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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Modeling a generic TRISO-fueled heat pipe microreactor using SCALE: Depletion, transportation criticality, and shielding

This paper demonstrates the applicability of the SCALE code system to tristructural-isotropic (TRISO)-fueled heat pipe microreactors through depletion, transportation criticality, and shielding analyses of a generic reference design. The study conducted supports US Nuclear Regulatory Commission code readiness efforts for advanced non–light-water reactor technologies and is intended as a code capability demonstration rather than as an optimization of a specific microreactor design. The modeled reactor employs high-assay low-enriched uranium (HALEU) uranium oxycarbide (UCO) TRISO fuel and beryllium oxide (BeO) reflectors and operates at 7.5 MWth with a nominal lifetime of about 3 effective full power years. Representative cases for fresh and irradiated cores were selected to exercise SCALE methods relevant to reactor operation and post-irradiation transport. The discharged-core decay heat is approximately 6% of operating power immediately after shutdown. Transportation criticality calculations show that internal water ingress is the dominant reactivity effect, with fully flooded fresh core and irradiated core configurations remain above the subcriticality criterion, even with the available control mechanisms. Shielding calculations for a simplified transportation package indicate that normal-condition dose rates are governed mainly by shielding thickness and cooling time, whereas the breached hypothetical accident case is governed primarily by cooling time. Overall, the study shows that SCALE supports depletion, transportation criticality, and shielding evaluations efficiently for TRISO-fueled heat pipe microreactors within a single code system.

Criticality↗
DOE OSTI2022

Bioinspired synthesis of thermally stable and mechanically strong nanocomposite coatings

Abstract An innovative biomimetic method has been developed to synthesize layered nanocomposite coatings using silica and sugar-derived carbon to mimic the formation of a natural seashell structure. The layered nanocomposites are fabricated through alternate coatings of condensed silica and sugar. Sugar-derived carbon is a cost-effective material as well as environmentally friendly. Pyrolysis of sugar will form polycyclic aromatic carbon sheets, i.e., carbon black. The resulting final nanocomposite coatings can survive temperatures of more than 1150 °C and potentially up to 1650 °C. These coatings have strong mechanical properties, with hardness of more than 11 GPa and elastic modulus of 120 GPa, which are 80% greater than those of pure silica. The layered coatings have many applications, such as shielding in the form of mechanical barriers, body armor, and space debris shields. Graphical abstract

36 MATERIALS SCIENCE↗
DOE OSTI2023

Neutron Capture and Transmission Measurements and Evaluation of 54 Fe at the RPI LINAC [Slides]

This presentation covers neutron capture and transmission measurements and evaluation of 54 Fe (Iron) at the Rensselaer Polytechnic Insitute (RPI) Linear Accelerator (LINAC). Presentation conclusions find that 54 Fe is near its conclusion, covariance matrix generation is needed for both experiments before data is made available and released (Completion by April 2023). RPI will be conducting a full Resolved Resonance Region (RRR) evaluation following release of data (Completion by September 2023). Additionally, the neutron beam imager shows promise in providing an easy way to align samples in the most intense part of the neutron beam and there will be a new evaluation that will offer improvements in crit safety, shielding, and stellar applications.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

IAEA Photonuclear Data Library 2019

Here, photo-induced reaction cross section data are of importance for a variety of current or emerging applications, such as radiation shielding design and radiation transport analyses, calculations of absorbed dose in the human body during radiotherapy, physics and technology of fission reactors (influence of photo-reactions on neutron balance) and fusion reactors (plasma diagnostics and shielding), activation analyses, safeguards and inspection technologies, nuclear waste transmutation, medical isotope production and astrophysical applications. To address these data needs the IAEA Photonuclear Data library was produced in 1999, containing evaluated photo-induced cross sections and neutron spectra for 164 nuclides which were deemed relevant for the applications. Since the release of the IAEA Photonuclear Data Library however, new experimental data as well as new methods to assess the reliability of experimental cross sections have become available. Theoretical models and input parameters used to evaluate photo-induced reactions have improved significantly over the years. In addition, new measurements of partial photoneutron cross sections using mono-energetic photon beams and advanced neutron detection systems have been performed allowing for the validation of the evaluations and assessments of the experimental data. Furthermore, technological advances have led to the construction of new and more powerful gamma-beam facilities, therefore new data needs are emerging. We report our coordinated efforts to address these data needs and present the results of the new up-to-date evaluations included in the new updated IAEA Photonuclear Data Library consisting of 219 nuclides. The new library includes 188 new evaluations produced by the CRP evaluators, and one evaluation taken from the JENDL/PD-2016 library, while 20 evaluations were retained from the previous 1999 IAEA Photonuclear Data Library. In most of the cases, the photon energy goes up to 200 MeV. A total of 55 nuclides are new in this library reflecting the progress in measurements but also the developing data needs. In this paper we discuss the new assessment method and make recommendations to the user community in cases where the experimental data are discrepant and the assessments disagree. In addition, in the absence of experimental data, we present model predictions for photo-induced reaction cross section on nuclides of potential interest to medical radioisotope production.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Improvement and Verification of Online Cross Section Generation Capability of Griffin for TRISO-fueled Reactors

Griffin, a MOOSE-based reactor multiphysics code jointly developed by Idaho National Laboratory and Argonne National Laboratory under the DOE Office of Nuclear Energy’s NEAMS program, has pursued the development of an online multigroup cross section generation capability for a few years to enable high-fidelity, problem-dependent neutronics analyses of advanced thermal reactors. Recent advancements in Griffin’s online multigroup cross section generation capability have significantly improved the accuracy, robustness, and efficiency of self-shielding calculations for both prismatic and pebble-bed TRISO-fueled reactor applications. Key developments include a unified fuel self-shielding method applicable to both TRISO and annular compact/spherical shell fuel zone geometries; an advanced Dancoff Category-based Equivalence Theory using a bell function for non-fuel resonance treatment, achieving more than an order-of-magnitude speedup compared to the Tone method; an on-the-fly multigroup equivalence approach to mitigate group condensation errors; and a streaming correction method for pebble-bed homogenization. A proof-of-concept demonstration of on-the-fly group condensation with consistent P0 transport correction was also achieved. The method reproduced direct fine-group solutions with excellent accuracy (eigenvalue errors within 10 pcm and pin-power differences within 0.5%), but due to performance limitations of the current fixed-source solver, improvements to solver efficiency will be addressed in future work. Verification tests were performed on graphite-moderated TRISO-fueled two-dimensional core benchmark problems representing gas-cooled microreactors, heat pipe-cooled microreactors, gas-cooled pebble-bed reactors, and fluoride salt-cooled high-temperature reactors. Across all cases, Griffin showed excellent agreement with Serpent2 continuous energy Monte Carlo solutions: eigenvalue errors within 200 pcm, pin-power root-mean-square errors within 2%, and control rod and drum worth errors less than 2%. It should be noted that, for the benchmark problem, cross section generation contributed less than 3% of the total simulation times. These results demonstrate that Griffin’s online cross section generation capability delivers accurate and efficient reactor physics solutions across a wide spectrum of TRISO-fueled advanced reactor designs. With further improvements to the fine-group fixed-source solver and planned extensions to depletion, transients, and coupled neutron–gamma transport, Griffin will be well-positioned to become a powerful and comprehensive tool for advanced reactor analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Correlating electronic properties with M-site composition in solid solution Ti y Nb 2- y CT x MXenes

High electrical conductivity is desired in MXene films for applications such as electromagnetic interference shielding, antennas, and electrodes for electrochemical energy storage and conversion applications. Due to the acid etching-based synthesis method, it is challenging to deconvolute the relative importance that factors such as chemical composition and flake size contribute to resistivity. To understand the intrinsic and extrinsic contributions to the macroscopic electronic transport properties, a systematic study controlling compositional and structural parameters was conducted with eight solid solutions in the Ti y Nb 2-y CT x system. Here, in particular, we investigated the different roles played by metal (M)-site composition, flake size, and d-spacing on macroscopic transport. Hard x-ray photoemission spectroscopy and spectroscopic ellipsometry revealed changes to electronic structure induced by the M-site alloying. Consistent with the spectroscopic results, the low- and room-temperature conductivities and effective carrier mobility are correlated with the Ti content, while the impact of flake size and d-spacing is most prominent in low-temperature transport. The results provide guidance for designing and engineering MXenes with a wide range of conductivities.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Surrogate modeling of Monte Carlo radiation transport with convolutional neural networks for shielding optimization

Here, we present a machine learning (ML)-based surrogate model using convolutional neural networks (CNN) designed to emulate the attenuation of neutron fields as they pass through various shielding materials. This model can compute the outgoing neutron flux almost instantaneously and achieves reasonable accuracy compared to traditional Monte Carlo (MC)-based codes, which are computationally intensive. This emulator alleviates the complexity of neutron radiation transport through shielding materials by reducing the dimensionality and enables shielding optimization for a known radiation environment. This optimization process, which would have taken an unrealistic timeline due to several complex radiation transport simulations, can now be achieved in minutes, thus increasing computational capabilities in radiation shielding assessment. We demonstrate the applications of this emulator in computing effective dose rates and optimizing shielding solutions for a heavy-ion accelerator facility, such as the Facility for Rare Isotope Beams, where secondary neutrons produced via beam interactions dominate the radiation environment.

accelerator shielding↗
DOE OSTI2023

Thermal Scattering Law Data Development for Paraffin Wax

Paraffin wax is often used as a nuclear moderator to slow down the fast neutrons in experimental critical assemblies [1]. It is a colorless and soft solid material that consists primarily of straight-chain alkanes (n-alkanes), which are hydrocarbons with the general formula CnH2n+2 [2-3]. The length of the hydrocarbon chain ranges from C20 to C30 and higher [2]. It is distinguished by its solid state at room temperature and begins to melt above approximately 310 K [4]. Paraffin wax is a commonly employed substance in the manufacture of shielding. One of its noteworthy characteristics is its ability to effectively absorb the neutrons. Also, it possesses a high macroscopic cross section, which enables it to efficiently moderate neutrons. As a result, paraffin wax is extensively utilized in various applications where moderation and shielding of neutrons are needed. For simulations, it is necessary to evaluate its thermal scattering law (TSL) and cross sections. Computationally, classical molecular dynamics (CMD) simulations provide the capability of simulating atomic details. For example, several unary, binary, and few multi component mixtures have been investigated of the paraffin model by using molecular dynamics simulations [5-12]. An assessment of thermal neutron scattering in a heavy paraffinic oil treated both as a solid and a viscous fluid containing 25% linear branched paraffin (C30H62), 35% one ring cycloalkane (C30H60), 15% two rings cycloalkane (C30H58), and 25% aromatic (C30H60) chains has been studied using CMD simulations for producing TSL data [13]. Nevertheless, there is lack of TSL and cross section data for paraffin wax as most of the reported analyses focus on the unary and binary mixture of n-alkanes, which is not consistent with actual paraffin wax [2]. In this work, we applied the equilibrium CMD simulations technique to explore the structure and dynamical properties of wax, which are fundamental input to calculate the TSL. A paraffin wax system was modeled using the CMD code LAMMPS (Large-scale Atomic/Molecular Massively Parallel Simulator) [14-15] with the semi-empirical COMPASS [16] force field. The density of state (DOS) was calculated from the normalized velocity autocorrelation function (VACF), which is the Fourier transform of the normalized VACF. The DOS was used for the calculation of the TSL and thermal scattering cross sections. The paraffin wax atomic system was constructed by using the MedeA material design platform [17], and was benchmarked using available properties (i.e., density, bond lengths, angles, diffusivity, and viscosity).

Nuclear Criticality Safety Program (NCSP)↗
DOE OSTI2024

The FLUKA code: Overview and new developments

The FLUKA Monte Carlo Radiation Transport and Interaction code package is widely used to simulate the interaction of particles with matter in a variety of fields, including high energy physics, space radiation, medical applications, radiation protection and shielding assessments, accelerator studies, astrophysical studies and well logging. This paper gives a brief overview of the FLUKA program and describes recent developments, in particular, improvements in the modelling of particle interactions and transport are described in detail. In addition, an overview of selected applications is given.

Ballarini, Francesca↗
DOE OSTI2022

Polyimide aerogels for ballistic impact protection

The ballistic performance of edge-clamped monolithic polyimide aerogel blocks (12 mm thickness) has been studied through a series of impact tests using a helium-filled gas gun connected to a vacuum chamber and a spherical steel projectile (approximately 3 mm diameter) with an impact velocity range of 150–1300 m s -1 . The aerogels had an average bulk density of 0.17 g cm -3 with high porosity of approximately 88%. The ballistic limit velocity of the aerogels was estimated to be in the range of 175–179 m s -1 . Moreover, the aerogels showed a robust ballistic energy absorption performance (e.g., at the impact velocity of 1283 m s -1 at least 18% of the impact energy was absorbed). At low impact velocities, the aerogels failed by ductile hole enlargement followed by a tensile failure. By contrast, at high impact velocities, the aerogels failed through an adiabatic shearing process. Given the substantially robust ballistic performance, the polyimide aerogels have a potential to combat multiple constraints such as cost, weight, and volume restrictions in aeronautical and aerospace applications with high blast resistance and ballistic performance requirements such as in stuffed Whipple shields for orbital debris containment application.

42 ENGINEERING↗
DOE OSTI2024

A General Model for the Ductility of Intermetallics Applied to Fe-Co Alloys

The mechanical properties, specifically ductility, of high performing soft magnets such as Fe-Co alloys are a limiting factor to their broader use in a number of systems. The understanding of the mechanical robustness in these materials is currently insufficient to be able to support the growing interest in applications such as magnetic shielding or electric motors. Fe-Co alloys provide the highest commercially available magnetic saturation and high magnetic permeability but have very poor ductility. The addition of vanadium to these alloys has allowed for significant commercialization and some ductility improvements, but the fundamental reasons for the observed improvements are not well understood. In most published work on mechanical properties in these alloys, the precise chemistry of the alloys investigated, often a critical aspect of intermetallics, is not reported or controlled and thermal history is unclear. This work creates a ductility model that is sensitive to changes in chemistry and can predict relative strain to failure as well as brittle fracture mode for intermetallics and is applied to Fe-Co alloys. Through the application of density functional theory (DFT), this model identifies defect stabilities, anti-phase boundary (APB) energies, the energy necessary to cross-slip, and cleavage energies and combines them through energetic competition to determine a relative failure strain. The model correctly predicts ductility improvements with the addition of vanadium as well as the transition from intergranular to transgranular cleavage, though more precise experiments are necessary to appropriately validate the various improvements observed.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Self-consistent modeling of tokamak edge plasma transport with lithium sources

Magnetic confinement fusion devices require effective heat and particle exhaust solutions on the divertor plates to operate sustainably, especially under reactor-relevant conditions. Liquid lithium divertors have been proposed to address two major challenges: control of excessive heat flux to plasma-facing components through vapor shielding and minimization of core plasma contamination from impurities. The National Spherical Torus Experiment-Upgrade (NSTX-U) will explore lithium as a divertor material due to its potential to meet both objectives. We present a self-consistent coupling framework between the plasma boundary transport code UEDGE and the lithium wall transport code Wall–Li to evaluate the feasibility and operational limits of lithium-based divertors. The model aims to optimize lithium sourcing levels to prevent core plasma contamination via fuel dilution while ensuring divertor protection through vapor shielding. This integrated framework, applicable to any tokamak with lithium sources, dynamically adjusts lithium sourcing based on local plasma conditions and surface temperature. The coupled model is tested using NSTX-like geometry and plasma conditions to assess its performance and reliability. Wall–Li calculates lithium fluxes from plasma-facing components, incorporating physical sputtering, thermally enhanced sputtering, and evaporation driven by surface temperature and ion flux. These fluxes are reintroduced into UEDGE as neutral lithium atoms, enabling simulation of their transport and distribution within the plasma. UEDGE computes plasma and neutral transport, surface heat flux, and iteratively feeds this information back to Wall–Li. A small time step is employed to ensure numerical stability and convergence, enabling accurate simulations over typical tokamak discharge durations. This integrated modeling approach provides a robust tool for identifying operational regimes that balance effective lithium sourcing with minimal core plasma contamination, offering critical insights for optimizing lithium-based divertor systems in current and future fusion devices.

Magnetic confinement fusion↗
DOE OSTI2021

Advances in Processing, Manufacturing, and Applications of Magnetic Materials

Magnetic materials are increasingly important for many green energy technologies. Probably, the best known of these are permanent magnets. They are used to supply a magnetic field and are widely used in actuators, motors, generators, data storage, and sensors. These are “hard magnets,” meaning that they retain a large permanent magnetization (difficult to be demagnetized), which is what makes them so useful in motors where an opposing magnetic field can be used to push against them. A coercivity (resistance to demagnetization) of ~400 Oe (32,000 A/m) or more is typically the threshold value for a hard magnet and is typical of isotropic sintered ferrite (SrFe 12 O 19 ) such as found in many “refrigerator magnets.” Comparably important soft magnets are useful for directing magnetic fields and can be very easily (de)magnetized. As such, they typically have a coercivity < 12.5 Oe (1000 A/m), and specialized very soft magnets can have coercivities < 0.00125 Oe (1 A/m). Some common applications include coil cores, transformer cores, and magnetic shielding. Additionally, they are important for many AC electrical applications such as inductors, filters, and resonators, particularly at high frequencies up to and including microwaves.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Flexible 3D printed silicones for gamma and neutron radiation shielding

Here, resin formulations for direct ink write (DIW) additive manufacturing (AM) were developed for facile customization to shield radiation signatures consisting of a range of gamma radiation and neutron energies in a variety of applications. The resin developed in this work has been formulated to accommodate very high loadings of shielding filler materials: up to 40 wt% bismuth (Bi) metal for gamma radiation shielding, and up to 60 wt% boron-10 ( 10 B) metal for neutron shielding. We report the first instance of a printable resin containing 60 wt% 10 B, which remains flexible after curing despite the high solids content. We demonstrate that these resins can not only tolerate formulation modulation for radiation profile-matched shielding, but they can also be printed in a variety of geometries such that mechanical properties of the printed shield are readily tuned to specific applications. In contrast, most existing shielding materials are rigid, bulky, and do not allow for the complex shapes and mechanical flexibility needed in high-usage applications. This work demonstrates the successful development of a tunable resin for additively manufactured, flexible radiation shielding material, containing high loadings of shielding filler, which can be utilized in harsh radiation environments associated with nuclear materials.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

SCALE 6.3 Validation: Spent Nuclear Fuel

This report is the fifth volume in a series documenting the validation of SCALE 6.3 with ENDF/B-VII.1 libraries for nuclear criticality safety, reactor physics, radiation shielding, and spent nuclear fuel applications. This fifth volume, which focuses on validating SCALE capabilities that impact spent nuclear fuel applications, provides an update of the similar validation reported for SCALE 6.2.4. The experimental data used as basis for validation herein consist of measurement data for nuclide inventories and decay heat, including the following: 1. radiochemical assay (RCA) measurements of nuclides important to burnup credit, decay heat, and radiation shielding in 205 light-water reactor (LWR) spent nuclear fuel samples that cover burnups ranging up to 80 GWd/MTU and initial fuel enrichments up to 4.9% 235U; 2. full-assembly decay heat measurements for 236 LWR assemblies with initial fuel enrichments ranging up to 4% 235 U, assembly burnups of 5–51 GWd/MTU, and decay times after fuel discharge in the 2- to 27-year range (of importance to spent nuclear fuel storage, transportation, and disposal); and 3. pulse fission irradiations for fissionable materials at cooling times of interest to severe accident analyses (<10 5 s).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Photonuclear Physics in SCALE

Radiation shielding modeling involves a wide range of applications, such as power reactors, critical experiments, portable sources, linear accelerators, and fusion systems. Particle energy ranges vary from thermal energies to fast energy spectra. MAVRIC, as the radiation shielding sequence of the SCALE code system [1] developed and maintained by Oak Ridge National Laboratory, is capable of modeling both neutron and photon interactions in matter and is widely used by the US Nuclear Regulatory Commission, Department of Energy, the National Nuclear Security Administration, and radiation shielding practitioners all around the world. Although most applications don’t involve highly energetic photons that can cause photonuclear reactions and create coupled chains of high energy, ionized electrons and bremsstrahlung radiation within a material medium, when present, these reactions can have a significant impact upon the downstream particle fluxes from incident photons above 20 MeV. For example, linear accelerators used for medical isotope production are examples of applications in which high-energetic photons are generated and used for driving or controlling other components of a complex facility. Radiation shielding analyses of such a facility would require accounting for any and all significant sources of radiation. Unfortunately, MAVRIC included no data and methods to model either photonuclear reactions or the bremsstrahlung photons until this work, which enabled accounting for both. Photonuclear reaction cross sections were imported from the Evaluated Nuclear Data File libraries, and corresponding methods were implemented in the SCALE code system; see Section 2 for how the nuclear data are imported and used. Bremsstrahlung photons were modeled by assuming thick-target approximation whereby all generated electrons and positrons due to photon interactions are assumed to be absorbed within the same material in which they were born, yielding a quick particle transport simulation method by using fast lookup tables for generating subsequent gammas; see Section 4 for details. A suite of verification models was used to test implemented methods and associated data. Tests yielded satisfactory results for the new features. Enhancements will be released as part of the next major SCALE release (v7.0), which is anticipated to be deployed in 2025, with beta releases available in the interim.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

SCALE 6.3 Validation: Spent Nuclear Fuel

This report is the fifth volume in a series documenting the validation of SCALE 6.3 with ENDF/B-VII.1 libraries for nuclear criticality safety, reactor physics, radiation shielding, and spent nuclear fuel applications. This fifth volume, which focuses on validating SCALE capabilities that impact spent nuclear fuel applications, provides an update of the similar validation reported for SCALE 6.2.4.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗