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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Effects of PLA/PHB Blend Ratio and Wood Flour Loading on the Melt Rheology and Thermomechanical Properties of Biobased Polymer Composites

The development of sustainable, biobased polymer composites is crucial for reducing supply chain dependence on fossil fuels. A major challenge lies in understanding and predicting the processability of these sustainable material alternatives, which directly impacts large-scale manufacturing. Here, this study systematically investigates the effect of poly(lactic acid) (PLA)/polyhydroxybutyrate (PHB) blend ratios and wood flour (WF) loadings on the thermal, rheological, and mechanical properties of the composites. Composites were prepared via melt compounding and characterized by using differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), oscillatory shear melt rheology, tensile testing, and scanning electron microscopy. Results showed that increasing the PHB content lowered the glass transition temperature, while higher PHB and WF loadings decreased thermal stability. Complex viscosity decreased with increasing PHB in 0 and 10% WF, whereas it increased in 20% and 30% WF for PHB-rich blends (80/20 and 70/30 PLA/PHB) due to stronger filler interactions with flexible PHB chains. At 30% WF, PHB-containing composites also showed elastic dominance with G′ higher than G″. This behavior was confirmed by van Gurp–Palmen plots, where the phase angle decreased with increasing PHB and WF loadings. WF addition also shifted Cole–Cole plots away from semicircular terminal relaxation, while time–temperature superposition validity was maintained, confirming that polymer chain dynamics continue to dominate melt behavior. Young’s modulus increased with WF loading, while tensile strength and toughness decreased due to weaker interfacial adhesion between the matrix and the filler. Interestingly, 10% WF systems containing PHB exhibited the highest toughness among all of the filled systems, indicating synergistic reinforcement from WF fillers and PHB-induced ductility. Overall, melt rheology effectively captured the internal structure and stiffness of the composites as the formulation changed, reinforcing the concept that the PLA/PHB ratio and WF content can be tuned to adjust melt elasticity and balance mechanical property trade-offs for targeted applications.

melt rheology↗
DOE OSTI2025

Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds

The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard U model (DFT+U) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+U calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. In conclusion, the DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

History of Melt-Processing System at SRNL

Capabilities of Melt Treatment Technology • Research Reactor Aluminum SNF (can extend to SS and Mg Clad Fuel) Aluminum Cladding Alloy • Typical thickness = 38 mils • Boehmite layer > 50 µm • Fuel Core • U -Alx, U308-Al; U3Si2-Al • Fabricated by cast extrusion or powder metallurgy • Enrichment = 20% to > 90% U235 • Typical Reactor Burnup = 5% to 55% • Thermal Output: <100 W, after 10 Years in storage

Adams, Thad M.↗
DOE OSTI2020

Review of thermophysical property methods applied to fueled and un-fueled molten salts

Recent investigation into designs of molten salt reactors (MSR) requires accurate data of candidate salts. Review of the literature shows lacking thermophysical (TP) property data for coolants and fuel-bearing salts of Advanced High Temperature Reactors (AHTR). Measurement uncertainties of current data are even more deficient, and this affects ongoing research where MSR modeling and testing are being performed. Melting temperature, viscosity, density, specific heat capacity, and thermal and electrical conductivity measurements are often not arranged together or discussed in great detail, especially for fuel-bearing salts. Raw data from key findings spanning over 70 years for LiF, LiF-BeF 2 , LiF-NaF-KF, KCl-MgCl 2 , KF-ZrF 4 , and NaF-KF-MgF 2 have been organized and presented here. Brief description of methods used to obtain these results, in addition to the TP data of some of these salts as fuel carriers, has also been discussed. Thermal conductivity data are particularly weak due to older, high-error measurement techniques. More recent measurement systems to evaluate salt thermal conductivity show promise for obtaining consistent absolute values. This work concludes that measurement systems such as transient hot-wire method, forced Rayleigh scattering, and laser flash analysis provide the most promise to fill these data gaps and reduce current uncertainty in thermal conductivity measurements of both fueled and un-fueled molten salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Radiological Characterization of Uranium Decay Products in LEU U-10Mo Before and After Heat Treatment

Fuel for the U.S. high-performance research reactor fleet is undergoing significant development as the United States moves away from using highly enriched uranium dispersion fuels. The proposed fuel is a high-density, low-enriched uranium (LEU, 19.75 wt% U 235) alloyed with ten weight percent molybdenum fuel (known as LEU U 10Mo). The LEU U 10Mo fuel samples studied in this report were made by down-blending highly enriched uranium with a master alloy, which is cast from molybdenum rods and depleted uranium, via vacuum induction melting and then casting.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Modeling Time-Dependent Surrogates of Additive-Manufactured Nuclear Fuels Processes

Additive manufacturing (AM) technology is being increasingly adopted in a wide variety of application areas for its ability to rapidly produce, prototype, and customize designs. Recently, a hybrid AM technique was successfully developed at Idaho National Laboratory (INL) to manufacture nuclear fuels [1]. Despite the advantages, this AM technique needs optimization due to defects from a highly complex melting and sintering process. The complex metallurgical phenomena during AM processes are strongly related to parameters such as applied laser power, traveling speed, and scan style, which could lead to differences in density, residual stress, crystallographic texture, and mechanical properties. In addition, stochastic variations in laser energy interaction and associated multiscale/multiphysics phenomena cause variations in microstructure evolution and mechanical properties. Currently, researchers at INL are focusing on developing a comprehensive modeling framework, leveraging INL’s simulation tools MOOSE/MARMOT/BISON/RAVEN [2-4] to describe all steps of this AM process across multiple length scales. Although this advanced framework plays a critical role in enabling enhancements to traditional trial and error approaches for design and optimization of nuclear fuel materials, it remains computationally intense, limiting its use in sensitivity and optimization analysis. In this case, an accurate and inexpensive surrogate becomes an effective tool for providing a tractable approximation of the underlying underline physics. Surrogate models generally not based on the physics of a system are purely mathematical models used to capture the relationships between specific system inputs and outputs. Popular approaches, including neural networks [5], response surfaces [6], and subspace-based reduced order models [7], have been applied to a wide range of disciplines, such as nuclear reactor design, aerospace design and automotive design. In this summary, we employ advanced time-dependent surrogate models such as high-dimensional model representation (HDMR) [8] and physics-informed deep neural network (PINNs) [9] to accelerate the design and optimization of AM process.

42 ENGINEERING↗
DOE OSTI2026

Physicochemical evolution of uranium nitride kernel microstructure with varying carbon distribution for advanced TRISO fuel forms

Uranium nitride (UN) has emerged as a fuel candidate for advanced nuclear reactor concepts due to its superior uranium density, thermal conductivity, and high melting temperature. However, the fabrication route for converting UO 2 to UN is complex and difficult to standardize. Although the chemistry of this conversion process is well-studied, more insight into the physicochemical dynamics of this conversion using advanced characterization techniques can help further our understanding of this material system. This work leveraged thermogravimetric analysis (TGA), X-ray diffraction (XRD), and nondestructive 3D X-ray computed tomography (XCT) to characterize dynamic microstructural changes in the UO 2 → UCO → UN fabrication pathway for two kernels with a varying carbon distribution in the starting composition. TGA and XRD were used to quantify changes in the mass, density, and chemical composition of the two kernels, while three-dimensional image processing and segmentation of XCT data were used to quantify the volume, surface area, and spatial distribution of features within each kernel for multiple steps along the fabrication pathway. The analysis indicates distinct differences between the two kernels that are correlated to downstream conversion efficiency. In conclusion, this work is among the first to perform 3D quantification of physicochemical evolution during UN conversion, providing quantitative correlation between processing, properties, and expected fuel performance.

Nuclear fuel↗
DOE OSTI2021

A first principles framework to predict the transient performance of latent heat thermal energy storage

Thermal energy storage (TES) is increasingly recognized as an essential component of efficient Combined Heat and Power (CHP), Concentrated Solar Power (CSP), Heating Ventilation and Air Conditioning (HVAC), and refrigeration as it reduces peak demand while helping to manage intermittent availability of energy (e.g., from solar or wind). Latent Heat Thermal Energy Storage (LHTES) is a viable option because of its high energy storage density. Parametric analysis of LHTES in terms of dimensionless numbers is highly desired as a tool to model LHTES systems. One approach is to develop a model equation so as to minimize the error between the model and data obtained from experiments or simulations. While this approach can produce an accurate correlation applicable within the range of data used for its creation, it does not provide physical understanding of the rate-limiting process controlling the transient behavior of the device. In this paper we present an alternative approach whereby the potential rate-limiting processes are identified from first principles and then the key process is determined as a function of time as a LHTES device is charged. For example, in a simple geometry, the melt-fraction can be expected to vary linearly in time if the heat transfer rate is limited by natural convection of the phase changing material and we show it scales with the PCM Grashof number as $Gr^1_p$ and PCM Prandtl number as $Pr_p^{(1/3)}$. On the other hand, if surface area of solid PCM limits the heat transfer rate, the melt fraction increases asymptotically to reach full melting. The existence of these linear and asymptotic regions and the $Gr^1_pP r^{1/3}_p$ shape of the melt fraction curve is verified using our database of 64 simulations. Of practical importance in designing LHTES devices is the melt fraction at which the heat transfer rate ceases to be limited by convection, after which the heat storage rate deteriorates. For our geometry, this is found to be about 90%. This test case of our methodology shows the value of our approach, that predicting heat storage rate based on the rate-limiting physical phenomenon as a function of time is an effective approach to modeling LHTES devices.

25 ENERGY STORAGE↗
DOE OSTI2022

Silicon alloying enhances fast heating rate combustion of aluminum particles

Alloys and composites of aluminum (Al) have shown promise in regulating and enhancing particle combustion for energy generation applications. Recent work on aluminum-silicon (Al-Si) spherical alloy particles has shown improved combustion at low heating rates through enhanced diffusion accompanying a lower melting temperature. This study extends reactivity analysis to higher heating rates, comparing oxidation of Al-Si with Al. Flame speeds of Al-Si powder mixed with molybdenum trioxide (MoO 3 ) powder (Al-Si+MoO 3 ) exhibited a faster transition to steady propagation relative to Al+MoO 3 . Bomb calorimetry experiments revealed up to 5.8% greater early temperature rise for Al-Si powder. Rapid steady propagation for Al-Si mixture and faster temperature rise for Al-Si particles were attributed to accelerated kinetics evidenced in thermal equilibrium analysis of the mixtures using a differential scanning calorimeter (DSC). Larger DSC exotherms in the early stages of oxidation (i.e., 480–720 °C) correlated with early heat release and promoted steady flame propagation for Al-Si+MoO 3 compared with Al+MoO 3 . Furthermore, in multiple heating rate DSC studies, Al-Si+MoO 3 ignited at a lower heating rate (i.e., 15 ⁰C/min) than Al+MoO 3 (i.e., 20 ⁰C/min). Both mixtures ignited at temperatures less than 625 °C which is above the melting temperature of Al-Si (574 °C) and below Al (659 °C). Thus, alloy particle fuels with enhanced diffusion-controlled kinetics promote steady flame propagation and show promise for energy generation applications. This is especially promising for technologies driven by enhanced condensed phase combustion such as some propellant additives or applications dependent on reliable burn rates such as pyrotechnic time delay formulations and primers.

36 MATERIALS SCIENCE↗
DOE OSTI2026

A robust alloy design (RAD) strategy for next-generation (IV) nuclear fission reactors

Next-generation nuclear reactors demand structural materials capable of withstanding extreme conditions, including high temperatures, intense neutron flux, and corrosive environments. Multi-Principal Element Alloys (MPEAs) have emerged as promising candidates due to their exceptional radiation tolerance, thermal stability, and compositional flexibility. This study introduces a versatile and customizable Robust Alloy Design (RAD) strategy for systematically designing MPEAs for GEN-IV reactor fuel cladding. The RAD framework integrates nuclear-relevant selection criteria, empirical parameter assessments, and high-throughput CALPHAD simulations to efficiently narrow compositional space and identify stable alloys. A unified RAD score developed for the first time, combines key performance metrics, including fuel-clad chemical interaction (FCCI), neutron absorption cross-section (NAC), valence electron configuration (VEC), and melting point factor (MPF), into a flexible ranking system adaptable to reactor-specific priorities. Among 724 candidates, V555(5Al–5Cr–5Fe–85V) emerged as the top alloy, validated experimentally with a homogeneous single-phase BCC microstructure and superior mechanical properties (nano-indentation: 3.389 ± 0.258 GPa; Vickers hardness: 240 ± 6.7 HV), significantly outperforming Zircaloy-4 and V-4Cr-4Ti. Importantly, the RAD strategy is not limited to nuclear applications; its customizable weighting system enables scalability to other extreme environments. This adaptability positions RAD strategy as a versatile tool for advanced materials design across multiple industries.

Alloy design↗
DOE OSTI2021

Automatic Differentiation in MetaPhysicL and Its Applications in MOOSE

Efficient solution via Newton’s method of nonlinear systems of equations requires an accurate representation of the Jacobian, corresponding to the derivatives of the component residual equations with respect to the degrees of freedom. In practice these systems of equations often arise from spatial discretization of partial differential equations used to model physical phenomena. These equations may involve domain motion or material equations that are complex functions of the systems’ degrees of freedom. Computing the Jacobian by hand in these situations is arduous and prone to error. Finite difference approximations of the Jacobian or its action are prone to truncation error, especially in multiphysics settings. Symbolic differentiation packages may be used, but often result in an excessive number of terms in realistic model scenarios. An alternative to symbolic and numerical differentiation is automatic differentiation (AD), which propagates derivatives with every elementary operation of a computer program, corresponding to continual application of the chain rule. Automatic differentiation offers the guarantee of an exact Jacobian at a relatively small overhead cost. In this work, we outline the adoption of AD in the Multiphysics Object Oriented Simulation Environment (MOOSE) via the MetaPhysicL package. We describe the application of MOOSE’s AD capability to several sets of physics that were previously infeasible to model via hand-coded or Jacobian-free simulation techniques, including arbitrary Lagrangian-Eulerian and level-set simulations of laser melt pools, phase-field simulations with free energies provided through neural networks, and metallic nuclear fuel simulations that require inner Newton loop calculation of nonlinear material properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Phase Behavior of the Ternary NaCl-PuCl 3 -Pu Molten Salt

There is a gap in our understanding of the behavior of fused and molten fuel salts containing unavoidable contamination, such as those due to fabrication, handling, or storage. Therefore, this work used calorimetry to investigate the change in liquidus temperature of PuCl 3 , having an unknown purity and that had been in storage for several decades. Further research was performed by additions of NaCl, making several compositions within the binary system, and summarizing the resulting changes, if any, to the phase diagram. The melting temperature of the PuCl 3 was determined to be 746.5°C, approximately 20°C lower than literature reported values, most likely due to an excess of Pu metal in the PuCl 3 either due to the presence of metallic plutonium remaining from incomplete chlorination or due to the solubility of Pu in PuCl 3 . From the melting temperature, it was determined that the PuCl 3 contained between 5.9 to 6.2mol% Pu metal. Analysis of the NaCl-PuCl 3 samples showed that using the Pu rich PuCl 3 resulted in significant changes to the NaCl-PuCl 3 phase diagram. Most notably an unreported phase transition occurring at approximately 406°C and a new eutectic composition of 52.7mol% NaCl–38.7mol% PuCl 3 –2.5mol% Pu which melted at 449.3°C. Additionally, an increase in the liquidus temperatures was seen for NaCl rich compositions while lower liquidus temperatures were seen for PuCl 3 rich compositions. It can therefore be concluded that changes will occur in the NaCl-PuCl 3 binary system when using PuCl 3 with excess Pu metal. However, melting temperature analysis can provide valuable insight into the composition of the PuCl 3 and therefore the NaCl-PuCl 3 system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Finite temperature properties of uranium mononitride

Uranium mononitride (UN) is a promising nuclear fuel that combines the advantageous properties of readily used UO 2 and uranium alloys, such as high melting temperature and high uranium density, and thermal conductivity, respectively. A better understanding of UN behavior at operating temperatures can be obtained from finite temperature data, such as elastic properties. To get this information, ab initio molecular dynamics (AIMD) simulations were performed at five different temperatures using constant volume (NVT) and constant pressure (NPT) ensembles. Initially, the performance of PBE functional in reproducing experimental crystallographic properties and magnetic ordering is assessed. The finite temperature phonon dispersions are calculated using NVT simulation results, which show a softening of the phonon modes with increasing temperature. The NPT results are used to obtain the thermal expansion of UN and finite temperature electronic properties. The calculated thermal expansion is compared with our measurements using neutron diffraction. Additionally, the temperature dependent elastic properties of UN are evaluated using the strain-stress method in AIMD simulations, indicating that UN becomes softer and more compressible with increasing temperature. Also, the calculated Young’s modulus slope is in very good agreement with the experiment. The finite temperature heat capacity and electronic thermal conductivity are calculated from AIMD simulations, which are in better agreement with the experiment than the heat capacity and thermal conductivity calculated using the structures relaxed at 0K. Finally, the thermal diffusivity from AIMD has opposite temperature dependence compared to experimental results, which we argued comes from the underestimated electronic thermal conductivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of a TRISO assessment case for BISON based on transient experiments in the NSRR

The tristructural isotropic (TRISO) fuel assessment and validation database currently available in BISON is based mainly on steady-state irradiation and high-temperature furnace testing. Transient assessment cases are needed to support ongoing U.S. industry efforts to design and deploy commercial reactors that utilize TRISO fuels. Most available historical transient tests involving TRISO fuels were characterized by power densities, temperatures, and energy depositions that were highly conservative with respect to typical high-temperature gas-cooled reactor (HTGR) accident conditions. Nonetheless, assessment of BISON's predictive capabilities, adaptation of material properties toward high-temperature and high-particle-power regimes, and development of a systematic approach to validating BISON for TRISO transient applications now motivate the modeling of historical transient tests. This work describes the development of 1-D thermal models of transient experiments conducted at the Nuclear Safety Research Reactor (NSRR) in Japan, with BISON predictions of energy deposition, UO2 melting onset, and molten volume fractions being compared against experimental measurements. Melting is accounted for by defining an effective specific heat capacity for UO2—one that leverages the material's heat of melting. BISON's predictions of energy deposition and molten volume fraction are in good agreement with the experimental data from the low-energy-deposition tests; however, BISON tends to overpredict melting at higher energy depositions. Though overly conservative compared to the operating conditions expected for near-term TRISO-fueled reactor applications, these simulations effectively exercise BISON TRISO models over a wider range of conditions than those encompassed by the existing assessment database. Potential contributors to the observed discrepancies are noted and additional future developments proposed.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Analysis and Thermal Property Investigations into Ternary Actinide Chloride Salt Systems Containing UCl3 and PuCl3

While regulators, the scientific community, and MSR developers still lack access to literature data on the thermal properties of clean fuel salts, even less information is available on the properties of fuel salts containing impurities. It is essential to understand, benchmark, and predict crucial data on the changes in thermal properties of fuel salt systems due to impurities arising from moisture, corrosion, and reactor operation (i.e., fission products). This research focuses on two actinide fuel salts (1) to investigate a worst-case scenario buildup of actinide fission product in a NaCl-UCl3 eutectic fuel salt and (2) to investigate NaCl-PuCl3 eutectic salt after 1000 hours of operation in a natural circulation flow loop flow to determine if corrosion or atmospheric (moisture/oxygen) products are present. For the first salt, a conservative assumption or worst-case scenario, for fission product buildup in a fuel salt was investigated by adding PuCl3 to eutectic 67 mol% NaCl – 33 mol% UCl3 salt resulting in a ternary salt having a composition of 61 mol% NaCl – 30 mol% UCl3 – 9mol% PuCl3. Addition of PuCl3 to eutectic NaCl-UCl3 resulted in a ternary salt that had a higher melting temperature than either the NaCl-PuCl3 or NaCl-UCl3 binary eutectic mixture. Addition of PuCl3 also resulted in an increase in density which was expected. The second salt was extracted from a micro loop. The composition of the fuel (primary) salt prior to flow loop operations was determined to be 64 mol% NaCl – 36 mol% PuCl3, however, the post-flow loop salt showed increased levels of MgCl2 and NaCl changing the salt composition to 10 mol% MgCl2 – 63mol% NaCl – 26mol% PuCl3) indicating the primary salt interacted with the rinse salt. Analysis of the post flow loop salt detected low concentrations of Al, Ni, Co, Nb, and Zr, most likely corrosion products from the flow loop material of construction. Contamination of the fuel salt (with the rinse salt NaCl-MgCl2) decreased the density by approximately 10% and reduced the onset of melting temperature by 50 °C, from 451 °C to approximately 400 °C. Results from the fission product simulated salt (61 mol% NaCl – 30 mol% UCl3 – 9mol% PuCl3) and the corrosion product salt (10 mol% MgCl2 – 63mol% NaCl – 26mol% PuCl3) will be included in two separate manuscripts for submission to peer-reviewed journals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

FIDES-II/P2M Simulation Exercise on AN3 and AN10 bump tests: main results

This paper presents simulations of bump tests conducted as part of the P2M (Power to Melt and Maneuverability) project within the OECD/NEA FIDES-II framework (Framework for Irradiation Experiments). To prepare fuel performance codes for analyzing the P2M experiments, an international Simulation Exercise (SE) was initiated under the P2M project. The objective of this exercise was to model the AN3 and AN10 bump tests carried out in the 1980s at the DR3 reactor (3rd RISØ Fission Gas Project), during which instrumented fuel rodlets (equipped with thermocouples and pressure sensors) were subjected to prolonged power transients and power dips. The ongoing simulation exercise involves seven organizations from industry, research, and regulatory bodies, each using different fuel performance codes. This paper provides detailed descriptions of the test cases and simulation results from six fuel performance codes (ALCYONE, BISON, FAST, FEMAXI, FRAPCON, and TRANSURANUS), with a particular focus on fission gas release kinetics.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Synthesis and Thermophysical Property Determination of NaCl-PuCl3 Salts

Currently, a knowledge gap exists in the available data and understanding of thermophysical properties relating to fresh fuel salts, especially those containing plutonium. Determination of this data is necessary for the design and construction of test reactors, as well as the licensing of future commercial molten salt reactors. Thermophysical properties such as melting temperature, salt stability, density, and heat capacity were determined on synthesized eutectic NaCl-PuCl3 and a more sodium rich composition containing 25 mol% PuCl3. These measurements document the baseline properties of the salt as a function of temperature for future experiments on irradiated fuel salt which will provide a holistic perspective on the change of thermophysical properties during reactor operations. It was determined that the NaCl-PuCl3 ingot synthesized for this study contained 63.4 mol% NaCl, 36.3 mol% PuCl3, and was 99.7% pure. Upon heating it was shown that the NaCl- PuCl3 eutectic was stable at temperatures of 800°C. The onset of melting occurred at 541°C and the enthalpy of fusion was determined to be 140.7 ± 8.4 J/g. Specific heat capacity measurements showed a slightly decreasing trend with respect to temperature in the liquid phase ranging from 0.67 to 0.57 J/g·K, with an average value of 0.637 ± 0.03 J/g·K (104 ± 5 J/mol·K) between 500 to 720°C. Three independent trials of the molten NaCl-PuCl3 eutectic salt found the density to be ?(T) = 3.8589 – 9.5342·10-4 T(°C), validated between 500 to 800°C. In addition to salt synthesis and experimentally determining thermodynamic properties, Ab Initio molecular dynamic (AIMD) simulations were used to calculate density and heat capacity values.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗