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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Fuel Assembly and Irradiation Parametric Study for Extended-Enrichment and High-Burnup Light-Water Reactor Spent Nuclear Fuel in Dry Storage Casks and Transportation Packages

There is an increased interest in operating commercial light-water reactors (LWRs) in the United States with improved economics that would result from longer fuel cycle lengths, fewer and shorter refueling outages, and fewer fuel assemblies requiring storage at the back end of the fuel cycle. To support this, fuel discharge burnups, as well as initial 235 U enrichments, must be higher than those used in current commercial LWRs. The typical upper limit considered for assembly average burnup in this report is 75 gigawatt-days (GWd) per metric ton of uranium (MTU), as opposed to the current typical upper bound of approximately 62 GWd/MTU. The upper limit considered for initial 235 U enrichment is 8 weight percent (8 wt %), as opposed to the current regulatory limit of 5 wt %. The enrichment range from 5 to 8 wt % is referred to in this report as extended enrichment . To investigate the effect of high burnup and extended enrichment conditions on dose rates and burnup credit for dry storage casks and transportation packages, a fuel assembly and irradiation parametric study was performed. The conclusions from this study will assist U.S. Nuclear Regulatory Commission staff in reviewing applications for dry storage casks and transportation packages that contain high-burnup and extended enrichment fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Comprehensive analysis of disruption mitigation methods using gas and pellet-like injections in ITER-like Tokamaks

Abstract Inert-gas shielding could be an effective mechanism for protection of plasma facing surfaces (PFS) against plasma particles impact and photon radiation heat loads during transient events in fusion devices. Neutral gas injection is one promising way to mitigate erosion of tokamak components and contamination. The objective of this work is to study and optimize mitigation methods using neutral gas and pellet-like injections to decrease the heat load to the divertor surfaces and to prevent vaporization of the various internal surfaces due to transient events in ITER-like devices. The integrated self-consistent models implemented in the HEIGHTS package was used for detailed analysis of the potential secondary plasma generation from the injected inert gas, its radiative characteristics, and shielding effectiveness. We varied the density, size, and location of an argon gas cloud to minimize the disruption energy deposited into the divertor components. We also investigated innovative ways for minor changes in ITER-like internal design to mitigate disruptions. We found the optimum parameters to fully protect ITER tokamak surfaces from erosion and vaporization during plasma instabilities. This preliminary analysis showed that using Ar gas injection methods could lead to enhancement in components lifetime in ITER-like and future DEMO devices with minor design changes.

Science & Technology - Other Topics↗
DOE OSTI2025

The SHMS 11 GeV/c spectrometer in Hall C at Jefferson Lab

The Super High Momentum Spectrometer (SHMS) has been built for Hall C at the Thomas Jefferson National Accelerator Facility (Jefferson Lab). With a momentum capability reaching 11 GeV/ c , the SHMS provides measurements of charged particles produced in electron-scattering experiments using the maximum available beam energy from the upgraded Jefferson Lab accelerator. The SHMS is an ion-optics magnetic spectrometer comprised of a series of new superconducting magnets which transport charged particles through an array of triggering, tracking, and particle-identification detectors that measure momentum, energy, angle and position in order to allow kinematic reconstruction of the events back to their origin at the scattering target. The detector system is protected from background radiation by a sophisticated shielding enclosure. The entire spectrometer is mounted on a rotating support structure which permits measurements to be taken with a large acceptance over laboratory scattering angles from 5.5° to 40°, thus allowing a wide range of low cross-section experiments to be conducted. Finally, these experiments complement and extend the previous Hall C research program to higher energies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

LBNF Material Activation Study

This poster presents a study on material activation at the Fermilab Long-Baseline Neutrino Facility (LBNF), focusing on understanding how high-energy beams interact with surrounding materials to produce radioactive isotopes. Using a simplified model of the LBNF-20 bunker and the FLUKA Monte Carlo simulation tool, the project quantifies isotope production at key locations inside and outside the shielding structures. The study employs a two-step simulation process, tracking primary and secondary particles, and explores the impact of different geometric and material configurations on activation rates. Results include preliminary comparisons of simulated activation rates and suggest methods for refining geometry, improving accuracy, and efficiently simulating future configurations. This work informs the design and operation of the LBNF facility, aiding in radiation safety and shielding optimization. Future efforts will focus on sample iteration, automated simulations, and detailed residual isotope analyses to further enhance the understanding of material activation in high-energy physics environments.

Pietraszek, Jacob [Unlisted, US]↗
DOE OSTI2026

Aluminum Based Solvent-Free Organic–Inorganic Hybrid Materials

In emerging materials, molecular hybrids are especially promising, as they have molecular level mixing of the organic and inorganic components, producing homogeneous materials without interfaces that can deteriorate properties. However, the current methods of manufacturing molecular hybrids are based on solution processing, which is impractical for bulk materials such as may be used for optically clear radiation and electromagnetic shielding components or photonics. Here we examine molecular hybrids composed of aluminum isopropoxide (AIP) and epoxy resins aiming to understand the molecular scale chemistry and manufacturability of these hybrid materials. DSCmonitored cure revealed the ideal cure temperature for these materials is 160−170 °C and demonstrated that an AIP concentration of 16.7 wt % maximizes the extent of reaction. Kinetic analysis of the curing reaction showed the Sestak−Berggren autocatalytic model is effective at temperatures over 140 °C but the reaction has diffusion limitations at a temperature of 120 °C. Mechanical testing with custom resin molds revealed a decrease in properties of the bulk samples with increasing AIP content due to an increase in defects but further testing with nanoindentation demonstrated comparable or improved mechanical properties of AIP-epoxy hybrids compared to epoxy resin with a standard hardener. Ultimately, this work lays the foundation for hardener-free epoxy-aluminum inorganic/organic hybrids and presents opportunities to expand on properties for specific applications such as thermal conductivity, optical clarity, and dielectric constant.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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 OSTI2024

Special Nuclear Material Hold-up Measurement

Facilities that process special nuclear material (SNM) generally have a variety of equipment, shielding, and nearby radiation sources, which can pose challenges when performing nondestructive analysis (NDA) within the facility, as shown in Figure 1. In this figure, the left image features two individuals performing a holdup measurement of a large duct at a posting, utilizing a detector probe attached to a yardstick to reduce some aspect of the measurement uncertainty. Note that the distance to the holdup within the pipe from the posting cannot be accounted for. The right image similarly features several individuals guiding a holdup measurement of a ducting pipe several meters overhead, by attaching a detector probe to a long handle and lining that up with a posting on the outside of the pipe. SNM processing begets SNM holdup, and this difficulty with accurate NDA can result in unaccounted accumulation of SNM holdup within a process or area. Normal SNM processing operations can accumulate holdup within filters, pumps, pipes, ducts, other equipment, and facility support systems, as demonstrated by Figure 2. It is important for deactivation and decommissioning (D&D) gloveboxes and their ancillary equipment, material accountability, criticality safety, facility operations, waste management, radiation safety, and security to accurately characterize the location, composition, and quantity of this holdup to minimize these associated risks and potential diversion pathways. Current NDA techniques for assay of holdup are expensive, time consuming, and have high uncertainty.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Geant4 Visualization

Geant4 offers versatile visualization tools to study detector geometry, particle trajectories, and hits. These tools support both interactive and batch modes (e.g., vis.mac). With a unified interface, Geant4 integrates various graphics systems such as G4VisExecutive, OpenGL, Qt, HepRApp, DAWN, VTK, and ToolsSG. Advanced examples, including the ICRP145 Human Phantoms and the gMocren tool, highlight its applications in medical research and simulation and as well as radiation protection and shielding studies. These features make Geant4 a robust framework for exploring particle interactions and supporting domain-specific use cases.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2025

Simulation of the radiological impact during selected space travel scenarios using the Monte Carlo code FLUKA

Radiation is one of the major challenges of space exploration and can negatively impact both biological and electronic systems, particularly in the case of long-term journeys or if the spaceship features inadequate shielding. Here, in this work, the cumulative dose levels from prompt radiation in the spacecraft are quantified alongside the residual dose contributions arising from activation of vessel components. The radiological impact was assessed for various space exploration scenarios, considering the same spaceship model featuring three shielding design variants. In each scenario, the radiation environment was generated with the Monte Carlo particle transport and interaction code FLUKA. These results can be used to quantify the contribution of prompt and residual dose in spacefaring ventures and help determine optimal radiation shielding needed to mitigate the overall radiological impact on both astronauts and equipment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

SCALE Shielding Calculations for Advanced Reactor Accident Scenarios

The study results presented in this paper demonstrate the capabilities of the SCALE computer code for non–light water reactor (non-LWR) radiation source term and shielding calculations in support of the U.S. Nuclear Regulatory Commission non-LWR fuel cycle demonstration project. Representative non-LWR types, including the sodium-cooled fast reactor (SFR), the molten salt reactor (MSR), and the heat pipe microreactor (HPMR), were analyzed to evaluate dose rates associated with postulated accident scenarios (SFR and MSR), reactor operation (MSR and HPMR), and fuel self-protecting characteristics (HPMR). New features were implemented in SCALE depletion codes to better simulate MSR operation.

MSR↗
DOE OSTI2026

Evaluation of Radiation Transport Through a Nuclear-Grade Sandwich Composite

The development and deployment of microreactors are often synonymous with their transportability; however, designing a shield which is appropriate for both transportation and operations is challenging. A recent approach is the use of composite style shield which combines the reactor pressure vessel and biological shielding into a single unit to enhance the shields effectiveness without degrading the structural properties. To assess the viability of a nuclear grade sandwich composite (NGSC), and understanding of how radiation transports through the shield needs to be understood to select the appropriate materials. This work examines a two-layered NGSC of varying thicknesses with two primary shielding materials (tungsten-tetraboride cermet and boron carbide). To assess the NGSC, quantities of interest including neutron attenuation, photon attenuation, energy deposition, displacements per atom, helium generation, and dose are examined. Altogether, it is found that a combination of both tungsten-tetraboride and boron carbide are necessary to act as an appropriate shield, where the order in which these materials are place in the composite are important to their effectiveness.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Uncertainty Quantification for Neutron Shield Using Convolutional Neural Networks

Uncertainty quantification from radiation transport calculations was conducted using a Bayesian inference approach. A surrogate model, using a convolutional neural network, was employed to emulate the neutron fluence, which was simulated with a Monte Carlo radiation transport model. This allowed for a computationally cheap approach to evaluate input parameters and to sample their corresponding posterior probability distributions. Experimental data from the literature were employed to perform uncertainty quantification studies for concrete shields. As a result, the method is a nonintrusive approach that enables studies with multiple input parameters and can be applied to any radiation transport model.

Bayesian inference↗
DOE OSTI2024

Radiation Dose Modeling for Niowave’s Accelerator Driven Uranium Target Assembly 3

Molybdenum-99 is a high-value radionuclide commonly used for medical purposes within the United States. The National Nuclear Security Administration (NNSA) seeks to reliably produce the radioisotope 99 Mo without the use of highly enriched uranium. NNSA’s Office of Material Management and Minimization (M3) provides funding and government laboratory expertise to private companies to expedite the production process domestically and currently funds designs that use low-enriched uranium or other 99 Mo production pathways. Several production designs are being explored across the industry, including uranium fission and photonuclear conversion of 100 Mo targets. Niowave Inc. seeks to produce 99 Mo via a high-energy electron accelerator that strikes a lead-bismuth eutectic target that ultimately produces a consistent neutron flux. The neutron flux then interacts in a subcritical reactor core configuration to produce fission in low-enriched or natural uranium targets. These fissionable targets are then processed to extract 99 Mo. The purpose of this work is to estimate the neutron and photon dose response across Niowave’s proposed facility for worker safety during operation. Owing to the size of the proposed Niowave facility and necessary shielding, unbiased Monte Carlo radiation transport is impractical, and variance reduction methods are required. This work focuses on the weight window variance reduction method to produce high confidence dose response results within a Monte Carlo radiation transport code. Specifically, an adjoint-informed weight window methodology was created to improve the dose response estimates for accelerator-driven subcritical reactor designs. This adjoint-informed methodology was implemented for Niowave’s proposed design and improved dose results at far-field locations across the facility. Acceptable dose rate contours for the proposed facility were generated across the facility and are presented in this work.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Scintillator based nuclear photovoltaic batteries for power generation at microwatts level

A nuclear photovoltaic battery uses scintillator to convert radiation into visible light, which is then collected by a photovoltaic (PV) cell to generate electricity. If the radiation is gamma-rays emitted from external sources, the battery may also be referred as gammavoltaic battery. In this study, a polycrystalline CdTe solar cell was optically coupled with a 2.0 cm × 2.0 cm × 1.0 cm Gadolinium Aluminum Gallium Garnet (GAGG) scintillator, and the resulting device was tested using intense gamma radiation fields from a Cs-137 (1.5 kRad/h) and a Co-60 (10 kRad/h) irradiator. Measurements with Cs-137 provided a maximum power output (P max ) of ~288 nW, with a short-circuit current density (J sc ) of ~1.22 μA/cm 2 and an open-circuit voltage (V oc ) of ~0.34 V. In contrast, Co-60 irradiator gave a P max of 1.5 μW, with a J sc of ~4.73 μA/cm 2 and a V oc of ~0.38 V. The CdTe was also paired with a Lutetium-Yttrium Oxyorthosilicate (LYSO) crystal and tested with the Cs-137 source. The experiment presents a scalable option to reach to higher power outputs by harvesting gamma radiation fields in many cases where high radiation field demands heavy shielding and is often regarded as unwanted waste.

25 ENERGY STORAGE↗
DOE OSTI2025

Macromolecules & Manufacturing Science

Outline • SRNL Overview • Mission overview • Polymers enabling the mission • R&D Highlights • Polymers in radiation environments • Tooling in shielded cells • Packaging for nuclear material shipments • Polymers supporting tank waste remediation • Ref electrode • Epoxy and polymer grout • Polymers for fusion energy • Deuterium labelling • Polymers for additive manufacturing • Coalescence and blends: experimental and predictive • Process modelling and sorting through big data (peregrine and latticeJ)

Chatham, Camden [Savannah River National Laborator↗
DOE OSTI2025

MARIE: A Python-Based Framework for Comprehensive Fuel Recycling Modeling

One of the most pressing challenges to the continued deployment of nuclear energy systems is in the ultimate management and disposition of discharged fuel assemblies. While reprocessing and recovery of valuable materials from UNF assemblies has been considered as part of an overall strategy for minimization of the volume of reactor-based wastes to be managed, the deployment of commercial-scale reprocessing facilities presents an enormous economic challenge. The MARIE software package has been developed as a means of confronting this challenge. Representing components of a generic fuel reprocessing operation as individual physical processes, MARIE is designed as a modular framework intended to allow for analysis and cost-optimization for a hypothetical reprocessing facility while realistically accounting for the physical characteristics of the used fuel source term, such as decay heat, activity, and radiation dose (informing corresponding shielding requirements). Capabilities supported by MARIE include head-end operations such as fuel shearing, voloxidation, and dissolution; generic solvent extraction operations informed by available open-literature data; a suite of unit operations intended to represent electrochemical processing of used fuel assemblies (i.e., oxide reduction, electrorefining, and electrowinning); and finally, accounting for both costs and physical features of discharged waste streams, which can be used to inform follow-on analyses such as the feasibility of deep-borehole disposal of HLW. This paper presents an overview of the MARIE software capabilities, including how individual unit operations are implemented to enable a larger-scale optimization of a hypothetical reprocessing operation on aspects such as cost and recovery of valuable materials.

Skutnik, Steve [ORNL] (ORCID:000000016441135X)↗
DOE OSTI2025

Benefits and challenges of vibroacoustic process monitoring to support operators in nuclear research facilities

According to the International Atomic Energy Agency in 2024, global nuclear energy capacity is projected to increase by 2.5 times by 2050 in their high-case scenario. Research on the laboratory scale of innovative processes within nuclear facilities is ongoing to improve nuclear fuel cycle capabilities. Vibroacoustic process monitoring of these techniques has potential to inform operators regarding process specific metrics, predictive maintenance, and anomalous event detection. Additionally, this type of monitoring has the potential to aid in nuclear safeguards. Challenges regarding vibroacoustic monitoring in these environments include high temperature, high radiation, limited access, and shielded equipment limiting sensor type and placement. Microphones, accelerometers, and temperature sensors were deployed both inside and near these environments during operation to determine environment driven limitations, sound attenuation of containment enclosures, process specific metrics, and future potential of vibroacoustic monitoring in these environments. Overall, this work highlights the benefits and limitations of vibroacoustic process monitoring in nuclear research facilities.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A high-temperature Rutherford Backscattering Spectrometry apparatus for in situ material characterization

A new methodology for high-temperature Rutherford Backscattering Spectrometry (HT-RBS) has been developed to enable in situ material characterization at elevated temperatures. A 3.5 MeV proton beam penetrates a 10-µm-thick 316L stainless steel foil mounted on a graphite substrate, with backscattered signals detected using an HT-RBS system. Conventional semiconductor detectors, primarily based on silicon, suffer significant performance degradation at temperatures higher than ~ 60 °C due to increased leakage current and noise, leading to signal distortion and failure. Here, to preserve spectral quality, a 5 µm aluminum foil shields the detector from thermal radiation, allowing reliable operation up to 900 °C at the target. A rotatable shutter provides additional thermal isolation during data collection pauses. In situ measurements of areal density changes of 316L stainless steel were conducted to validate the technique, revealing consistency with the known thermal expansion coefficient. The method facilitates seamless switching between irradiation and analysis, enabling continuous studies. This approach supports in situ investigations of diffusion, void swelling, creep, and corrosion, offering a versatile tool for advanced materials research.

36 MATERIALS SCIENCE↗