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DOE OSTI2024

The AGN-201 Digital Twin: A test bed for remotely monitoring nuclear reactors

Research reactors have historically provided researchers and scientists with a means for testing and understanding the workings of nuclear phenomena. With the lack of new research reactors built in the past decades, it is important now more than ever to utilize and provide evidence for the usefulness of research reactors. Here, this work explores the use of the AGN-201 research reactor at Idaho State University as a test bed for developing a digital twin (named the AGN-201 DT) to realize remote monitoring for nuclear reactors. The goal of the AGN-201 DT is to monitor the AGN-201 reactor and detect when undeclared events take place to provide information for a monitoring agency. The AGN-201 DT was able to detect (without a priori knowledge) when multiple undeclared experiments were placed in the core using on-the-fly machine learning and reactor physics analysis. The AGN-201 reactor provided a test bed for developing, deploying, and testing a digital twin for monitoring nuclear reactors.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2015

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Meta-Analysis of Advanced Nuclear Reactor Cost Estimations

Supporting Data can be downloaded at: https://gain.inl.gov/content/uploads/4/2024/06/INL-RPT-24-77048-R1.xlsx Nuclear energy is a critical cornerstone of the current United States clean energy supply and may play a larger role in the future in support of a transition to a net-zero economy. The current fleet of nuclear reactors predominantly consists of large light-water reactors (LWRs), while many of the reactor designs under consideration are smaller and/or different technologies. Because these new designs have not yet been built, there is a high degree of uncertainty associated with their cost. This complicates energy-planning efforts because cost projections are not always standardized, consistent, and centralized in an easily accessible location. To help support energy planning in the US, this report provides advanced nuclear cost ranges using a transparent methodology along with other relevant information that can be used to help support decision making and energy planning. The purpose of this work was to conduct a methodical process for cost evaluation using only public information that was vetted with the end-goal to provide reference cost projections for nuclear energy. To provide a solid basis for these values, the approach and assumptions are explicitly laid out throughout the report allowing any user of the data to challenge or reconsider them. Because future US nuclear-reactor costs are still unknown due to little recent observed data, the report opted to compile a comprehensive list of bottom-up estimates and evaluate averages/trends within the data to identify reference ranges. This was deemed preferable to opining on the robustness or validity of one cost estimation versus another. To that end, the work evaluated thousands of lines of cost subaccounts from several bottom-up cost estimates. A wide variety of different reactor types captured in the data are of various sizes and technologies. Some of these reactors will be representative of advanced reactors under development while others will not. Thus, the results here are dependent on the data that are available and the accuracy of the estimates that are used. Each bottom-up estimate was reviewed to determine whether it was complete. Incomplete data sets were corrected to ensure an adequate basis of cross-comparison. The report is not without limitations and should be interpreted as an initial step to develop cost ranges for nuclear technology. Ultimately, future work can build upon the methodology with refined cost estimates to reduce uncertainty. US-based overnight capital cost (OCC) estimates were compiled from extensive data sets into ranges for both large and small reactor sizes for 2030. To project the cost declines over time, learning rates were sampled from literature sources. No SMRs were previously built; hence, learning rates based on bottom-up approaches (e.g., by quantifying the impact stemming from fabrication of different components, modular work, site construction, commissioning) were prioritized. For larger reactors, actual learning rates from deployments were used to project future costs (adjusted to account for standardization or lack thereof between designs). Other costs included are fixed and variable operations and maintenance costs. The final variables were capacity factors and ramp rates to support energy planning.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Proposed Classifications of Remote Operations for Nuclear Reactors Based on Physical and Cybersecurity Considerations

The incorporation of remote operations into reactor operations is a topic of high interest among advanced and small modular reactor (A/SMR) vendors, with some considering it essential to the success of their business models. However, remote operations are a concept novel to the nuclear industry. While various technical aspects of remote operations have been explored, a significant gap remains in understanding the security implications of integrating remote operations into reactor designs, particularly concerning the security requirements for remote-operations facilities and infrastructure. This report aims to address this gap by first defining classes of remote operation based on the extent of remote access to reactor control systems and grounded in the existing regulatory framework with compatible terminology. Secondly, the report outlines the physical and cybersecurity requirements applicable to remote-operations facilities and infrastructure at each defined class. These requirements are based on existing licensing frameworks provided by 10 Code of Federal Regulations (CFR) Part 50 and 10 CFR Part 52, as well as the upcoming A/SMR licensing framework in the proposed Part 53. The assessment focuses specifically on security regulations, such as 10 CFR Part 73, which includes provisions for both cybersecurity (§ 73.54) and physical security (§ 73.55). This report proposes five classes of remote reactor operations. Class 1 involves remote monitoring only, with no control over reactor systems. Class 2 allows for the remote issuance of allowlisted commands to the reactor facility. Class 3 extends control to non-safety-significant, non-safety-related, or not important to safety systems and equipment. Class 4 permits remote control of safety-significant systems. Finally, Class 5 allows remote control of safety-related systems. It is important to note that these classes were defined purely with functionality in mind, without considering the practicality or feasibility of implementation for each class under current or upcoming regulatory guidance. The intention behind this approach is to enable an assessment of which security requirements apply to each class, allowing readers to evaluate the implementation possibilities for their specific use cases. Following the definition of remote-operation classes, the report assesses the specific physical and cybersecurity requirements applicable to the remote-operations facility and infrastructure within each defined class. This includes defining the types and locations of operators that are possible at each class of operation and, based on operator type and location, as well as functionality within each class, outlining the physical and cybersecurity requirements. By detailing the security requirements by class, the report provides readers with the information needed to determine the type of security program they may need to implement for their desired concept of operation. The next contribution of this report was to assess the practicality of implementing each proposed class of remote operations based upon the security requirement assessment. In short, three of the five proposed remote-operation classes were found to possibly have a practical path forward to implementation under the U.S. regulatory framework. Class 1 remote operations are currently in use in the U.S. while Class 2 and 3 remote operations may be logistically possible to implement under the U.S. regulatory framework. The final two Classes, 4 and 5, would likely be logistically difficult, if not infeasible to implement within the current U.S. physical- and cybersecurity regulatory framework. Given the results of the feasibility assessment, an example architecture is proposed for both Class 2, remote allowlisted commands, and Class 3, remote control of non-safety systems as well as security implication assessments of each architecture. These example implementations are not meant to be prescriptive in terms of how Class 2 or Class 3 remote operations should be deployed; instead, they are intended to be informative to stakeholders on how Class 2 or Class 3 could potentially be applied in order to inform their system design. An example architecture for Class 1 remote monitoring was not provided as Class 1 in already in use in U.S. nuclear operations. Example architectures for Class 4 and Class 5 were not provided due to their assessment of being likely infeasible to implement. The final contribution is an assessment of the physical- and cybersecurity implications of introducing autonomous operations into an A/SMR. What was found was that the security implications can be separated into two cases. Autonomous operations supported by SSCs located only at the reactor site, and autonomous operations supported by SSCs outside of the reactor site. For the first case, the introduction of autonomous systems will likely not change the facility’s requirement to comply with existing cyber and physical security regulation

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Monolithic U-10Mo Fuel Development for U.S. High Performance Research Reactors

To support the U.S Department of Energy (DOE), National Nuclear Security Administration (NNSA) mission to reduce the threat posed by the civilian use of highly enriched uranium (HEU) worldwide, the US High Performance Research Reactor (US HPRR) fuel development program is pursuing development and qualification of a new high-density monolithic U-Mo fuel to facilitate conversion of five higher power research reactors located in the U.S. (ATR, HFIR, NBSR, MIT and MURR) and one critical facility (ATR-C) from HEU to low enriched uranium (LEU). These reactors require higher uranium fuel density than is achievable with dispersion fuel systems used to convert lower power research reactors. The down selected fuel system consists of U-10Mo alloy fuel foils having a thin Zr diffusion barrier interlayer, clad in 6061 Al alloy by hot isostatic pressing. In order to support fabrication development, fuel performance evaluations, generic fuel qualification through the Nuclear Regulatory Commission (NRC), and eventual reactor licensing, a series of irradiation experiments and testing campaigns are being planned. The tests are designed to provide data needed for fabrication process down selection, to establish acceptable fuel performance under normal operating conditions and anticipated transients, and to demonstrate successful scale-up to prototypic reactor-specific fuel element designs. A comprehensive testing program is being executed as part of the US HPRR fuel development program to develop and qualify a new monolithic UMo fuel system to allow conversion of five high power research reactors and one critical facility within the US in support of the NNSA mission to convert research reactors worldwide from HEU to LEU fuels. A significant outcome of this program will be the qualification and eventual licensing of a new plate-type fuel system for use in research reactors through the NRC for the first time in nearly 3 decades. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Design and implementation of safety channels for the Idaho State University AGN-201m research reactor control console - 216

The purpose of the work described in this paper is to design, implement, and install new safety channels for the control console of the Idaho State University (ISU) AGN-201m Research Reactor. These channels will replace the current vacuum-tube driven system, which has been in operation for over 50 years at the ISU nuclear engineering laboratory. Due to the aging issues associated with this technology an entirely new control console is needed. There are 3 safety channels which serve as the monitors and safety systems for adequately shutting down the reactor should any limits be broken. These 3 safety channels work harmoniously with other systems within the new control console to ensure appropriate operation of the reactor. The safety channel replacements are composed of three complementary signal processing systems whose circuits monitor the operating parameters of the reactor. The key reactor information is shown on several independent meters corresponding to each of the three safety channels. Finally, the reactor trip points are set to correspond with the standards promulgated by the Nuclear Regulatory Commission (NRC) to ensure safe operation of the ISU AGN-201m Research Reactor. The interconnections between each of these systems ensure that if any Channel fails or otherwise breaks a limit the entire system will initiate a scram and the reactor will become adequately sub-critical. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Biogenic coal-to-methane conversion can be enhanced with small additions of algal amendment in field-relevant upflow column reactors

Subsurface coal environments, where biogenic coal-to-methane conversion occurs, are difficult to access, resulting in inherent challenges and expenses for in situ experiments. Previous batch reactor studies provided insights into specific processes, pathways, kinetics, and engineering strategies, but field-relevance is restricted due to limited substrate availability or byproduct accumulation that may influence reactions or metabolisms. In this study, continuous-flow column reactors were used to overcome some batch limitations, improve the understanding of in situ conditions, and increase field-relevance for subsurface engineering technology development. The bench-scale reactor system was constructed to investigate the addition of algal amendment for enhancing microbial coal-to-methane conversion previously developed in batch systems. Four reactor columns were packed with coal and inoculated with a microbial consortium from the same Flowers-Goodale coal bed. Two reactors were amended with 13 C-labeled algal amendment on day 0, and two were unamended. On day 61, one previously amended and one previously unamended reactor were re-amended. Produced gases were captured in a gas trap, and CH 4 and CO 2 were quantified. The reactor amended twice produced 1712.6 µmol CH 4 (4.6% as 13 CH 4 ). The reactor amended only on day 0 produced 1485.5 µmol CH 4 (2.6% as 13 CH 4 ). The reactor amended only on day 61 produced 278.9 µmol CH 4 (3.9% as 13 CH 4 ). The reactor with no amendment produced no measurable gases for the duration of the 172-day experiment. Amendment increased the rate of coal-to-methane conversion and total gas production; most of the produced gases were due to coal conversion with only small contributions (<7%) from amendment conversion.

09 BIOMASS FUELS↗

DOE OSTI2015

Consumption of the electric power inside silent discharge reactors

An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodes in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Numerical Simulation Comparison of Two Reactor Configurations for Chemical Looping Combustion and Chemical Looping With Oxygen Uncoupling

Chemical looping with oxygen uncoupling (CLOU) is a carbon capture technology that utilizes a metal oxide as an oxygen carrier to selectively separate oxygen from air and release gaseous O 2 into a reactor where fuel, such as coal, is combusted. Previous research has addressed reactor design for CLOU systems, but little direct comparison between different reactor designs has been performed. This study utilizes Barracuda-VR® for comparison of two system configurations, one uses circulating fluidized beds (CFB) for both the air reactor (AR) and fuel reactor (FR) and another uses bubbling fluidized beds for both reactors. Initial validation of experimental and computational fluid dynamic (CFD) simulations was performed to show that basic trends are captured with the CFD code. The CFD simulations were then used to perform comparison of key performance parameters such as solids circulation rate and reactor residence time, pressure profiles in the reactors and loopseals, and particle velocities in different locations of the reactor as functions of total solids inventory and reactor gas flows. Here, using these simulation results, it was determined that the dual CFB system had larger range for solids circulation rate before choked flow was obtained. Both systems had similar particle velocities for the bottom 80% of particle mass, but the bubbling bed (BB) obtained higher particle velocities as compared to the circulating fluidized-bed FR, due to the transport riser. As a system, the results showed that the dual CFB configuration allowed better control over the range of parameters tested.

30 DIRECT ENERGY CONVERSION↗

DOE OSTI2022

A Forward Feedback Control Scheme for a Solar Thermochemical Moving Bed Counter-Current Flow Reactor

Abstract Pelletized thermochemical energy storage media has a potential for long-duration energy storage. Production of solid-state energy storage media can be done within a cavity chemical reactor that captures concentrated solar radiation from a solar thermal field. The temperature stability of a solar reactor is directly influenced by the solar flux intercepted. This paper presents a low-order physical model to simulate the dynamic response of temperature inside a tubular plug-flow reactor prototype. Solid granular particles are fed to the reactor from the top whereas a counter-current flowing gas enters the reactor from the bottom. An in-house code was developed to model transient heat transfer of the reactor wall, gas, and moving particles. The model was preliminarily validated with packed beds for different temperature ranges and two gas flowrates. Dynamic response of the reactor temperature is simulated for different input power and gas/particle flowrates. The results show that the system response can be controlled efficiently by utilizing input power (solar flux) as a control parameter. A conventional proportional integral (PI) controller is designed to control the temperature inside the reactor and to maintain it during the solar flux intermittency. The controller parameters are tuned using the Ziegler–Nichols method to ensure optimal system response. The results show that the feedback control model is successful in tracking different reference reactor temperatures within a reasonable settling time of 30 min and eliminated overshoot. This study can be extended to include a hybrid reactor with a multi-input, multi-output variable system.

Energy & Fuels↗

DOE OSTI2023

Informing and Engaging the Future Workforce on Generation IV International Forum Reactor Systems

Generation IV reactors are six nuclear reactor designs, Sodium Fast reactor (SFR), Very High Temperature Reactor (VHTR), Gas cooled Fast Reactor (GFR), Molten Salt Reactor (MSR), Lead Fast Reactor (LFR), and Supercritical Water Reactor (SCWR), that are considered to be the most promising in light of various criteria based on the following objectives: • continuation of the progress made by Generation III water reactors in terms of competitiveness and safety; • more effective use of uranium resources; • less radioactive waste, especially high-level, long-lived waste; • greater protection against malicious acts and the diversion or theft of nuclear materials. The GIF Education and Training Working Group (ETWG) launched two major initiatives in 2016 and in 2021 to not only inform and educate but also to engage the future workforce in support of these reactor systems. This paper will present an overview of the GIF EWTG webinar series as well as the 2021 Pitch your Gen IV Research Competition which is followed by the 2023 Pitch your Gen IV Research. Competition

Paviet, Patricia D.↗

NASA NTRS1987

Small space reactor power systems for unmanned solar system exploration missions

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

Bloomfield, Harvey S.↗

NASA NTRS1961

Some Nuclear Calculations of U-235-D2O Gaseous-Core Cavity Reactors

The results of a multigroup, diffusion theory study of spherical gaseous-core cavity reactors are presented in this report. The reactor cavity of gaseous U235 is enclosed by a region of hydrogen gas and is separated from an external D2O moderator-reflector by a zirconium structural shell. Some cylindrical reactors are also investigated. A parametric study of spherical reactors indicates that, for the range of variables studied, critical mass increases as: (1) Fuel region is compressed within the reactor cavity, (2) moderator thickness is decreased, (3) structural shell thickness is increased, and (4) moderator temperature is increased. A buckling analogy is used to estimate the critical mass of fully reflected cylindrical reactors from spherical results without fuel compression. For a reactor cavity of a 120-centimeter radius uniformly filled with fuel, no structural shell, a moderator temperature of 70 F, and a moderator thickness of 100 centimeters, the critical mass of a spherical reactor is 3.1 kilograms while that of a cylinder with a length-to-diameter ratio of 1.0 (L/D = 1) is approximately 3.8 kilograms and, with L/D = 2, 5.9 kilograms. For the range of variables considered for U235-D2O gaseous-core cavity reactors, the systems are characterized by 95 to 99 percent thermal absorptions, with the flux reaching a maximum in the moderator about 10 to 15 centimeters from the reactor cavity.

Ragsdale, Robert G.↗

NASA NTRS2018

The Kilopower Reactor Using Stirling TechnologY (KRUSTY) Nuclear Ground Test Results and Lessons Learned

The Kilopower nuclear ground testing nicknamed KRUSTY (Kilopower Reactor Using Stirling TechnologY) was completed at the Nevada Nuclear Security Site (NNSS) on March 21, 2018. This full scale nuclear demonstration verified the Kilopower reactor neutronics during startup, steady state, and transient operations in a space simulated environment. This was the first space reactor test completed for fission power systems in over 50 years and marked a turning point in NASA's nuclear program. The completed reactor power system design incorporated flight prototypic materials and full scale components in an effort to study the reactor dynamics at full power and significantly reduce follow on risk of a future flight demonstration. This design provided a unique opportunity for the power system to simulate several expected and unexpected mission scenarios that allowed the designers to verify that the reactor dynamics could tolerate many worst case conditions regarding reactor stability and control. The dynamic changes imposed on the reactor validated the ability of the reactor to load follow the power conversion system and passively control the fuel temperature and overall system stability. With successful completion of the KRUSTY experiment, the NASA/DOE team will evaluate the lessons learned throughout the project and apply them towards a flight demonstration of a Kilopower reactor.

Gibson, Marc A.↗

NASA NTRS2018

Kilopower Reactor Using Stirling TechnologY (KRUSTY) Nuclear Ground Test Results and Lessons Learned

The Kilopower nuclear ground testing nicknamed KRUSTY (Kilopower Reactor Using Stirling TechnologY) was completed at the Nevada Nuclear Security Site (NNSS) on March 21, 2018. This full scale nuclear demonstration verified the Kilopower reactor neutronics during startup, steady state, and transient operations in a space simulated environment. This was the first space reactor test completed for fission power systems in over 50 years and marked a turning point in NASA's nuclear program. The completed reactor power system design incorporated flight prototypic materials and full-scale components in an effort to study the reactor dynamics at full power and significantly reduce follow on risk of a future flight demonstration. This design provided a unique opportunity for the power system to simulate several nominal and off-nominal mission scenarios that allowed the designers to verify that the reactor dynamics could tolerate many worst case conditions regarding reactor stability and control. The dynamic changes imposed on the reactor validated the ability of the reactor to load follow the power conversion system and passively control the fuel temperature and overall system stability. With successful completion of the KRUSTY experiment, the NASA/DOE team will evaluate the lessons learned throughout the project and apply them towards a flight demonstration of a Kilopower reactor.

Gibson, Marc A.↗

NASA NTRS2018

Kilopower Reactor Using Stirling Technology (KRUSTY) Nuclear Ground Test Results and Lessons Learned

Gibson, Marc↗

NASA NTRSDate not supplied

Kinetic Model Development of the Carbon Vapor Deposition Reactor to Predict Performance Versus Design

To enable complete closure of a Sabatier-based atmosphere revitalization system where oxygen recovery is predicated by carbon dioxide reduction to methane and water, there must be a means of recovering the hydrogen from the byproduct methane. Honeywell’s carbon vapor deposition (CVD) reactor is a potential technology that provides a thermochemical means by which methane decomposes at high temperatures (> 1,000 °C) into hydrogen, solid carbon, and a small percentage of hydrocarbons. This advanced methane pyrolysis reactor utilizes a carbon fiber substrate to capture and immobilize the solid pyrolytic carbon rather than allowing continuation toward undesirable soot. The CVD reactor is a highly dynamic system whose characteristic geometry changes while operating. The reactor model was developed to provide predictive performance assessment capabilities at different operating conditions, geometries, and configurations to aid in technology development. Modeling of the CVD reactor considered the multitude of simultaneous and interdependent physicochemical processes that are entailed, including: (1) the homogeneous gas-phase pyrolysis reactions, (2) the heterogeneous deposition reactions, and (3) the transport of gaseous species in the reactor domains. A 1-D CVD reactor model was developed based on a general packed-bed, axial-dispersion plug flow reactor model with a solid substrate balance to account for carbon deposition. The model includes 59 species and 243 reactions constituting the methane pyrolysis reaction network, carbon deposition, and interconversions of oxygenate species. The reactor model was compared to brassboard test data, where it was shown to be able to capture the pertinent physicochemical phenomena as evidenced by its agreement with short-duration run data (i.e., looking at different temperature and flow rate conditions), as well as trend agreement with the brassboard reactor’s long-duration test data. Recommendations are provided on further model development steps as well as its use cases.

Air Revitalization↗

DOE OSTI2017

ADVANCED TEST REACTOR (ATR) CONVERSION TO LOW ENRICHED URANIUM (LEU) FUEL

The United States Department of Energy (DOE), National Nuclear Security Administration (NNSA), Office of Material Management and Minimization (MMM) is working to convert research reactors globally from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. MMM and its predecessor programs have converted or verified the shutdown of 88 HEU research reactors and isotope production facilities, using LEU fuel developed in the 1980s. However, there is a small set of high performance research reactors which require a new high density LEU fuel for conversion, including six U.S. high performance research reactors (USHPRRs), which include the Advanced Test Reactor (ATR) and the ATR Critical (ATRC) Facility at the Idaho National Laboratory (INL) near Idaho Falls, ID; University of Missouri Research Reactor (MURR) at Columbia, MO; Massachusetts Institute of Technology Reactor (MITR) at MIT in Cambridge, MA; High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) near Oak Ridge, TN; and National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) in Gaithersburg, MA. These reactors will share the same uranium-molybdenum (U-Mo) fuel plate design. The U-Mo fuel meat foil will have a zirconium (Zr) diffusion barrier with aluminum cladding. Three different application methods for the Zr diffusion barrier layer are being pursued – co-rolled, electroplating and plasma spraying. Fabrication of the higher density LEU U-Mo fuel has presented several challenges. Because the U-Mo fuel fabrication process uses the existing stockpile of 93% highly enriched uranium (HEU) to be blended with molybdenum and depleted uranium (DU) to form 19.75% LEU, cost benefits of enriching to 19.75% will not be realized. Homogeneity of the U-235 and molybdenum in the metal foil is a challenge due to U-Mo casting process results in variations between adjacent cast U-Mo plates. LEU will not require less U-235 as previously believed due to self-shielding properties of U-238 requiring additional U-235 to be loaded in each fuel element. Although the overall dimensions of the ATR fuel element remain unchanged, the weight of the LEU ATR fuel element is 72% greater than a HEU ATR fuel element. During waste recovery and recycling efforts, separation of aluminum cladding from the HEU fuel meat is a relatively simple chemical dissolution, however, separation of the Zr diffusion barrier from the U-Mo fuel meat is significantly more challenging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗