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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Fuel Salt Synthesis for the Molten Chloride Reactor Experiment: Scale-up, Operations, and Production Update

Over the previous 5 years, Idaho National Laboratory (INL) has been working with Southern Co. and TerraPower on the Advanced Reactor Demonstration Program (ARDP) funded Molten Chloride Reactor Experiment (MCRE) project. As part of this effort, INL has developed a fuel synthesis process to produce the NaCl-UCl3 fuel salt that MCRE will need for operation. Dr. Phillips will present on process development and scale-up testing and results, as well as provide an update on the current status of fuel production for MCRE. To date, the fuel synthesis process has been demonstrated at full scale using depleted uranium, and the equipment necessary for production of the fuel for MCRE has been installed in the Fuel Manufacturing Facility (FMF) at the INL’s Materials and Fuels Complex (MFC). The NaCl-UCl3 produced to-date has been shown to be of 99.99% purity or greater, and be within tolerance for all relevant parameters. Overall process efficiency in terms of uranium utilization has been demonstrated to be above 90%. The synthesis operation has also been shortened to allow for completion of the reaction within a single 10 hour working shift. Consequently, the process developed is expected to be capable of meeting all project objectives for efficiency, purity, scale, and scheduling. Production of NaCl-UCl3 fuel salt for MCRE is projected to begin during the Summer of 2025.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

SAM Enhancements and Model Developments for Molten-Salt-Fueled Reactors

To support the development and utilization of the SAM code for molten-salt-fueled reactor (MSR) safety analysis and licensing, an effort was devoted to enhancing code capabilities and developing reference models for the MSR primary loop. A reference standard problem of a prototypical reactor design is foundational to NRC to verify the adequacy of computer codes and evaluation models for a specific reactor type. A thermo-fluid model of the Molten Salt Reactor Experiment (MSRE) has been developed. The MSRE primary loop model consists of a 2-D core region and external core components in 1-D or 0-D. Both the steady-state and a transient scenario are simulated. The porous medium model is utilized for both the molten salt fluid channels and the moderator matrix in the MSRE core. The use of the porous medium model for the MSRE core is verified with the higher-fidelity simulation results using 1-D representations of the fluid channels and 3-D modeling of core structures. To further enhance SAM multi-scale simulation capabilities for MSRs, a mass transport model is developed and implemented in SAM multi-dimension flow module for modeling of species transport such as delayed neutron precursors in MSRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Continuous removal of fission products from molten-salt-fueled reactors

A method based on separation of volatilized molten salt (MS) components, including by mass, is being developed to extract fission products (FPs) from operating molten-salt reactors. The initial application of this method is for accelerator-driven subcritical reactors fueled by (fluorinated molten salt) spent nuclear fuel (UNF) from any past, present, or future reactor. The actinides remain in the subcritical reactor to produce profitable energy and be transmuted while the extracted FPs can be buried without long-lived actinides such that a geologic repository is not necessarily needed to close the nuclear fuel cycle. By removing neutron-absorbing FPs and operating sub-critically, where the restrictive link between operation and criticality is broken, it is possible to envision complete burnup of the UNF fuel. The game-changing feature of continuously processing the molten salt inside the reactor while the reactor operates eliminates the need for a separate reprocessing plant. This feature also simultaneously improves the neutronics of the reactor, increasing the burnup of the fuel and extending its useful life for generating energy. Nuclear nonproliferation and need for geologic repositories are addressed by keeping actinides inside the reactor containment until they are consumed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Assessment of Technical Information Needs and Considerations for Front-End Activities for Molten Salt Fuel Types

This report details an assessment of the hazards and safety impacts related to front-end operations of molten salt reactors (MSR) as well as an assessment of existing safety guidance. Publicly available information was evaluated to make recommendations to the U.S. Nuclear Regulatory Commission (NRC) for guidance development and information gathering needs. Five different MSR technologies were evaluated along with information from 1960’s about the Atomic Energy Commission programs for the Aircraft Reactor Experiment and Molten Salt Reactor Experiment. Primary areas of evaluation include fuel cycle facilities, transportation and at reactor fuel mixing operations up to reactor startup. In general, there was difficulty synthesizing publicly available information due to the early stage of the MSR development and proprietary information associated with the reactor technology companies analyzed. There are a few specific recommendations from this report. One overarching paradigm shift that the NRC needs to evaluate is the possibility of collocating fuel cycle facilities with operating MSRs. Although there is no specific prohibition to this in the NRC regulations, the collocation of the fuel cycle facilities with the operating MSR presents a novel approach compared to the light water reactor baseline. More guidance and research into thorium operations and regulation is needed. The hazards associated with the stability of stored fuel constituents need to be evaluated. Some reactor fuel cycles use beryllium, and its use and the associated hazards need further analysis. NRC-approved designs for transportation packages for shipping molten salt fuels and factory-built and fueled reactors are not available.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

ENYGF 2023 - Design of Temperature-controlled Fuel-salt Irradiation Experiment

Idaho National Laboratory (INL) is developing a Molten salt Research Temperature-controlled Irradiation (MRTI) experimental capsule for the Neutron Radiography (NRAD) reactor. The experiment is envisaged to be versatile and able to host a variety of salt-wall material combinations. The first test is designed to contain actinide-bearing chloride salt, which previously has not been subjected to a neutron field. The main objectives of the experiment are to produce irradiated salt samples for Post Irradiation Examination (PIE), to evaluate salt behavior under irradiation, and to test in-situ sensors/instrumentation. The final design for the experiment completed in 2022, and fabrication of the experiment is underway. The salt is double encapsulated providing insulation and redundancy in the case of leakage. A gas-gap between the inner capsule and the outer container enables tuning of thermal conduction to the NRAD coolant by altering its thickness and its gas-mixture. An inert gas mixture ratio in the gas-gap ensures the configurations reached a targeted average salt temperature of 600oC, while avoiding salt freezing or exceeding material limits. A resistive heater placed within a thermowell and controls the temperature of the experiment (both when the reactor is on or off). The selected configurations can contain over 10 cm3 of fuel-bearing chloride salt, which is sufficient for the purposes of the post-irradiation examination (PIE) and provides sufficient volume for two submerged thermocouple probes to monitor salt temperature. The capsule wall material is Inconel 625, and weld qualifications for the assembly have been developed. A glove-box laser weld technique was refined as part of this process to ensure a leak-tight assembly for the salt-bearing assembly and reduce the risks of impurities permeating into the salt. This work was supported through the INL Laboratory Directed Research & Development (LDRD) Program under DOE Idaho Operations Office Contract DE-AC07-05ID14517.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Structural Changes in Molten Salt Fuel and/or Waste Stream Compounds Cs 2 UCl 6 and Cs 2 UO 2 Cl 4 from Room Temperature to Melting and Related Materials [Slides]

Perovskite phases have been explored in recent literature as waste forms for long term storage of radioactive waste from molten salt reactors specifically. Synthesis of new halide perovskite phases can inform these efforts as well as molten salt reactor chemistry in general. Inorganic perovskites have also shown promise as new scintillating materials for detection of X-rays and gamma rays. Modifying the structure and chemistry of these compounds, perhaps with organic components, could be used to synthesize new scintillating compounds or modify the properties of existing ones. Both U 6+ and U 4+ readily form compounds with a striking similarity to vacancy ordered perovskites (Cs 2 UO 2 Cl 4 and Cs 2 UCl 6 ). Each of these compounds is relevant to molten salt reactor chemistry and/or waste streams. These phases could from in uranium chloride salt systems or waste streams: Cs being a fission product.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Criticality Safety S/U based USL Calculation for UCl3-NaCl Fuel Salt Operations

Recent experimental data has shown inconsistencies with the 35Cl(n,p) cross-section. The cross-section uncertainty data does not account for the recent data. The 1s relative uncertainty is assumed to be 100% for the 35Cl(n,p) cross section over all energies above 0.017 MeV TerraPower and LANL have recently done cross-section measurements and developed new cross-sections for 35Cl.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

Electroanalytical Sensors for Liquid Fueled Fluoride Molten Salt Reactor

ThorCon is developing a thermal thorium / uranium molten salt fueled reactor that uses a fuel salt consisting of NaF - BeF 2 - ThF 4 - UF 4 - UF 3 , which uses 19.75% enriched uranium. It is not a breeder reactor but requires regular additions of fissile material. On-site fuel processing is limited to adding make-up fuel, and adding beryllium to maintain redox balance in the fuel salt. The fluoride salt is an excellent flux that will strip the protective oxygen away from all metal surfaces it contacts. In order to protect the metal surfaces in the primary loop, no free fluorine is allowed. Maintaining a ratio of UF 4 to UF 3 provides a chemical buffer that can absorb any free fluorine released during the fission process before it can harm the structure. Periodic additions of metallic beryllium to the fuel salt will react with UF 4 to form BeF 2 and UF 3 , which will maintain the desired UF 4 /UF 3 ratio. A means to measure the UF 4 /UF 3 ratio is important to maintain the fuel salt in a non-corrosive state and could also provide additional insight into the corrosion behavior and chemical constitution of the system. A promising measurement technique is based on electrochemistry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Multiphase Species Transport Modeling for Molten Salt Reactors in the System Analysis Module: Generation, Decay, Deposition, and Extraction of Insoluble Fission Products

With the increase of interests in the design and deployment of advanced reactor systems, a desire for simulation tools supporting system analysis of reactor operation and safety is rising. Molten salt reactors (MSRs), one of the advanced reactor systems, utilize liquid fused salt fuel as both coolant and fuel. During operation, MSR generates insoluble fission products, including noble metals and gases. The buildup of these species in fuel salt presents safety concerns as they may deposit on surfaces of critical components and produce excessive decay heat, causing the failure of system components. Timely removal of these noble metals and gases would ensure the safe operation of the reactor system. The dynamic nature of salt fuel system, involving the generation, decay, deposition, and extraction of noble metals and gases, calls for robust species transport models to facilitate system analysis and monitoring, and design of efficient species removal components. This paper concentrates on the development of a computational framework for species transport, consisting of multiphase transport model formulation, mass transfer between phases, numerical implementation in MOOSE environment, verification through Method of Manufacture Solutions (MMS) and validation against experimental data from the Molten Salt Reactor Experiment (MSRE). Integrating this framework into the System Analysis Module (SAM) code further enhances SAM’s capabilities for advanced reactor analysis in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Modeling a Thermal-Spectrum LEU-fueled Molten Salt Reactor Co-fueled with Thorium in SCALE 6.3.1 and Serpent-2

Interest in the development of advanced nuclear fission reactors for commercial electricity production has risen in recent years, with Generation IV reactor designs offering numerous advantages in safety, efficiency, fuel cycle sustainability, and waste management. In particular, thorium-fueled molten salt reactors (MSRs) are considerably promising for enhancing fuel cycle sustainability in their ability to breed fissile 233U fuel from thorium, a presently untapped and widely abundant resource. A novel MSR fuel cycle concept, the ”Sourdough” refueling and waste management strategy, has previously been demonstrated with a traditional uranium-based fuel cycle in a thermal-spectrum MSR operating with low enriched uranium (LEU) fuel with favorable neutronic performance. However, the ability to use this unique fuel cycle approach with thorium-based molten salt fuels has not yet been studied. In this work, the Sourdough fuel cycle was implemented in a small, thermal-spectrum MSR fueled with high assay low enriched uranium (HALEU) and fertile 232Th for breeding 233U fuel. Relevant neutronic data, including fuel and isothermal temperature feedback behavior, was studied using the SCALE 6.3.1 and Serpent-2 code systems, and keff data was measured during simulated depletion at 400 MWth. The Sourdough fuel cycle concept is shown to perform favorably with a thorium-fueled MSR model, thus warranting further study into its use in other MSR designs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Fuel Qualification for Molten Salt Reactors

This report supports the development and review of an efficient methodology or process for liquid salt fuel system qualification. The report describes the technical issues encountered when developing an adequate understanding of the fuel salt’s chemical and physical behavior under both normal and accident conditions. The concepts presented in this report align with the fundamental safety functions (FSFs) approach of the more general non-light water reactor (nonLWR) fuel qualification guidance provided in NUREG-2246. However, because liquid salt fuel has substantial technical differences from solid fuels, this report describes tailored approaches for achieving the common safety intents when using liquid salt. The purpose of this report is to illustrate how variations in fuel salt properties impact achievement of FSFs at liquid-salt fueled molten salt reactors (MSRs). The focus on FSFs supports methodology compatibility with both current and developing licensing processes. Customized fuel qualification guidance for liquid salt–fueled Nuclear Power Plants (NPPs) are also provided in this report to supplement the generic non-LWR guidance recently developed by the NRC. As with the broader non-LWR fuel qualification, this report focuses on identification and understanding of fuel life-limiting failure and fuel salt property degradation mechanisms that occur as a result of irradiation during reactor operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Disruptive thermal-spectrum molten salt breeder reactor fuel cycle technologies

Multiple, disruptive fuel salt processing technologies for thermal-spectrum molten salt breeder reactors (TS-MSBRs) have been conceived of and brought to laboratory levels of maturity since the termination of the historic U.S. government program. This paper describes these technologies and the role that they could serve in decreasing the remaining technical risks for TS-MSBR deployment. In conclusion, the focus of the paper is on describing how the technologies could integrate into a dramatically simpler, proliferation resistant fuel salt processing system as well as identifying remaining development hurdles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗