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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

A Brief Review of the Impact of Neutron Irradiation Damage in Tungsten and Its Alloys

Neutron irradiation poses a substantial challenge in the development and application of tungsten (W) and its alloys, predominantly in the framework of nuclear fusion and fission environments. Although W is well-acknowledged for its unique properties like its high melting temperature and higher resistance to sputtering, transmutation products, such as Re and Os, form and impact the alloy properties as a result of neutron irradiation. This transmutation effect accompanied by significant microstructure damage due to neutron irradiation can lead to the significant degradation of mechanical properties. This review surveys the literature focusing on the microstructural modifications post-irradiation and its impacts on the irradiation hardening. This review provides insights into the elaborative understanding on the neutron radiation damage on W and W alloys by exploring the microstructural evolution and hardness changes post-irradiation. The gaps and future opportunities for understanding neutron radiation damage in W are briefly summarized.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Interpretation of Ion Irradiation and Neutron Irradiation Damage in Additively Manufactured 316 Stainless Steel using Multiscale Modeling

The accelerated adoption of nuclear energy necessitates advanced manufacturing technologies, such as additive manufacturing, to meet heightened supply chain requirements and support innovative reactor technologies. Due to the unique microstructural characteristics of additively manufactured materials under distinct solidification conditions, comprehensive evaluation of their performance in reactor environments is essential. The Advanced Materials and Manufacturing Technologies program under the Department of Energy's Office of Nuclear Energy focuses on understanding the irradiation performance and damage evolution of laser powder bed fusion 316 stainless steel, with an emphasis on integrating ion and neutron irradiation data to accelerate the development and qualification of materials for advanced nuclear reactor applications. While ion irradiation is a cost- and time-effective method, modeling and simulation are required to interpret the data for the broader range of irradiation conditions encountered in advanced reactors. In fiscal year 2025, integrated multiscale modeling and simulations were conducted to assess irradiation damage in additively manufactured 316 stainless steel. Key outcomes include predictions of chromium enrichment at grain boundaries, nickel enrichment at dislocation cell walls and void surfaces, and heterogeneous void evolution under ion and neutron irradiation conditions. Cluster dynamics simulations revealed the coarsening of voids at high irradiation temperatures and the suppression of void growth by high network dislocation density, while also demonstrating significant growth and coarsening of voids and self-interstitial atom loops at low dose rates. Machine learning-accelerated atomistic simulations highlighted the impact of the local environment and chromium concentration on vacancy diffusivity, providing key insights on the influence of composition on void swelling and radiation-induced segregation. Additionally, molecular dynamics simulations demonstrated the presence of defect production bias and a significant effect of carbon content on defect cluster behavior. These combined efforts aim to predict the performance of additively manufactured materials under various reactor conditions, supporting their qualification for nuclear reactor applications by interpreting ion irradiation data. This report underscores the potential of integrated multiscale modeling to analyze ion irradiation data in the effort to accelerate the qualification of additively manufactured materials for nuclear reactor components.

316 stainless steel↗
DOE OSTI2021

Emulation of neutron damage with proton irradiation and its effects on microstructure and microchemistry of Zircaloy-4

This work assesses the potential of proton irradiation to simulate the neutron damage to the matrix and laves phase Zr(Fe,Cr) 2 precipitates in Zircaloy-4. Isothermal proton irradiation has been performed on Zircaloy-4 samples at irradiation temperatures ranging from 250 to 350 °C. Two-step proton irradiation was also performed to enhance the amorphization of and iron loss from the laves phase Zr(Fe,Cr) 2 precipitates. The irradiated microstructures, including dislocation loops and rafts near SPPs, were observed in proton irradiated Zircaloy-4, which are consistent with neutron irradiated material at a similar damage level. The amount of irradiation-induced hardening after proton irradiation was similar to post neutron irradiated data. The significant amorphization of the SPPs and concurrent Fe redistribution observed on neutron irradiated materials can be effectively emulated using a two-step proton irradiation on Zircaloy-4. Hence, the neutron irradiation effect on Zircaloy-4 can be mostly captured using the two-step proton irradiation described in this study.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Damage development in neutron-irradiated concrete in a test reactor: Hygro-thermal and mechanical simulations

This paper reports the development of a 3D mesoscale hygro-thermal-mechanical simulation approach to predict damage in concrete irradiated in a test reactor. This framework, developed in MOOSE, considers the effects of elevated temperature, moisture content, and high neutron fluence (energy threshold, E > 0.1 MeV) on the mortar and aggregates separately. The first-stage simulation implements hygro-thermal analysis to determine the temperature and RH inside the specimen as a function of imposed radiation energy. These are used as inputs to the second stage, which considers radiation-induced volumetric expansion (RIVE) of aggregates, and creep, shrinkage, and stress-strain response of mortar to predict the expansion, stresses, and damage in specimens made using different coarse aggregates and subjected to different irradiation times. The irradiation time-dependent damage in the mortar is expressed using an isotropic damage parameter. This multi-physics model serves as a predictive tool for damage quantification in concrete due to neutron irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Neutron capture of UO 2 targets prepared by spin-coating assisted combustion synthesis

Two uranium dioxide (UO 2 ) targets of (414 ± 23) nm and (1092 ± 93) nm thicknesses were prepared on 6061 aluminum alloy and puratronic grade aluminum backing materials. The targets were deposited with a novel method combining spin coating and solution combustion synthesis (SCS). The target layers consisted of small (3–7 nm) UO 2 grains and uniformly distributed ultra-small (1–3 nm) pores. The prepared targets were tested at the Los Alamos National Laboratory’s LANSCE facility for neutron irradiation damage and suitability for neutron capture experiments. The samples showed no signs of target material loss after the irradiation. However, irradiation caused a significant increase in the grain size (4–10 nm), as well as upward mass diffusion and coalescence of the pores due to the thermal spikes. The magnesium in the aluminum 6061 alloy backing also diffused into the UO 2 layer during neutron irradiation. The structural changes in the target after the irradiation do not affect the data from neutron capture. As a result, the new method can be used more broadly to prepare other actinide targets for nuclear physics experiments.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Nanoscale lattice strains in self-ion implanted tungsten

Developing a comprehensive understanding of the modification of material properties by neutron irradiation is important for the design of future fission and fusion power reactors. Self-ion implantation is commonly used to mimic neutron irradiation damage, however an interesting question concerns the effect of ion energy on the resulting damage structures. The reduction in the thickness of the implanted layer as the implantation energy is reduced results in the significant quandary: Does one attempt to match the primary knock-on atom energy produced during neutron irradiation or implant at a much higher energy, such that a thicker damage layer is produced? In this work, we address this question by measuring the full strain tensor for two ion implantation energies, 2 MeV and 20 MeV in self-ion implanted tungsten, a critical material for the first wall and divertor of fusion reactors. A comparison of 2 MeV and 20 MeV implanted samples is shown to result in similar lattice swelling. Multi-reflection Bragg coherent diffractive imaging (MBCDI) shows that implantation induced strain is in fact heterogeneous at the nanoscale, suggesting that there is a non-uniform distribution of defects, an observation that is not fully captured by micro-beam Laue diffraction. At the surface, MBCDI and high-resolution electron back-scattered diffraction (HR-EBSD) strain measurements agree quite well in terms of this clustering/non-uniformity of the strain distribution. However, MBCDI reveals that the heterogeneity at greater depths in the sample is much larger than at the surface. This combination of techniques provides a powerful method for detailed investigation of the microstructural damage caused by ion bombardment, and more generally of strain related phenomena in micro-volumes that are inaccessible via any other technique.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effects of neutron irradiation on hydrogen isotope permeation on W-Re

Hydrogen and deuterium permeation behavior for neutron damaged W and W-3 %Re was evaluated using a Plasma Driven Permeation device. For neutron damaged W-3 %Re at 773 K and 1473 K, precipitation of Re and Os was observed by atom probe tomography. Neutron irradiation introduced defect damage throughout the sample and enhanced H and D diffusion toward downstream side. The existence of irradiation damage by neutron irradiation was found to increase hydrogen isotope permeation. Moreover, the existence of Re has suppressed the formation of irradiation defects for hydrogen isotope trapping sites. Further, the recovery of irradiation defects by increasing the irradiation temperature led to the structural changes due to the diffusion of Re and Os, which also affected the hydrogen isotope permeation behavior.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Design, construction, and testing of no-insulation small subscale solenoids for compact tokamaks

Fusion energy systems studies (FESS) for next-step devices based on the most promising magnetic configurations indicate that high magnetic fields and high current density for magnet coil systems may reduce device size and lower the cost. High current density and radiation resistant fusion magnets are particularly beneficial for low cost, low aspect ratio compact reactor designs such as Fusion Nuclear Science Facility (FNSF), Fusion Pilot Plant (FPP) of low-aspect ratio spherical tokamak (ST) or compact stellarators. Unlike typical high field magnets for nuclear magnetic resonance (NMR), magnetic resonance imaging (MRI), or high energy physics research applications, neutron irradiation damage to organic insulations in the coil winding pack is a critical issue for next generation fusion reactors where orders of magnitude higher neutron fluence than those in present experimental reactors such as ITER are expected. Moreover, a high coil winding pack current density is needed for high field magnets in low cost, compact radial build next step fusion reactors. The slow current charging time, however, is an issue in fully non-insulated coils. In this work, we present the design, construction and testing of subscale Nb3Sn solenoids, up to half the diameter of the central solenoid (CS) for National Spherical Torus Experiment (NSTX), with and without inter-layer insulation for enhancing radiation resistance while improving the winding pack current density and coil performance. The major radius for the NSTX/NSTX-Upgrade (reference for compact ST) is 0.854/0.934 meter, and 4.8 meters for FNSF. The magnetic field for NSTX/NSTX-U at the plasma center is 0.6/1 Tesla and 9 Tesla for FNSF. The inner diameter of the central solenoid coil is 0.2/0.4 meter for NSTX/NSTX-U and 1.2 meters for FNSF. The current charging and discharging behavior of the prototype solenoids was investigated to quantify performance advantages of a simplified coil fabrication without error-prone vacuum pressure impregnation (VPI) for the removal of epoxy organic insulation. Scalability of the no insulation concept for fusion can be addressed via engineered coating for novel insulation on contact resistance or intra-layer no insulations. Radiation resistance and coil winding efficiency were significantly improved in the no insulation coils for next step compact fusion reactors.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2021

Technological aspects in blanket design: Effects of micro-alloying and thermo-mechanical treatments of EUROFER97 type steels after neutron irradiation

Presently available data on neutron irradiation damage raise doubts on the feasibility of using EUROFER97 steel for a water-cooled starter blanket in a DEMO reactor, since the ductile-to-brittle transition temperature (DBTT) increases significantly for irradiation temperatures below 350°C. The additional DBTT shift caused by H and He transmutation can only be estimated based on very few results with isotopically tailored EUROFER97 steel. Conservative calculations show that the DBTT of EUROFER97 steel could exceed the operating temperature in water-cooled starter blankets within a relatively short time period. This paper presents results from a EUROfusion funded irradiation campaign that was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. The paper compares ten newly developed reduced activation ferritic-martensitic (RAFM) steels irradiated to a nominal dose of 2.5 dpa at 300°C. The post-irradiation experiments using Small Specimen Test Technology included hardness, tensile, and fracture mechanics tests combined with fractography and microstructure analysis are presented. Results show that micro-alloying EUROFER97-type steels influenced the mechanical properties but a dominating impact on irradiation damage resistance could not be identified. In contrast, specific thermo-mechanical treatments lead to better DBTT behavior. Discussion about irradiation response to heat treatment conditions is also given. Here, despite requiring data also at high dpa values, the results indicate that with these modified materials an increased lifetime and potentially also an increased operating temperature window can be achieved compared to EUROFER97.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Novel effects of grain size and ion implantation on grain boundary segregation in ion irradiated austenitic steel

Nuclear reactor environments are extreme in nature due to the combination of exposure to corrosive coolant, mechanical stresses, and neutron irradiation damage. This combination of environments can lead to irradiation induced/assisted precipitation, segregation, and embrittlement coupled with enhanced creep and stress corrosion cracking. While nanostructured materials are of interest in a variety of applications due to their unique properties including high mechanical strength, for nuclear applications nanostructured materials have the added benefit of irradiation resistance due to the high volume density of sinks for irradiation-induced defects. In this study, we show unique radiation-induced segregation (RIS) behavior in nanocrystalline 304 stainless steel (SS304). Typical RIS in austenitic steels includes Ni enrichment and Cr depletion within the grain boundaries. In nanostructured SS304 this RIS is reduced. Furthermore, interestingly, some grain boundaries in the nanocrystalline SS304 in this study were found to be enriched in Cr after Fe 2+ ion irradiation at 500 °C up to 50 displacements per atom (dpa). To understand this unique observation, lattice-based atomic kinetic Monte Carlo simulations were performed to investigate the effects of grain size, thermal segregation, as well as injected interstitials (introduced as Fe self-ions). The results indicate that the Cr enrichment at grain boundaries may be caused by combined effects of grain size, self-ion injection and preferential Cr diffusion via the interstitial mechanism. In conclusion, the findings also suggest a possible difference in RIS between ion and neutron irradiations.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Equibiaxial Flexure Strength of a Superfine-Grained Nuclear Graphite

The strength of advanced graphite is reported in accordance with ASTM D7846-16 using a two-parameter Weibull distribution for uniaxial strength testing. The rule of thumb for Weibull distributions is to use a minimum of 30 strength measurements to have a high level of confidence in the Weibull characteristic strength and the Weibull modulus. These large sample sets for statistical confidence are easily obtained for as-manufactured graphite, but determination of the Weibull two-factor parameters on graphite that has been exposed to neutron irradiation is nearly impossible due to irradiation testing restraints on specimen size and space limitations in test reactors for the accommodation of specimens. These restrictions have resulted in irradiation programs that use sub-sized specimens to measure uniaxial strength change, but the specimen geometries still hinder the number of replicate specimens for each irradiation condition. In contrast, this work used an advanced ceramic standard, ASTM C1499-05, as the foundation for equibiaxial strength testing of a superfine-grained nuclear graphite to investigate the irradiation-induced strength change and any changes to the two-parameter Weibull distribution. The goal was to investigate the effect of specimen size on the two-parameter Weibull distribution parameters in support of the small specimens and geometry desired for irradiation. A study of the effect of a reduced sample population was also undertaken to investigate the reliability of using the two-parameter Weibull distribution to study changes to the materials properties caused by neutron irradiation damage. The results of this work suggest that these equibiaxial specimens would be ideal for statistically significant studies of the effects of irradiation on the Weibull characteristic strength (but not the modulus) and for providing a method for surveillance specimen campaigns for future operating commercial reactors.

Campbell, Anne↗
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 OSTI2019

Neutron irradiation-induced microstructure damage in ultra-high temperature ceramic TiC

Titanium carbide (TiC) is an ultra-high temperature ceramic with potential as a structural material candidate for advanced reactor concepts. However, the irradiation tolerance of TiC is not well understood. Here, we reveal the key irradiation damage microstructure degradation processes in TiC using mixed spectrum neutron irradiations at dose of ~2 displacements per atom (dpa) at temperatures of ~220, 620, and 1115 °C, combined with state-of-art microstructure characterization using transmission electron microscopy (TEM) and high-resolution TEM (HRTEM). The helium injection (~65 atomic parts per million) produced by neutron transmutation also occurred in the sample. TiC was observed to form irradiation-induced interstitial-type dislocation loops and He-stabilized cavities. At 220 and 620 °C, the analysis of the electron diffraction patterns, rel-rod imaging and HRTEM revealed that the dislocation loops were faulted Frank loops with Burgers vector b Frank = 1/3<111> lying on {111} planes. A detailed Burgers vector identification performed by the g.b technique revealed that the dislocation loops forming at 1115 °C were unfaulted, edge-type, with Burgers vectors b = a<100> and a/2<110> with corresponding {100} and {110} habit planes. Using continuum mechanics, we estimated the critical radius at which a faulted dislocation loop transitions to a perfect loop to be 9 nm. Further, no amorphization occurred in TiC under-investigated irradiation conditions while macroscopic swelling under point defect swelling regime was observed. He-stabilized cavities were detected at 1115 °C, but not at lower temperature irradiation. These observations indicate the onset of long-range vacancy migration in TiC occurs between 620 and 1115 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Microstructure and mechanical properties of friction stir weld performed on neutron-irradiated 304L steel with helium

As nuclear power plants (NPPs) approach or exceed 40–60 years of service, it may become necessary to repair damaged neutron-irradiated components to prolong longevity. However, fusion welding repair of irradiated steels and metallic materials is challenging because of helium-induced cracking. This study used friction stir welding (FSW) to address helium-related issues, such as helium bubble formation and grain boundary cracking. The microstructure, helium-induced degradation, and mechanical properties of a friction stir weld produced on neutron-irradiated 304L stainless steel with approximately 5.2 appm of helium were characterized. The analysis focused on variations in grain size, texture, and the morphology of helium-induced damage in the friction stir weld. Mechanical properties were characterized on the irradiated base metal (BM) and metallurgical zones of the friction stir weld: the stir zone (SZ), the thermo-mechanically affected zone (TMAZ), and the heat affected zone (HAZ). Only minor scattered porosity in the SZ and TMAZ and a few short microcracks (below 20 μm in length) in the TMAZ were observed, indicating limited helium-induced degradation. Tensile tests revealed good mechanical properties and fractography analysis demonstrated predominantly ductile fracture. In conclusion, the results highlight the immediate and substantial benefits of the FSW approach for repairing or joining helium-containing irradiated materials in NPPs.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A Study of the Radiation Tolerance of CVD Diamond to 70 MeV Protons, Fast Neutrons and 200 MeV Pions

We measured the radiation tolerance of commercially available diamonds grown by the Chemical Vapor Deposition process by measuring the charge created by a 120 GeV hadron beam in a 50 μm pitch strip detector fabricated on each diamond sample before and after irradiation. We irradiated one group of samples with 70 MeV protons, a second group of samples with fast reactor neutrons (defined as energy greater than 0.1 MeV), and a third group of samples with 200 MeV pions, in steps, to (8.8±0.9) × 1015 protons/cm2, (1.43±0.14) × 1016 neutrons/cm2, and (6.5±1.4) × 1014 pions/cm2, respectively. By observing the charge induced due to the separation of electron–hole pairs created by the passage of the hadron beam through each sample, on an event-by-event basis, as a function of irradiation fluence, we conclude all datasets can be described by a first-order damage equation and independently calculate the damage constant for 70 MeV protons, fast reactor neutrons, and 200 MeV pions. We find the damage constant for diamond irradiated with 70 MeV protons to be 1.62±0.07(stat)±0.16(syst)× 10−18 cm2/(p μm), the damage constant for diamond irradiated with fast reactor neutrons to be 2.65±0.13(stat)±0.18(syst)× 10−18 cm2/(n μm), and the damage constant for diamond irradiated with 200 MeV pions to be 2.0±0.2(stat)±0.5(syst)× 10−18 cm2/(π μm). The damage constants from this measurement were analyzed together with our previously published 24 GeV proton irradiation and 800 MeV proton irradiation damage constant data to derive the first comprehensive set of relative damage constants for Chemical Vapor Deposition diamond. We find 70 MeV protons are 2.60 ± 0.29 times more damaging than 24 GeV protons, fast reactor neutrons are 4.3 ± 0.4 times more damaging than 24 GeV protons, and 200 MeV pions are 3.2 ± 0.8 more damaging than 24 GeV protons. We also observe the measured data can be described by a universal damage curve for all proton, neutron, and pion irradiations we performed of Chemical Vapor Deposition diamond. Finally, we confirm the spatial uniformity of the collected charge increases with fluence for polycrystalline Chemical Vapor Deposition diamond, and this effect can also be described by a universal curve.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Statistical analysis of displacement damage in small devices from neutron and ion irradiation

Modern semiconductor devices, such as gate-all-around nanosheet field-effect transistors (GAA NS FETs), are smaller than displacement damage cascades from fission neutrons. In this regime, device failure may occur through low-probability single events, rather than by parametric degradation previously seen in larger devices. Here, we present a statistical model that predicts the probability of a damage event in a small device and the probability distribution of the magnitude, i.e., number of displacements within the device, from each event. The model is developed first for neutron irradiation and then for energetic ion irradiation. The model is consistent with results from recent experiments in which lithium-ion irradiation produced stepwise increases in subthreshold current in GAA NS FETs.

Wampler, W. R.↗