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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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Criticality Safety SU based USL Calculation for UCl3-NaCl Fuel Salt Operations

The Molten Chloride Reactor Experiment (MCRE) is a fast spectrum, molten salt fueled reactor that is planned to be constructed and operated in the Laboratory for Operation and Testing in the U.S. (LOTUS) testbed, formerly known as the ZPPR cell, at Idaho National Laboratory (INL). MCRE will provide valuable data to support design, licensing, and operation of full scale commercialized molten salt reactor designs.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Development of Two-Step Method for Fast Chloride MSR Neutronics

This work presents neutronics models of a small and large fast-spectrum molten chloride-salt reactor. The models are similar to designs being pursued by industry, and they may serve as generic preconceptual and simplified neutronics models that provide information for decision making in licensing-related areas. Here, the two models were created using Serpent, a Monte Carlo neutron transport code, and Moltres, a neutron diffusion core simulator tool. Specifically, this study focused on exploring the applicability of diffusion theory to fast molten salt reactor (MSR) models, the capabilities of an open-source, MSR-oriented simulation tool (Moltres), and optimal energy-group structures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Temperature-Dependent Morphological Evolution during Corrosion of the Ni-20Cr Alloy in Molten Salt Revealed by Multiscale Imaging

Understanding the mechanisms leading to the degradation of alloys in molten salts at elevated temperatures is significant for developing several key energy generation and storage technologies, including concentrated solar and next-generation nuclear power plants. Specifically, the fundamental mechanisms of different types of corrosion leading to various morphological evolution characteristics for changing reaction conditions between the molten salt and alloy remain unclear. Here, in this work, the three-dimensional (3D) morphological evolution of Ni–20Cr in KCl–MgCl 2 is studied at 600 °C by combining in situ synchrotron X-ray and electron microscopy techniques. By further comparing different morphology evolution characteristics in the temperature range of 500–800 °C, the relative rates between diffusion and reaction at the salt–metal interface lead to different morphological evolution pathways, including intergranular corrosion and percolation dealloying. In this work, the temperature-dependent mechanisms of the interactions between metals and molten salts are discussed, providing insights for predicting molten salt corrosion in real-world applications.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Management of Alkali and Alkaline Earth Fission Products in Used Pyroprocessing Salt

Pyroprocessing of spent nuclear fuel (SNF) involves dissolving metallic fuel into a molten salt electrolyte (typically eutectic LiCl-KCl) and then preferentially depositing actinides onto inert cathodes. Subsequent operations include drawdown of residual actinides and lanthanides from the electrolyte prior to re-using the salt. The recovered actinides are recycled and the recovered lanthanides are disposed as waste. Alkali and alkaline earth metal fission products in the fuel, such as Cs, Sr and Ba, dissolve into the salt during electrorefining. The concentrations of these elements buildup over time in the molten salt electrolyte, which may change the freezing point. The radioactive decay of 137 Cs and 90 Sr (half life 30 and 29 years) generates significant heat and produces strong ionizing radiation fields (β and γ). The increasing heat load and radioactivity as these elements build up in the molten salt requires frequent replacement and disposal of the electrolyte salt. Alternatively, the salt can be treated to remove these and other elements and then recycled to the electrorefiner. An effective strategy to manage these alkali and alkaline earth metal fission products in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. Alkali and alkaline earth metal fission products are extremely stable in molten salt as chlorides--even more stable than the LiCl-KCl eutectic base salt--making them challenging to remove. They are not removed during drawdown operations to recover residual actinides and lanthanides and a separate operation is required to sufficiently purify the salt for reuse. This work is focused on selecting a method for separating Cs, Sr and Ba from the salt recovered from the lanthanide drawdown operation prior to recycling the cleaned salt back to the electrorefiner. Not addressed in this work is the management of the waste stream produced by the separation. This report summarizes the issues to be addressed when developing removal strategies for cesium, strontium, and barium and reviews existing methods to identify suitable methods and any technological gaps in their application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Novel Electrowinning Reactor for the Energy-Efficient, Low- Cost Production of Rare Earth Metals

This project developed a novel neodymium (Nd) electrowinning reactor for energy-efficient electrowinning of Nd metal. Throughout this project we have developed an alternative chloride based molten salt electrolysis process. Our process lowers the specific electrical energy consumption compared to the state of the art, while producing reusable chlorine gas and eliminating direct CO2 and PFC emissions. Facilities required for implementing the project were setup and designs were finalized, and a standard operating procedure for safe operation of high temperature electrolysis cells was written. The electrowinning reactor was designed and constructed. Electrolysis experiments confirmed the ability to reproducibly electrowin Nd metal on a Mo cathode. The current efficiency for Nd electrowinning was measured as a function of applied current density in the presence of the separator. Successful electrowinning of Nd sponge at high current densities (200 mA/cm2 and above) at a current efficiency >80% was demonstrated using multiple techniques. Stable Nd electrowinning up to 10h at 250 mA/cm2 was demonstrated. All of these design advancements were used to develop a techno-economic and life cycle assessment model that demonstrated that our process could be operated at cost of less than $0.20/kg-Nd (~30% lower compared to state of the art when comparing electrolysis energy cost) with a >20% total reduction in global warming potential compared to the state of the art while generating no direct CO2 or perfluorocarbon emissions.

42 ENGINEERING↗
DOE OSTI2022

Interaction of beryllium with 316H stainless steel in molten Li 2 BeF 4 (FLiBe)

The increased demand in renewable energy resources has led to a renewed interest in nuclear reactors including a new generation of molten salt reactors. One of the issues encountered in handling of molten chloride or fluoride salts is the inability of metals to form and retain a protective oxide surface layer, generally chromium oxide. The low chromium content, nickel-based alloy INOR-8, later marketed as Hastelloy N, was specifically developed to contain molten fluorides. This alloy had some deficiencies, but its composition resulted in it incurring far less chromium depletion by the molten fluoride salts than alloys like 316 stainless steel. For the next generation of molten salt reactors, an alloy with pressure vessel code approval and higher temperature capability than Hastelloy N is required. Consequently, several reactor designers have reportedly chosen to build the reactor containment vessel from 316H stainless steel. In order to minimize corrosion of the stainless steel by the molten fluoride salt (2LiF-BeF 2 ), it is proposed to add beryllium metal to the salt to react with impurities and provide a means to lower the oxidation potential thus making the salt less corrosive. Even so, there is a concern whether the beryllium would react with components of the stainless steel. To address this concern, static capsule tests were conducted in which selected amounts of beryllium were added to capsules containing 2LiF-BeF 2 (FLiBe) salt and 316H stainless steel and Hastelloy N samples. Following exposure, the samples were cleaned and examined using a variety of analysis techniques including, optical microscopy, SEM-EDS, XPS, LIBS, XRD and EPMA. Tensile samples were also exposed in capsules, and those were subjected to tensile testing. Notably, It was found that for the level of beryllium additions used in this study, intermetallic compounds were formed which could be detrimental to the long-term performance of the 316H stainless steel.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Mechanochemical Synthesis of High-Purity Anhydrous Binary Alkali and Alkaline Earth Chloride Mixtures

A direct synthesis route for high-purity, anhydrous binary salt mixtures has been developed. This atom efficient, solvent-free process is easily scalable, with the potential to produce salt mixtures that meet the purity standards required for industrial heat transfer and nuclear applications. The essence of the methodology lies in mechanochemical synthesis of carnallite precursors that can mitigate the hydrolysis of MgCl 2 ·6H 2 O under direct heating. Each dehydrated salt carnallite was then analyzed for purity and oxide content through subsequent powder X-ray diffraction, and strong acid titration. In conclusion, this process presents a more effective alternative route compared to previous methods for obtaining low-oxide, high-purity chloride salt mixtures.

Alkali↗
DOE OSTI2023

First-of-a-Kind Fuel-bearing Molten Chloride Irradiation Experiment

Molten salt reactors (MSRs) have garnered increasing attention recently with several demonstration efforts on the way. A key challenge to the licensing basis for these reactors is the lack of experimental data on fueled salts. This is expected to be crucial to the safety evaluation and licensing basis of reactors of this type deployed in the future. While capability for irradiating molten salts has been reestablished in the recent decade, no enriched fuel irradiation capability has been developed and tested as of yet. A new experiment vehicle under development at Idaho National Laboratory (INL) is presented here. The Molten-salt Research Temperature-controlled Irradiation (MRTI) experiment was developed to host enriched uranium bearing salt samples to be irradiated at a test reactor within the lab complex. One of the key scientific objectives is to provide irradiated salt samples for post-irradiation examination (PIE) to study the impact of fission product generation and neutron/gamma radioactivity on the salt solution and salt-facing wall material. This paper provides a detailed overview of the mechanical design of the experiment, followed by an overview of the fabrication and assembly of an initial prototype vehicle (with non-fuel-bearing salt). A summary of the key analyses conducted as a part of the performance and safety evaluation is then provided. Lastly, an overview of the test conducted in prototypic out-of-pile (non-neutron) environment are shown. These evaluations provide the foundation for a planned irradiation of an enriched uranium-bearing chloride salt sample in the near term. The upcoming irradiation will contain 13 cm3 of UCl3-NaCl salt (93% enrichment) generating around 20 W/cm3 of fission energy during irradiation and a temperature range that can be contained between bounds of 525-900°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

ALLOY SELECTION AND C-276 CODE DESIGN VALUE EXTENSION FOR ADVANCED MOLTEN SALT TECHNOLOGY TEST FACILITIES EXPERIMENTATION

Molten halide salts are being considered as working fluids for nuclear and concentrated solar power applications. High temperature molten fluoride and chloride salts are known to preferentially attack and deplete Cr in alloys, which leads to the use of high-Ni lower-Cr alloys in test facilities for advanced Molten Salt Technology. Alloys 600 and C-276 are two commercially available Ni alloys that have moderate Cr contents and are qualified to the maximum temperature of 649°C and 677°C (1,200°F and 1,251°F), respectively, in the Boiler and Pressure Vessel Code. Both alloys have good corrosion resistance to acids, are resistant to stress-corrosion cracking, and have long track records of use in the chemical industry. Comparatively, Alloy C-276 has a similar Cr content but much higher allowable stresses for temperatures greater than 600°C (1,112°F). Therefore, it has been considered as a structural material for test facilities that require operations at 700°C (1,292°F) or greater to develop high-temperature Molten Salt Technology. To meet the requirements, the current code design values were extended based on some experimental data. Analysis showed that above current Codified maximum temperature, strength of the alloy is mainly controlled by creep rupture life under the average stress, although the Sc creep rate criterion is close to the Favg.Savg rupture criterion. This paper presents the intended test facilities and the design requirements, alloy selection considerations, literature review, data analysis, and proposed allowable stress extension based on some creep test data for C-276 at temperatures greater than 677°C (1,251°F). Further research activities are also briefly mentioned.

Ren, Weiju↗
DOE OSTI2024

Removal and Deactivation of Bond Sodium from Fast Reactor Blanket Materials

The disposition of sodium-bonded spent nuclear fuel and blanket materials in a repository is complicated by the presence of sodium metal that is used as a thermal bond between the uranium metal fuel and blanket slugs and their cladding. The concern is that the metallic sodium could react with water, producing explosive hydrogen gas, or could exhibit a pyrophoric character. Thus, experimental studies were performed to investigate and demonstrate the removal and deactivation of bond sodium from blanket material in a dry environment. Specifically, bond sodium was removed from unirradiated Fermi-1 blanket elements and an assembly via a melt-drain-evaporate process using elevated temperature and reduced pressure. The effectiveness of sodium metal removal from the blanket materials and their associated cladding was =99.9998%, based on post-test quantitative analyses. The separated sodium metal was collected and subsequently deactivated by reacting it in a molten state with a controlled addition of ammonium chloride particles atop a molten salt medium. In this process sodium chloride is formed and assimilates into the salt pool. The subsequent deactivation of the bond sodium produced a solid ingot of sodium chloride, potassium chloride, lithium chloride, and cesium chloride that was devoid of sodium metal based on post-test analyses of the salt product. Both the sodium removal and deactivation operations were conducted within a dry inert atmosphere enclosure. The results of this study substantiate a path forward for the disposition of sodium-bonded blanket materials, including 34 metric tons heavy metal in irradiated Fermi-1 blanket material currently stored at Idaho National Laboratory.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Moltensaltpropnet

MoltenSaltPropnet is a physics-informed machine learning framework that aims to predict the thermophysical properties of molten fluoride and chloride salt mixtures, which are crucial for the design and safety of Generation IV molten salt reactors. The code processes data from the Molten-Salt Thermal Properties Database (MSTDB-TP) and the Janz compendium, converting critically evaluated correlations into fast, differentiable surrogate models for density, viscosity, thermal conductivity, and heat capacity across 448 distinct salt systems. The implementation consists of several key components: 1. Data Curation: The code parses and cleans the raw data, normalizing elemental mole fractions and extracting relevant regression coefficients for various thermophysical properties. 2. Feature Engineering: It generates fixed-length numerical descriptors that encapsulate the composition and temperature, incorporating polynomial interaction terms and dimensionality-reduction techniques to optimize model performance. 3. Coefficient Learning: Four different machine learning architectures are employed: a deep residual network (ResNet), a Kolmogorov–Arnold network (KAN), a sparsity-inducing neural network (SNN), and classical regression models. Each model learns to predict coefficients that define the temperature-dependent correlations for the thermophysical properties. 4. Property Reconstruction: The predicted coefficients are used to compute temperature-dependent property values, ensuring positivity and monotonic trends through a composite loss function that enforces physical constraints. 5. User Interface: An open-source web application enables users to filter the database, train task-specific models, and visualize the results, allowing for rapid exploration of candidate salt mixtures. MoltenSaltPropnet bridges the gap between limited experimental data and high-fidelity reactor simulations, providing a powerful tool for researchers in the field of molten salt reactors and advanced nuclear energy systems.

Retamales, Mauricio Eduardo Tano [Idaho National L↗
DOE OSTI2024

FY24 Progress Report on Viscosity and Thermal Conductivity Measurements of Nuclear Industry Relevant Chloride Salts: An Experimental and Computational Study

As presented in this report, experimental and computational techniques were performed to assess the viscosity and thermal conductivity of key alkali and actinide chloride mixtures for molten salt reactor developers. These mixtures were pure LiCl, NaCl-KCl, LiCl-NaCl, LiCl-KCl, LiCl-NaCl-KCl, and NaCl-UCl 3 . Experimental measurements of viscosity were performed with a rolling ball viscometer, whereas experimental measurements of thermal conductivity were performed with a variable gap apparatus. Additional benchmarking work was performed using both property measurement systems to prepare for x-ray radiography in stainless-steel crucibles for viscosity and to ensure that calibration methods were accurate for thermal conductivity before assessing the NaCl-UCl 3 system. Validation data for the NaCl-UCl 3 in literature are minimal. Details on the calibration methods, salt measurement processes, and sources of error and uncertainty are discussed in detail for both property measurements. The computational methods described herein involved ab-initio molecular dynamics (AIMD) calculations using CP2K. The calculations were performed for the LiCl-KCl-NaCl and NaCl-UCl 3 systems. These calculations not only provided thermophysical property estimations for comparison to experimental data, but they also allowed for the determination of diffusion coefficients, coordination numbers, and radial distribution functions to provide insight into ion mobility and local coordination environments, which is linked to macroscopic property trends.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Corrosion testing needs and considerations for additively manufactured materials in nuclear reactors

Metal additive manufacturing holds significant promise as an enabling technology for the 21st century nuclear energy industry. Metal additive manufacturing (MAM) can allow the fabrication of novel materials and innovative component designs that are not achievable through conventional manufacturing. Due to its very different fabrication methods, as-fabricated MAM components are characteristically different from conventionally manufactured components. MAM materials exhibit very different microstructures from conventional cast or wrought materials. For example, austenitic stainless steels fabricated by laser powder bed fusion exhibit columnar grain structures, dislocation cell structures, and melt pool fingerprints. In addition, MAM fabrication may result in defects such as porosity, incomplete processing of the feedstock (e.g., lack of fusion in melt-based methods), and oxide inclusions. Heat treatments may further evolve the microstructure, microsegregation, and stresses within the component. Furthermore, MAM components have a rough surface with feature sizes on the order of the feedstock material, as opposed to smooth surfaces resulting from conventional machining and forming operations. As a result, the corrosion behavior of MAM components will likely be significantly different from that of conventionally formed components. Corrosive environments for structural materials within advanced reactor environments include molten fluoride and chloride salts, liquid sodium and lead-bismuth, and high-temperature helium. The Advanced Materials and Manufacturing Technologies program within the Department of Nuclear Energy in the United States Department of Energy is assessing the unique concerns of MAM component corrosion and testing methodologies in advanced nuclear reactor environments. We discuss these concerns and testing strategies in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Cesium Removal from Surrogate Pyroprocessing Salt by Electrodeposition

Active metals in used nuclear fuel dissolve into the salt during pyroprocessing and are not removed by electrorefining or drawdown operations. The buildup of 137 Cs over time increases the heat load and ionizing radiation level of the salt such that it must be replaced frequently, resulting in a significant amount of salt waste. An effective means of managing cesium in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. A previous report summarized issues that must be addressed when developing a removal strategy and assessed the suitability of existing methods and remaining technological gaps to their application (Rose and Thomas 2023). Cesium is extremely stable in molten salt as a chloride—even more stable than the LiCl-KCl eutectic base salt used for pyroprocessing fuel—which makes removing cesium a challenge. However, sufficiently strong atomic interactions occur between active metal species and liquid metals that make the electrodeposition of active metal fission products into liquid metal electrodes energetically favorable. The feasibility of recovering cesium from LiCl-KCl pyroprocessing salt through electrodeposition into liquid metals is eing assessed by identifying potentially effective liquid metals and performing tests to determine the effectiveness of electrodepositing cesium from a LiCl/KCl salt into these liquid metals. Previous studies investigating the electrodeposition of Sr 2+ , and Ba 2+ into zinc, cadmium, bismuth, lead, tin and antimony have shown that alkali and alkaline earth metals can be electrodeposited at liquid metal cathodes (Kim et al., 2018). The removal of Ba 2+ and Sr 2+ was measured to be more efficient than the removal of monovalent cations due to the greater thermochemical driving force for alloying those elements with the liquid metal (Jang et al. 2022). Because the equilibrium potentials are dependent on the interactions of the active metal in the liquid metal, it is likely that the other alkali metals Li + , and K + , will deposit from LiCl/KCl salt with the Cs + . Therefore, application of this method to recover active metals from pyroprocessing salt will benefit from the use of a liquid metal and set of operating conditions that sufficiently increase the reduction potential of cesium to remove cesium from the waste salt with an acceptable amount of co-deposited lithium and potassium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Internal Insulation and Corrosion Control of Molten Chloride Storage Tanks

The U.S. Department of Energy Gen3 CSP program aims to develop integrated systems capable of delivering high temperature thermal energy to advanced power cycles [1]. The National Renewable Energy Laboratory (NREL) is leading a multi-national team to validate the potential of a chloride-based molten-salt system that uses a ternary blend of MgCl2/KCl/NaCl to provide higher temperature thermal energy storage capability (Fig. 1a). The economic viability of this concept is dependent on internally insulating the hot and cold storage tanks to enable the use of low-cost steel shells (Fig. 1b).

14 SOLAR ENERGY↗
DOE OSTI2023

First-of-a-Kind Fuel-bearing Molten Chloride Irradiation Experiment

As a dozen MSR developers in the U.S. work toward an aggressive commercialization timeline, many of their fueled-salts—notably, chloride-based compositions—have never been irradiated. Licensing and operating these reactors requires an understanding of (1) the source term, radiation chemistry, and gas generation; (2) unanticipated irradiation-induced corrosion effects; and (3) the impact of burnup on thermophysical properties, all of which can be deduced through irradiation testing. Idaho National Laboratory (INL) is ideally suited to leverage its expertise in chloride salt chemistry, as well as its Neutron Radiography Reactor (NRAD) facility in the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC), to conduct the world’s first fuel-bearing chloride salt irradiation and fill in the knowledge gaps pertaining to salt chemistry under irradiation. Molten-salt Research Temperature-controlled Irradiation (MRTI) consists of a salt-containing capsule that is internally heated inside of a secondary containment. The experiment must rely on resistive heating to melt the salt before irradiation (to avoid the impact of radiolysis) and once the reactor is turned on and fission reactions commence in the salt, the resistive power can then be reduced. Salt-immersed thermocouples coupled to a controller allow for the heater power to be adjusted as needed to meet experimental objectives. The bulk of the experimental results will be achieved through Post-Irradiation Examination (PIE). At which point a range of different measurements are anticipated to assess the salt/plenum/wall composition, the capsule corrosion rate, and the evolution of salt properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗