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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Potential Disposition Options for a Salt Fueled Molten Salt Reactor At the INL (A White Paper)

The molten salt reactor (MSR) design concept has been the focus of significant research and development efforts, both in the U.S. and around the world. There are generally two types of MSRs, salt fueled MSRs that have the fuel dissolved within the salt and solid fueled MSRs that only use the molten salt as a coolant. The range of conceptual designs and the flexibility of operations of MSRs results in a wide range of potential waste streams from salt fueled MSRs; all of which require safe and secure storage and eventual disposition. This white paper will 1) survey the range of salt fueled MSR designs, 2) recognize the implications of design, on the resulting waste streams, and 3) identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of these streams. Research in MSR waste streams is young and the focus of intense research that requires more funding and time to become mature enough to fully support design decisions. Once a specific design is selected, a more detailed management and disposition effort must be performed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Molten Salt in the Advanced Test Reactor [Slides]

The Advanced Test Reactor (ATR) is one of the world’s most important resources for fuel qualification. Prior to this project, ATR could not support molten salt irradiation experiments. This work expands ATR’s capability to support fuel qualification for molten salt reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Experimental Investigations into the Corrosion of Alloy 625 Using NaCl-PuCl3 Molten Salt in a Natural Circulation Microloop

Molten salt reactors (MSRs) can potentially revolutionize the nuclear industry by providing a path to a near-zero nuclear waste fuel cycle, contributing to more sustainable energy sources. As a plethora of MSR developers in the United States work toward an aggressive commercialization timeline, many of their fueled-salts—notably, chloride-based compositions—have limited operational testing with nuclear material. Licensing and operating these reactors require an understanding of corrosion effects on reactor materials of construction under operational conditions. The TerraPower Molten Chloride Fast Reactor (MCFR) is a liquid-fueled chloride-salt fast reactor which has received notable interest from the utility sector based on its desirable economic characteristics. The reactor operates at low pressure but does not require the use of highly reactive chemicals, leading to a reduced use of concrete and steel during construction. Additionally, liquid fuel allows for inherently stable behavior and natural circulation during a loss-of-site-power scenario. MCFR can be refueled while operating which makes it compatible with variable generation sources such as wind and solar. MCFR is a breed-and-burn in-situ reactor that does not implement any chemical processing or separations in the fuel cycle. Only mechanical filtration of noble metals and off-gassing of noble gases are utilized while the actinides stay mixed with the fuel at all times. The MCFR will require technology development to reach commercialization. With a breed-and-burn in-situ reactor like MCFR, the transmutation of fertile U-238 to fissile Pu-239 allows for much greater fuel utilization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Preliminary results of molten salt corrosion of high entropy alloys manufactured at INL

Materials development is needed to support the design and deployment of advanced nuclear reactors that will operate at higher temperatures and in harsher environments than the current light water reactor fleet. High entropy alloys have been identified as a class of alloy that could address the needs of advanced reactors, such as the molten salt reactors, due to their potential strength, radiation resistance, and corrosion resistance. This project summarizes the work done in the year 2024 for the Additive Manufacturing of High Entropy Alloys for Nuclear Applications Project. This part of the work has focused on molten salt corrosion of high entropy alloys which were developed and manufactured at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Cradle to grave: the importance of the fuel cycle to molten salt reactor sustainability

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A Molten Salt Electrochemical Process for the Preparation of Cost-Effective p-Block (Coating) Materials

Solar energy applications rely heavily on p-block elements and transition metals. Silicon is, by far, the most commonly used material in photovoltaic cells and accounts for about 85% of modules sold presently. Of late, thin film photovoltaic cells have gained momentum because of their higher efficiencies. Most of these thin film devices are made out of just five elements, namely, cadmium, tellurium, selenium, indium, gallium and copper. The present manuscript describes an elegant and inexpensive molten salt-based electrolytic process for fabricating a tellurium-coated metallic substrate. A three-electrode set up was employed to coat iridium with tellurium from a molten bath containing lithium chloride, lithium oxide and tellurium tetrachloride (LiCl-Li 2 O-TeCl 4 ) at 650 °C for a duration ranging from 30 to 120 min under a galvanostatic mode. The tellurium coating was observed to be thick, uniform, smooth and homogeneous. Additionally, the deposited tellurium did not chemically react with the iridium substrate to form intermetallic compounds, which is a good feature from the standpoint of the device’s performance characteristics. The present process, being generic in nature, shows the potential for the manufacture of both the coated substates and high-purity elements not just for tellurium but also for other p-block elements.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Quantitative Analysis of Fission-Product Surrogates in Molten Salt Chloride Aerosols

This work demonstrates laser-induced breakdown spectroscopy (LIBS) applied to a stream of aerosolized salt from molten eutectic LiCl-KCl. We demonstrate analytical capabilities to track fission-product surrogates of Cs, Sr, Pr, and Nd simultaneously, with application to monitor salts in pyroprocessing schemes and molten salt reactors. This work demonstrates limits of detection using LIBS on the order of 100 μg/g, which proves potentially applicable to monitoring fission-product concentrations in pyroprocessing applications. Additionally, this work explores fundamental aspects of plasma temperature and plasma electron density of the aerosolized species during LIBS with a specific focus on potential non-uniform plasma conditions in the aerosol.

74 - ATOMIC AND MOLECULAR PHYSICS↗
NASA NTRS2023

Long Life Molten Salt Battery for NASA Venus Application

A development program is being conducted at Advanced Thermal Batteries (ATB) to create a low discharge rate, long life, molten salt battery for a NASA Venus surface probe. Battery goals are 60 days continuous operation at +25V +0/-6 volts and -25V +6/-0 Volts under severe environments of 465°C, 92 atm pressure, and corrosive sulfur dioxide in a super critical carbon dioxide atmosphere. Conveniently, molten salt thermal battery electrochemistry starts to operate efficiently at 330°C. However, a major barrier to direct application of existing thermal battery technology is the high self-discharge rate, resulting in lifetimes in only minutes, not days. The best performance to date, which meets the required voltage regulation, has been a lithium silicon (LiSi) alloy anode discharging through all three voltage plateau versus an iron monosulfide (FeS) cathode. In order to perform reproducibly, additional engineering controls within the battery design are required to limit cell–to–cell parasitic discharge mechanisms. Results of these trials have improved the efficiency and manufacturability of the design.

Thermal Battery↗
DOE OSTI2020

Tritium Transport Phenomena in Molten-Salt Reactors

In this work, we review phenomena relevant to tritium transport in molten-salt reactors, which produce tritium from lithium and beryllium salts at significantly higher levels than other reactor types. Modeling of such phenomena began following MSRE operations, and these early models attempted to predict measured tritium distributions in the MSRE, accounting for turbulent mass-transport processes (using established heat transfer correlations), permeation through a variety of metal structures such as heat exchanger tubes, and transport to and from bubbles introduced into the salt by gas sparging. The models reasonably reproduced the MSRE data, but did so best when the permeability of structures was reduced by about a factor of 1,000. This issue does not appear to have been conclusively resolved, and all of the more recent attempts to model tritium transport in molten salts appear to make use of the same methodology. MSRE remains, however, essentially our only source of integral tritium transport data relevant to MSRs. Here, we generalize the MSRE approach to permeation in order to include potential rate-limiting effects at interfaces, as well the effects of added hydrogen. Appropriately non-dimensionalized, this system of equations identifies two dimensionless numbers whose relative values clearly delineate the conditions under which mass transport, surface effects, permeation, and hydrogen swamping are expected to become rate-limiting. We also describe the preliminary conceptual design of a forced convection FLiBe loop, into which tritium would be introduced for the purpose of providing validation data for such a model. The primary purpose of this is to investigate the coupled transport phenomena described above and identify those that are rate-limiting in MSRs. Additional test section configurations are described that would address other transport phenomena relevant to MSRs, including bubbly flows and graphite interactions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

CNRS Molten Salt Reactor Benchmark Analysis Using Griffin-Pronghorn Coupled Multi-physics Code System

Molten salt reactors (MSRs) with flowing fuel have the unique feature of utilizing the fuel salt for heat generation and extraction at the same time, since the fuel salt is circulating through the whole primary loop. This movement of fuel salt results in a partial decay of the delayed neutron precursors (DNPs) outside the core and the corresponding redistribution of the DNPs in the active region of the core. To capture this effect, neutronics and thermal-hydraulics (T/H) codes need to be able to handle the movement of DNPs and their decay [1, 2]. Idaho National Laboratory (INL) is actively working on developing the neutronics and T/H codes to model the MSRs with flowing fuel. As one of the efforts, the neutronics code Griffin [3] and the T/H code Pronghorn [4] that are built upon the MOOSE framework [5] have been extended to handle the flowing fuel involving the drift of DNPs, and this capability is being tested and verified with MSR benchmark problems including the CNRS benchmark problem [6]. Additionally, a multi-physics analysis of the molten salt reactor experiment (MSRE) is being conducted to verify and validate the codes against available experimental data [7]. This work presents the analysis of the CNRS benchmark problem using the Griffin-Pronghorn coupled multi-physics code system of INL. Both steady-state and transient problems are analyzed, and the results are compared with those of other participants presented in [6]. The following section provides a description of the benchmark, along with input parameters and observables for each step. Then, the results for each step are presented in comparison with other participants’ results, followed by summary and conclusion.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

High Temperature Ternary Chloride Molten Salt Pump

This work investigates the design criteria considering thermal, structural, and material selection of a ternary chloride (MgCl2-NaCl-KCl) molten salt pump that is capable of running at temperatures above 720 °C. In this work, preliminary testing of salt-wetted hydrodynamic bearings was done to investigate different material pairings including Colmonoy grades, Stellite grades, and a novel Nickel Tungsten Carbide alloy (HybrimetTM NiWC3b) used as a High Velocity Oxygen Fuel (HVOF) coating. The coating has shown to withstand the corrosive nature of the molten salt. The pump wetted components are made of Inconel 625 and features a carbon ring seal (CIRPAC Seal) flooded with a N2 gas flow on the overhead to keep salt vapors from interacting with the external ambient air. A comprehensive pump test program has been completed. Over 90 hours run time pumping hot chloride salts at temperatures up to 720 °C in which enough data was captured to produce performance curves at 3 different operating speeds. This research showcases a promising design for use in the Concentrated Solar Power (CSP) Gen 3 liquid-pathway and sets the ground for further development and use of the HybrimetTM NiWC3b as a competitive alternative to available superalloys.

14 SOLAR ENERGY↗
DOE OSTI2022

Transient convective delayed neutron precursor of {sup 235}U for the molten salt reactor experiment

This work investigates the delayed neutron precursors from the fission of {sup 235}U in molten salt reactors. The six delayed neutron groups predict the spatial distribution of the advective transport for radioactive mass transfer in the molten-salt reactor experiment (MSRE). The Mole code, using the framework of the Multiphysics Object-Oriented Simulation Environment (MOOSE), was used to compute the delayed neutron concentration of the whole system. This approach can be used for analysis of reactivity in circulating conditions. The effect of delayed neutron concentrations on the static or dynamic behavior of the system in MSRE was analyzed. Using this approach, the pattern and validation of delayed neutron for a range of parameters of interest are reported and discussed herein. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multi-Physics Coupled Seismic Safety Analysis of Molten-Salt Reactors

Pool-type Molten Salt Reactors (MSRs) are a promising advanced nuclear reactor concept relying on passive physical behavior to provide increased safety characteristics. The high heat capacity, high boiling point, unpressurized liquid salt contains the nuclear fuel and as it passes through the vessel achieves criticality and produces nuclear power. The salt then carries away the heat produced along with the delayed neutron precursors towards heat exchangers and primary pumps before entering the core. This forms a highly coupled multiphysics problem between neutronics and fluid flow, which makes modeling these reactors challenging. The safety/licensing case heavily relies on modeling and simulation to prove their safety, and significant effort has been expanded by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to develop the appropriate simulation tools. Their safety/licensing will require a thorough analysis of their behavior during earthquake-based transients. Earthquakes are a transient condition commonly experienced in many regions, and are safely survived by dozens of reactors every year. Nonetheless, significant conservatisms were introduced with limited predictive modeling and simulation available. The first-of-a-kind capability developed by this research will allow for engineers to analyze ranges of design-basis and beyond-design basis accident scenarios to assess the integrity of the reactor vessel and surrounding system, possibly allowing for limiting the costly conservatisms in the design. This projects uses MASTODON, Griffin, and Pronghorn in the NEAMS ecosystem to model the coupled multiphysics problem posed by an earthquake in a molten salt reactor. It marches through a number of number of coupling schemes, from uncoupled simulations to tightly two-way coupled simulations. The appropriate level of coupling for these simulations will then be determined based on the accuracy in the quantities of interest (QoI) and computational effort involved in each scheme. The QoIs chosen for this paper include the mass flow rate across the heat exchanger (for Pronghorn simulations), power output of the reactor (for Griffin simulations), and the maximum Von Mises stress (for MASTODON simulations). This summary paper reports on the current progress of this project, with 2D tightly coupled multiphysics results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Demonstrate and De-Risk an Internally Lined Tank Design for Molten-Salt Thermal Energy Storage Benefitting CSP and Nuclear-Energy Sectors

The LIMOSA project extends the Gen3 CSP "Liquid Pathway" project to evaluate 720 degrees C thermal energy storage based on molten chloride salts. While SETO selected the Particle Pathway for Gen3 demonstration, LIMOSA was created to resolve a major risk factor with molten chloride salts - the design of the salt storage tank. The project team designed and assembled a test system leveraging experience from Oak Ridge National Lab and the Liquid Pathway Phases 1 and 2. The key innovation in the project is development of a graphite-foil barrier to protect insulation from salt intrusion and wetting.

chloride salt↗
DOE OSTI2024

Research Highlight - Dispatch Optimization, System Design and Cost Benefit Analysis of a Nuclear Reactor with Molten Salt Thermal Storage

We highlight our work from our previous publication titled "Dispatch Optimization, System Design, and Cost Benefit Analysis of a Nuclear Reactor with Molten Salt Thermal Storage" published in MDPI Energies 2022 (doi.org/10.3390/en15103599). Variable renewable energy availability has increased the volatility in energy prices in most markets. Nuclear power plants, with a large ratio of capital to variable costs, have historically operated as base load energy suppliers but the need for more flexible operation is increasing. We simulate the techno-economic performance of a 950 MWt nuclear power plant, based on the Westinghouse lead-cooled fast reactor, coupled with molten salt thermal storage as a method for flexible energy dispatch. We use the System Advisor Model to model the nuclear reactor thermal power input and power cycle operating modes. We combine this robust engineering model with a mixed-integer linear program model for optimized dispatch scheduling. We then simulate the coupled nuclear and thermal storage system under different market scenarios with varying price volatility. We find that the coupled plant outperforms the base plant under markets where energy price peaks fluctuate by a factor of two or more about the mean price. We show that a calculated power purchase agreement price for the plant improves by up to 10% when operating under California energy market conditions. Sensitivity analysis on the thermal storage cost shows that the optimal design remains unchanged even when doubling costs.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Corrosion Testing Of Additively Manufactured Stainless Steel 316H In Molten Salt Environments

The development of a new ASTM standard for evaluating the corrosion resistance of additive manufactured (AM) stainless steel (SS) 316H in chloride molten salts is critical for the use of these materials in extreme environments such as molten salt reactors (MSRs). This report pertains to a work package of the Advaned Materials and Manufacturing Technologies (AMMT) developing a systematic methodology to link changes in AM fabrication parameters, namely surface finishing, porosity, microstructure, and chemical heterogeneity, to corrosion performance in NaCl-MgCl2 salt, a proposed secondary coolant for MSRs. During Fiscal Year 2024, Idaho National Laboratory investigators in the AMMT program, utilized SS316H bars, fabricated with laser bed powder fusion at Los Alamos National Laboratory, to establish and optimize the workflow for evaluating these process-to-performance relationships. Standard practices for specimen preparation were established using these specimens, with particular focus on descaling and sectioning methods that align with ASTM guidelines. A comprehensive experimental design for static corrosion testing was developed, with pre- and post-exposure analysis utilizing optical microscopy and scanning electron microscopy. So far standard descaling techniques were optimized, one static corrosion test was conducted on the LANL AM SS316H specimens, and pre- and post-corrosion practices were established. In addition, the work package yielded a review paper on corrosion testing gaps for AM materials in nuclear applications and submitted a proposal for a rapid-turnaround experiment to the Nuclear Science User Facilities Program to investigate the combined effects of proton irradiation and corrosion on AM SS316H. This work package establishes a foundation for evaluating processing-to-performance relationships for AM SS316H in harsh conditions, contributing to the safe and efficient design of components for next-generation nuclear reactors. The creation of a standardized methodology and the generation of relevant publications and future research pathways represent significant strides towards integrating AM materials into critical applications where corrosion resistance is paramount.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Overview of Molten Salt Chemistry

This is a presentation to be given at an INL workshop. "Crossroads of Nonproliferation and Safeguarding Technologies for Implementation in Molten Salt Reactors", December 6-7, 2023. The presentation discusses historical molten salt reactor technologies and basic salt chemistries. Nothing in the presentation is sensitive or proprietary.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Multi-Physics Investigation of a Natural Circulation Molten Salt Micro-Reactor that Utilizes an Experimental In-Pile Device to Improve Core Physics and System Thermal-Hydraulic Performance.

INL employee PhD Dissertation - The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. This thesis investigates the performance benefit of a new device invented by the doctoral candidate. The device is referred to as a Wrapped Helix around an Inclined Plane or WHIP. The WHIP is protected under a provisional patent filed with the USA Patent Office on 28 September 2021 under application number 63/261,776, BEA docket number BA-1254. The WHIP device can be located in-core or near-core to promoted enhanced thermal-hydraulics and neutronics performance. While the WHIP can be employed in a variety of solid or liquid fuel designs, this thesis investigates the device’s benefit in the application of a natural-circulation, micro-molten salt nuclear battery concept (MsNB). This thesis will specifically investigate the temperature coefficient of the MsNB fuel (FLiNaK) using novel temperature sensitivity techniques unique to Serpent particle transport code, evaluate the WHIP’s thermal-hydraulic and neutronic performance effects using both established (STAR-CCM+) and novel (Python code developed to estimate the circulation effective delayed neutron fraction, ßef f ) analytical and numerical methods, evaluate the neutron noise behavior of the MsNB and how the WHIP may alter the character of the MsNB’s transfer function, and how the WHIP affects the autonomous, load-following performance under transient power conditions using Python code developed by the candidate. The sum body of this work, in part, has been published in three journal articles as the timing of the provisional patent process has allowed. Results show that creative utilization of WHIP engineering design and function reduces reactor system volume, fuel loading, control/stability in the buoyant regime

buoyant flow↗