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At least 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Draft Summary Report on Irradiated Fuel Handling and Management for LOTUS

The National Reactor Innovation Center (NRIC) has conceptualized the design of the Laboratory for Operation and Testing in the United States (LOTUS) test bed to provide the United States Department of Energy (DOE) with the infrastructure necessary to make advanced reactor designs available for commercial developers. LOTUS will provide a test bed to developers with the capabilities of supporting a wide range of experiment design possibilities. Upon completion of the developers’ operations and experiments within the test bed, the irradiated fuel, reactor components, and other experiment materials must be removed. Idaho National Laboratory (INL) possesses significant capabilities for radioactive material handling such as casks, carts, and forklifts. However, given the unique environment presented by the NRIC-LOTUS test bed, located inside the Zero Power Physics Reactor (ZPPR) Perimeter Intrusion Detection and Assessment System (PIDAS) area at the Materials and Fuels Complex (MFC) and the complexity of novel removal activities of recently operated reactor experiments through the new proposed access tunnel. The efficacy was not apparent for existing equipment to provide all the needed capability. To bridge the potential gaps in cask designs, storage, and transportation, NRIC requested the development of trade studies for the transfer, handling, and storage requirements of irradiated fuel salts and other radioactive materials. NRIC directed Boston Government Services, LLC (BGS) to perform the trade studies and develop a report analyzing alternatives. In addition to the BGS reports, the Idaho National Lab’s (INL), provided by the first potential user’s Advanced Reactor Development (ARD) team, prepared a feasibility study for the storage of specific irradiated fuel within the existing ZPPR vault.This summary report is intended to present the trade studies, options, and alternatives that were investigated. The maturity level of LOTUS, the reactor, fuel salt containers, gloveboxes, and reactor testing campaign and concept of operations were not at a level sufficient to base critical decisions on. This report is not intended to present a final recommendation. The final recommendations for fuel storage location, transport, handling equipment, and operations will be made in FY 2024 and will be based on known materials, test campaign requirements, funding, and final analysis of the fuel and equipment to be used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1996

Total Solar Irradiance Variability: A Review

Observations of total solar irradiance from space within the last two decaades convinced the skeptics that total irradiance varies over a wide range of periodicities: from minutes to the 11-year solar activity cycle. Analyses based on these space-borne observations have demonstrated that the irradiance variations are directly related to changes at the photosphere and the solar interior.

solar irradiance solar luminosity irradiance model↗
DOE OSTI2021

Understanding TRISO Coated Particle Neutron Irradiation Behavior:Evolution of Advanced Micro Analysis and Electron Microscopy Approaches

Objectives and motivation• Provide data for fuel qualification in support of reactor licensing• Establish a domestic commercial TRISO fuel fabrication capability Approach • Focus is on developing and testing UCO TRISO fuel –Develop fuel fabrication and QC measurement methods, first at lab scale and then at industrial scale –Perform irradiation testingover a range of conditions (burnup, temperature, fast neutron fluence) –Perform post-irradiation examination and safety testingto demonstrate and understand performance during irradiation and during accident conditions –Develop fuel performance modelsto better predict fuel behavior –Perform fission product transport experiments to improve understanding and refine models of fission product transport

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Laterally Modulating Carrier Concentration by Ion Irradiation in CdO Thin Films for Mid‐IR Plasmonics

This report demonstrates tunable carrier densities in CdO thin films through local ion irradiation, providing lateral control of mid‐IR optical properties. Ion‐solid interactions produce donor‐like defects that boost electron concentrations from the practical minimum of 2.5 × 10 19 cm −3 to a maximum of 2.5 × 10 20 cm −3 by metered ion exposure. This range is achieved using He, N, Ar, or Au ions at 1–2.8 MeV; when normalized by displacements per atom, all ion species produce comparable results. Since CdO is well‐described by the Drude model, irradiation‐tuned carrier densities directly alter the infrared dielectric function, and in turn, mid‐infrared optical properties. Further, it is demonstrated that by combining irradiation with traditional lithography, CdO films expose to ions in the presence of 3‐µm thick, patterned photoresist exhibit lateral carrier density profiles with ≈400‐nm resolution. Scanning near‐field optical microscopy reveals sharp optical interfaces with almost no companion contrast in surface morphology, microstructure, or crystallinity. Finally, CdO lateral homostructures supporting surface plasmon polaritons (SPPs) are demonstrated whose dispersion relation can be tuned through periodic patterning in a monolithic platform by simple nanofabrication. Numerical simulations show these polaritons result from strong coupling between excitations at CdO plasma frequencies and SPPs supported by the platinum substrate.

cadmium oxide↗
DOE OSTI2025

Effects of Galactic Irradiation on Thermal and Electronic Transport in Tungsten

The impact of irradiation on the thermal and electronic properties of materials is a persistent puzzle, particularly defect formation at the atomic and nanoscales. This work examines the nanoscale effects of low-energy irradiation on tungsten (W), focusing on defect-induced modifications to thermal and electronic transport. Using the site-projected thermal conductivity (SPTC) method [A. Gautam et al. PSS-RRL, 2400306, 2024], bulk and twin-grain boundary W with vacancy defects based on the Norgett–Robinson–Torrens displacements per atom model is analyzed. SPTC provides a detailed prediction of postcascade spatial thermal conductivity distribution. Electronic conductivity activity is estimated using the “N 2 method” [K. Nepal et al. Carbon, 119711, 2025] to explore the consequences of vacancies and grain boundaries, highlighting the defect-dependent nature of charge transport behavior. These findings offer high-resolution insights into irradiation-driven transport phenomena, with implications for space-exposed materials and nanoscale thermal/electronic management.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2024

Evolution of dislocation loops in irradiated α-Uranium: An atomistically-informed cluster dynamics investigation

An atomistically-informed mean field cluster dynamics model has been developed to investigate the nucleation and growth of prismatic loops in irradiated a-Uranium. TEM analysis of neutron irradiated a-Uranium shows the evolution of self-interstitial atom and vacancy loops on (010) and (100) crystallographic planes, respectively, resulting in an anisotropic lattice swelling of its face-centered orthorhombic crystal. To provide model parameters, the crystallography of loops and the binding energy of point defects to these loops were studied using an angular dependent EAM potential and classical molecular dynamics (MD) simulations. Furthermore, using the bond-boost hyperdynamics method, the anisotropic diffusion of interstitials and vacancies in a-Uranium was investigated. Here, the mechanisms of point defect diffusion and the associated migration energies were reported and compared with previous DFT studies. The energetics and kinetic quantities mentioned above were used in the cluster dynamics model to predict the cluster density as a function of dose rate, dose and temperature and the results were compared to the reported neutron irradiation experiments. The model predictions reveal an accumulation of small sized vacancy loops along with a population of large and growing self-interstitial loops, which closely corresponds to the TEM observations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Formation of carbon homonuclear bonds in β-SiC under neutron irradiation at various temperatures and neutron doses

To elucidate radiation defect processes in SiC, Raman spectroscopy was systematically applied to high-purity, polycrystalline β-SiC that was neutron irradiated at a range of temperature and dose conditions. The analysis specifically focused on formation of carbon homonuclear bonds by irradiation; these bonds were indicated by D and G bands and amorphous carbon peaks. Intensity of the carbon peaks relative to SiC peaks significantly decreased in the case of high temperature and/or high neutron dose of 500 °C to 29 displacements per atom (dpa) and about 800 °C to 1.38 and 29 dpa. The absence of carbon bond peaks under those conditions was explained by growth of stoichiometric defect clusters, consistent with previous atomistic simulations on SiC defect stability. The lack of Raman bands associated with carbon clusters under high-temperature and high-dose radiation conditions accounts for the resistance of SiC to phase separation under irradiation. The findings further suggest that material compositions and chemical properties that are inherently resistant to chemical disordering under high-dose radiation conditions are indicative of the long-term durability of ceramic compounds in radiation environments.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Fast neutron irradiation capability in existing thermal test reactors

In today’s nuclear industry, momentum towards the design, licensing, and construction of advanced nuclear demonstration plants, including fast reactors, is at a remarkably high level. However, there are currently no dedicated fast spectrum irradiation test facilities in the United States to support the development of fast spectrum technologies. As a result, a unique situation is developing where most of these plants will likely be designed by leveraging historic nuclear material technologies, but where the further optimization and advancement is impeded by the lack of fast neutron irradiation test facilities. While these circumstances present a challenge, there are some near-term opportunities that, if seized, can still help develop advanced fast reactor materials to a meaningful level of readiness to support future commercial fast reactors. Here, in this paper, we assess the feasibility of using thermal neutron filtering materials in existing experiment positions in the Advanced Test Reactor (ATR) at Idaho National Laboratory and the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory to simulate fast reactor test environments for nonfuel test specimens. Items investigated include the incident neutron flux (both fast and thermal), the total neutron fluence and cumulative atom displacements, helium production rate due to thermal neutron capture in nickel, and the potential impact that the thermal neutron filter material has on the cycle length of a given reactor. It is concluded that while HFIR provides the highest fast flux of all the options investigated, it is limited in the amount of thermal neutron filtering material that can be introduced into an experiment position without significantly affecting the operation of the reactor. Irradiation in Outboard-A positions in the ATR was found to be the most realistic near-term experiment avenue due to having ample space for several capsules in a moderately fast flux.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Irradiation at Ultra-High (FLASH) Dose Rates Reduces Acute Normal Tissue Toxicity in the Mouse Gastrointestinal System

Preclinical studies using ultra-high dose rate (FLASH) irradiation have demonstrated reduced normal tissue toxicity compared with conventional dose rate (CONV) irradiation, although this finding is not universal. We investigated the effect of temporal pulse structure and average dose rate of FLASH compared with CONV irradiation on acute intestinal toxicity.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2025

Coupled cluster and dislocation dynamics modeling of microstructure evolution in irradiated materials

We develop here a coupled cluster and dislocation dynamics framework to study the microstructure evolution of irradiated materials. The framework not only accounts for the three dimensional diffusion of radiation-generated clusters, but also their interaction with dislocation networks and the resultant climb motion of discrete dislocations within finite crystals. The framework is solved with a superposition solution scheme, and is applied to investigate the evolution of the irradiation-induced dislocation loops in zirconium (Zr), considering the effects of various bias factors including the diffusion anisotropy difference (DAD) of interstitials and interstitial clusters, the dislocation bias of defects to discrete dislocation segments, and the production bias of defects from the radiation cascade. We find that the DAD is the most critical factor influencing the kinetics of the loop evolution in Zr, while the recombination/interaction of mobile defects can induce a strong spatial dependence of the loop evolution together with the DAD. Here, the method is also adopted to study the evolution of interstitial $\langle$a$\rangle$ and vacancy $\langle$c$\rangle$ dislocation loop ensembles consistent with the microstructure observed during irradiation-induced growth of Zr. Our findings not only reveal the spatial dependence of the size and ellipticity of the dislocation loops, but also suggest a limit on the anisotropy factor of interstitials to reproduce the co-growth of $\langle$a$\rangle$ and $\langle$c$\rangle$ loops in zirconium, in good agreement with experimental observations and other simulation results.

Bias factors↗
DOE OSTI2020

Thermal conductivity measurement of the interaction layer between UMo and Al produced by high-energy heavy ion irradiation

In this work, we report the first direct thermal conductivity measurement results for Al-UMo interaction layer (IL), which is typically observed in UMo/Al dispersion fuel plates under irradiation. The investigated IL was formed by irradiating Al coated UMo substrate using 80 MeV iodine ions at 180 degrees C up to 3.03 x 10 17 ions/cm 2 fluence. Microstructural characterization indicated that the induced IL is amorphous with an approximately (U 0.8 ,Mo 0.2 )Al 5.3 stoichiometry, which is similar to that formed under in-pile irradiation. Focused ion beam (FIB) was used to prepare nine specimens of various lengths from the IL that could be suspended across a microfabricated device for thermal conductivity measurement. The measured thermal conductivity values of the IL were significantly lower than the values for both the original UMo fuel and the Al. The successful measurement of the Al-UMo IL provides valuable information for the development and qualification of UMo/Al dispersion fuels for research and test reactor conversion applications, and further demonstrates the promising capabilities of utilizing the suspended bridge method in nuclear fuel research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Cluster dynamics modeling of irradiation growth in single crystal Zr

Irradiation growth of single-crystal Zr is modeled with a reduced set of cluster dynamics equations. Nucleation and growth of basal and prismatic dislocation loops are both accounted for in the model. Equations are developed for the time evolution of single point defects, a limited number of point defect clusters, interstitial and vacancy loop nucleation rates, as well as for the growth of vacancy loops on basal planes and interstitial loops on prismatic planes. Reduction of the usual infinite hierarchy of cluster dynamics equations to the simple set studied here is justified on the physical basis of the stability of small loops once nucleated. This simplified cluster dynamics model with a small number of adjustable parameters avoids the complexity of explicit representation of higher order point defect clusters. The model shows consistency with experimental observations of the following aspects: (1) the growth rates of Zr crystals along the a- and c-axes; (2) the onset dose for breakaway irradiation growth; (3) the saturation dislocation loop densities of vacancy and interstitial loops; (4) the effects of cold work and temperature on irradiation growth.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiation effects on Al 0.3 CoCrFeNi and CoCrMnFeNi high-entropy alloys, and 316H stainless steel at 500 °C

To evaluate the potential of high entropy alloys (HEAs) for nuclear applications, two HEAs, Al 0.3 CoCrFeNi and CoCrFeMnNi, and a conventional Type 316H stainless steel (SS) were irradiated with 1 MeV krypton ions at 500 degrees C up to 1 dpa, and examined in-situ with a transmission electron microscope (TEM). After irradiation, a high density of ordered L1 2 nanoparticles was observed in Al 0.3 CoCrFeNi. In contrast, no phase transformation was observed in CoCrFeMnNi and 316H SS. In the thin foil regions of TEM samples, stacking-fault tetrahedra were observed in the HEAs. In the thick foil regions, interstitial dislocation loops were observed for all alloys, and the areal loop density increased linearly with foil thickness. The Al 0.3 CoCrFeNi had the largest loop size (the lowest density), followed by the CoCrFeMnNi and then the 316H SS. The higher loop nucleation rate in the 316H SS was attributed to carbon content. Finally, the degree of irradiation hardening was slightly lower for the HEAs than for the 316H SS, which is a promising sign for the nuclear application of HEAs at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Dislocation loop evolution in F/M steel T91 under in-situ ion irradiation: Influence of the presence of initial dislocations

Ferritic/Martensitic (F/M) steel T91 was irradiated in-situ to 4 and 10 dpa at 470 °C using 1 MeV Kr 2+ . Additionally, the microstructure evolution under irradiation was followed and characterized by in-situ Transmission Electron Microscopy (TEM), in terms of dislocation loop Burgers vector, average size and density as a function of dose. Furthermore, the influence of the presence of dislocations in the pre-irradiated matrix was investigated by comparing the radiation induced defect evolution in grains with and without initial dislocations.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiation-induced swelling of pure chromium with 5 MeV Fe ions in the temperature range 450–650 °C

A surface coating using pure chromium has been proposed to increase the accident tolerance of Zircaloy cladding in pressurized water reactors. However, there is not much irradiation experience with Cr alloys and especially pure Cr. In the present study, pure chromium was irradiated with 5 MeV Fe ions to 50 peak dpa (displacements per atom) at temperatures of 450, 500, 550, 600 and 650 °C. Then irradiation at the peak swelling temperature of 550 °C was conducted to 50, 100, 150 peak dpa. Swelling at 50 dpa was observed over the entire temperature range studied, 450–650 °C, but appeared to be decreasing strongly at the temperature boundaries of the experiment. After an initial transient of rapid swelling, chromium was observed to swell at a rate of ~0.03–0.04%/dpa (up to 120 local dpa), which is much lower than pure Fe at 0.2%/dpa. This low swelling rate was found to be relatively insensitive to dpa rate, which varied by a factor of ~2 over the depth of data collection. Here, swelling was observed to begin quickly with an incubation period less than 10 dpa. Self-organization in the form of void ordering was observed to be developing at 50 dpa, becoming better defined with increasing dose. The void alignment direction is determined to be the $\langle$111$\rangle$ axial direction.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Effects of helium on irradiation response of reduced-activation ferritic-martensitic steels: Using nickel isotopes to simulate fusion neutron response

Understanding the effects of helium on microstructures and mechanical properties of reduced-activation ferritic-martensitic steels is important to use of these steels in fusion reactor structures. The 9Cr-2WVTa steels were doped with 58Ni and 60Ni isotopes at 2 weight percent to control the rate of transmutation helium generation. The samples were irradiated in the High Flux Isotope Reactor to ~24 displacements per atom at nominal temperatures of 300, 400, and 500°C, producing 228 and 7 atomic parts-per-million helium in the 58 Ni- and 60 Ni-doped samples, respectively. Transmission electron microscopy revealed a variety of precipitates and the radiation-induced dislocation loops and cavities (voids or helium bubbles). Tensile tests of the irradiated samples at the irradiation temperatures showed radiation-induced hardening at 300°C and radiation-induced softening at 400°C. Analysis indicates that the hardening primarily originated from the loops and cavities. The 58 Ni-doped samples had greater strengthening contributions from loops and cavities, leading to higher hardening with lower ductility than the 60 Ni-doped samples. The greater helium production of 58 Ni did not show pronounced reductions in ductility of the samples.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Formation of multicomponent alloy particles in doped ceria under I 2+ ion irradiation and thermal annealing

Multicomponent alloys resembling the e-phase particles in nuclear fuel have the potential to immobilize 99 Tc in a stable waste form. Formation of such multicomponent alloys under extreme conditions is studied in a non-activated model system: CeO 2 doped with 2 wt% Mo, 1.5 wt% Ru, 0.75 wt% Pd, 0.5 wt% Re and 0.25 wt% Rh. The doped CeO 2 films were irradiated with I 2+ ions (610°C, 10 16 and 5×10 16 I 2+ /cm 2 ). For selected samples post-irradiation heat treatment was conducted (900°C, 1100°C). Analytical TEM revealed the formation of 5-20 nm precipitates containing Pd, Mo, and Re sometimes associated with cavities. Iodine was found to segregate to grain boundaries and cracks. After heat treatment at 1100°C, the CeO 2 matrix had recrystallized. Precipitates containing Pd, Mo, Re, and Ru were observed near the interface with the polycrystalline YSZ substrate. It follows that Pd is the most mobile element, followed by Mo and Re. Finally, while irradiation promotes precipitation, high-temperature effects are more significant.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Heavy ion irradiation response of an additively manufactured 316LN stainless steel

Additive manufacturing has become an appealing technique to fabricate three-dimensional metallic materials and components for nuclear reactors. However, response of additively manufactured alloys to high-dose heavy ion irradiations at elevated temperatures is still not well understood. Here, an additively manufactured 316LN austenitic stainless steel with high-density solidification cells was irradiated using 3.5 MeV Fe ion to a peak dose of 220 dpa at 450 °C. Microscopy studies show a lower Frank loop density and smaller size in the additively manufactured sample compared with its cold worked counterpart, and the cellular structures may largely suppress the formation of perfect loops and dislocation networks and reduce the magnitude of solute segregations comparing with high angle grain boundaries. In conclusion, the present work advances the understanding on the high-temperature irradiation response of additively manufactured steels for nuclear reactor applications.

36 MATERIALS SCIENCE↗