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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

6cr

A series of model Fe–Cr alloys containing 3–18 at.% Cr was neutron irradiated at a nominal temperature of 563 K to 1.82 dpa. Solute distributions were analyzed by atom probe tomography, which revealed α′ precipitation for alloys containing more than 9 at.% Cr. Both the Cr concentration dependence of α′ precipitation and the measured matrix compositions are in agreement with the recently published Fe–Cr phase diagrams. An irradiation-accelerated precipitation process is strongly suggested. Irradiation was carried out in the Advance Test Reactor (ATR) at Idaho National Laboratory. A series of six Fe–Cr alloys of nominal compositions 3, 6, 9, 12, 15 and 18 at.% Cr was irradiated at a neutron fluence (E > 1 MeV) of 1.1 × 1021 n cm−2 at 563 ± 15 K and to a damage level of 1.82 displacements per atom (dpa). Nominal neutron flux and dpa rate are 2.3 × 1014 n cm−2 s−1 and 3.4 × 10−7 dpa s−1, respectively. The microstructures of the Fe–Cr alloys were studied by atom probe tomography (APT) using a Cameca 4000X HR instrument. APT specimens were prepared by a standard lift-out process using a Quanta 3D 200i dual beam scanning electron microscope ensuring that the analyses were performed away from grain boundaries.

Bachhav, Mukesh↗
DOE OSTI2022

Comparison of Irradiated and Unirradiated Graphite Oxidation Performance

This work examines the oxidation behavior of NBG-25 graphite using irradiated specimens and (not irradiated) companion specimens from the Advanced Graphite Creep (AGC) Experiment. Irradiated and companion specimens were quartered, split into four small samples, to enable four oxidation test runs for each (0.5 inch diameter by 0.25 inch tall) piggyback button obtained from the AGC Experiment inventory. These split samples were oxidized in air in a thermogravimetric analyzer (TGA) and benchmarked against observations with separately sourced same-grade specimens of three geometries oxidized either in the TGA or in a vertical furnace built and operated to satisfy the specifications of ASTM D7542. The study considers both irradiation damage and relief of damage by thermal annealing. A range of isothermal oxidation temperatures were tested for specimens exposed to a similar irradiation environment, nominally 6.5 dpa at 650°C. To assess the dose dependency of observed oxidation behavior, specimens with a range of irradiation exposures (up to 6.8 dpa) were tested at the single oxidation temperature of 650°C. In the conventional analysis (rate determined over the 5-10% mass loss range) any annealing effects appear to be negligible, while irradiation to ~6.5 dpa may double or triple the subsequent oxidation rate. However, examination of rate with extent of reaction (over incremental ranges from initial onset up to 5% mass loss) illustrates that (at least for ~6.5 dpa) irradiation initially inhibits oxidation. Slower rates of oxidation than companion split samples are clearly indicated across all oxidation temperatures tested up to ~0.5% mass loss. Over the range of 0.5-1% mass loss, regardless of normalization strategy, there is no statistically meaningful difference in oxidation rate. However, beyond 0.5% mass loss oxidation of the irradiated sample becomes progressively faster than the companion sample (up to the 10% mass loss level observed).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Investigation of irradiation damage and heat deposition: a comparative analysis for HEU-to-LEU conversion in HFIR

The planned conversion of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel requires detailed evaluation of experiment-relevant parameters to ensure continued performance for materials testing and isotope production. Here, this study presents the first comprehensive assessment of displacements per atom (dpa) and heat deposition rates in target materials within the HFIR flux trap with both HEU and candidate LEU core configurations. Seven analyses were conducted to evaluate key performance metrics, including fast neutron flux distribution, cross section response functions, cross section data, and local dpa and heat deposition rates using mesh- and cell-based tallies. Simulations employed Shift, Monte Carlo N-Particle (MCNP), and the HIFR Controller (HFIRCON) tool suite for high-fidelity transport and depletion modeling. The LEU designs—using U 3 Si 2 -Al dispersion fuel and operating at 95 MW—were compared to the current 85 MW HEU configuration. Results show that while the candidate LEU cores exhibit higher dpa rates due to a harder spectrum and extended cycle lengths, they also demonstrate reduced heat deposition rates in irradiation experiments, primarily due to increased gamma self-shielding from higher 238 U content in the core. These findings confirm that LEU conversion can maintain HFIR’s materials irradiation capabilities but may require redesigning existing experimental hardware.

HEU↗
DOE OSTI2022

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Thermal diffusivity recovery and defect annealing kinetics of self-ion implanted tungsten probed by insitu transient grating spectroscopy

Tungsten is a promising candidate material for plasma-facing armour components in future fusion reactors. A key concern is irradiation-induced degradation of its normally excellent thermal transport properties. In this comprehensive study, thermal diffusivity degradation in ion-implanted tungsten and its evolution from room temperature (RT) to 1073 K is considered. Five samples were exposed to 20 MeV self-ions at RT to achieve damage levels ranging from 3.2 x 10 -4 to 3.2 displacements per atom (dpa). Transient grating spectroscopy with in-situ heating was then used to study thermal diffusivity evolution as a function of temperature. Using a kinetic theory model, an equivalent point defect density is estimated from the measured thermal diffusivity. The results showed a prominent recovery of thermal diffusivity between 450 K and 650 K, which coincides with the onset of mono-vacancy mobility. After 1073 K annealing samples with initial damage of 3.2 x 10 -3 dpa or less recover close to the pristine value of thermal diffusivity. For doses of 3.2 x 10 -2 dpa or higher, on the other hand, a residual reduction in thermal diffusivity remains even after 1073 K annealing. Transmission electron microscopy reveals that this is associated with extended, irradiation-induced dislocation structures that are retained after annealing. A sensitivity analysis shows that thermal diffusivity provides an efficient tool for the assessing total defect content in tungsten up to 1000 K.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Degradation of electrical resistivity of tungsten following shielded neutron irradiation

A major challenge for heat transfer in nuclear materials is to ensure thermal mobility after high amounts of neutron irradiation. Tungsten is widely selected as a heat transfer material in fusion reactors. In metals, thermal conductivity is dominated by electrons’ ability to transfer energy. Neutron irradiation generates point defects, clusters, and solid transmutation (e.g.rhenium and osmium in tungsten), which inhibit electron motion. The purpose of this work is to quantify the irradiation-induced change in electron mobility and deconvolute transmutation and microstructural effects on observed changes to electron mobility. Single and polycrystalline tungsten were fast neutron irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory to doses between 0.2 and 0.7 displacements per atom (dpa) and temperatures from 500 °C to 1000 °C. Grain growth was observed in all samples. Microstructure and transmutation were quantified. The geometric orientation of samples with elongated grains has been shown to affect electrical resistivity. A mathematical model was developed and used to deconvolute solid-solution transmutation, grain, and temperature-dependent lattice effects on resistivity. At ~0.4 dpa at ~590 °C, the combined resistivity degradation due to voids, vacancies, interstitials, and dislocations is estimated to be greater than the contribution from solid solution Re transmutation, which is greater than the contribution from grain boundaries. At doses of ~0.7 dpa at ~750 °C, solid solution Re contributions are greater than all other effects combined. As a result, this work establishes a basis to predict the effects of irradiation temperature and transmutation on thermal properties of tungsten and highlights the importance of irradiation temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2025

MX precipitate behavior in an irradiated advanced Fe-9Cr steel: Helium effects on phase stability

As part of an ongoing series aimed at optimizing Fe-9Cr reduced activation ferritic/martensitic (RAFM) alloys for fusion energy systems, this study explores MX precipitate behavior under dual-ion irradiations, specifically examining correlations between helium transmutation and irradiation-induced damage. Utilizing single and dual-beam ion irradiation, the research explores the combined effects of helium (10–25 appm He/dpa), temperature (400–600 °C), and damage levels (15–100 dpa) on the microstructural evolution of CNA9 steel, a variant of Castable Nanostructured Alloys (CNAs). The study demonstrates that helium co-implantation hinders radiation-enhanced coarsening of MX-TiC precipitates at 500 and 600 °C, maintaining MX-TiC precipitate stability at moderate damage levels (15 dpa) but failing to prevent complete precipitate dissolution at higher damage levels (≥50 dpa) when irradiated at 500 °C. Here, a generalized precipitate stability model suggests that helium-induced suppression of diffusion alters the balance between recoil resolution and back diffusion for MX-TiC precipitates, enhancing the current understanding of precipitate behavior under damage and transmutation simulated dual-ion irradiation conditions.

Characterization↗
DOE OSTI2021

Radiation effects on stress evolution and dimensional stability of large fusion energy structures

Here we assess the effects of neutron irradiation on the deformation and stress evolution of large-scale fusion energy structures. This is accomplished through non-linear finite element structural analysis of the coupled thermal and mechanical fields at the Beginning-Of-Life (BOL), at 45 dpa, and at 90 dpa. Radiation effects include volumetric swelling and the influence of radiation on the mechanical properties. The system studied here is a large section of a full inboard module of an integrated structure comprising the First Wall and Blanket (FW/B) of a Dual Cooled Lithium-Lead (DCLL) energy conversion unit in the Fusion Nuclear Science Facility (FNSF). The structural material is the ferritic/martensitic steel F82H. We analyze several radiation effects phenomena that can lead to significant impact on the mechanical design and lifetime of the structure. These include volumetric swelling effects on the assembly and disassembly of modules and the impact of spatial gradients in neutron damage and helium generation on deformation, stress evolution, and plastic strain accumulation. We show that the stress state is strongly influenced by the accumulation of swelling strains, radiation hardening and softening, and by spatial gradients in displacement damage and helium gas generation. Several key regions in the FW/B structure are identified where severe plastic strains accumulate and may be potential sites for failure, especially during cyclic reactor operations. We finally show that minimum gaps between Inboard (IB) blanket sectors must be present during assembly so as to avoid excessive stress and strain in FW/B modules critical regions and raising chances of failure. A range of structural displacements are shown, from as low as 4 mm at BOL for free side walls to as high as 46 mm at 90 dpa for fully constrained side walls.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

In-situ microstructure observation of oxidized SiC layer in surrogate TRISO fuel particles under krypton ion irradiation

In accidental scenarios of high temperature gas-cooled reactors, both oxidation of and irradiation to the SiC layer in tri-structural-isotropic (TRISO) fuel particles can change the microstructure and integrity of the fuel elements. Here, in the present study, microstructure and defect evolution in the oxidized SiC layer of surrogate TRISO fuel particles under 1.2 MeV krypton ion irradiation was observed by in-situ transmission electron microscopy. The SiC layers oxidized in water vapor at 1200 °C were irradiated azt room temperature and 800 °C and at damage levels of 0.28–11.2 dpa, respectively. SiC and SiO 2 were found to still be in their crystal structures at the damage level of 11.2 dpa at 800 °C, while SiC was observed to have been amorphized at only 0.56 dpa irradiation at room temperature. The defect number density at 800 °C was an order of magnitude lower than that in the sample irradiated at room temperature. Also, crystalline SiO 2 had higher radiation resistance compared to SiC. A defect reaction rate theory was utilized to understand the fundamental defect evolution process and irradiation resistance difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

In-situ TEM study of Kr ion irradiation tolerance of SiFeOC nanocomposite

Here, in this work, ion irradiation of polymer derived SiFeOC nanocomposite was carried out using 1.2 MeV Kr ions at room temperature and 600°C. The starting composite was composed of Fe 3 Si, SiC, SiOC, SiO 2 , and graphitic C. In-situ TEM investigations show uniform distribution of nano-crystalline Fe 3 Si and SiC phases in the amorphous SiOC matrix. During ion irradiation, the SiOC bulk microstructure and interfaces between Fe 3 Si or SiC crystallites and the SiOC matrix remain defect-free and demonstrate outstanding ion irradiation resistance. At room temperature, the crystalline domains are stable up to 2 dpa. At 600°C, Fe 3 Si crystallites are more stable than SiC; SiC crystallites are stable up to 4 dpa while the Fe 3 Si crystallites are stable up to 10 dpa. These crystallites also coalescence and amorphize simultaneously during ion irradiation. The exceptional tolerance to defect formation and irradiation of the SiFeOC nanocomposite provides important guidance to developing irradiation resistant fuels for advanced gas cooled reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Layer dissolution in accumulative roll bonded bulk Zr/Nb multilayers under heavy-ion irradiation

In this work, the heavy-ion irradiation behavior of bulk zirconium-niobium multilayered composites was investigated up to large doses. Multilayers with an average individual layer thicknesses ranging between 15 and 80 nm were synthesized by accumulative roll bonding technique. The irradiation was performed with a defocused 7 MeV Zr 2+ ion beam at 500 °C. The maximum dose achieved was ~145 dpa at the depth of ~1.5 μm from the irradiated surface. Sub-surface microstructural damage and the chemical redistribution were characterized by transmission electron microscopy and energy dispersive spectroscopy, respectively. Irrespective of the layer thicknesses, the irradiation condition caused layer instability and the extent of damage varied with the dose levels. Doses lesser than ~60 dpa caused layer fragmentation and greater than ~60 dpa resulted in layer dissolution. The chemical mixing of layers occur to a depth of ~1 μm, consuming multiple bi-layer periods. Despite the elevated irradiation temperature (500 °C) and a slightly positive heat of mixing (+6 kJ/mol), no phase separation was observed and single-phase was retained in the mixed region. The results demonstrate that chemical mixing was facilitated by the liquid phase miscibility of Zr and Nb, which propelled the interdiffusion within the thermal spikes towards mixing.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Radiation induced amorphization of carbides in additively manufactured and conventional ferritic-martensitic steels: In-situ experiments on extraction replicas

In-situ irradiations using 1 MeV Kr 2+ ions in a transmission electron microscope were performed on extraction replica samples containing M 23 C 6 carbides and MX carbonitrides from two ferritic-martensitic (FM) steels: (i) 9%Cr-1%W-TaV based Eurofer97 and (ii) 9%Cr-1%Mo-VNb (all in wt.%) based additively manufactured (AM) Grade91 steel. The irradiations were performed between 100 and 773 K up to a maximum dose of ~2.4 displacements per atom (dpa). The M 23 C 6 carbides are highly susceptible to radiation induced amorphization (RIA), while the MX type nanoprecipitates are highly amorphization resistant across the entire irradiation temperature range. Between 100 and 423 K, RIA of M 23 C 6 carbides occurs very rapidly with critical amorphization doses ranging between ~0.35 and 0.9 dpa, increasing to higher doses at 573 K. Complete amorphization of the M 23 C 6 carbides up to doses of ~2.4 dpa is not possible at 773 K under the present irradiation conditions. The critical amorphization dose of M 23 C 6 carbides increases nearly exponentially with irradiation temperature. The critical temperature for the crystalline-to-amorphous phase transformation (T c→a ) of M 23 C 6 carbides irradiated as replica samples, i.e. without the surrounding metal matrix, was estimated to be ~812 K. Comparing the present results with neutron irradiation data on bulk samples reveals a decrease in the critical amorphization dose for M 23 C 6 and an increase of T c→a , highlighting the effect of dose rate on the amorphization behavior that is qualitatively consistent with literature on other non-metals. Changes in the minor chemistry of M 23 C 6 , such as presence or absence of W, V, Nb, Mo, seems to have little effect on the amorphization behavior. Here, in-situ irradiations on extraction replica samples provide a novel pathway to explore radiation tolerance of nanoprecipitates in nuclear structural materials

36 MATERIALS SCIENCE↗
DOE OSTI2022

Emulation of neutron-irradiated microstructure of austenitic 21Cr32Ni model alloy using dual-ion irradiation

Here, in this study, the capability of heavy-ion irradiation to emulate neutron irradiation was demonstrated on an austenitic 21Cr32Ni type ternary model alloy. The model alloy used in this study is chemically analogous but compositionally simpler than of alloy 800H, which is a candidate austenitic Ni alloy which has been proposed for use in Generation IV reactors. The microstructure of the 21Cr32Ni model alloy irradiated in the BOR-60 fast reactor to 17.1 dpa and 35 dpa at ~380°C was characterized using transmission electron microscopy (TEM). The 17.1 dpa BOR-60 irradiated microstructure was then compared with the microstructure of the same material developed under dual-ion (DI) irradiation using various He/dpa ratios between 0.1 and 16.6 appm He/dpa in the temperature range of 430°C-500°C. The results showed that both neutron and DI irradiation of 21Cr32Ni model alloy produced dislocations in the form of a dislocation network as well as {111}-type faulted dislocation loops, cavities, and radiation-induced Ni enrichment at radiation-induced sinks. When the dose and the He/dpa ratio were kept similar to those in neutron irradiation, DI irradiation of the 21Cr32Ni model alloy at 460°C resulted in over-nucleation of small cavities and in a high density of faulted dislocation loops compared to those observed in the fast-neutron irradiated alloy of the same heat irradiated at ~380°C. The optimal condition for reproducing the neutron-irradiated microstructure was DI irradiation at 460°C and 0.1 appm He/dpa. In that case, the faulted loop and cavity size distributions in the BOR-60 irradiated 21Cr32Ni model alloy samples closely matched with those measured in the DI irradiated 21Cr32Ni model alloy sample. The fact that the He/dpa is an order of magnitude smaller than the helium generation rate for fast neutron irradiation, stops over nucleation and allows for the development of a similar microstructure as for neutron irradiation.

21Cr32Ni model alloy↗
DOE OSTI2022

Heavy ion irradiation effects on CrFeMnNi and AlCrFeMnNi high entropy alloys

Co-free but Al-included medium/high entropy alloys (M/HEAs) have gained increasing interests due to their lower cost and the potential to tune the multi-phase microstructure. The irradiation response of two Co-free HEAs, face-centered cubic (FCC) CrFeMnNi with limited Cr enriched α' phase and body-centered cubic (BCC) AlCrFeMnNi with B2s phase and nanoprecipitates were explored. Ion irradiations using 5 MeV Fe 2+ ions were performed at 500°C to a peak fluence of 50 and/or 100 displacements per atom (dpa). In dual-phase AlCrFeMnNi, there was no significant radiation induced segregation or chemical intermixing at the coherent matrix (FeCrMn-rich)/second phase (AlNi-rich) boundaries. In CrFeMnNi, limited voids were only detected at the peak damage location of ~ 50 dpa. On the other hand, voids were widely distributed in AlCrFeMnNi: under 50 and 100 dpa irradiation conditions, voids were found with larger dimension and denser distribution in the FeCrMn-rich matrix, smaller and slightly lower density in an AlNi-rich second phase. In addition, the diameter of the FeCMn-rich nanoprecipitates didn't reveal any tendency of dissolution or growth. This is correlated with their superior structural stability against irradiation. Significant radiation-induced hardening (increases from 3.8 ± 0.2 GPa to 4.7 ± 0.6 GPa) was measured in CrFeMnNi, but only ~ 4% hardness increase (from 7.4 ± 0.8 GPa to 7.7 ± 0.4 GPa) was noted in AlCrFeMnNi. Finally, in addition to the radiation-induced defects, such as voids, dislocation loops and point defects, other factors, such as chemical short-range ordering may play an important role.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Comparative characterization of mixed spectra and thermal neutron shielded irradiated tungsten

The effect of mixed spectra and thermal neutron shielded irradiation on tungsten was evaluated with plasma exposure in the tritium plasma experiment followed by thermal desorption spectroscopy, X-ray photoelectron spectroscopy, and transmission electron microscopy. The two different irradiation campaigns were performed at the High Flux Isotope Reactor to 0.39–0.74 displacement per atom (dpa) in the 894–1379 K temperature range. A neutron spectrum influence on the void size and void number density was not observed. However, a strong correlation was found between void size and void number density with temperature, but not with dpa in the limited dpa range of this study. Thermal neutron shielding significantly reduced the transmutation to Re+Os. Higher irradiation temperature will lead to larger voids with lower number density, which reduces deuterium retention. In conclusion, grain growth was also observed for high-temperature irradiation of over ~1300 K within the limited grains visible in the transmission electron microscopy specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Discerning the effect of various irradiation modes on the corrosion of Zircaloy-4

Using proton irradiation, this study investigates the individual influence of several factors on the corrosion kinetics of Zircaloy-4 in a hydrogenated water environment simulating a Pressurized Water Reactor (PWR). Using both simultaneous irradiation-corrosion and autoclave corrosion, we separately examine: (i) the effect of pre-irradiation on modifying the structure of the material, (ii) the impact of irradiation on creating defects in the growing oxide layer during corrosion, and (iii) the influence of irradiation on increasing the corrosion potential through radiolysis during corrosion. To replicate neutron-irradiated microstructure, two proton pre-irradiation schedules were employed: Schedule 1 (isothermal irradiation at 350 °C to 5 dpa) to simulate high-temperature PWR conditions, and Schedule 2 (two-step process: irradiation to 2.5 dpa at -10 °C followed by 2.5 dpa at 350 °C) to simulate lower temperature PWR and Boiling Water Reactor (BWR) conditions. Long-term autoclave corrosion testing over 360 days at 320 °C revealed no significant difference between unirradiated samples and those pre-irradiated according to either schedule, with all samples exhibiting sub-cubic kinetics within the pre-transition regime. Irradiated samples underwent Simultaneous Irradiation Corrosion (SIC) tests, corroding in 320 °C water while being irradiated with protons. Corrosion was found to accelerate in all SIC-tested samples relative to autoclave conditions, with the greatest increase observed in non-pre-irradiated samples. Pre-irradiation with either schedule resulted in a slower corrosion rate compared to non-pre-irradiated regions under SIC conditions. The degree of radiolysis observed in the SIC tests surpassed typical PWR conditions, approaching levels found in BWRs. Radiolysis products were identified as the primary contributors to accelerated corrosion, corroborated by radiolysis bar tests. Furthermore, these findings underscore the intricate interactions between irradiation, corrosion, and water chemistry in determining Zircaloy-4 corrosion kinetics within nuclear reactor environments.

Pressurized Water Reactor↗
DOE OSTI2024

Radiation damage study of POCO ZXF-5Q graphite for neutrino production targets using 4.5 MeV helium ions

To address the challenges of increased beam power and target survivability associated with next-generation particle production beam lines, high dose, high-energy proton beam conditions are simulated using irradiation from low-energy ion beams. A low-energy ion irradiation study of POCO ZXF-5Q graphite under conditions similar to those of the NuMI NT-02 neutrino production target at the Fermi National Accelerator Laboratory is reported. Helium ion irradiation was performed at 100 ∘ C to a maximum damage level of 0.9 displacements per atom (DPA). Irradiation induced hardening, swelling of the irradiated region, inter-plane lattice expansion, and intraplane lattice contraction with increasing ion fluence was observed using micromechanical (nanoindentation, atomic force microscopy) and electron microscopy (high-resolution imaging, selected area diffraction) characterization. Similar changes were also observed in post irradiation examination of the NT-02 target indicating that ion irradiation can be a valuable tool for estimating radiation damage in proton beam targets. Caution must be exercised though, because the hardening, lattice alteration, and swelling occur to different magnitudes for a given damage level. The observed hardening and embrittlement were greater for ion irradiated graphite. For He ion irradiated samples the lattice spacing changes were smaller at low damage levels (78% less expansion and 71% less contraction at 0.1 DPA) and larger at high damage levels (38% more expansion and 5% more contraction at 0.9 DPA) relative to that observed in the NT-02 target. The magnitude of swelling was 8.5× greater under ion irradiation which is influenced by the differing damage gradients and inclusion of implanted He ions in the region of interest.

43 PARTICLE ACCELERATORS↗