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DOE OSTI2017

AGR-5/6/7 Fuel Specification

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) program is developing and qualifying fuel for a first of a kind high temperature gas-cooled reactor (HTGR) and associated systems to generate electrical power and produce hydrogen or process heat. One of the critical technology development areas of the program is nuclear fuel development and qualification. Fuel qualification is needed to support licensing, operation, and near term deployment of an HTGR reactor.5 Toward this end, the Advanced Gas Reactor (AGR) program includes manufacturing fuels for a series of irradiation experiments in the Advanced Test Reactor (ATR) at INL.5 The fifth, sixth, and seventh irradiation experiments, called “AGR 5/6/7,” will use fuel manufactured entirely in engineering scale equipment to demonstrate the performance of fuels fabricated in a production environment. This fuel will be composited from multiple low enriched uranium carbide/oxide (LEUCO) kernel forming runs, multiple tristructural isotropic (TRISO) coating runs, and multiple overcoating runs. The fuel will be composited into lots as kernels, TRISO particles, and compacts. Previous AGR fuel specifications1,2,8,9 provide the basis for most of the following fuel specifications, which are specific to the AGR 5/6/7 experiment, and may not be immediately applicable to fuel produced for an HTGR reactor. Fuel specifications will be influenced by the final reactor design.

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DOE OSTI2017

Experimental Capability for Preparation and Purification of Molten Salts to Support Advanced Reactor Concepts, Nuclear Hybrid Energy Systems and Dynamic Grid Energy Storage Concepts

High-temperature multi-fluid, multi-loop test facility to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications is being developed at Idaho National Laboratory (INL). In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. The main objectives to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers at prototypical operating conditions, flow, and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for light water reactors at prototypical conditions. Fluoride salt coolants are eutectic binary or tertiary mixtures of fluoride salts with melting points in the range of 320 to 500 deg. C. Fluoride high temperature reactors have reactor outlet temperatures of 600 deg. C or higher for high-efficiency power generation or process heat applications. To address research needs related to salt coolants experimental capability has been added to study thermodynamic properties such as melting point temperature, heat capacity, and thermal conductivity.

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DOE OSTI2026

Survey of emerging nuclear data needs for nonproliferation applications with advanced reactors

Nuclear science plays a key role in non-proliferation activities supporting advanced reactor technologies. Nuclear data underpin predictions and interpretations of nuclear material behavior and signatures in reactor fuel production, use, transport, and storage. Advanced reactors provide new challenges compared to the current fleet of thermal fission reactors. This report consolidates reported nuclear data needs from representative workshops, conferences, and publications, identifying six themes for recommended future investments supporting non-proliferation and safeguards applications. While also identified as data needs, major fission product evaluations and (α,n) reactions were omitted as there are ongoing activities producing new data under NA22/Objective O. Each theme is summarized below with example data and association with the nonproliferation mission for advanced fuels and reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Phase II Development of the Surveillance Test Articles to Improve the Design, Fabrication, and Testing

Advanced reactors, such as the molten salt reactor (MSR), require materials that will withstand harsher environments than the materials used for lower temperature water reactors. Materials used in MSR construction need not only withstand elevated temperature, temperature cycling, and neutron radiation, but must be able to withstand the corrosive molten salt environment. Significant materials research has been driven by the materials needs of the MSR due to the harsh environment and the material data requirements to support licensing. Limited operation experience with MSRs has made this challenging work. Information is very limited on materials degradation due to irradiation, molten salt corrosion, elevated temperature, and the resulting fatigue, creep, and creep-fatigue loading during operation. While efforts are underway to better understand the effects of this harsh environment on construction materials, the Advanced Reactors Technologies (ART) Program has been working to develop materials surveillance test articles that could be used in a materials surveillance program and allow for the collection of information on the materials degradation during plant operation and could support timely licensing of these advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

RELAP5-3D Modeling in the OECD-NEA HTTF Benchmark

Prismatic gas-cooled reactors are a technologically mature reactor concept of interest for near- to mid-term deployment. To accelerate the deployment of these reactors, the DOE's Advanced Reactor Technologies Gas-Cooled Reactor campaign is spearheading a thermal hydraulics code validation benchmark based on the High Temperature Test Facility at Oregon State University. This presentation provides an introduction to that benchmark and an overview of some of the RELAP5-3D validation activities stemming from that benchmark.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Preliminary Procedures and Acceptance Criteria for in-situ Structural Materials Surveillance for MSR

This report describes the initial development of the technical basis for a materials surveillance technology that, when fully developed and validated, can be used by stakeholders to develop and implement a materials surveillance program to manage materials degradation during reactor operations. The procedures described in this report aim to bound the detrimental effects of the reactor environment on the creep-fatigue life of the reactor components through materials surveillance. Specifically, the report describes procedures for selecting critical locations within the reactor components for surrogate materials surveillance and the sizing and placements of passively-actuated materials surveillance test articles under development by the US Department of Energy, Advanced Reactor Technologies Program. These test articles apply cyclic thermomechanical load to a surrogate sample of the component structural material passively through thermal expansion mismatch within the articles, without requiring any penetrations in the reactor coolant boundary. The procedures can be used to size materials surveillance test articles so that they will fail before the corresponding structural component. These articles would be exposed to the component operating environment, located to experience a bounding radiation fluence and representative coolant exposure. They can be monitored during reactor operations to ensure that they have not failed under the combination of creep-fatigue damage and environmental degradation. The report describes the key technical decisions and the corresponding rational for using these procedures to provide advanced warning of any impending material failure. An appendix describes the technical basis and the procedure for sizing a family of passively-actuated test articles. Finally, the report describes future work needed to complete the technical basis underlying this materials surveillance technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Accelerated statistical failure analysis of multifidelity TRISO fuel models

Statistical nuclear fuel failure analysis is critical for the design and development of advanced reactor technologies. Although Monte Carlo Sampling (MCS) is a standard method of statistical failure analysis for fuels, the low failure probabilities of some advanced fuel forms and the correspondingly large number of required model evaluations limit its application to low-fidelity (e.g., 1-D) fuel models. In this paper, we present four other statistical methods for fuel failure analysis in Bison, considering tri-structural isotropic (TRISO)-coated particle fuel as a case study. The statistical methods considered are Latin hypercube sampling (LHS), adaptive importance sampling (AIS), subset simulation (SS), and the Weibull theory. Using these methods, we analyzed both 1-D and 2-D representations of TRISO models to compute failure probabilities and the distributions of fuel properties that result in failures. The results of these methods compare well across all TRISO models considered. Overall, SS and the Weibull theory were deemed the most efficient, and can be applied to both 1-D and 2-D TRISO models to compute failure probabilities. Moreover, since SS also characterizes the distribution of parameters that cause TRISO failures, and can consider failure modes not described by the Weibull criterion, it may be preferred over the other methods. Finally, a discussion on the efficacy of different statistical methods of assessing nuclear fuel safety is provided.

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DOE OSTI2021

Optimization Methodology of Pebble Bed HTGR Start-Up and Running-in Strategy

In recent years interest in advanced reactor technologies has increased significantly. However, the methods used for analysis of traditional nuclear reactors are insufficient to consider all the different and varied advanced reactor designs without further development. One promising advanced reactor design is the pebble bed reactor (PBR). PBRs possess unique operational and fuel cycle features that require the development of specific analysis methodologies to adequately design and analyze the systems. There is a need in PBR research for a capability to analyze and optimize the process of transitioning from the start-up reactor core to the equilibrium core (known as the “running-in” of the reactor). The start-up of a PBR and the transition to the equilibrium core is a complex, multi-physics challenge that has not yet been well researched and has many opportunities for design, analysis, and optimization of the process. In this research, a methodology is defined to consider the potential strategies in PBR start-up and run-in to the equilibrium core. Multiple candidate software are considered, and their pros and cons are discussed for the PBR-specific application in the methodology. A preliminary software selection for the physics engine is made, and initial verification of key modules is performed. Software to support optimization of the reactor run-in strategies through reduced order modeling (ROM) and machine learning are considered. Challenges for a full implementation of the methodology are discussed. Additional code selections and verification of models relevant to the application are needed before full demonstration of the methodology can be achieved and an optimal strategy determined.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Initial Demonstration of New Griffin Technologies for Simulating the Running-In Phase of Pebble Bed Reactors

Griffin is a reactor multiphysics modeling application based on MOOSE (Multiphysics Object-Oriented Simulation Environment) and specifically targeting transient modeling of advanced reactors. Griffin has been used recently to model pebble-bed reactors for the Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research and the Advanced Reactor Technology program. This modeling work has focused thus far on the direct calculation of equilibrium cores. This report documents an initial demonstration of a new running-in simulation capability. The new running-in capability is verified using the existing direct equilibrium core calculation capability. A simplified pebble-bed reactor model is then used to demonstrate the running-in simulation capability. This demonstration shows that Griffin is able to simulate years of operation during the running-in phase efficiently with each depletion step taking only several seconds. Two new technologies are also presented in this report which have been developed in Griffin that will be essential for improved accuracy both of the direct equilibrium core computation and the new running-in simulation capability. The first technology is an online cross section generation capability specifically targeted for pebble-bed reactors. This will improve the accuracy of the depletion calculation as the cross sections are generated at the exact core status. This also avoids the difficult step of pre-generating a separate standalone multigroup cross section set. Secondly, a newly implemented discretization for discontinuous finite element method (DFEM) SN transport in cylindrical (RZ) coordinates, which can be solved efficiently using the existing SN sweep solver, is discussed and some results are shown demonstrating the usefulness of the additional accuracy transport provides over a diffusion approximation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Two-Dimensional Neutronic and Fuel Cycle Analysis of the LEU-Fueled Transatomic Power Molten Salt Reactor

A recent Third Way report detailing $1.3 billion in private investment in advanced reactor technology includes several liquid-fueled molten salt reactor (MSR) concepts. Interest in these MSR concepts is driven by the enhanced safety, economic, and promising fuel cycle benefits of these advanced reactor concepts. The US Department of Energy (DOE) Office of Nuclear Energy established the Gateway for Accelerated Innovation in Nuclear (GAIN) to provide private companies pursuing innovative nuclear energy technologies with access to the technical support necessary to move toward commercialization. GAIN activities in 2016 included competitive selection of projects for vouchers to fund applicant-specific work at national laboratories. One of these GAIN small business voucher awards was to Transatomic Power Corporation (TAP) to work with Oak Ridge National Laboratory (ORNL) to perform neutronic and fuel cycle analysis and design of TAP's innovative MSR concept. The initial phase of this project is to perform two-dimensional (2D) analysis with ORNL-developed tools to verify TAP-calculated metrics and to develop an understanding of the effects of fuel feed and isotopic removal rates to prepare for the more complex three-dimensional (3D) analysis in coming phases. Design information was based on the TAP white papers wherever possible. This summary discusses the results of this initial phase of the collaboration between ORNL and TAP.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

AGR-2 PIE at Oak Ridge National Laboratory, Rev 5

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under potential accident conditions. PIE work on fuel from the second test irradiation, Advanced Gas Reactor-2 (AGR-2), began at INL in July 2014. This work scope includes Oak Ridge National Laboratory (ORNL) providing technical input, performing PIE testing and analysis, and contributing expertise to this effort.

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DOE OSTI2020

AGR-2 PIE at Oak Ridge National Laboratory

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under potential accident conditions. PIE fuel work from the final test irradiation (Advanced Gas Reactor [AGR]-5/6/7) is expected to commence at INL in early 2021, but the PIE preparations work began in FY2016. The work scope in this statement of work includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities; technical input to the moisture/air-ingress furnace design, fabrication, and equipment qualification; and technical support for development of equipment and techniques for planned PIE evolutions.

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DOE OSTI2021

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

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DOE OSTI2026

Effect of Surrogate Fission Products on Thermal Conductivity and Viscosity of NaCl-UCl 3

Thermophysical properties of salt systems relevant to molten salt reactors (MSRs) are experimentally measured in support of the US Department of Energy Office of Nuclear Energy (DOE-NE) MSR campaign within the Advanced Reactor Technology (ART) Program. These property measurements also support the development of the thermophysical arm of the Molten Salt Database (MSD) within DOE-NE’s Nu clear Energy Advanced Modeling and Simulation (NEAMS) Program. Several US Department of Energy (DOE) laboratories, including Oak Ridge National Laboratory (ORNL), have been capitalizing on modern methodologies and sample characterization techniques to conduct thermophysical property measurements to support these programs and to provide MSR developers and modelers with more recent and higher-quality data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the last in the series of test irradiations, Advanced Gas Reactor (AGR)-5/6/7, will begin at Idaho National Laboratory (INL) in approximately September 2020, but because of the complexity of the experiment preparations for the PIE are going to begin in FY2016. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities, providing technical input to the moisture furnace design, providing technical support for development of equipment and techniques for planned PIE evolutions, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

INL ART TDO AGR-5/6/7 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the last in the series of test irradiations, Advanced Gas Reactor (AGR)-5/6/7, will begin at Idaho National Laboratory (INL) in approximately September 2020, but because of the complexity of the experiment preparations for the PIE are going to begin in FY2016. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities, providing technical input to the moisture furnace design, providing technical support for development of equipment and techniques for planned PIE evolutions, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Initial Design of SABRE Fueled Molten Salt Experiment Irradiation Vehicle

Molten Salt Reactors (MSRs) are emerging as promising advanced reactor technologies, utilizing molten salts for both fuel and primary cooling. These reactors may offer advantages such as passive safety, enhanced economic viability, and efficient waste reprocessing while operating at high temperatures and low pressures, which leads to increased system efficiency and reduced mechanical stress on containment structures. However, challenges arise from the complex chemistry and high corrosion rates of fueled molten salts, alongside the volatility of certain fission products during irradiation. The accumulation of these fission products can alter the fuel salt chemistry, affecting corrosion potential and radioactive source terms. Therefore, understanding these phenomena under neutron irradiation is essential for future MSR designs. The SABRE (Salt and Actinide Burnup in a Reactor Environment) experiment, currently in conceptual design at Idaho National Laboratory (INL), aims to investigate these challenges by conducting a drop-in capsule experiment in the Advanced Test Reactor (ATR). The primary objectives include achieving a minimum burnup of 2 GWd/MTU and characterizing fission products while studying corrosion behavior in molten salt systems. This innovative experiment will leverage recent advancements in ATR capabilities, allowing for the safe irradiation of molten salts, thereby advancing the understanding and qualification of MSR technology for future nuclear power generation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗