Search NASA⌕ Search

SEARCH · Search NASA

Results for “irradiation testing”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Quantitative assessment of Ni + and He + ion irradiation damage in a tungsten heavy alloy under the simulated nuclear fusion environment

A 90W-7Ni-3Fe (wt.%) tungsten heavy alloy has been sequentially Ni + and He + ion irradiated at 700 °C to simulate the high temperature irradiation environment of a fusion reactor interior. W/Ni–Fe-W dual-phase alloys have been proposed to serve as plasma facing materials and require detailed investigation of their behavior under fusion relevant conditions to assess their overall applicability. To evaluate material performance under five years of simulated fusion reactor service, microstructural characterization of the nanoscale defect distribution has been performed on both constituent phases, revealing peak swelling in the W phase of approximately 0.03%. The γ-phase (Ni–Fe-W) is found to swell approximately 0.68% under the same irradiation conditions, indicating significant cavity formation and growth. Additionally, a novel multi-projection imaging approach has been applied to determine the extent of damage segregation along the dual-phase W-to-γ interface and exposes that these interfaces act as sink sites for the accumulation of cavities. Interphase boundaries are noted to possess an 11.8% areal coverage of defects along the boundary plane, primarily on the γ-phase side of the boundary. The accumulation of cavities at these interphase boundaries is anticipated to adversely affect overall material toughness, and this work reveals a pressing need for mechanical property testing of irradiated W–Ni-Fe dual-phase alloys.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Transient testing of uranium silicide fuel in zircaloy and silicon carbide cladding

Using the newly restarted Transient Reactor Test (TREAT) facility at Idaho National Laboratories, fresh (unirradiated) U 3 Si 2 fuel pellets were tested under reactivity-initiated accident conditions in dry helium capsules. These tests were performed with total energy depositions sufficient to induce melting of the U 3 Si 2 fuel pellets based on available thermophysical property data. The instrumented experiments utilized in-pile thermocouples and infrared pyrometers to measure the cladding temperature during the transient, which are in reasonable agreement with predicted peak cladding temperatures using a simple BISON fuel performance model. This same model is then used to predict expected cladding stresses during the transients in areas where pellet cladding interaction is expected. It was shown that for transient energy depositions up to 528J/g rod like geometry of the test articles was maintained. While it is acknowledged that more confirmatory testing is required in more prototypic environments, these first of their kind experiments have demonstrated that unirradiated U 3 Si 2 fuel and unirradiated SiC-SiC claddings can likely retain a “rod like” geometry for transient energy depositions up to 528J/g.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

MFC Transient Testing and Fuel Fabrication Capabilities

MFC Transient Testing Capabilities and Fuel Fabrication capabilities are summarized in this slide deck. Slides are included for TREAT and TREAT capabilities, EFF, FASB, FMF, ZPPR, AFF, and the Fuels Fabrication Modernization Capability. First use of this slide deck is for presentation at Westinghouse Electric Corporation, but there will be other uses later.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Analysis of Design Alternatives of Actively Cooled RF Window for MPEX

The Materials Plasma Exposure eXperiment (MPEX) has been designed as a linear plasma divertor simulator in order to address plasma material interaction (PMI) science for next-generation fusion devices. It will have the capability to test neutron irradiated samples with plasma fluxes of greater than 10 24 m -2 s -1 . It is expected to operate steady state for up to 10 6 s to consider PMI affects through reactor end of life. The conceptual design of MPEX was completed in 2019, with preliminary design having begun in 2020. The plasma source for MPEX is a helicon antenna, where the energized helical antenna sits outside of the vacuum in order to minimize impurities in the plasma. It is expected to receive up to 200 kW of continuous power, and so the antenna and the window must be actively cooled. The water-cooled copper antenna has been operated at full power on the Proto-MPEX device (which is a test facility to demonstrate the plasma source and heating systems). The water-cooled window, however, is a novel component that must meet numerous competing requirements. It requires a low dielectric loss to allow the Radio Frequency (RF) power to create the plasma within the vacuum boundary. It must be structurally robust to handle the significant heat flux from the plasma and any heat from dielectric coupling. It must be compatible with the coolant (preferably water). It requires a vacuum seal that minimizes impurities into the plasma and does not compromise the structural integrity of the window. Two window designs have been tested. Results from these tests, where temperatures are measured and heat fluxes inferred from infrared camera data, have been correlated with thermal-structural simulations. When these simulations are extrapolated to the full power steady-state heat fluxes that are expected in MPEX, the designs do not appear to have the necessary structural robustness. This study explores design alternatives for the MPEX helicon antenna window, presents analysis results for several of the alternatives, and shows a viable solution that satisfies the requirements for MPEX operation.

Tipton Jr, Joseph↗
DOE OSTI2023

Mechanical Properties of Irradiated U-10wt%Mo Alloy Degraded by Porosity Development

A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States’ high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo fuel and the data produced includes bending strength and Young’s modulus. Although the material behaves in a brittle manner due to the developed porosity, a general trend of strength and modulus reduction are found as fission density increases. The data produced is evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Chemical Thermodynamic Modeling of Molten Salts to Support Off-Gas Abatement Systems

The reprocessing of used nuclear fuel by any means will liberate gaseous fission products, such as hydrogen ( 3 H), carbon ( 14 C), noble gases ( 85 Kr), and halogens ( 129 I), from the irradiated fuel. These elements will distribute through chemical processing operations and partition into process off-gas streams, and the specific volatile release fractions will be dictated by the chemical and physical properties of the system. A recent assessment found that there were significant knowledge gaps regarding the release quantities of volatile radionuclides from individual unit operations. These knowledge gaps limited the ability to determine what dedicated off-gas treatment technologies could be required for electrochemical-based reprocessing facilities. Unfortunately, experimental efforts to quantify release fractions are limited by the challenges associated with performing experiments using irradiated fuel. This report documents preliminary thermodynamic predictions of iodine and tritium release from chloride-based molten salts as part of an effort to better direct resources toward those experiments (both simulant and irradiated) that will be of the greatest impact. It was predicted that less than 0.5% of tritium was expected to be released and that nearly all of that amount would be released as H 2 . Thermochemical data for hydrogen (H 2 ) are of high fidelity, and no additional validation is recommended. Less than 0.05% of iodine was predicted to be released, with the primary volatile species being Cs 2 I 2 . Unlike the tritium predictions, the data underpinning the iodine release predictions are of low quality. It is recommended that a limited experimental program be dedicated to expanding the current physical property and thermodynamic property data for iodine in electrochemical processing and molten salt conditions, including vapor pressure measurements for the dimerized species predicted to comprise the majority of iodine release. Validating and improving the thermodynamic properties used in predictive modeling can reduce the need for expensive testing with irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Design of a 10 MeV beamline at the Upgraded Injector Test Facility for e-beam irradiation

Electron beam irradiation near 10 MeV is suitable for wastewater treatment. The Upgraded Injector Test Facility (UITF) at Jefferson Lab is a CW superconducting linear accelerator capable of providing an electron beam of energy up to 10 MeV and up to 100 µA current. This contribution presents the beam transport simulations for a beamline to be used for the irradiation of wastewater samples at the UITF. The simulations were done using the code General Particle Tracer with the goal of obtaining an 8 MeV electron beam of radius (3-σ) of ~2.4 cm. The achieved energy spread is ~74.5 keV. The space charge effects were investigated when the bunch charge is varied to be up to 1000 times and the results showed that they do not affect the beam quality significantly.

Li, X.↗
DOE OSTI2021

Design of a 10 MeV beamline at the Upgraded Injector Test Facility for e-beam irradiation

Electron beam irradiation near 10 MeV is suitable for wastewater treatment. The Upgraded Injector Test Facility (UITF) at Jefferson Lab is a CW superconducting linear accelerator capable of providing an electron beam of energy up to 10 MeV and up to 100 µA current. This contribution presents the beam transport simulations for a beamline to be used for the irradiation of wastewater samples at the UITF. The simulations were done using the code General Particle Tracer with the goal of obtaining an 8 MeV electron beam of radius (3-σ) of ~2.4 cm. The achieved energy spread is ~74.5 keV. The space charge effects were investigated when the bunch charge is varied to be up to 1000 times and the results showed that they do not affect the beam quality significantly.

Li, X.↗
DOE OSTI2022

A concurrent irradiation-mechanics multiscale coupling model

The vast majority of our current knowledge regarding the basic mechanisms controlling irradiation effect on mechanical properties is based almost entirely on the results of post-irradiation experiments or theoretical models. However, the concurrent effects of irradiation, mechanical stress, and thermal damage on the failure phenomena of materials and components remain largely unexplored due to its internal multiscale-multiphysics coupling nature. Here, we present here a concurrent irradiation-mechanics multiscale coupling model. The concurrent evolutions of nanoscale irradiation defect clusters, microscale dislocation configurations, and mechanical responses are well captured based on coupling cluster dynamics, discrete dislocation dynamics, and the finite element methods using an effective time marching scheme. Model predictions of defect densities and size are in general agreement with experimental observations. Irradiation hardening is shown to take place also in samples undergoing concurrent irradiation-mechanical loading, similar to samples tested post-irradiation. However, the occurrence of plastic flow localization and dislocation channel formation is not accompanied with apparent yield drop (softening) under concurrent irradiation-mechanical loading conditions, which is different from the post-irradiation case.

42 ENGINEERING↗
DOE OSTI2023

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

Here, this paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO 2 . The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB 2 Integral Fuel Burnable Absorber (IFBA) and B 4 C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd 2 O 3 -doped UO 2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Beyond interpolation: Physics-inspired gating transformers for extrapolating irradiation conditions to novel nuclear fuels

The qualification of advanced nuclear fuels relies on irradiation experiments in test reactors that emulate commercial conditions. Designing these tests requires accurate prediction of key irradiation quantities, particularly heat generation rate and burnup, yet obtaining them typically involves computationally expensive multi-step simulation workflows. We propose a physics-inspired gating transformer (PIGT) that integrates an inverse-square, distance-based attenuation into the encoder representation to bias attention toward physically relevant spatial relationships while retaining data-driven flexibility. Using MiniFuel irradiation data from the High Flux Isotope Reactor at Oak Ridge National Laboratory, we benchmark against ensemble methods, feedforward and recurrent networks, convolutional models, and standard transformers. While baseline models perform well under interpolation, they exhibit a pronounced generalization gap when evaluated on fuels not included in the training set. The proposed model consistently improves extrapolative accuracy and stability, yielding the strongest performance on unseen fuel configurations. These results indicate that a lightweight physics structure embedded within attention mechanisms can substantially improve robustness, enabling more reliable surrogate predictions to accelerate the design of nuclear fuel irradiation experiments.

Fuel qualification↗
DOE OSTI2020

Construction status of the Mu2e crystal calorimeter

The Mu2e experiment at Fermilab will search for the charged-lepton flavor violating neutrino-less conversion of a negative muon into an electron in the field of an aluminum nucleus. The Mu2e detector is composed of a tracker, an electromagnetic calorimeter and an external veto for cosmic rays. The calorimeter plays an important role in providing excellent particle identification capabilities and a fast online trigger filter, while aiding the track reconstruction capabilities. Calorimeter requirements are to provide a large acceptance for 0~10 MeV electrons and reach: i) a time resolution better than 0.5 ns; ii) an energy resolution better than 10%; and iii) a position resolution of 1 cm. The calorimeter consists of two disks, each one made of 674 pure CsI crystals. Each crystal is readout by two large area 2×3 arrays of UV-extended SiPMs of 6×6 mm2 dimensions. A large scale prototype (Module-0) has been tested at an electron beam. We report here the tests done to finalize the calorimeter design, the results obtained with Module-0 and the status of production. At this time, the performance characteristics of 85% of the crystals and all of the SiPMs have been measured. The calorimeter engineering drawings have been completed and the large mechanical components are under fabrication. Analog and digital electronics have been prototyped and tested with irradiation dose. Their serial production is being organized. The calorimeter assembly phase is planned for mid-2020.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Irradiation Experiments on High Efficiency Nanowire Solar Cells Including Tilted Incidence Angle

Nanowire based solar cells have promising potential to become the next generation of space solar cells. In this study we corroborate their high radiation tolerance by irradiating high efficiency GaAs (efficiency > 15 %) nanowire solar cells with 1 MeV electrons at a fluence of 10 16 e - /cm 2 . Additionally, we theoretically show that the damage in nanowire solar cells is expected to be slightly dependent on the incidence angle and the geometry of the nanowire array. However, experimentally, two sets of GaAs nanowire solar cells have been irradiated with 1 MeV electrons at normal incidence and with a 25° tilted angle exhibiting a similar degradation.

high specific power↗
DOE OSTI2021

Impact of Coating Defects on Performance of Coated Zirconium Cladding

Research on accident tolerant fuels (ATF) has started after the Fukushima accident [1–3]. While efforts have been expended on both fuel and cladding ATF concepts, the bulk of work has been devoted to improved cladding. The overarching goal of these approaches is to extend the coping time available during a severe accident before any event would result in release of radioactivity to the public. The most basic ATF cladding concept is obtained by applying a thin coating of highly corrosion-resistant material on the surface of a licensed zirconium cladding alloy. This thin coating is intended to not interfere with the neutronic or mechanical performance of the base cladding under normal operating conditions. Different coating materials, thicknesses, coating processes, process parameters, and testing methods have an impact on the microstructure and mechanical properties and therefore on the results of the applied investigation methods. These challenges have motivated an initial focus on demonstrating that the presence of coatings do not perturb the critical performance benchmarks of uncoated material. Ongoing lead test assembly irradiations of coated zirconium concepts in commercial reactions is intended to establish baseline performance in this regard in the coming years.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Natrium Demonstration Reactor Support [Abstract]

TerraPower, LLC (TerraPower, Participant) and other private industry partners endeavor to design, license, construct, and operate a sodium-cooled fast-spectrum nuclear reactor technology demonstration plant called Natrium. This demonstration plant is supported by the U.S. Department of Energy (DOE) through the Advanced Reactor Demonstration Program (ARDP; DE-FOA-0002271). TerraPower, together with its technology co-developer GE Hitachi Nuclear Energy (GEH) and engineering and construction partner Bechtel, submitted a proposal under the program’s Advanced Reactor Demonstration Pathway for its Natrium reactor and energy system and recently received an award. TerraPower is partnering with Battelle Memorial Institute, the Management and Operating Contractor of Pacific Northwest National Laboratory (PNNL, Contractor) under the Natrium project to provide critical research outcomes necessary to demonstrate the reactor technology. Over the expected five-year timeframe of the project, PNNL will provide TerraPower and its partners with vital support in the areas of post-irradiation examination (PIE) of specimens irradiated in test reactors. These efforts will be combined and managed as a program titled “Natrium Demonstration Reactor Support.”

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗