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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Effects of Proton Irradiation on GaN Vacuum Electron Nanodiodes

Gallium nitride (GaN)-based nanoscale vacuum electron devices, which offer advantages of both traditional vacuum tube operation and modern solid-state technology, are attractive for radiation-hard applications due to the inherent radiation hardness of vacuum electron devices and the high radiation tolerance of GaN. Here, we investigate the radiation hardness of top-down fabricated n-GaN nanoscale vacuum electron diodes (NVEDs) irradiated with 2.5-MeV protons (p) at various doses. We observe a slight decrease in forward current and a slight increase in reverse leakage current as a function of cumulative protons fluence due to a dopant compensation effect. The NVEDs overall show excellent radiation hardness with no major change in electrical characteristics up to a cumulative fluence of 5E14 p/cm 2 , which is significantly higher than the existing state-of-the-art radiation-hardened devices to our knowledge. In conclusion, the results show promise for a new class of GaN-based nanoscale vacuum electron devices for use in harsh radiation environments and space applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Neutronics Analysis of Shielding Material Alternative to Tungsten Carbide in the FESS-FNSF Facility

Neutron transport calculations have been performed on advanced shielding materials. Metal hydrides and borohydrides were evaluated to find an alternative to tungsten carbide (WC), which is used in the in-vessel components. The study was conducted using a 22.5° sector and a detailed 360° geometry of the Fusion Energy System Studies-Fusion Nuclear Science Facility (FESS-FNSF) using OpenMC and FENDL-3.2b nuclear data library. Here, the neutronics analysis in this article was concentrated on calculating total nuclear (neutron and photon) heating at the magnet and the radiation damage of the inboard vacuum vessel (IBVV). For example, yttrium hydride (YH 2 ) and vanadium hydride (VH 2 ) showed lower radiation damage values compared to WC and other metals. Whereas alternative shielding materials did not show a significant change in the magnet nuclear heating.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Performance evaluation of pulse shape discrimination capable organic scintillators for space applications

Scintillators with pulse-shape discrimination (PSD) capability are of great interest to many fields in the scientific community. The ability to discern a gamma ray from a neutron using PSD varies between different types of scintillator materials and dopants. Here a new generation of organic scintillator materials with PSD capability were studied to determine their radiation hardness to neutron and gamma-ray radiation. The PSD capability, average pulse shapes, and light output of four types of organic scintillator were characterized before and after neutron and gamma-ray irradiation. The main goal of this investigation is to study the effects of long-term irradiation that may be experienced in space applications on the light output and particle discriminating capabilities of each material. EJ-270, EJ-276, organic glass, and Stilbene were tested. Damage due to neutron irradiation (displacement damage) was not observed in any of the scintillators up to 2.56 x 10 11 n/cm 2 , except for Stilbene which showed a small (12%) decrease in light output. All scintillators presented some light output reduction after gamma-ray irradiation (total ionizing dose), with reductions of 17% (EJ-276 and OGS), 32% (EJ-270), and 42% (Stilbene) observed immediately after 100 kRad.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Radiation-Induced Modifications in Copper Oxide Growth

Radiation-induced effects and their influence on oxidation processes were evaluated for their use as a forensic tool for special nuclear material (SNM). A beam of 10 MeV Au³⁺ ions was used to mimic the accumulation of microstructural damage from self-irradiation through the decay of radionuclides. Several copper samples were irradiated as suitable surrogate materials at 200 °C with a flux of 1x10¹² ions/cm·s to damage levels of 5, 10, and 15 displacements per atom (dpa). This corresponds to about 50, 100, and 150 years, respectively, of accumulated α-decay damage in a PuGa alloy assuming a damage rate of 0.1 dpa/year. After irradiation, all samples were exposed to an accelerated aging process induced by thermal treatment at 350 °C for 1 hour in air. This resulted in the growth of a mixed oxide layer (Cu₂O and CuO) which was characterized in detail using several complementary analytical techniques: Scanning Electron Microscopy, Raman spectroscopy, Synchrotron X-ray diffraction (transmission mode), and Grazing Incidence X-ray diffraction. The oxide layer growth of irradiated Cu at 350°C is distinctly modified as a result of the ion irradiation. Most notably, the growth of the CuO phase is suppressed with increasing radiation damage on the Cu substrate, and structural changes occurred in the Cu₂O phase. These results indicate that damage from self-irradiation over time can cause quantifiable modifications in the oxidation process of metals that could be harnessed for their use as a novel forensic tool.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Effect of sink strength on coherency loss of precipitates in dilute Cu-base alloys during in situ ion irradiation

In situ irradiations with 1 MeV Kr ions at 50~613 K up to a fluence of 6.25 × 10 14 ions/cm 2 (~1.25 displacements per atom, dpa) have been performed on pre-aged dilute Cu-0.9%Co, Cu-0.9%Fe and Cu-0.8%Cr alloys containing uniform matrix dispersions of coherent precipitates in order to study the effects of initial precipitate sink strength, damage dose and irradiation temperature on radiation-induced coherency loss of precipitates. Coherent precipitates with different point defect sink strengths (2πNd, where N and d are the precipitate density and diameter) were used in this work to examine potential differences in atomic relaxation during absorption of point defects. In all cases, irradiation to low doses (<~1 dpa) was very effective at inducing loss of precipitate coherency. At low sink strengths (~10 13 m -2 ), loss of precipitate coherency could be induced for doses ~0.01 dpa. This suggests there might be an efficient preferential medium-range strain-induced bias for absorption of interstitial defects due to tensile strains emanating from the undersized precipitates, which induces relatively rapid loss of coherency at low precipitate sink strengths. High precipitate sink strength (~10 14 m -2 ) conditions were relatively resistant to the radiation-induced loss of coherency (requiring higher doses approaching ~1 dpa) and this might be due to nearly equal numbers of interstitial and vacancy defects arriving at the precipitate interface for such high sink strength conditions. The precipitate coherency loss was observed to have a weak dependence on irradiation temperature. Molecular dynamics simulations confirm a strong effect of precipitate sink strength on the probability of interstitial absorption at precipitates.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Radiation Effects Related to Repaired BWR Core Shrouds

The material properties of the core shroud and supporting tie rod components are important factors for Boiling Water Reactor (BWR) life extension considerations. Given the variability of nickel and boron content found in 304 stainless steel shroud plates, and XM-19 and Inconel X-750 tie rod materials, this paper combines the latest spectral information in the vicinity of the core shroud to calculate hydrogen and helium gas production limits as well as atomic displacement damage spanning 80 full-power years (FPY) using the SPECTER code. Results indicate that, whereas boron has little effect on the atomic displacement damage, helium production from boron can be important and must be included in weldability assessments. Since the boron concentration is uncertain, helium generation is calculated assuming a composition range based on specifications and reported measurements. The results show that helium production caused by transmutations in the 304 SS shroud material are dominated by the concentration of boron and will have a significant impact on weldability after only a few years of service. For tie rods, the susceptibility to irradiation-assisted stress corrosion cracking (IASCC) has been assessed by determining helium production in Inconel X-750 IASCC specimens irradiated in the Advanced Test Reactor (ATR) reactor and correlating with the helium produced in a BWR. As a tie rod material in a high thermal neutron flux environment, Inconel X-750 is more likely to be IASCC sensitive during service in a BWR due to helium than the XM-19 stainless steel alloy because of a greater nickel content and boron concentration. Weldability of the core shroud plate material, IASCC of the tie rod materials, and the gamma photon contribution to radiation damage are discussed.

radiation damage in materials, BWR↗
DOE OSTI2026

Comparison of DeePMD, MTP, GAP, ACE and MACE Machine‐Learned Potentials for Radiation‐Damage Simulations: A User Perspective

Accurate and efficient interatomic potentials are essential for molecular dynamics (MD) simulations of radiation damage, gas diffusion, and phase stability in complex ceramics such as LiAlO 2 , especially under extreme conditions relevant to tritium production. Here, we evaluate the performance of six machine-learned interatomic potentials (MLIPs), moment tensor potential (MTP), Gaussian approximation potential, deep potential (DeePMD), atomic cluster expansion (ACE), message-passing ACE (multilayer atomic cluster expansion (MACE) pretrained) and MACE (trained from-scratch), all trained on the same density functional theory dataset with inclusion of tritium. The MLIPs are benchmarked against traditional Buckingham and ReaxFF potentials in terms of energy accuracy, density predictions, thermal equilibration behavior, threshold displacement energy (E d ), tritium diffusivity, and computational cost. Among the models, MTP shows the best overall balance between efficiency and accuracy, with low force and energy errors and realistic E d values for Li and Al. The ACE and MACE (pretrained and trained from scratch) models exhibit high E d (>200 eV) and unphysical pair interactions. DeePMD underestimates Ed due to overly repulsive behavior even at equilibrium distances. All models over-estimate tritium diffusion but the pretrained MACE model behaves well during tritium-diffusion simulations up to 500 K, maintaining diffusivities in the physically consistent 10 −11 m 2 /s range. Finally, we quantify the computational cost of each potential in large-scale atomic/molecular massively parallel simulator, finding that only MTP is more efficient than traditional empirical potentials, while others are significantly more expensive. These findings explain the trade-offs between accuracy and computational cost in MLIP development and provide essential guidance for use in high-throughput radiation damage and gas diffusion simulations in nuclear ceramics.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2025

Proton radiation effects in indium oxide using cascade molecular dynamics simulations

Metal oxide (MO) semiconductors, characterized by their wide band gaps and notable charge transport properties, are promising candidates for electronic applications in extreme environments, including near-Earth space. However, atomistic simulations of radiation–matter interactions in MOs remain challenging due to the limitations of existing interatomic potentials, which often fail to capture both the short-range repulsive forces essential for radiation damage modeling and the long-range electrostatic effects governing defect evolution. In this work, we develop a customized interatomic potential tailored for radiation damage simulations in indium oxide (In 2 O 3 ) as a model system, a representative MO material. Our potential integrates the Ziegler-Biersack-Littmark potential to accurately describe short-range interactions with Buckingham and Coulombic potentials to account for long-range forces. We perform molecular dynamics simulations of low-energy proton irradiation using this custom potential. We employ the primary knock-on atom (PKA) cascade method to study atomic displacements and primary defect formation. Simulations were conducted for 1 keV proton irradiation in four randomly chosen directions, and PKA-driven defect analyses at 5, 10, and 15 keV to examine the effects of direction and energy level on damage generation. Our results provide insight into the impact of irradiation direction and energy level on the cascade evolution and defect formation mechanisms.

Atomistic simulations↗
DOE OSTI2022

Effect of differently oriented interlayer phases on the radiation damage of Inconel-$\mathrm{Ni}$ multimetallic layered composite

Multimetallic layered composites (MMLCs) have shown an excellent potential for application under extreme environments, e.g., accident-tolerant fuel cladding, because of their low oxidation tendency and high corrosion resistance. Interfacial phases or complexions in nanocrystalline materials accelerate the annihilation of defects and enhance the radiation resistance of materials, making MMLCs with engineered interlayer phases compelling to deploy in extreme conditions. However, implementation of MMLCs in full capacity remained a challenge due to a lack of fundamental understanding of the underlying mechanisms governing the characteristics of the interface between the metallic layers. The precise role of interlayer phases in MMLCs and their interaction with defects, specifically under extreme conditions, is still unexplored. Pursuing atomistic simulations for various Inconel-Ni MMLCs model materials, we revealed accelerated defect mobility in interlayers with larger crystalline misorientation and the inverse relationship between the interface sink strength to the misorientation angle. Furthermore, we found a linear relation between interlayer misorientation angle with the density of radiation-induced defects and radiation enhanced displacements. So, our results indicate that radiation-induced material degradation is accelerated by the higher defect formation tendency of MMLCs with a high-angle interlayer interface.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Damage buildup in multilayer hexagonal boron nitride films under Ar ion bombardment

Ion irradiation can be used to modify properties of two-dimensional hexagonal boron nitride (2D-hBN). However, the behavior of multilayer 2D-hBN under ion bombardment remains essentially unexplored. Here, in this study, we systematically study the damage buildup in 40-nm-thick multilayer 2D-hBN films irradiated at room temperature with 500 keV Ar ions with doses up to 0.64 displacements per atom. Transmission electron microscopy, soft X-ray absorption spectroscopy, and Raman spectroscopy are employed to study effects of ion irradiation. Results reveal a monotonic buildup of damage that can be described by the phenomenological zero cascade overlap model. Compared to the behavior of other group-III nitride semiconductors, radiation damage in hBN is influenced by polymorphism, involving transitions between bonding configurations of hexagonal, cubic, and amorphous phases.

Seo, M. [Lawrence Livermore National Laboratory (L↗
DOE OSTI2022

Implications of Microstructure in Helium-Implanted Nanocrystalline Metals

Helium bubbles are known to form in nuclear reactor structural components when displacement damage occurs in conjunction with helium exposure and/or transmutation. If left unchecked, bubble production can cause swelling, blistering, and embrittlement, all of which substantially degrade materials and—moreover—diminish mechanical properties. On the mission to produce more robust materials, nanocrystalline (NC) metals show great potential and are postulated to exhibit superior radiation resistance due to their high defect and particle sink densities; however, much is still unknown about the mechanisms of defect evolution in these systems under extreme conditions. Here, the performances of NC nickel (Ni) and iron (Fe) are investigated under helium bombardment via transmission electron microscopy (TEM). Bubble density statistics are measured as a function of grain size in specimens implanted under similar conditions. While the overall trends revealed an increase in bubble density up to saturation in both samples, bubble density in Fe was over 300% greater than in Ni. To interrogate the kinetics of helium diffusion and trapping, a rate theory model is developed that substantiates that helium is more readily captured within grains in helium-vacancy complexes in NC Fe, whereas helium is more prone to traversing the grain matrices and migrating to GBs in NC Ni. Our results suggest that (1) grain boundaries can affect bubble swelling in grain matrices significantly and can have a dominant effect over crystal structure, and (2) an NC-Ni-based material can yield superior resistance to irradiation-induced bubble growth compared to an NC-Fe-based material and exhibits high potential for use in extreme environments where swelling due to He bubble formation is of significant concern.

36 MATERIALS SCIENCE↗
NASA NTRS2010

Method for Detecting Perlite Compaction in Large Cryogenic Tanks

Perlite is the most typical insulating powder used to separate the inner and outer shells of cryogenic tanks. The inner tank holds the low-temperature commodity, while the outer shell is exposed to the ambient temperature. Perlite minimizes radiative energy transfer between the two tanks. Being a powder, perlite will settle over time, leading to the danger of transferring any loads from the inner shell to the outer shell. This can cause deformation of the outer shell, leading to damaged internal fittings. The method proposed is to place strain or displacement sensors on several locations of the outer shell. Loads induced on the shell by the expanding inner shell and perlite would be monitored, providing an indication of the location and degree of compaction.

Youngquist, Robert↗
DOE OSTI2020

Radiation Effects in Graphite

Graphite has been a core structural material and neutron moderator since the first man-made chain reaction experiment in Chicago. This article will discuss the requirements of what makes graphite a good solid moderator. The manufacture and properties of some currently available near-isotropic and isotropic grades are described. The major features of a graphite moderated reactors are briefly outlined. Displacement damage and the induced structural and dimensional changes in graphite are described. Recent characterization work on nano-carbons and oriented pyrolytic graphites that have shed new light on graphite defect structures are reviewed, and the effect of irradiation temperature on the defect structures is highlighted. Changes in the physical properties of nuclear graphite caused by neutron irradiation are reported. Finally, the importance of irradiation induced creep is presented, along with current models and their deficiencies.

Campbell, Anne↗
DOE OSTI2020

From suppressed void growth to significant void swelling in NiCoFeCr complex concentrated solid-solution alloy

Void swelling can result in dimensional instability and undermine the safe operation of nuclear reactors. Current strategies to inhibit void swelling mainly focus on enhancing defect absorption and recombination by introducing high-density defect sinks. Complex concentrated solid-solution alloys (CSAs), including high-entropy alloys, can withstand severe radiation damage due to their inherent chemical complexity without interfaces. However, the underlying mechanisms for void suppression in CSAs are far from clear. In this research, we studied the void evolution with respect to irradiation depths, doses, and temperatures in equiatomic NiCoFeCr under 3 MeV Ni ion irradiations. At relatively low doses (16 and 54 displacements per atom, dpa), voids form mainly outside of the ion-damaged region, and void formation in the peak damage region is suppressed, leading to negligible swelling. However, with further increase of dose (86 up to 250 dpa), significant void growth occurs in the peak damage region and extended dislocation lines dominate instead of short dislocation lines and loops formed at lower doses. From 500 to 700 °C, the dislocation density decreases while dislocations grow. Although the overall void swelling increases dramatically from 500 to 580 °C at 54 dpa, void growth in the peak damage region is still suppressed. The transition from suppressed void growth to significant void swelling is attributed to dislocation evolution and local chemical inhomogeneity (enrichment of Fe/Cr in the matrix) at higher doses. Our study shows that controlling element diffusion and defect evolution through tuning chemical complexity can further enhance the swelling resistance of CSAs.

36 MATERIALS SCIENCE↗
NASA NTRS2003

Recent radiation test results at JPL

This paper documents recent TID (including ELDRS) and proton damage test results obtained by JPL. Unusual test results, such as abnormally low or high failure levels or unusual failure or response mechanisms, are emphasized.

total ionizing dose↗
DOE OSTI2021

HFIR SiC-SiC Composite Clad Tube Bowing Test: Pre-Irradiation Characterization

This report describes the pre-irradiation characterization of the SiC-SiC composite clad tube bowing experiment that will be irradiated in the High Flux Isotope Reactor (HFIR). There are concerns that SiCSiC fuel cladding in light water reactors could undergo bowing because of the non-uniform fast neutron flux profiles. This experiment is combined with the SiC-SiC channel box experiment, which has similar concerns regarding control blade movements in boiling water reactors. The HFIR experiment aims to validate the predicted deformation and stresses in the composite clad specimen after the exposure of fast neutron flux gradients. Significant radial fast neutron flux gradients that exist in the permanent reflector of HFIR were thoroughly characterized by using detailed 3D neutronic calculations. The 3D displacement damage dose rate profile and the resulting volumetric swelling in SiC were used as inputs to structural analyses to determine the predicted deformation that will affect the specimens. The tube specimens were characterized by using traditional dimensional inspection and surface profilometry to provide detailed information regarding the pre-irradiation condition as the first step to evaluate the radiation-induced deformation. Furthermore, fine engraving markers were inscribed along all outer surfaces of the specimen and mapped by using a digital microscope and a multidimensional stage. This allows marker spacings to be accurately measured and compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. The experiment successfully completed two HFIR cycle irradiations, cycles 492 and 493, which finished on July 25, 2021.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Probing the Damage Recovery Mechanism in Irradiated Stainless Steels Using In-Situ Microcantilever Bending Test

Single crystalline microcantilevers are fabricated from the base metal and heat-affected zone (HAZ) of a laser welded, neutron irradiated austenitic stainless steel, for scanning electron microscope (SEM) in-situ bending. In the HAZ, cantilevers exhibit higher yield point and lower crack tip blunting displacement than in the base metal and unirradiated archive specimen. These results suggest that radiation-induced defects harden the base metal, whereas the HAZ exhibits annealing of defects leading to mechanical softening. Dislocation nucleation ahead of the crack tip is responsible for ductile blunting behavior and provides a pathway to mitigating helium-induced cracking during weld repairs of irradiated materials.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Simulation-based methodology to assess the lattice defects creation as energy storing process

In nuclear industry, damages caused to in-core structural materials by high-energy neutrons have been studied for years. Recently, the possibility of storing energy through the creation of defects in crystalline materials via heavy particle bombardment was investigated. To evaluate the potential of the Wigner effect as energy-storing process, a graphite-moderated Molten Salt Reactor was adopted as a test case. Neutronics simulations performed with OpenMC Monte Carlo code allowed estimating the amount of energy released during fission events, the fraction that is deposited in graphite, and the fraction that has the potential to cause radiation damages. Additionally, molecular dynamics simulations were then performed to study the evolution of cascades triggered by high-energy neutrons. As a major outcome, a more accurate estimate of the radiation damage with respect to the traditional approaches was obtained, and the fraction of the deposited neutron energy that can be stored through long-lived defects was quantified.

25 ENERGY STORAGE↗