li & + " + !

!  ! (() = 0.8 s)  ! + * + +

+ " + #

13 of 146 10

TRITIUM PRODUCTION: MSR n Lithium

  • Natural lithium is 92.4% 7Li and 7.6% 6Li
  • 7Li has a much smaller tritium-producing cross-section than 6Li (as will be shown)
  • Lithium enrichment utilized to reduce 6Li due to tritium concerns
  • 99.995% 7Li enrichment is typical, further enrichment may be cost prohibitive n Establishing Equilibrium
  • If a molten salt contains both Li and Be, the existing 6Li contained in the salt will be consumed by neutron interactions, but new 6Li is created from neutron interactions with beryllium
  • If salt only contains Be, 6Li concentration will build over time until an equilibrium is reached 14 of 146 11

TRITIUM PRODUCTION: MSR 15 of 146 Source: K. Dolan, "Tritium Thermal Desorption Testing of Nuclear Graphite Irradiated at Fluoride-Salt-Cooled High-Temperature Reactor Conditions," Thesis, Massachusetts Institute of Technology, 2018 12

TRITIUM PRODUCTION: PB-FHR EXAMPLE Effective Full Power Years 16 of 146 Source: J. Stempien, "Tritium Transport, Corrosion, and Fuel Performance Modeling in the Fluoride Salt-Cooled High Temperature Reactor (FHR)," Dissertation, Massachusetts Institute of Technology, 2015 13

TRITIUM PRODUCTION: RATE COMPARISON n Example MSR Concepts:

  • Molten Salt Breeder Reactor (MSBR): A 1000MWe, FLiBe salt-fueled MSR concept studied extensively by ORNL in the 1970s following the operation of MSRE
  • Pebble-Bed Fluoride Salt-Cooled High-Temperature Reactor (PB-FHR): A solid fuel, FLiBe salt-cooled FHR design developed by the University of California-Berkeley, which serves as the basis for much recent FHR research Normalized Tritium Production Rate Reactor (Ci/GWe/yr)a Type Control Fuel Coolant Moderator Elements PWR 11,000 - 25,000 300 - 1,000 1,000 BWR 11,000 - 25,000 b 3,000 - 5,000 HWR 14,000 - 20,000 50,000 600,000 - 2,400,000 1,000 MSBR 730,000 b b PB-FHR b 2,100,000/720,000c b b a Unit is curies of tritium produced per GWe during an approximate operating year b Negligible or unknown.

c Beginning of life/Steady-state 17 of 146 14

TRITIUM PRODUCTION:

SUMMARY

Key Point For MSRs that contain lithium or beryllium within the molten salt, it is possible to generate tritium at rates far exceeding current U.S.

LWR systems (on a per GWe basis) due to neutron interactions with 6Li. In addition, tritium generated through this pathway will be present within the molten salt and not contained within fuel or control elements 18 of 146 15

TRITIUM TRANSPORT: CHEMICAL FORM Molten Salt 19 of 146 16

TRITIUM TRANSPORT: CHEMICAL FORM Molten Salt 20 of 146 17

TRITIUM TRANSPORT: CHEMICAL FORM Molten Salt 21 of 146 18

TRITIUM TRANSPORT: CHEMICAL FORM Molten Salt 22 of 146 19

TRITIUM TRANSPORT: CHEMICAL FORM n Tritium Fluoride (TF)

  • Likely form of tritium born from 6Li reactions
  • Low permeability through structural materials
  • A powerful oxidizer and principle cause of corrosion in MSRs n TF Corrosion
  • Unlike LWRs, corrosion products are soluble in salt, which then expose underlying metal
  • Measures must be taken to reduce (in the chemical sense) TF before it interacts with structure
  • Multiple techniques available for redox control but all reduce TF to molecular HT/T2 n Molecular HT/T2
  • Highly permeable through structural materials at the operating temperatures of MSRs, increasing likelihood of tritium escaping the reactor system 23 of 146 20

TRITIUM TRANSPORT: CHEMICAL FORM Key Point The production of tritium within the molten salt is inextricably tied to corrosion concerns due to the formation of TF, a powerful oxidizer.

Corrosion control strategies will likely result in the reduction of TF to a molecular hydrogen form (HT/T2), which are highly permeable in structural materials at the operating temperatures of most MSR designs 24 of 146 21

TRITIUM TRANSPORT: BARRIERS n Similarities and Differences

  • Salt-fueled and salt-cooled MSRs share some of the same tritium barriers and transport phenomena n Notational Diagrams
  • Following diagrams outline high-level transport and retention pathways n Importance of Graphite
  • Experience with MSRE demonstrated high tritium retention within core graphite
  • High specific surface area of graphite offers many bonding sites for tritium
  • Tritium can be liberated from graphite at high temperatures (above normal operating temperatures)
  • Many factors influence graphite retention capabilities, such as form and irradiation history
  • In general, nuclear grade graphite has lower retention than activated forms of carbon due to the annealing process, which is necessary for irradiation stability in core 25 of 146 22

TRITIUM TRANSPORT: BARRIERS (PB-FHR) 26 of 146 23

TRITIUM TRANSPORT: BARRIERS (FUEL-SALT MSR) 27 of 146 24

TRITIUM TRANSPORT: BARRIERS 28 of 146 25

TRITIUM TRANSPORT: CHEMICAL FORM Key Points

  • Due to the large quantity of tritium in the system and the mobile chemical form, tritium control and removal strategies are necessary to prevent the relocation of tritium to areas outside of the reactor system and potentially to the environment
  • Graphite within thermal MSR systems likely offers an initial retention mechanism 29 of 146 26

TRITIUM TRANSPORT: CONTROL n Coatings

  • Use of coatings or barriers that have low hydrogen/tritium permeability
  • Most historical tritium coatings are not compatible with molten salts (oxides, aluminum)
  • Coatings may need to be placed on surfaces not in contact with molten salt 30 of 146 27

TRITIUM TRANSPORT: CONTROL n Permeators

  • Use a combination of high and low permeability materials to direct tritium transport
  • Use of low-pressure purge gas or vacuum can encourage tritium removal in certain areas of system
  • Can be integrated into a double-wall heat exchanger Permeator Tritium Double-wall Heat Exchanger Removal System Tritium Removal System 31 of 146 28

TRITIUM TRANSPORT: CONTROL n Gas Sparging/Stripping

  • Bubbling an inert gas, such as helium, through the molten salt encourages the movement of tritium from the salt to the sparge gas
  • Technique dependent on contact surface area between gas and salt
  • Also can invert the process by spraying salt through a gas volume or vacuum 32 of 146 29

TRITIUM TRANSPORT: CONTROL n Adsorber Bed

  • Utilizes a bed (such as spheres) made of material with high tritium retention
  • Could use activated carbon rather than nuclear grade graphite since placed away from the core
  • Once saturated, spheres could be removed and stored or heated to liberate tritium 33 of 146 Source: C. Forsberg et al., "Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward," Nuclear Technology, vol. 197, no. 119-139, 2017 30

TRITIUM TRANSPORT: MONITORING AND STORAGE n Monitoring

  • Infeasible to directly measure tritium concentration in molten salt, due to self-shielding of low energy beta emission from the salt
  • Instead, tritium concentrations likely derived from the tritium removal system, such as the off-gas stream
  • Flow-through detectors are needed due to low energy beta, although alternative approaches are being explored (optical spectroscopy) n Storage
  • For CANDUs, removed tritium is stored as a metal hydride (tritide)
  • Metals, such as titanium, form metal hydrides when exposed to hydrogen/tritium and can retain incredible amounts of hydrogen (densities greater than that of liquid hydrogen)
  • Metal hydrides are stable at room temperature and pressure, but the process is reversible and tritium can be liberated if heated above 500ºC
  • Other storage avenues possible, such as within low-water cement 34 of 146 31

TRITIUM TRANSPORT: CONTROL Key Point Numerous tritium control and removal concepts exist, with varying levels of technology readiness. An MSR tritium control strategy will likely include multiple components or systems to both retain tritium within the salt and remove it at designated locations.

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MSR Tritium MSRE Phenomena Experience

  • Production
  • Design
  • Transport
  • Operation
  • Control
  • Tritium Modeling and Regulatory Simulation Considerations
  • Requirements
  • Quantitative Limits
  • Current Capabilities
  • Regulation
  • Assessment 36 of 146
  • Assessment

MOLTEN SALT REACTOR EXPERIMENT (MSRE) n Design

  • Built at Oak Ridge National Laboratory
  • Salt-Fueled: LiF-BeF2-ZrF4-UF4 (99.9923% 7Li)
  • 7.34 MWth, no power conversion
  • Graphite moderated
  • FLiBe secondary system n Operation
  • Operated 1965-1969 (~17,000 critical hours)
  • Used both 235U and 233U at different stages
  • Gas space of primary pump used for off-gas system 37 of 146 34

MSRE: TRITIUM EXPERIENCE n Tritium Balance

  • During final MSRE runs, a study performed to examine tritium transport
  • Through measurements of reactor systems, the study attempted to determine where the produced tritium was going
  • Tritium production was estimated through neutronic calculation and compared to measured quantities 38 of 146 35

MSR Tritium MSRE Phenomena Experience

  • Production
  • Design
  • Transport
  • Operation
  • Control
  • Tritium Modeling and Regulatory Simulation Considerations
  • Requirements
  • Quantitative Limits
  • Current Capabilities
  • Regulation
  • Assessment 39 of 146
  • Assessment

MODELING AND SIMULATION n Functional Requirements

  • The final report outlines functional requirements for the modeling and simulation of tritium in MSRs based on production and transport phenomena
  • The functional requirements aid in the identification of necessary code capabilities n Code Survey
  • The current code landscape was examined
  • Multiple MSR tritium analysis stand-alone codes or packages currently under development
  • Development of data for code validation is a need recognized by the MSR industry n Tritium Production Assessment
  • To gauge current capabilities, a trial calculation was performed of tritium production in MSRE
  • Utilized MCNP 6.2, ORIGEN-S/COUPLE Tritium Production Rate: 235U Fuel Calculation Method (Ci/d)

Single Flow Passage Model 27.1 Whole Core Model 40 of 146 29.0 ORNL Estimate 31.7 37

MSR Tritium MSRE Phenomena Experience

  • Production
  • Design
  • Transport
  • Operation
  • Control
  • Tritium Modeling and Regulatory Simulation Considerations
  • Requirements
  • Quantitative Limits
  • Current Capabilities
  • Regulation
  • Assessment 41 of 146
  • Assessment

REGULATION: QUANTITATIVE CONSTRAINTS n Multiple Regulatory Sources

  • Almost all constraints are dose or dose-derived
  • Only exception is tritium release to sewers, not shown in table (limit of 5 Ci per year)
  • Some constraints are cumulative across all releases or all beta releases 42 of 146 See report (ANL/NSE-20-15) for table footnotes 39

REGULATION: EFFLUENTS n 10 CFR 50.34a and 50.36a

  • Applicants must identify design objectives to keep effluent releases to unrestricted areas ALARA
  • ALARA, in this context, allows for the consideration of the state of technology and economics in relation to public health and safety and public benefits of atomic energy
  • Appendix I limits meet these objectives n Applicant must describe:
  • Equipment utilized to achieve ALARA requirements
  • Estimate of annual liquid and gases effluent releases
  • Description of packaging, storage, and shipment of waste from treating effluents n Expectations
  • Licensee shall be guided by past experience, which indicates the typical releases are only a small percentage of 10 CFR 20.1301 limits 43 of 146 40

REGULATION: MINIMIZATION OF CONTAMINATION n 10 CFR 20.1406

  • Applicants shall describe how facility design and procedures will minimize, to the extent practical, generation of waste and contamination of the facility and the environment n To the extent practical
  • RG 4.21 provides guidance, other competing concerns, such as the implication to safety systems and the overall cost should be considered. Thus the minimization of facility contamination must be considered in the context of overall facility safety.

n RG 4.21 Guidance

  • Utilizes a risk-informed, performance-based approach
  • Minimizing facility contamination through use of SSC and operational procedures
  • Minimizing environmental contamination through understanding of radionuclide transport and use of a conceptual site model
  • Facilitation of decommissioning considered in the design process
  • Minimizing generation of waste, however NRC recognizes the constraints and competing factors to waste minimization 44 of 146 41

REGULATION: OTHER FACTORS n 10 CFR 20.1701: Restricting Internal Exposures

  • Licensee shall use, to the extent practical, process and engineering controls to control radioactive material in air n 10 CFR 50 - Appendix A: GDCs
  • RG 1.232 found no need to modify effluent GDCs for non-LWRs n Others
  • Assessment of tritium in PRA as part of Licensing Modernization Project (LMP) process
  • Storage of removed tritium, DC/COL-ISG-013/014
  • Monitoring effluents: RG 1.21, 4.1, 4.15, 1.109 45 of 146 42

REGULATION: ASSESSMENT FINDINGS Key Points

  • Limits on tritium release to the environment are primarily dose- or concentration-based, rather than centered on cumulative activity released. This is essentially a performance-based system, which is not LWR-specific and could allow MSR vendors the necessary flexibility to develop tritium control strategies.
  • Current regulation requires a description of the systems and procedures in place to limit radioactive releases, including an estimate of predicted effluents during operation. This would encompass tritium control strategies and systems.
  • Regulation and guidance on the release of radioactive effluents to the environment permits the use of a risk-informed performance-based evaluation to minimize releases to the extent practical. Although there may be subjectivity in the determination of practicality, the diversity in MSR designs and tritium control strategies likely makes generic guidance on this issue difficult.

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SUMMARY

n Tritium in MSRs

  • For MSRs that contain lithium or beryllium in the salt, the production of tritium must be considered
  • Due to corrosion concerns, tritium will be converted to a mobile molecular form
  • There are many options available for the control and removal of tritium
  • Development of modeling tools and validation data is an ongoing project n Regulation
  • Current regulatory environment appears adequate to address tritium concerns in MSRs
  • Generally performance-based dose limits on tritium release
  • Existing requirements for license applicants to minimize releases and describe the strategies and systems utilized to control releases
  • Flexibility to consider plant operation and economics when developing control strategies 47 of 146 44

QUESTIONS?

48 of 146

Discussions on Mechanistic Source Term Methodologies and Associated Information 49 of 146 1

INTRODUCTION NRCS Vision and Strategy and the development of mechanistic source terms for non-LWRs

  • Development of sufficient computer codes and tools Staff interactions with ACRS
  • Related to mechanistic source term (MST) methodologies
  • Expanding guidance for developing MSTs
  • Expectations for the technical adequacy in using MST
  • Tools for staff confirmatory analysis NEIMA requirement
  • Evaluation on developing and implement guidance for the resolution of issues relating to the use of MST 50 of 146 2

INTRODUCTION (Contd)

Development of final reports

  • SAND2020-0402, Simplified Approach for Scoping Assessment of Non-LWR Source Terms
  • INL/EXT-20-58717, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities Path forward
  • Use INL and SNL reports as additional aid in resolving MST issues, and for developing design-specific MST methodologies
  • Methods, results, and conclusions of the staffs pilot studies and use of MELCOR will be publicly shared 51 of 146 3

August 20, 2020 Kurt Vedros Andrea Alfonsi Paul Humrickhouse Hongbin Zhang Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities 52 of 146

Objective

  • Document written as a project for NRC team: INL/EXT-20-58717
  • Develop a risk-informed, performance-based, and technology-inclusive approach to determine source terms for dose-related assessments at advanced nuclear facilities to
1) support the NRCs Non-LWR Vision and Strategy Near-Term Implementation Action Plans (ADAMS Accession No. ML16334A495) and,
2) the NRCs response to the Nuclear Energy Innovation and Modernization Act (NEIMA) Public Law No: 115-439, of January 2019 53 of 146

F-C target curve (NEI 18-04).

54 of 146

Definitions

  • Source Terms for Advanced Reactors: the quantities, timing and other characteristics of radionuclides released from the facility to the environment.
  • Non-Mechanistic Source Terms Methodology: adopt conservative approaches and assumptions based on known physical and chemical principles.
  • Mechanistic Source Terms Methodology: consider design-specific scenarios and use best-estimate models with uncertainty quantification for a range of licensing basis events to be used for the design and licensing of advanced nuclear technologies.

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Illustration of radionuclides retention and removal process for one non-LWR concept 1

Mechanistic source terms can be correlated using1:

, , = ( , , ) ( , , ) ( , , ) ( , , )

1SAND2020-0402 56 of 146

Illustration of radionuclides retention and removal process for one non-LWR concept where:

, , is the total release to the environment of radionuclide over the entire release duration time (t)

( ) is the initial fission product inventory at the time of the reactor accident for radionuclide

( , , ) is the fraction of release of radionuclide from fuel system boundaries to the fuel matrix

( , , ) is the fraction of release of radionuclide from fuel matrix to primary system

( , , ) is the fraction of release of radionuclide from primary system to leak path

( , , ) is the fraction of release of radionuclide from leak path to the environment 57 of 146

Technology-inclusive source terms determination methodology components 58 of 146

Technology-inclusive source terms determination methodology 59 of 146

Step 1: Identify Regulatory Requirements 60 of 146

Step 2: Identify Reference Facility Design

  • The developer defines the reference facility design
  • Identifies:

All foreseeable facility system operating modes Barriers Engineered safety features within barriers

  • SSCs of these systems, or needed for these systems 61 of 146

Step 3: Define Initial Radionuclide Inventories

  • Determine equilibrium radionuclide inventories (or appropriate values if equilibrium conditions are not achieved for a particular plant design) in all plant systems (e.g., fuel, barrier 1, barrier 2, etc.) during normal steady-state operation.

Description is provided of initial inventories

  • e.g., equilibrium nominal end of life 62 of 146

Step 4. Perform Bounding Calculations

  • These bounding calculations are performed to determine the dose consequences of the releasing radionuclide inventories identified by the previous step for the maximum credible accident (MCA)

The MCA is postulated as a nuclear accident that would result in a potential hazard that would not be exceeded by any other accident considered credible during the lifetime of the facility.

  • Demonstrate compliance with the established regulatory criteria. If criteria met, proceeds to documentation.

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Step 5. Conduct SHA and Perform Simplified Calculations

  • Conduct a SHA (FMEA, STPA, or equivalent) to identify potential SSC failure modes that lead to radionuclide releases, as well as to identify a spectrum of postulated LBEs.

Consider the behavior of the barriers after SHA and determine dose consequence by using simplified methods.

  • Simplified methods are still bounding calculations based on proven physical properties.
  • Inventory release to environment is modified from MCA by behavior of design barriers identified in SHA.
  • If criteria met, proceeds to documentation.

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Step 6. Consider Risk-informed System Design Changes

  • Consider a system redesign to include additional barriers or SSCs as identified by hazard analysis, which will either return to Step 3 or proceed to Step 7.

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Step 7. Establish Adequacy of MST Simulation Tools

  • Identify any gaps from MST simulation tools criteria1:

The performance of the reactor and fuel under normal and off-normal conditions is sufficiently well understood to permit a mechanistic analysis.

The transport of fission products can be adequately modeled for all barriers and pathways to the environs, including specific consideration of containment design. The calculations should be as realistic as possible so that the values and limitations of any mechanism or barrier are not obscured.

The events considered in the analyses to develop the set of source terms for each design are selected to bound severe accidents and design-dependent uncertainties.

  • Develop and complete analytic and testing programs to fill identified gaps in available MST simulation tools.

1 SECY-93-092 66 of 146

Step 8. Select Initial List of LBEs and Conduct PIRT

  • Develop initial list of LBEs which may not be complete but are necessary to develop the basic elements of the safety design
  • Conduct Phenomena Identification and Ranking (PIRT) exercise to identify safety-significant phenomena for the LBEs
  • Assess importance, knowledge level, and status of modeling for each phenomenon:

67 of 146

Step 9. Develop and Update PRA Model

  • PRA is used to model LBEs in a probabilistic manner.
  • Utilize the PRA group of analyses that inform the logic model which informs consequence modeling.
  • Static PRA is used for design and regulatory decisions.
  • Dynamic PRA can be used to validate the outcome of sequence end states.
  • Adhere to the most current Non-LWR PRA Standard (ASME/ANS RA-S-1.4-2013) when any conflicts are encountered between standards.

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Step 10: Identification or revision the list of LBEs

  • The identification process:

needs to be considered as an integral part of the overall design process and, should be re-iterated since its selection informs the design requirements of safety-related and non-safety-related SSCs Identification Process LBEs are LBEs Selection of reviewed at updated if Initial Event the end of the design set the design changes phase 69 of 146

Step 11: Select LBEs to Include Design Basis External Hazard Level for Source Term Analysis

  • A set of design basis external hazard levels (DBEHLs) will be selected to form an important part of the design and licensing basis:
  • As supported by methods, data, design, site information, and supporting guides and standards, these DBEHLs:

will be informed by a probabilistic external hazards analysis and will be included in the PRA using design features that are incorporated to withstand these hazards

  • Other external hazards not supported by a probabilistic hazard analysis will be covered by DBEHLs that are determined using traditional deterministic methods.

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Step 12: Perform LBEs Source Term Modeling and Simulation

  • The source term assessment needs to characterize the generation, release, transport, and retention of fission product and activation radionuclides
  • The process for the development of modeling and simulation tools for non-LWR applications is like LWR applications.

X-energy plan for source term characterization 71 of 146

Step 12: Perform LBEs Source Term Modeling and Simulation: Source term evaluation model for non-LWRs

  • Technology-inclusive because it relies on the same codes with the suite of physics models needed for the different non-LWR technologies.

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Step 13. Review LBEs List for Adequacy of Regulatory Acceptance

  • Develop a final list of LBEs.

Review current

  • Safety classifications of SSCs Are any end results changes desired before the final list?
  • Changes to increase F-C target criteria margin
  • Reduction of uncertainties in LBE frequencies or consequences
  • Limit restrictions on siting or emergency planning
  • etc
  • If the final list is not complete, go back to Step 6.

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Step 14. Document Completion of Source Term Development

  • Prepare a documentation covering methods used, source term calculations and results and submit to the NRC for approval.

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BACK UP SLIDES 75 of 146

Step 12: Perform LBEs Source Term Modeling and Simulation: Source term assessment software requirements

  • Reactor Physics Computer Models:

Calculate radionuclide inventories and power distributions in the design.

  • Fuel Performance Computer Models:

Calculate thermal and stress histories for fuel and identify fuel failure and radionuclide release

  • System Analysis Computer Models:

Calculate the progression of accident and radionuclide transport.

Requires boundary conditions from fuel performance analysis.

  • Radionuclide Transport Models (linked to system analysis models):

Calculate radionuclide release and transport within the reactor and surrounding structures.

Calculate radionuclide transport from the reactor to the EAB and transport in the atmosphere (plume dispersion).

  • Dosimetry Computer Models (linked to radionuclide transport models):

Calculate doses within and outside the site boundaries during normal operation and accident conditions. Used to determine whether the plant design meets offsite dose limits and criteria and risk goals.

  • Uncertainty Assessment Computer Models:

Categorize the uncertainties associated with the events source terms and select the most impactful ones to be considered. 76 of 146

1 Advanced-Reactor Source Term -

Pilot Studies Advanced Reactor Stakeholder Meeting August 20, 2020 77 of 146

Non-LWR Evaluation Model 78 of 146 2

3 Project Objectives Develop an understanding of non-LWR beyond-design-basis-accident behavior

  • Provide insights for regulatory guidance
  • Facilitate dialogue on the staffs approach to assessing source term Demonstrate application of MELCOR and SCALE
  • Develop publicly available input models - available upon request
  • Code distribution handled separately 79 of 146

4 Project Stages

1. Select design
2. Develop input deck
3. Select scenarios
4. Perform calculations and refine input deck
  • Full-plant decks have been developed for heat pipe and gas-cooled reactors
  • Salt-cooled reactor input deck in preparation
  • Results shown here are preliminary to illustrate approach
5. Public workshop 80 of 146

5 INL Design A Heat Pipe Reactor 81 of 146

6 INL Design A - Reactor vessel and core nodalization Postulated fission product release pathways Release from gap Release to between fuel cells secondary system and heat pipe (bypass reactor confinement)

Release from gap between fuel cells Release to reactor confinement Adjacent Fuel Element Adjacent Fuel Element Release to heat pipes 82 of 146

7 INL Design A - Reactor building nodalization CV1000 (Environment) Building leakage based on BWR reactor building values FL5010 (Upper Leakage)

FL5025 (Lower Leakage) CV5010 (Reactor Building Floor 2)

FL5015 FL5020 CV5005 (Reactor Building Floor 1)

FL5000 FL5005 (Reactor Cavity Flow) (Reactor Cavity Flow)

Natural Convection Natural Convection UP UP CV5000 (Reactor Cavity) Includes natural Ground circulation flow into the reactor cavity 83 of 146

INL Design A SCALE model

~1 m core diameter

  • Design features

- 1134 annular hexagonal UO2 fuel elements (19.75% 235U)

- Fast neutron spectrum

  • Modeling strategy

- Flux was evaluated assuming a fixed control drum configuration

- Isotopic inventory evaluated at full power over core life

  • Radionuclide inventory and decay heat data provided for MELCOR model 1.5 m active height 84 of 146 8

9 INL Design A - Demo calculations

  • Reference case o Initiator trips secondary heat removal o Control rod insertion o Thermal radiation from the reactor vessel o Natural circulation flow through the reactor cavity
  • Adiabatic case o No convective or radiative heat transfer from the vessel 85 of 146

10 INL Design A - Peak fuel temperatures Fuel cladding &

1800 HP melting Reference Range of HP creep rupture Adiabatic failures (1425 +/- 100 K) 1600 1400 Temperature (K)

HP creep rupture Potential range of 1200 responses is design-and scenario-specific 1000 800 600 0 24 48 72 96 120 144 168 192 216 240 Time (hr) 86 of 146

11 INL Design A - Iodine and cesium release 1.E+00 Iodine release from fuel 1.E-01 Cesium release from fuel Iodine release to environment Cesium release to environment Fraction of initial inventory (-)

1.E-02 The Reference case did not have a release 1.E-03 Adiabatic 1.E-04 Case 1.E-05 1.E-06 0 24 48 72 96 120 144 168 192 216 240 Time (hr) 87 of 146

12 INL Design A - Peak fuel temperatures radial nodalization sensitivity 1700 Range of HP creep rupture 1600 failures (1425 +/- 100 K) 1500 1400 Temperature (K) 1300 1200 1100 1000 Reference (6-ring)

Adiabatic (6-ring) 900 Reference (15-ring)

Adiabatic (15-ring) 800 700 0 6 12 18 24 30 36 42 48 Time (hr) 88 of 146

13 PBMR-400 reactor and core Ring 1 Ring 2 Ring 3 Ring 4 Ring 5 Ring 6 Ring 7 Ring 8 Heat Structures Inlet plenum 129 116 229 126 329 136 429 146 529 156 629 166 729 176 829 186 128 228 170 328 170 428 170 528 170 628 170 728 170 828 170 127 227 327 427 527 627 727 827 Legend 126 115 226 326 426 526 626 726 826 125 135 145 155 165 125 225 325 425 525 625 725 825 Blue Cells: CVH 124 224 324 424 524 624 724 824 Red Cells: Core 123 223 323 423 523 623 123 723 823 Black Cells: Heat Structure Ring 1: Radius=1.000 m 122 222 322 422 522 622 122 722 822 114 124 134 144 154 164 Ring 2: Radius=1.170 m 121 221 321 421 521 621 121 721 821 Ring 2: Radius=1.340 m 120 220 320 420 520 620 120 720 820 Ring 3: Radius=1.510 m Ring 4: Radius=1.680 m 119 219 319 419 519 619 119 719 819 Ring 5: Radius=1.850 m 118 218 318 118 418 518 618 118 718 818 Pebble bed 117 113 217 123 317 133 417 143 517 153 617 117 163 717 817 Ring 6: Radius=2.436 m Ring 7: Radius=2.606 m 171 181 Ring 8: Radius=2.731 m 116 216 316 416 516 616 116 716 816 115 215 315 415 515 615 715 815 114 214 314 414 514 614 714 814 113 112 213 122 313 132 413 142 513 152 613 162 713 813 112 212 312 412 512 612 712 812 111 211 311 411 511 611 711 811 110 210 310 410 510 610 710 810 109 209 309 409 509 609 709 809 108 111 208 121 308 131 408 141 508 151 608 161 708 808 107 207 307 407 507 607 707 807 106 206 306 406 506 606 706 806 105 205 305 405 505 605 705 805 104 204 304 404 504 604 704 804 630 Inlet flow Outlet plenum 103 110 203 110 303 110 403 110 503 110 603 110 703 803 102 202 302 402 502 602 702 802 182 182 101 201 301 401 501 601 701 801 To CV-100 Outlet flow

[P.J. Venter, M.N. Mitchell, F. Fortier, PBMR reactor design and development, in: Proceedings from the 18th International Conference on Structural Ring 1 is the graphite Pebble bed is Ring 7 is the outer graphite reflector Mechanics in Reactor Technology (SMiRT 18), Beijing, China, Aug. 2005]

inner reflector in Rings 2-6 Ring 8 is inlet flow risers & reflector 89 of 146

14 PBMR-400 vessel and reactor building Reactor building RCCS adapted from the nodalization CV-999 (Environment)

RCCS (not part of from MHTGR PBMR-400 design)

CV-83 (Reactor Building Upper)

Environment CV-51 (Cavity 1)

RCCS RCCS Outlet inlet Plenum Plenum GROUND GROUND CV-82 (Reactor RCCS Building Lower) CV-50 CV-75 Riser 4 Conc (Cavity 2)

Walls (4 panels, 3 CV-80 (Reactor CV-81 RCCS with 5' thick, Building Cavity) (Steam Generator CV-74 1 for 2.5' Compartment)

Riser 3 RCCS Down thick) comer RCCS CV-73 Riser 2 RCCS CV-72 Riser 1 Core barrel Rings 1- 8 Core barrel Vessel wall (previous slide) Vessel wall RCCS RCCS Heat structure

[Ducknor, Nuclear Engineering and Technology, 49, 360-372, 2017] CVH volume 90 of 146

[DOE-HTGR-86-024-Vol.1]

PBMR-400 SCALE model

  • Design features

- Fueled by graphite pebbles containing UO2-bearing TRISO fuel particles

- Pebbles circulate multiple passes through the core to achieve a high burnup

  • Modeling strategy

- Analysis focused on understanding axial & radial power shape and the neutron spectrum for depletion calculations

- Facilitate depletion calculations via pre-calculated Origen reactor data libraries

  • Radionuclide inventory and decay heat data provided for MELCOR model PBMR-400 SCALE geometry

& neutron flux profile 91 of 146 15

16 PBMR-400 - Demo Calculations Depressurized loss-of-forced circulation (DLOFC) accident

  • Large recirculation pipe break o Reactor trip o Secondary system trips & isolates o Passive reactor cavity cooling system (RCCS) available
  • Reference case includes nominal heat transfer to the RCCS
  • Vessel to RCCS heat transfer sensitivity o Heat transfer coefficient to air in the RCCS varied from 0 to 5 W/m2 K
  • RCCS blockage sensitivity o Natural circulation air flow area into the RCCS decreased by 90%, 99%, and full blockage TRISO fission product release model o Diffusivity data from IAEA TECDOC-978, Appendix A o Fuel failure fraction is user-specified - temperature dependent curve 92 of 146

17 PBMR-400 - DLOFC results Peak fuel temperature sensitivity to the RCCS Peak fuel temperature sensitivity to RCCS blockage heat transfer coefficient p

2000 2000 1750 1750 1500 1500 Temperature (°C) Temperature (°C) 100% RCCS blockage HTC = 0 W/m2-K 99% RCCS Blockage 1250 HTC = 1 W/m2-K 1250 90% RCCS Blockage HTC = 2 W/m2-K No Cavity Blockage (5 W/m2-K)

HTC = 3 W/m2-K 1000 HTC = 4 W/m2-K 1000 HTC = 5 W/m2-K 750 750 500 500 0 24 48 72 96 120 144 168 0 24 48 72 96 120 144 168 Time (hr) Time (hr) 93 of 146

18 PBMR-400 - DLOFC reference case results Release from the pebbles to the coolant Release to the environment 1.E+00 1.E+00 1.E-01 Xe, I 1.E-01 Cs Ba Fraction of initial inventory (-) Fraction of initial inventory (-)

1.E-02 Xe, I 1.E-02 Cs Ag 1.E-03 Ba 1.E-03 Ag 1.E-04 1.E-04 1.E-05 1.E-05 1.E-06 1.E-06 1.E-07 1.E-07 0 24 48 72 96 120 144 168 0 24 48 72 96 120 144 168 Time (hr) Time (hr) 94 of 146

19 Fluoride-Salt-Cooled High-Temperature Reactor (FHR) 95 of 146

FHR SCALE model

  • Design features

- TRISO particles with UCO fuel kernel (19.9% 235U) in graphite pebbles

- 236 MWth core with approx.

470,000 fuel pebbles &

218,000 graphite pebbles

- FLiBe salt coolant

  • Modeling strategy

- Fixed pebble positions (no buoyancy effects)

~5.7 m 3.0 cm

  • Radionuclide inventory and decay heat data provided for MELCOR model Graphite Fuel pebble pebbles Fuel pebbles 96 of 146 20

MELCOR fission product release model Vaporization and bubble Radionuclides grouped into 6 forms as found in the burst release (see Vol. 3) Molten Salt Reactor Experiments at ORNL Salt droplets with soluble & insoluble FP from bursting bubbles Soluble Insoluble colloidal Insoluble gas/liq. Insoluble surface Gases /Vapors Initial (salt-seeking)

Gases (Xe, Kr, T) and suspension interphase colloid deposit Kr, Xe, Cs, I State Form 1 volatiles (Cs, I) Form 2 Form 3 Form 4 Form 5 Transitions by Insoluble Insoluble Insoluble Agitation Mass Transfer Solubility colloid Interface Wall Adds More and Temp.

limit? migrate Colloid Deposit Bubbles Changes yes up to the limit Radionuclide- Pebbles with contaminated intact or failed molten salt TRISO Bubble Film Rupture Releases Aerosol Gas Form 6 Release 97 of 146

22 Concluding remarks and next steps Preliminary working input models

  • INL Design A - November 2020
  • PBMR-400 - November 2020
  • FHR model - March 2021 Followed by public workshops New computer code versions will be released with updated phenomenological models 98 of 146

Break Meeting/Webinar will resume shortly Telephone Bridgeline: (888) 810-4937 Passcode: 8854397#

99 of 146

Annual Fee Regulations for Non-Light Water Reactors August 20, 2020 100 of 146

©2020 Nuclear Energy Institute

Current Annual Fee Regulations Annual fees outlined in 10 CFR Part 171, governed by OBRA-90 Variable fee structure established for light-water SMRs in June 2016 Currently, annual fees not technology-inclusive and apply only to LWRs Timely consideration necessary given micro-reactor COL application docketed by NRC and more non-LWR developers in pre-application discussions with the NRC 101 of 146

©2020 Nuclear Energy Institute 2

Goals to Consider in Fee Rule Change Urgent need for annual fee regulations for non-LWRs; important for investment decisions Meet NEIMA requirements (FY 2021 and beyond)

Regulatory costs shared fairly and equitably among large and smaller-scale reactor facilities, as well as among various technologies Reasonable relationship to cost of regulatory services Ensure continued protection of public health and safety 102 of 146

©2020 Nuclear Energy Institute 3

Preferred Annual Fee Rule Approach Expand the SMR variable fee structure to include non-LWRs Basis for light-water SMR variable annual fee is equally applicable to non-LWRs Maximum, minimum and variable fees are appropriate for large &

SMR non-LWRs Address disproportionate impacts to micro-reactors Current minimum fee too high for micro-reactors; causes disproportionate impacts and overestimates regulatory costs Three options considered:

1. Amend variable fee structure
2. Fee cap to avoid disproportionate impact 103 of 146
3. Separate fee structure for micro-reactors ©2020 Nuclear Energy Institute 4

Evaluation of Disproportionate Impact Thermal Power 5 10 30 50 75 100 Rating (MWt)

Current Annual

$134,650 $134,650 $134,650 $134,650 $134,650 $134,650 Fee Annual Plant $554,800 $1,109,600 $3,328,800 $5,548,000 $8,322,000 $11,096,000 Generating Cost Annual Fee as Percent of Annual 24.27% 12.14% 4.05% 2.43% 1.62% 1.21%

Plant Generating Cost

  • All numbers are preliminary estimates; calculations use generating cost of $40/MWh1 for micro-reactors, 95% capacity factor 104 of 146

©2020 Nuclear Energy Institute 5 1NEI Report, Cost Competitiveness of Micro-Reactors for Remote Markets (April 15, 2019).

Evaluation of Options to Address Disproportionate Impact

1. Amend variable fee structure Re-align minimum fee to micro-reactor range (100MWt)*

Use current variable fee rate to extend down; or Set new minimum based on reduced regulatory costs

2. Fee cap to avoid disproportionate impact Create fee cap based on power level for those micro-reactors who would experience disproportionate impact (annual fee > 3% of annual generating cost)*

Reactors with thermal power ratings less than 40.5 MWt pay

$3,330/MWt; reactors 40.5MWt - 250MWt pay minimum fee*

105 of 146

  • All numbers are preliminary estimates

©2020 Nuclear Energy Institute 6

Fee Cap to Avoid Disproportionate Impact Thermal Power 5 10 30 40.5 75 100 Rating (MWt)

New Annual Fee $16,650 $33,300 $99,900 $134,650 $134,650 $134,650 Annual Plant $554,800 $1,109,600 $3,328,800 $5,548,000 $8,322,000 $11,096,000 Generating Cost Percentage of Annual Cost Under 24.27% 12.14% 4.05% 2.43% 1.62% 1.21%

SMR Structure Percentage of Annual Cost Under 3.00% 3.00% 3.00% 2.43% 1.62% 1.21%

Fee Cap

  • All numbers are preliminary estimates; calculations use generating cost of $40/MWh1 for 106 of 146 micro-reactors, 95% capacity factor ©2020 Nuclear Energy Institute 7 1NEI Report, Cost Competitiveness of Micro-Reactors for Remote Markets (April 15, 2019).

Evaluation of Options to Address Disproportionate Impact

3. Separate fee structure for micro-reactors Similar to separate SMR fee structure, create separate micro-reactor fee structure within power reactor fee class Annual fee proportionate to ~1.2% of estimated annual generating cost, to remain fair and equitable to current fleet (Part 171 annual fees constitute an average of ~1.2% of annual generating costs for current fleet)*

Micro-reactors (less than 100MWt) pay $1,360 /MWt*

107 of 146

  • All numbers are preliminary estimates

©2020 Nuclear Energy Institute 8

Separate Fee Structure For Micro-Reactors Thermal Power 5 10 30 50 75 100 Rating (MWt)

New Annual Fee $6,800 $13,600 $40,800 $68,000 $102,000 $134,650 Annual Plant $554,800 $1,109,600 $3,328,800 $5,548,000 $8,322,000 $11,096,000 Generating Cost Percentage of Annual Cost Under 24.27% 12.14% 4.05% 2.43% 1.62% 1.21%

SMR Structure Percentage of Annual Cost Under 1.23% 1.23% 1.23% 1.23% 1.23% 1.21%

New Structure

  • All numbers are preliminary estimates; calculations use generating cost of $40/MWh1 for micro-reactors, 95% capacity factor 108 of 146

©2020 Nuclear Energy Institute 9 1NEI Report, Cost Competitiveness of Micro-Reactors for Remote Markets (April 15, 2019).

Path Forward Release of NEI position paper on non-LWR annual fees, end of August Develop non-LWR annual fees; costs for developing advanced reactor regulatory infrastructure fee-exempt under NEIMA2 (until 2031)

Use future operating experience of SMRs and non-LWRs to:

Verify the expectations that advanced reactors require less regulatory service due to improved safety and simplicity Refine the SMR and micro-reactor annual fees as detailed information becomes available 109 of 146

©2020 Nuclear Energy Institute 10 2 See Section 102 (b)(1)(B)(iii) of the Nuclear Energy Innovation and Modernization Act, Public Law 115-439

10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors August 20, 2020 110 of 146 1

Background

  • Advance Notice of Proposed Rulemaking, Approaches to Risk-Informed and Performance-Based Requirements for Nuclear Power Reactors, dated May 4, 2006 (71 FR 26267)
  • NRCs Vision and Strategy report (12/16) for non-light-water reactors and related implementation action plans identified a potential rulemaking to establish a regulatory framework
  • Nuclear Energy Innovation and Modernization Act (NEIMA; Public Law 115-439) signed into law in January 2019 requires the NRC to complete a rulemaking to establish a technology-inclusive, regulatory framework for optional use for commercial advanced nuclear reactors no later than December 2027
  • Periodic Stakeholder Meeting - October 10, 2019 111 of 146 2

Background - NEIMA (1) ADVANCED NUCLEAR REACTORThe term advanced nuclear reactor means a nuclear fission or fusion reactor, including a prototype plant with significant improvements compared to commercial nuclear reactors under construction as of the date of enactment of this Act, (9) REGULATORY FRAMEWORKThe term regulatory framework means the framework for reviewing requests for certifications, permits, approvals, and licenses for nuclear reactors.

(14) TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORKThe term technology-inclusive regulatory framework means a regulatory framework developed using methods of evaluation that are flexible and practicable for application to a variety of reactor technologies, including, where appropriate, the use of risk-informed and performance-based techniques and other tools and methods.

112 of 146 3

SECY-20-0032, Rulemaking Plan

  • SECY-20-0032, Rulemaking Plan on Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors, dated April 13, 2020
  • Proposing a new 10 CFR part that could address performance requirements, design features, and programmatic controls for a wide variety of advanced nuclear reactors throughout the life of a facility.
  • Focus the rulemaking on risk-informed functional requirements, building on existing NRC requirements, Commission policy statements, and recent activities (e.g., SECY-19-0117)
  • Awaiting Commissions Staff Requirements Memorandum; including schedule goals 113 of 146 4

Technology Inclusive Regulatory Framework Project Life Cycle Requirements Functional System Construction Operation Retirement Definition Design Design

  • Fundamental Safety Functions
  • Prevention, Mitigation, Testing Surveillance Performance Criteria Maintenance (e.g., F-C Targets)
  • Normal Operations Configuration (e.g., effluents) Control
  • Other Design Changes Plant/Site (Design, Construction, Configuration Control)

Clarify Controls Analyses (Prevention, Mitigation, Compare to Criteria) and Distinctions Between Plant Documents (Systems, Procedures, etc.)

LB Documents (Applications, SAR, TS, etc.)

114 of 146 5

Example - Possible Layout

  • General Provisions
  • Technology-Inclusive Safety Objectives o Regulatory limits, safety goals
  • Design Requirements
  • Siting
  • Construction and Manufacturing Requirements
  • Requirements for Operation
  • Decommissioning Requirements
  • Applications for Licenses, Certifications and Approvals
  • Maintaining and Revising Licensing Basis Information
  • Reporting and Administrative Requirements 115 of 146 6

NRC Staff White Paper

  • The NRC staff developed a white paper (ADAMS ML20195A270) to support discussions with ACRS and other stakeholders
  • Soliciting information that:
1) Defines the scope of stakeholder interest in a rulemaking to develop a technology inclusive framework for advanced nuclear reactors,
2) Identifies major issues and challenges related to technology-inclusive approaches to licensing and regulating a wide variety of advanced nuclear reactor designs,
3) Supports prioritizing and developing plans to resolve identified issues within the rulemaking for the wide variety of advanced nuclear reactor designs, and
4) Supports the development of the proposed rule and related guidance.
  • Staff receptive to feedback on any aspect of developing a technology-inclusive regulatory framework to support the regulatory objective, whether or not in response to a question listed in this white paper or future solicitations.

116 of 146 7

Part 53 Rulemaking Objectives

1) Provide reasonable assurance of adequate protection of the public health and safety and common defense and security at reactor sites at which advanced nuclear reactor designs are deployed, to at least the same degree of protection as required for current-generation light water reactors;
2) Protect health and minimize danger to life or property to at least the same degree of protection as required for current-generation light water reactors;
3) Provide greater operational flexibilities where supported by enhanced margins of safety that may be provided in advanced nuclear reactor designs;
4) Ensure that the requirements for licensing and regulating advanced nuclear reactors are clear and appropriate; and
5) Identify, define, and resolve additional areas of concern related to the licensing and regulation of advanced nuclear reactors.

117 of 146 8

Questions for Public Feedback

1. Regulatory Objectives o Appropriate, understandable, achievable ?
2. Scope and Types of Advanced Nuclear Reactors o Limit to advanced reactors as defined in NEIMA?
3. Technical Requirements versus Licensing Process o Limit to regulations related to technical standards?

o Alternative licensing processes?

4. Performance Criteria o Technology-inclusive performance criteria?
5. Risk Metrics o Include risk metrics in the regulations?
6. Facility Life Cycle o How could new Part 53 align with facility life cycle
7. Definitions o Should Part 53 use existing definitions 118 of 146 9

Questions for Public Feedback

8. Performance-Based Regulation o How to incorporate performance-based concepts?
9. Identifying Levels of Protection o Differentiate requirements for adequate protection and safety improvements?
10. Integrated Approach to Rulemaking o How to integrate safety, security, emergency preparedness?
11. Consistency with Historical Standards o Use of existing standards (e.g., safety goals)?
12. Quality Standards o Recognize alternatives to Appendix B?
13. Stakeholder Documents, Standards, Guidance o Stakeholder interest in preparing guidance?
14. Other Issues?

119 of 146 10

Path Forward

  • Awaiting Commission Decision on Rulemaking Plan (SECY-20-0032)
  • Some stakeholders recommending accelerating schedule from rulemaking plan/NEIMA o See Letter dated May 14, 2020 from Senator Barrasso, Chairman Committee on Environment and Public Works (ML20136A164),

and Response dated June 17, 2020 from Chairman Svinicki (ML20155K912)

  • Accelerating schedule would result in need to have more active stakeholder engagement during 2021
  • Public meeting dedicated to developing Part 53 tentatively scheduled for September 17th o White paper (ADAMS ML20195A270) provides possible topics 120 of 146 11

Part 53 Rulemaking Marc Nichol Senior Director New Reactors August 20, 2020 121 of 146

©2019 Nuclear Energy Institute

1. Regulatory Objectives Establish a regulatory framework for new reactors that:
  • Provides reasonable assurance of adequate protection of the public health and safety and common defense and security
  • Is risk-informed, performance-based, and technology-inclusive
  • Is clear, flexible and efficient
  • Enables efficient foreign licensing of NRC approved designs Utilizes a rulemaking process that:
  • Starts with only the necessary legal requirements (e.g., AEA) as a blank-sheet approach
  • Considers all known, and unknown, reactor technologies
  • Benefits from lessons-learned through near-term licensing of new reactors 122 of 146

©2019 Nuclear Energy Institute 2

2. Scope and Types of Reactors
3. Type of Requirements Part 53 should be more inclusive, not less inclusive
  • All new reactor applications
  • All types of applications
  • All uses and applications
  • All power levels Address requirements based on needs
  • Technical requirements - complete redesign
  • Administrative requirements - improve efficiency, potential for some to be eliminated
  • Process requirements - utilize 123 Part of 146 50 and 52, improve/add additional flexibility and efficiency ©2019 Nuclear Energy Institute 3
12. Quality Assurance Appendix B
  • Innovative thinking when created in early 1970s
  • Only used by US nuclear industry
  • Shrinking supply chain ISO-9001
  • Achieves equivalent level of quality with Appendix B
  • Utilized world-wide by millions Benefits of using ISO-9001
  • Access to larger supply chain (higher quality)
  • Informed by broad experience (best practices)
  • Adoption of standards (more efficient) 124 of 146

©2019 Nuclear Energy Institute 4

9. Levels of Protection
7. Definitions Requirements should contain
  • High level standard: reasonable assurance of adequate protection
  • Inclusive performance objectives
  • Flexible for different licensing approaches Guidance could include
  • Technology specific acceptance criteria Need to create the Part 53 safety paradigm before addressing terms and definitions
  • Construct of demonstrating reasonable assurance of adequate protection E.g., rethink: design basis, safety-related, defense-in-depth 125 of 146
  • Balance of deterministic and probabilistic methods ©2019 Nuclear Energy Institute 5

Industrys Activities

1. Evaluation of Atomic Energy Act
  • Statutory requirements relevant to Part 53
  • Statutory requirements may need to be modified
2. Envision a new Part 53 safety paradigm
  • Create a new bridge from AEA to Part 53
  • Consider scope of reactor technologies
  • Promote flexibility and efficiency
  • Evaluate international regulatory paradigms
3. Evaluate existing regulatory framework to identify what should be new for, and what could be incorporated into, Part 53
  • Scope (e.g., security, decommissioning) 126 of 146
  • Regulatory precedent (e.g., risk metrics, performance criteria) ©2019 Nuclear Energy Institute 6

QUESTIONS?

127 of 146 By Third Way, GENSLER

U.S. Nuclear Industry Council Comments regarding Part 53 at NRC Stakeholders Meeting Cyril W. Draffin, Jr.

Senior Fellow, Advanced Nuclear U.S. Nuclear Industry Council 20 August 2020 128 of 146

Overall Comments

  • USNIC welcomes opportunity to engage with NRC to develop Part 53
  • will actively participate in NRC Part 53 discussion in September 2020
  • USNIC providing NRC with 50 comments addressing each of the 14 issues that the NRC raised in their July 2020 NRC Staff White Paper
  • 14 issues NRC identified is a good start for Part 53 planning
  • only a few of USNIC specific comments presented in these slides (due to time)
  • Goal should be to craft a flexible Part 53 process that is so well defined that developers want to use it over existing Parts 50 and 52
  • Part 53 should be technology inclusive 1 l U.S. Nuclear Industry Council Aug 2020 - Part 53 129 of 146

Specific Comments (on selective issues)

1. Regulatory Objectives
  • NRC regulatory objectives for Part 53 are generally good
2. Scope
  • Scope should be inclusive of all future applications and technologies.
  • Scope should be graded approach to facilitate First-Of-A-Kind reviews but flexible enough to accelerate nth of a kind reviews
  • Part 53 should be available to all Advanced Reactors technologies, but Advanced Reactor developers should not be compelled to use
3. Licensing Process
  • Should address licensing, administrative, procedural, reporting and inspection matters for Advanced Reactor applications
  • Goal to meet adequate protection standards, but in way that focuses on public health and safety and avoids unnecessary burden 2 l U.S. Nuclear Industry Council Aug 2020 - Part 53 130 of 146

Specific Comments (on selective issues)

9. Levels of Protection
  • NRC should not use the development of this rule to rachet up requirements
  • May be helpful to identify what prior regulations have been justified as cost-effective safety improvements
10. Integrated Approach
  • Desirable to apply risk informed approaches to safety, security and emergency preparedness (as Commission did recently for Emergency Planning Zones)
12. Quality Assurance
  • Part 53 provides opportunity for NRC to take a fresh look at Appendix B and NQA-1 Program
  • Level of quality of commercially available components meets and frequently exceeds prior nuclear standards without the need for the overly burdensome reporting requirements
  • Alternative approaches such as ISO 9000 series and commercial dedication programs should be considered
14. Other issues
  • When available, we look forward to understanding timeline for the Commission to review and vote on Part 53 SECY paper 3 l U.S. Nuclear Industry Council Aug 2020 - Part 53 131 of 146

Closing Comments

  • USNIC believes today is first step on interactive approach to developing an effective and useful Part 53
  • USNIC welcomes opportunity to continue the dialog with NRC staff to achieve rule that is fully effective in meeting the Adequate Protection Standard-- but does in a way that allows Advanced Reactors to be developed, licensed, and deployed in a manner that avoid unnecessary burden --

and enables the deployment of these important contributors to avoiding carbon emissions 4 l U.S. Nuclear Industry Council Aug 2020 - Part 53 132 of 146

U.S. Nuclear Industry Council Contacts For questions contact Cyril W. Draffin, Jr. Jerey S. Merri"eld Senior Fellow, Advanced Nuclear, Chairman, US Nuclear Industry Council U.S. Nuclear Industry Council Advanced Reactors Task Force Cyril.Draffin@usnic.org U.S. NRC Commissioner (1998-2007)

Je.Merri"eld@pillsburylaw.com 5 l U.S. Nuclear Industry Council Aug 2020 - Part 53 133 of 146

Status - Spent Fuel Reprocessing Rulemaking Jonathan Marcano, P.E.

NMSS/DFM August 20, 2020 134 of 146

Background

  • In 2013, the Commission directed the staff to develop a reprocessing rule focused on light water reactors (SRM-SECY-13-0093).

- Limited scope to resolving Gap 5 (of 21) - safety and risk analysis.

- Engage DOE to assess ongoing activities.

- Regulatory basis for rule due 3/31/2021.

  • Between 2013-2016, the NRC staff worked to develop a draft regulatory basis for Gap 5.
  • In 2016, NRC suspended the work.

- NRC budgetary constraints.

- Apparent lack of commercial interest in constructing and operating a reprocessing facility.

2 135 of 146

Background

  • On March 4, 2020, NRC held a public meeting to discuss status of the proposed rulemaking and to obtain stakeholder input.

- Staff informed stakeholders that a limited scope rulemaking would cost approximately $2.4 million dollars.

- Assess interest regarding continuation of rulemaking.

  • On May 28, 2020, the Nuclear Energy Institute (NEI) and American Nuclear Society sent letters encouraging the NRC to assess the needs of advanced reactors prior to discontinuing efforts on the proposed rulemaking.

3 146 136 of

Current State

  • NRC staff assessed the interest from the Advanced Reactor community and engaged with DOE to determine the need to continue rulemaking activities.

- Some designers have the capability to eventually source their fuel from the spent fuel of other reactors.

- NRC staff is not aware of any definitive vendor interest in pursuing reprocessing activities in the near future (next decade).

- No near-term industry or DOE initiatives are currently planned or undergoing associated with reprocessing of spent light water reactor fuel or potential efforts to reprocess spent HALEU fuel for reuse in advanced reactors.

  • NEI working group to assess community interest.

4 137 of 146

Next Steps

  • NRC staff plans to inform the Commission of its recommendation regarding any proposed rulemaking for spent fuel reprocessing on or before 3/31/2021.
  • In the future, NRC staff encourages early interactions from developers on anticipated needs or activities involving reprocessing.

5 138 of 146

Contacts and References

  • Email feedback to Jonathan.Marcano@nrc.gov and Tom.Boyce@nrc.gov.
  • References

- March 4, 2020, Public Meeting Summary (ADAMS Accession No. ML20077K144).

- Letter from the Nuclear Energy Industry (ADAMS Accession No. ML20154K554).

- Letter from the American Nuclear Society (ADAMS Accession No. ML20154K530).

- SRM-SECY-13-0093, Reprocessing Regulatory Framework -

Status and Next Steps, dated November 4, 2013 (ADAMS Accession No. ML13308A403).

6 139 of 146

Overview of the Oak Ridge National Laboratory Report on Preparing and Reviewing a Molten Salt Non-Power Reactor Application William B. Kennedy Project Manager Non-Power Production and Utilization Facility Licensing Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities U.S. Nuclear Regulatory Commission 140 of 146

Background

  • Under contract with NRC, Oak Ridge National Laboratory developed a report titled, Proposed Guidance for Preparing and Reviewing a Molten Salt Non-Power Reactor Application 141 of 146

Overview of the Report

  • An information resource for stakeholders interested in licensing of non-power MSRs
  • Based on NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors
  • Focuses on the technical information needed to apply NUREG-1537 to a non-power MSR licensing application 142 of 146

Overview of the Report

  • Covers topics including:

- Siting

- Design of structures, systems, and components

- Reactor description

- Reactor cooling systems

- Engineered safety features

- Instrumentation and control systems

- Auxiliary systems

- Radiation protection and waste management

- Accident analysis

- Technical specifications 143 of 146

Future Plans

  • The NRC staff intends to endorse the report for use by potential non-power MSR applicants by January 2021
  • Subsequently, the report will be incorporated in durable guidance (likely the next revision of NUREG-1537)
  • An update will be provided at the next advanced reactors planning meeting and any feedback on the report will be welcome 144 of 146

How to Get the Report

  • Available on the NRCs Agencywide Documents Access and Management System (ADAMS) at Accession No. ML20219A771
  • Contact me at william.kennedy@nrc.gov 145 of 146

Future Meeting Planning and Open Discussion 2020 Tentative Schedule for Periodic Stakeholder Meetings August 25 (GEIS for Advanced Reactors)

August 27 (TICAP, ARCAP, and Construction Permit)

September 17 (10 CFR Part 53)

September 24 (TICAP and ARCAP)

October 1 November 5 146 of 146

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