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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 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Neutron imaging of hydrogenous gases in porous-carbon materials around phase transitions

Focus herein is on hydrogenous gases interacting with porous-carbon materials on activated biochar at pressures ranging from 0 to 100 bar and temperatures from room (ambient) to cryogenic. Neutron radiography was conducted using a custom-made cell system to simultaneously image pure gas and gas interacting with the porous carbon. From neutron attenuation, the cross sections of methane and ethane and number concentrations were determined for different pressures and temperatures and compared with corresponding reference data. Sorption properties of the porous materials were investigated in situ with methane, ethane, and propane.

Ossler, Frederik [Lund University, Sweden]↗
DOE OSTI2026

Investigation of irradiation damage and heat deposition: a comparative analysis for HEU-to-LEU conversion in HFIR

The planned conversion of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel requires detailed evaluation of experiment-relevant parameters to ensure continued performance for materials testing and isotope production. Here, this study presents the first comprehensive assessment of displacements per atom (dpa) and heat deposition rates in target materials within the HFIR flux trap with both HEU and candidate LEU core configurations. Seven analyses were conducted to evaluate key performance metrics, including fast neutron flux distribution, cross section response functions, cross section data, and local dpa and heat deposition rates using mesh- and cell-based tallies. Simulations employed Shift, Monte Carlo N-Particle (MCNP), and the HIFR Controller (HFIRCON) tool suite for high-fidelity transport and depletion modeling. The LEU designs—using U 3 Si 2 -Al dispersion fuel and operating at 95 MW—were compared to the current 85 MW HEU configuration. Results show that while the candidate LEU cores exhibit higher dpa rates due to a harder spectrum and extended cycle lengths, they also demonstrate reduced heat deposition rates in irradiation experiments, primarily due to increased gamma self-shielding from higher 238 U content in the core. These findings confirm that LEU conversion can maintain HFIR’s materials irradiation capabilities but may require redesigning existing experimental hardware.

HEU↗
DOE OSTI2023

Modeling of fission gas diffusion and release for Gd 2 O 3 doped UO 2

Uranium dioxide (UO 2 ) is the primary nuclear fuel in light water reactors, and its excess neutronic reactivity can be controlled by adding burnable absorbers, such as Gd 2 O 3 . This burnable absorber has a large neutron absorption cross-section, lowering the high reactivity of the reactor's initial fuel load. However, there needs to be more understanding of how added Gd 2 O 3 influences the properties of UO 2 under irradiation. To understand the behavior of defects and fission gas in the UO 2 /Gd 2 O 3 system under irradiation, we use cluster dynamics modeling supported by density functional theory calculations. First, we calculate the formation energies of Gd point and cluster defects, and evaluate the temperature-dependent defect concentrations using the defect formation energies and entropies. We show that Gd is soluble in UO 2 , introducing a negative charge in the system. Using this information, we adapted the cluster dynamics code Centipede to model the influence of Gd on U self-diffusion and Xe diffusion in UO 2 with 10 wt% Gd 2 O 3 . Also, we analyzed the Xe diffusion as a function of Gd 2 O 3 concentration, showing that the Xe diffusivity is decreased, which means that the athermal diffusivity due to electronic stopping persists at higher temperatures. In conclusion, the decrease in Xe diffusion means that more Xe stays in the matrix, decreasing the Xe release, and lowering its influence of fission gas release on the thermomechanical properties of UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Structure evolution and tin redistribution during oxidation of Zircaloy-4 at 500°C

Zirconium (Zr) alloys are widely used as fuel cladding in nuclear power reactors due to their thermal stability, mechanical durability, corrosion resistance, and low neutron absorption cross-section. However, their performance is challenged by oxidation in reactor environments, making the study of Zr alloy corrosion behavior crucial for ensuring the safety, longevity, and economic viability of nuclear power systems. While the oxidation behavior of Zr-based cladding materials has been extensively studied since the 1950s, a mechanistic understanding into the relationship between structure evolution, solute element redistribution, and properties remains elusive. Valuable insights may be obtained through advanced experimental methods, such as in-situ and high resolution microscopy techniques. Here, in this study, the oxidation behavior of Zircaloy-4 at 500 °C in O 2 is characterized using a multimodal advanced characterization approach. Using in-situ X-ray diffraction, the phase evolution from metastable to stable oxides is tracked in real time. Complementary high-resolution techniques, including electron microscopy and atom probe tomography, reveal nanoscale insights into the microstructural changes and solute redistribution across the oxide/metal interface. Nanohardness mapping across the oxide/metal interface highlights localized mechanical property variations that may be linked to changes in microstructure and crystal structure within the oxide layer. These findings offer valuable insights into the microstructure and property evolution of Zircaloy-4 during oxidation, contributing to a better understanding of microstructural changes in Zr-based alloys under oxidative environments.

APT↗
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

Multi-phase equation of state of ultrapure hafnium to 120 GPa

Hafnium (Hf) is an industrially important material due to its large neutron absorption cross-section and its high corrosion resistance. When subjected to high pressure, Hf phase transforms from its hexagonal close packed α-Hf phase to the hexagonal ω-Hf phase. Upon further compression, ω-Hf phase transforms to the body centered cubic β-Hf phase. In this study, the high pressure phase transformations of Hf are studied by compressing and decompressing a well-characterized Hf sample in diamond anvil cells up to 120 GPa while collecting x-ray diffraction data. The phase transformations of Hf were compared in both a He pressure transmitting medium (PTM) and no PTM over several experiments. It was found that the α-Hf to ω-Hf phase transition occurs at a higher pressure during compression and lower pressure during decompression with a helium (He) PTM compared to using no PTM. Finally, there was little difference in the ω-Hf to β-Hf phase transition pressure between the He PTM and no PTM. The equation of state was fit for all three phases of Hf and under both PTM and no-PTM.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2020

Large-$N_c$ analysis of magnetic and axial two-nucleon currents in pionless effective field theory

We analyze magnetic and axial two-nucleon contact terms in a combined large-$N_c$ and pionless effective field theory expansion. These terms play important roles in correctly describing, e.g., the low-energy cross section of radiative neutron capture and the deuteron magnetic moment. In this work we show that the large-$N_c$ expansion hints towards a hierarchy between the two leading-order magnetic terms that matches that found in phenomenological fits. We also comment on the issue of naturalness in different Lagrangian bases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Measurement of neutron-proton capture in the $\mathrmSNO+ water phase

The SNO+ experiment collected data as a low-threshold water Cherenkov detector from September 2017 to July 2019. Measurements of the 2.2-MeV γ 's produced by neutron capture on hydrogen were made using an Am-Be calibration source, for which a large fraction of emitted neutrons are produced simultaneously with a 4.4-MeV γ. Analysis of the delayed coincidence between the 4.4-MeV γ and the 2.2-MeV capture γ revealed a neutron detection efficiency that is centered around 50% and varies at the level of 1% across the inner region of the detector, which to our knowledge is the highest efficiency achieved among pure water Cherenkov detectors. Finally, in addition, the neutron capture time constant was measured and converted to a thermal neutron-proton capture cross section of $336.3^{+1.2}_{-1.5}$ mb .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Inertial-confinement fusion-plasma-based cross-calibration of the deuterium-tritium γ-to-neutron branching ratio

The deuterium-tritium (D-T) γ -to-neutron branching ratio [ H 3 ( d , γ ) He 5 / H 3 ( d , n ) He 4 ] has been determined previously under inertial-confinement fusion (ICF) conditions and in beam-target based experiments. In the former case, neutron-induced backgrounds are mitigated compared to the latter due to the short pulse nature of ICF implosions and the use of gas Cherenkov γ -ray detectors. Furthermore, an added benefit of ICF based measurements is the ability to achieve lower center-of-mass energies as compared to accelerators. Previous ICF based experiments however report a large uncertainty in the D-T γ -to-neutron branching ratio of ≈ 48 % , which arises from the necessity of an absolute detector calibration and/or a cross-calibration against the D- He 3 γ -to-proton branching ratio. A more precise value for the branching ratio based on data taken at the OMEGA laser facility is reported here, which relies on a cross-calibration against the better known C 12 neutron inelastic scattering cross section. A D-T branching ratio value of ( 4.6 ± 0.6 ) × 10 - 5 is determined by this method.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Data Evaluation of Actinide Cross Sections: 230 Th and 231 Th

A new evaluation of the ENDL cross section set for Thorium (Z=90) is developed using the TALYS statistical model cross section code. The primary goal of this effort is to produce an evaluation that attempts to match as closely as possible fission cross sections developed through surrogate reaction techniques on actinide targets 230 Th and 231 Th. This evaluation effort and the processing needed to render its results into data libraries is a necessary step in making the efforts of nuclear experimentalists useful to the broad community of researchers engaged in simulations of nuclear fusion for basic and applied science. Another aspect, verification and validation against various AGEX experiments, is also presented. The end-product is an updated library that includes the latest measurements of fission cross sections for comparison against those measured via traditional techniques. All the steps in the evaluation, processing, validation and verification, and library release are described in the following sections. For completeness, the appendix contains all the parameters used in the TALYS cross section evaluation for neutrons incident on 230 Th and 231 Th.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Statistical Hauser-Feshbach model calculation for (α, n) reaction cross section and discrete level population

We calculate the α-particle induced reactions on 17,18 O, 19 F, and 23 Na, in the energy range 0 ≤ E α ≤ 10 MeV, with a particular attention to the branching ratios of populated discrete levels in the (α,n) channels. Since there are too many open channels to employ the R-matrix theory, we apply the statistical Hauser-Feshbach model to calculate these reaction cross sections and branching ratios. The branching ratio is defined as b n = $\frac{σ_n}{Σ_iσ_i + σ_c}$ , where σ n is the n-th level production cross section after the neutron emission, and σ c is the production of the continuum state. In the cases of our target nuclei and the energy range of interest, the residual nuclei of the neutron emission channel are always in their discrete states, so that σ c can be negligible.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2021

Fundamental Studies of Tritium Diffusivity in Pure and Defective Zircaloy-4 Getters (Tritium Science Program FY2020 Report)

Zirconium (Zr) alloys are extensively used as tritium (T) getters in nuclear reactors due to their low absorption cross section to thermal neutron, good mechanical properties at high temperature and resistant to corrosion in different environmental conditions. Zr alloys have better thermal properties than many other refractory alloys including stainless steel. The nuclear characteristic is simulated, and T is produced in tritium-producing burnable absorber rods (TPBAR) by the irradiation of neutron flux in pressurized water reactor (PWR). T thus produced diffuses through the pellets and is captured by the getter where it chemically reacts with Zr to form metal hydride (ZrTx). These hydrides are brittle and affect adversely the mechanical properties of alloys. In addition to that, mismatch in the lattice structures after hydride formation creates a stress which needs to be considered in component design and their life evaluation. Therefore, understanding the behavior of T and its species becomes significant as fuel burnup is increased, which leads to increase in hydrogen pickup and oxide formation. The transport of T and its species in pure Zr alloys, and in alloys with certain alloying elements and defects are important to understand in order to enhance the performance of the materials.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Fundamental Studies of Tritium Formation and Diffusivity in Pure and Defective Zircaloy-4 Getters

Zirconium (Zr) and its alloys are used as fuel cladding in nuclear reactors. These materials have their low absorption cross-section for thermal neutrons and have good mechanical and thermal properties. In addition, they are used as tritium getters in tritium-producing burnable absorber rods (TPBARs), in the form of nickel-plated Zircaloy-4 tubes. To produce 3 H, the breading blankets such as lithium aluminate are irradiated in pressurized water. 3 H produced in this process is captured by the getter and 3 H reacts chemically with Zr to form metal hydride (ZrTx). The precipitation of metal hydride results in not only the volume increase and hydrogen embrittlement of zircaloy but also changes in getter’s chemical properties. This degrades the getter’s performance. Therefore, understanding the behavior of 3 H (such as diffusion and solubility) in zircaloy is important to understand getter performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Baseline CGMF Calculations for Pu Isotopes

Updates were made to CGMF to run neutron-induced fission calculations on Pu isotopes of A =238, 239, 240, 241, and 242 (the release version of CGMF can calculate 239,241 Pu(n,f) and 238,240,242 Pu(sf) only). Here, we show the baseline calculations from CGMF for these neutron-induced fission reactions, including updates made for total kinetic energy systematics and fission barriers fit to be more consistent with the neutron-induced fission cross section.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

First-Principles Studies of Tritium Species Dissociability & Diffusivity Across the Interface of Nickel-Plated Zircaloy-4

Zirconium (Zr) and its alloys (Zircaloy-4) are widely used in nuclear reactors due to their low neutron adsorption cross-section and excellent corrosion resistance. In tritium-producing burnable absorber rods (TPBARs), the metal getter tube located between the cladding and the γ-LiAlO 2 pellets is composed of nickel (Ni)-plated Zircaloy-4, which is used to capture tritium ( 3 H) species (mainly 3 H 2 and 3 H 2 O) generated from γ-LiAlO 2 pellets during irradiation. The 3 H-related products transfer to the surface of metal Ni upon adsorption and dissociation to form new 3 H species and diffuse into the Zircaloy-4 getters to form metal hydrides (Zr 3 H x ). Therefore, exploring 3 H species ( 3 H 2 , 3 H 2 O) dissociation on the surface of Ni and diffusion across the interface of Ni-plated Zircaloy-4 getters can provide a better understanding of 3H species formation and transport from pellets into the getters.

36 MATERIALS SCIENCE↗
DOE OSTI2024

First-Principles Studies of Tritium Species Diffusivity Across the Interface of Nickel-Plated Zircaloy-4

α-Zr and its alloys are known as best 3 H getters due to their excellent corrosion resistance under chemically corrosive environment, low thermal neutron absorption cross-section and robust mechanical strength at high temperature. In this work, we conducted a systematic study on constructing and optimizing Ni and Zr surfaces, interfacing the stable Ni and Zr surfaces to create an optimal Ni-Zr interface, and an understanding the diffusion mechanisms for 3 H across Ni(111)-Zr(0001) interface with and without vacancies and impurities such as Sn and O.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Summary of FY25 activities on the project Realistic Reaction Evaluations for Fission Products Off Stability (RREFPOS)

Applications such nonproliferation, post-detonation forensics, spent-fuel assay, reactor burnup and design, as well as astrophysics, rely on the accurate description of the neutron interaction with unstable fission products. Current neutron-induced cross-section descriptions for these nuclei are either non-existent or based on simplified assumptions, leading to unquantified impacts on predicted cross-sections. The goal of the Realistic Reaction Evaluations for Fission Products Off Stability (RREFPOS) project is to develop a reproducible methodology leading to a more predictive and realistic new set of evaluated nuclear data, improving the reliability of applications involving fission products off stability. The current report documents the activities related to the RREFPOS project in its first year, corresponding to the Fiscal Year of 2025 (FY25).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1977

Comments on galactic evolution and nucleocosmochronology

Long-lived nucleochronologies are calculated for several recently proposed models of the chemical evolution of the Galaxy. Special attention is paid to the Re-187/Os-187 chronometer, for which important data have recently become available. It is found that although the rate of star formation does vary with time in the different models, the quantity that is related to the effective net rate of nucleosynthesis is constant for most of the physically plausible recent-evolution models examined. This constant net rate implies that for these models, the age of the Galaxy at the time the solar system formed is twice the mean age of the stable elements. This mean age can be estimated by the parameter 'delta max', in which case the age of the Galaxy is twice this parameter plus the age of the solar system. The present uncertainties in 'delta max' yield an age for the Galaxy of 7 to 20 billion years. However, this range could be significantly reduced by an accurate measurement of the half-life of Re-187 and more knowledge on the effect of stellar temperatures on the Os-186/Os-187 neutron-capture cross-section ratio. In fact, experiments which could be carried out in the next few years can reduce these uncertainties tremendously and enable an age determination to be made which might severely restrict cosmological models.

Hainebach, K. L.↗