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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Myths of nuclear graphite in World War II, with original translations

We re-examine a common narrative that experimental errors by Walther Bothe in 1941 led Germany to abandon graphite as a reactor moderator during World War II. Using document-based nuclear archaeology, we first show that both American and German scientists used an incorrect carbon scattering cross section, thereby undermining the accuracy of all wartime data, including their conclusions on carbon’s absorption. Moreover, we argue that the availability of exceptionally pure petroleum coke in the United States, rather than any academic breakthrough, decisively enabled their production of nuclear-grade graphite. In contrast, Bothe’s Siemens electrographite had more boron contamination than any graphites considered in Fermi’s experiments, rendering it genuinely impractical as a moderator. By reframing the decision to eschew graphite as a deliberate decision rather than a mere experimental oversight, we believe the German decision was a rational consequence of material constraints and wartime priorities.

42 ENGINEERING↗
DOE OSTI2025

Thermocouple and RTD Drift Due to Neutron Irradiation and Temperature for Space Nuclear Applications: A Review

Temperature measurement is a key factor in the safe and reliable operation of nuclear reactors. Thermocouples and resistance temperature detectors (RTDs) are among the most common and reliable tools deployed for temperature measurement, and some are capable of operating in extreme environments such as very high temperature and neutron radiation fields anticipated in nuclear thermal propulsion (NTP) or fission surface power. These devices are known to exhibit drifting behavior in their temperature readings throughout their operational lifetimes. Here, to help select appropriate instrumentation, it is important to identify the factors that influence drift as well as typical behavior in these environments. This work is focused on identifying contributing factors in the performance of thermocouples and RTDs. An extensive literature review is included that discusses several experimental results in different environments. Although individual instrument performance may vary with construction methods, materials selection, and vendors, this review aims to assist in the selection process and identify typical expected behaviors of several devices.

Chesser, Ryan J. [Oak Ridge National Laboratory (O↗
DOE OSTI2024

Atomistic Simulations for Thermophysical Properties of Uranium-Containing Halide Molten Salts

Characterizing the thermophysical properties in both fuel and coolant salts are critical in modeling, developing, process optimizing and utilizing molten salt reactors (MSRs), as these properties directly relate to operation metrics and can inform on the selection of candidate salts. The demand for consistent, accurate and publicly available thermophysical property data has become more apparent in recent years as interests have increased from molten salt reactor developers. There are a number of challenges in experimentally measuring properties such as thermal conductivity, viscosity, density and heat capacity , which have led to sparse and often times conflicting data points or molten salts in general. Additionally, there are a number of hazards to consider when synthesizing, storing, using, treating and disposing of molten salts. With the advances in computational capabilities over the last 10 years, the use of atomistic simulations can be implemented to support these efforts. The primary objective of this work is characterize the thermophysical transport properties in a number of molten chloride salts, and in particular NaCl-UCl 3 using ab-initio molecular dynamic (AIMD) simulations. In this binary salt the UCl 3 acts as the primary fissile material and NaCl acts as a carrier salt due with its’ high solubility for actinides A number of studies on the thermophysical properties of NaCl-UCl 3 have been published but there is not a vast amount of viscosity data for this system. In 1975, Desyatnik, et al published a study reporting dynamic viscosities that were calculated from kinematic viscosity measurements, and using the coefficients provided the viscosity in a 70:30 NaCl:UCl 3 mixture is 2.29 cP and 2.88 for a 60:40 mixture. Termini et al. recently reported viscosities in the range of 2.75 – 3 cP for the 63:37 NaCl-UCl 3 mixture in the same temperature range using rolling ball viscosity measurements. Computational viscosity of a similar mixture (64:36) can be obtained from the work Andersson et al. using the reported diffusion coefficients, and the hydrodynamic radius from the pair-radial distribution functions (RDFs). Using Eq (1) (vida infra), the viscosity would be 2.50 cP at 1100K. This is not to say that these values are incorrect due to the varying reported values, but aims to highlight the necessity of this work. The data reported in this ongoing work are computations on a 64:36 mixture of NaCl-UCl 3 at 987K. This work is likely to be expanded into varying concentrations of this mixture along with the inclusion of other salt candidate mixtures.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Local chemical ordering of a neutron-irradiated CrFeMnNi compositionally complex alloy

While ion-irradiation studies are a critical first step in studying compositionally complex alloys (CCAs) for nuclear applications, they do not capture all the microstructural changes occurring under the low irradiation dose rates and different particles’ scattering patterns experienced in a nuclear reactor setting. To explore these phenomena in reactor-relevant conditions for the first time in CCA, the single-phase solid-solution Cr 10 Fe 30 Mn 30 Ni 30 was neutron irradiated up to 6.61 displacements per atom at 395 and 579 °C. Irradiation-enhanced local chemical ordering (LCO) well beyond the range of short range ordering was observed, and is predicted to be the precursor to the precipitation of a coherent Ni-Mn L1 0 phase and a Cr-rich α’ phase, though TEM analysis did not indicate the presence of either in any irradiation condition. The line density of faulted dislocation loops decreased from 6.47 to 1.69 ∙ 10 15 m -2 from 3.43 to 6.61 dpa at 579 °C despite no appreciable faulted loop content in the unirradiated material. LCO is expected to increase the complexity of the energy landscape within this alloy, restricting interstitial point defect mobility and creating local regions of greater stacking fault energy. These contribute to the negative correlation between irradiation dose and faulted dislocation loop density in this study, as well as the lack of void swelling observed.

36 MATERIALS SCIENCE↗
DOE OSTI2025

ARC-DPA 1-MeV Neutron Fluence Equivalence Model for GaAs

Here, the legacy data used in the production of the American Society for Testing and Materials (ASTM) E722 standard for gallium arsenide (GaAs) were fit with the athermal recombination corrected-displacements per atom (ARC-DPA) model for the development of a novel 1-MeV neutron equivalent fluence metric. Improving on previous work, the coefficients for the ARC-DPA model were optimized using a conjugate gradient method, and the uncertainties in the observed damage in different benchmark neutron fields were sampled to quantify their contribution to the damage metric. The optimized parameters of −0.474 ± 0.03 for b arcdpa and 0.016 ± 0.002 for c arcdpa provided good agreement and indicated that the solution is not sensitive to the corresponding fluence uncertainties. Compared to the current 1-MeV neutron fluence equivalence standard for displacement damage in GaAs, our results indicate that for a thermal reactor neutron environment, displacement damage has been underestimated by around 25% and that the saturation value for the damage efficiency had been higher than previously modeled.

1-MeV equivalent fluence↗
DOE OSTI2025

Quality Assurance of the Irradiated EBR-II Metallic Fuels Data

The U.S. DOE-NE’s Advanced Reactor Technologies (ART) Fast Reactor Program (FRP) and the NE-4 Advanced Fuels Campaign (AFC) have jointly undertaken efforts to qualify post-irradiation examination (PIE) data for Experimental Breeder Reactor II (EBR-II) and Fast Flux Test Facility (FFTF) metallic fuels. Among the thousands of PIE datasets in the Metallic Fuels Irradiation & Physics Database (FIPD), data from selected EBR-II experiments have been prioritized as the most valuable to the U.S. nuclear industry for licensing activities related to metallic fuel-based advanced fast reactors. This report summarizes the general data qualification progress of the EBR-II and FFTF experiments, with focus on the selected EBR-II experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Serpent - Bison - THM Preliminary Multiphysics Modeling of a Nuclear Thermal Propulsion System Fuel Assembly

This work demonstrates the Monte Carlo neutronic and Thermo-Hydraulic coupling scheme using the Serpent code and MOOSE application Bison and Thermo Hydraulic Module. The coupling scheme is then applied to the reference BWX Technologies Nuclear Ther- mal Propulsion system at he fuel assembly level where it’s used to perform an analysis of the isothermal material coefficients and potential material reactivity worth. A method is developed to isolate which feedback effects should be considered for proceeding with reduced order deterministic neutronic modeling where branch off analysis must be con- ducted. The convergence behavior of the coupling scheme is demonstrated where it fol- lows the standard Picard iteration approach. Verification studies for the method of deduc- ing relevant feedback effects is also demonstrated.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Gas-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model Updates

This work presents the latest improvements to, and investigations performed with, the pebble-bed high-temperature gas-cooled reactor (PB-HTGR) reference plant models for the United States Nuclear Regulatory Commission. These models serve as the foundation for the future development of detailed design evaluation models based on license applications. The reference plant models have been developed with the Comprehensive Reactor Analysis Bundle, or BlueCRAB, which is the code suite proposed for non-light-water reactor systems safety analysis. It incorporates various simulation tools developed by the Nuclear Energy Advanced Modeling and Simulation program, including the Griffin code for reactor physics, the Pronghorn and SAM codes for core thermal fluids, the BISON code for solid conduction and fuel performance, and the SAM code for system analysis. The primary objective of the work that was performed was to assess BlueCRAB’s level of readiness for modeling a PB-HTGR. To do so, we first developed numerical models in BlueCRAB that include the key physics for this technology to ensure an adequate level of fidelity for modeling PB-HTGR core performance and for performing multiphysics simulations for equilibrium core conditions and different accident scenarios. Then we simulated transient scenarios, including depressurized and pressurized loss of forced cooling accidents, over-cooling, and control rod withdrawal events with delayed and prompt supercritical reactivity insertions. The analysis in this report includes comparisons of the 2D thermal fluid porous media models in Pronghorn and SAM, and comparisons of coupled SAM/Griffin/SAM and coupled Pronghorn/Griffin for depressurized and pressurized loss of forced cooling, over-cooling, and control rod withdrawal events. In addition, we compare 3D, 2D, and 0D/PKE neutronic models for the two control rod withdrawal scenarios with coupled Pronghorn/Griffin. The comparisons show that the BlueCRAB models lead to physically intuitive solutions for the scenarios examined. The changes in the various scalar and vector fields, such as neutron flux, power, temperature, density, pressure, and velocity, are within the expected ranges, and their distributions can be explained by the system response of the transients and the geometric and material variations. Several comparisons suggest that the porous media models in Pronghorn and SAM can lead to similar solutions, even though they are based on different methodologies. This work further highlights the need for flexible tools with various levels of fidelity to cover the breadth and depth of needs that may arise in future technical evaluations of the PB-HTGR. We believe that the BlueCRAB capabilities will be a significant asset for confirmatory analyses in order to resolve important safety questions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Commercial integration of advanced nuclear energy with Artificial Intelligence (AI): Possible implications

The integration of advanced nuclear technologies (both fission and fusion) with artificial intelligence (AI) presents unprecedented national security challenges and opportunities. As fusion energy approaches commercial viability alongside advanced Small Modular Reactors (SMRs), their integration with AI and Artificial General Intelligence (AGI) systems could fundamentally transform the global energy and AI landscapes — two pillars of national security. This document briefly examines how AI could accelerate nuclear energy development and deployment while altering existing power structures, a lot could be done to deepen the discussions. Simultaneously, it observes how nuclear-powered AI may expedite advances toward AGI and beyond. These issues are deeply interconnected and thus need to be examined as a whole and more comprehensively than what’s being summarized here. For instance, AI-powered autonomous operation of nuclear facilities could reduce human error but introduce new cybersecurity vulnerabilities and uncertainties. Further investigation would also address how AI-enhanced nuclear technologies might complicate proliferation concerns through advanced fuel cycle management, nuclear materials production and safeguard. The strategic advantage gained by first entities achieving successful AI-nuclear integration could reshape global and national security framework. Timely analysis of these implications may be crucial for policymakers seeking to harness these technologies' benefits while effectively mitigating their potential risks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

A Multiphysics Evaluation of Annular Uranium-Zirconium Metallic Fuels [Poster]

This study examines the performance of U-10Zr annular metallic fuel rodlets which were experimentally evaluated as part of the Advanced Fuels Campaign (AFC). The AFC mission is to develop novel fuel technologies and facilitate the implementation of those technologies by industry partners. A key objective is to improve steady-state and transient performance over current fuel types. The experiments of interest in this study included annular metallic U-Zr fuel rodlets within HT-9 cladding which were placed in SS-316 capsules and inserted in the Advanced Test Reactor (ATR). Certain mechanical and thermal conditions cannot be directly evaluated through experiments and fuel performance modeling is used to shed light on this evolution over time. In this study, BISON Multiphysics simulations are leveraged to investigate the state of the fuel system throughout and after the experimental conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Overview of the Neutron Radiography Reactor (NRAD) for Neutron Imaging and In-Core Experiment Capabilities at Idaho National Laboratory

NRAD is a 250-kilowatt TRIGA research reactor that first went online at INL in 1977. (TRIGA stands for Training, Research, Isotopes, General Atomics.) Historically, NRAD was utilized as a neutron radiography reactor that provides comprehensive, non-destructive information about the internal condition of irradiated nuclear fuel. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography at NRAD. In recent years, NRAD has gone through a transformation from the single purpose radiography reactor for which it was designed into a multipurpose research reactor, and expanding its in-core irradiation capabilities to support a broader mission for the US Department of Energy (DOE) Nuclear Energy (NE) programs, Basic Energy Science (BES) Programs, as well as Fusion Energy programs. NRAD is a designated user facility under the DOE Nuclear Science User Facility (NSUF) program, and is available for access for general public via a competitive proposal process. More information about NSUF and NRAD are available from the website: https://nsuf.inl.gov/Home/Facility/654.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Spent nuclear fuel receipt rate analysis within an integrated waste management system (IWMS) architecture that includes consolidated storage

A key parameter in analyzing the performance of an integrated waste management system (IWMS) architecture for the disposition of spent nuclear fuel (SNF) is the SNF receipt rate from reactor and other custodian sites. Receipt rate in this paper means how much SNF is accepted per year for transport in the IWMS from such sites. Introducing one or more federal consolidated interim storage facilities (CISFs) into the IWMS architecture can potentially accelerate the receipt rate profile over time relative to system architectures without a CISF. This raises the question of what an optimal SNF receipt rate profile for an IWMS architecture might be in view of practical constraints and desired system performance attributes and associated metrics. This paper describes a sensitivity study on SNF receipt rates and the associated results for a selected set of IWMS scenarios aimed at informing near-term planning for interim storage capabilities and transportation assets. Two different strategies for CISF operation while awaiting availability of a disposal system to receive SNF are compared: one that relatively quickly fills an initial CISF and then idles the transportation system; and another that aims for more continuous use of transportation assets and receipt capabilities at the CISF. This study examines cost considerations and other factors, such as the timing of clearing reactor sites of SNF, efficient use of capital assets, and some other metrics that might be important to a CISF host community. Based on the analysis, an initial approach is presented that targets a continuous receipt strategy while maintaining the flexibility to step up receipt capabilities to a reasonable degree when needed and beneficial, within overall system constraints.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Reproducible benchmark for the SNAP 8 experimental reactor at operating conditions

This work presents fully reproducible multiphysics benchmark models of the Systems for Nuclear Auxiliary Power (SNAP) 8 Experimental Reactor at operating conditions with coolant flow. Wet experiment (with coolant, at power) validation benchmarks are presented using both deterministic (Serpent-Griffin) and Monte-Carlo (OpenMC-Cardinal) multiphysics frameworks coupled with thermal-hydraulic solvers in MOOSE. Reactivity coefficient measurements including fuel temperature, isothermal temperature, and power coefficients show good agreement with experiments, with discrepancies within experimental uncertainty. Reactivity worth experiments for coolant, samarium, and xenon poisoning are reproduced with differences under 200 pcm. Comparison between Serpent-Griffin and OpenMC-Cardinal frameworks reveal multiphysics coupling introduces positive reactivity effects (100-200 pcm) compared to uniform temperature and density fields at nominal operating conditions. Comparison between Serpent-Griffin and reference Serpent solution shows that power distributions maintain consistent radial and axial peaking behavior. All models, assumptions, thermophysical and thermomechanical properties, and material definitions are thoroughly documented with cited references; model inputs and model generating scripts are stored in the snapReactors GitHub repository.

SNAP↗
DOE OSTI2025

REDOTHERM: a thermodynamic modeling framework for redox-based thermochemical processes

Two-step thermochemical redox cycles are being developed as a potential pathway for the production of hydrogen and syngas. While there are many possible reactor and system configurations, moving oxide systems are considered promising in terms of the redox thermodynamics, due to the potential implementation of a countercurrent system that can achieve higher performance compared to other configurations. There is a lack of a robust thermodynamic modeling framework in the field, with multiple models incorporating incorrect thermodynamic assumptions that violate the second law of thermodynamics. We present in this work REDOTHERM, an open-source system model for moving oxides that incorporates the correct thermodynamic limits, as well as various options for the system auxiliary units including product separation, heat recovery, and oxygen removal. The model is agnostic to the energy source, and could be used for solar thermal or other configurations. We highlight the uses of this model, presenting some of the tradeoffs and challenges in redox-active material selection and how they affect the entire thermochemical hydrogen production process. This model could be easily adapted and used for material exploration, system/reactor design, and technoeconomic analysis.

08 HYDROGEN↗
DOE OSTI2025

Investigation of a U(IV)/U(III) Thermodynamic Reference Electrode for High-Temperature Molten Fluoride Salts

While thermodynamic reference electrodes with known and stable potentials are common in traditional aqueous systems, the high temperature and corrosive environment of a molten fluoride salt makes achieving long term stability with a thermodynamic reference electrode challenging, especially at temperatures of 600°C or higher. In this work, a thermodynamic reference electrode consisting of U(IV)/U(III) in a boron nitride compartment was evaluated for use in FLiBe at temperatures ≥ 600°C. FLiBe used in the study was purified by AlphaTech's proprietary process and characterized by ICP-MS and square wave voltammetry. The free oxide concentration was <2 ppm. Using the purified FLiBe, the U(IV)/U(III) thermodynamic reference electrode was shown to provide a stable, well-defined, and reproducible potential for more than 600+ hours of use in different tests. Moreover, the thermodynamic reference electrode showed a consistent potential with no signs of failure, even after being cooled between tests and then reheated for reuse. Thus, the U(IV)/U(III) reference electrode is suitable for use in rigorous electrochemical studies in molten fluoride salts. As a result, it may be useful as a common standard, facilitating the advancement of nuclear applications such as isotope separation or online monitoring of reactor systems through improved certainty in the measurement of thermodynamic potentials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Supported-Single Nickel Atom Catalysts for the Methanation of Carbon Dioxide

Synthesis of twenty-seven bimetallic catalysts consisting of nickel and one of nine different dopants (B, Co, Cu, Fe, Mg, Mn, Sn, V, and Zn) supported on three different metal oxides (Al 2 O 3 , CeO 2 , and SiO 2 ) is carried out via organometallic grafting. The catalysts are evaluated for their activity and selectivity for the CO 2 methanation reaction at a feed ratio of H 2 /CO 2 of 4 at 300 °C in a high-throughput flow reactor system. After in situ pre-activation (500 °C in H 2 ), Ni/Co/CeO 2 exhibited high conversion (84.3%) and selectivity for methane (99.6%). Ni/Co/CeO 2 was characterized by high-resolution transmission electron microscopy (HRTEM), X-ray photoelectron spectroscopy (XPS), X-ray diffraction, H 2 -temperature-programmed reduction (H 2 -TPR), and CO 2 -temperature-programmed desorption (CO 2 -TPD). HRTEM showed the presence of single Ni and Co atoms on ceria after pre-reduction at 500 °C and after the methanation reaction at 300 °C for 15 h. XPS determined that the strong interaction between Ni, Co, and ceria increased after the reduction, leading to a charge transfer between Ni and Ce that created oxygen vacancies in ceria. Nickel was found to be Ni 2+ in the as-prepared material and was partially reduced in the presence of cobalt and after the activation in H 2 at 500 °C. The DFT results show that both nickel and cerium exhibit lower Bader charges in the Ni/Co/CeO 2 system, confirming that the presence of cobalt enhances the reduction of both Ni and Ce through electronic interactions. This indicates that single cationic Ni atoms are highly effective for the methanation reaction. In conclusion, the organometallic grafting technique is found to be efficient for synthesizing catalysts with highly homogeneous dispersed species at low metal loadings (0.16 wt % Ni–0.15 wt % Co), which leads to high turnover frequency (up to 248.7 h –1 ) and durability for methanation.

CO2 conversion↗
DOE OSTI2025

Fabrication and Characterization of MiniFuel Specimen

Burnable absorbers (BAs) are used in UO 2 fuels to tune the reactivity in a light water reactor (LWR) particularly at the beginning of life of the fuel. BAs are elements that have high neutron absorption cross sections and because of this property can prevent power peaking. The addition of BAs reduces the reactivity at the beginning of life, and this in turn allows for the increase of enrichments from > 5 % 235 U up to ~ 10 %. This also enables the increase of the burnup limit to 75 GWd/MTU and the cycle length to 24 months. The BA are eventually consumed under irradiation, and their concentration is diminished at roughly 10-20 GWd/MTU. The objective of this report is to fabricate UO 2 fuels that have been doped with BAs (Gd 2 O 3 in this case) and subsequently irradiate them in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. The test specimen that are fabricated include 4 and 8 weight % Gd 2 O 3 doped UO 2 to investigate the influence of the BAs on the thermophysical properties and microstructure of the fresh and irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Spent Nuclear Fuel Receipt Rate Analysis within an Integrated Waste Manage-ment System Architecture that Includes Consolidated Interim Storage

A key parameter in analyzing the performance of an integrated waste management system (IWMS) architecture for the disposition of spent nuclear fuel (SNF) is the SNF receipt rate from reactor and other custodian sites. The introduction of one or more federal consolidated interim storage facilities (CISFs) into the IWMS architecture can enable the receipt rate profile as a function of time to be accelerated relative to system architectures without a CISF. The question then arises as to what an optimal SNF receipt rate profile for an IWMS architecture might be in view of practical constraints and desired system performance attributes and associated metrics. This paper describes a sensitivity study on SNF receipt rates and the associated results for a selected set of IWMS scenarios aimed an in-forming near-term planning for interim storage capabilities and transportation assets. Two different strategies are compared, one that fills an initial CISF quickly and then idles the transportation system while a disposal system is prepared, and a second strategy that aims to provide a more continuous use of transportation assets and receipt capabilities at the IWMS while the disposal system is readied for SNF receipt. Cost considerations and other factors such as impact on timing of clearing reactor sites of SNF, efficient use of capital assets, and other metrics, including those which may be important to a CISF host community, are examined. Based on the analysis, an initial approach is presented targeting a continuous receipt strategy while having the flexibility to step up receipt capabilities to a reasonable degree when needed and beneficial within overall system constraints.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗