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At least 271 records · Page 15

Effect of proton irradiation temperature on persistent photoconductivity in zinc oxide metal-semiconductor-metal ultraviolet photodetectors

The electrical and structural characteristics of 50-nm-thick zinc oxide (ZnO) metal-semiconductor-metal ultraviolet (UV) photodetectors subjected to proton irradiation at different temperatures are reported and compared. The devices were irradiated with 200 keV protons to a fluence of 1016 cm−2. Examination of the x-ray diffraction (XRD) rocking curves indicates a preferred (100) orientation prior to irradiation, with decrease in crystal quality afterward. Additionally, peak shifts in XRD and Raman spectra of the control sample relative to well-known theoretical positions are indicative of tensile strain in the as-deposited ZnO films. Shifts toward theoretical unstrained positions are observed in the irradiated films, which indicates partial relaxation. Raman spectra also indicate increase in oxygen vacancies (VO) and zinc interstitial defects (Zni) compared to the control sample. Additionally, transient photocurrent measurements performed on each sample at different temperatures showed up to 2× increase in photocurrent decay time constants for irradiated samples vs the control. This persistent photoconductive behavior is linked to the activation of electron and hole traps near the surface, and to the desorption and reabsorption of O2 molecules on the ZnO surface under the influence of UV light. Using an Arrhenius model, trap activation energies were extracted and, by comparing with known energies from the literature, the dominant defects contributing to persistent photoconductivity for each irradiation condition were identified. The persistence of differences in photocurrent transients between different samples months after irradiation indicates that the defects introduced by the suppression of thermally activated dynamic annealing processes have a long-term deleterious effect on device performance.

Heuser, Thomas A. (ORCID:0000000225642400)↗

Thermal conductivity evaluation of ion irradiated Si 3 N 4 and ZrN ceramics using spatial domain thermoreflectance

Nitride ceramics have been investigated for different applications in the nuclear industry, such as space nuclear power, fusion reactor diagnostics and plasma heating, inert matrix fuels, and accident tolerant fuels. Although thermal conductivity remains one of the most important properties to track following irradiation, traditional techniques such as laser flash and xenon flash are limited to bulk sample characterization, which requires lengthy and cost-consuming neutron irradiation. This work used spatial domain thermoreflectance (SDTR) for the micrometer-scale measurement of thermal conductivity in 15 MeV Ni ion-irradiated silicon nitride and zirconium nitride from 1 to 50 dpa and 300 to 700 °C. Here, the SDTR-measured unirradiated thermal conductivity was found to be consistent with the published data on bulk samples. Electrically conductive ZrN exhibits modest reduction after irradiation which is minimal at the highest irradiation temperatures. In electrically insulating Si 3 N 4 , the reduction is more significant and unlike ZrN, the reduction remains significant even at a higher irradiation temperature. The thermal resistance evolution following irradiation was compared with lattice swelling, which was determined using grazing incidence x-ray diffraction, and radiation-induced defects were observed using transmission electron microscopy. A saturation value was observed between 15 and 50 dpa for thermal conductivity degradation in both nitride ceramics and a direct correlation with high-temperature defect recombination was observed, as well as the potential presence of additional carrier scattering mechanisms.

36 MATERIALS SCIENCE↗

Effects of irradiation damage on the hardness and elastic properties of quaternary and high entropy transition metal diborides

Multi-principal component transition metal (TM) diborides represent a class of high-entropy ceramics (HECs) that have received considerable interest in recent years owing to their promising properties for extreme environment applications that include thermal/ environmental barriers, hypersonic vehicles, turbine engines, and next-generation nuclear reactors. While the addition of chemical disorder through the random distribution of TM elements on the cation sublattice has offered opportunities to tailor elastic stiffness and hardness, the effects of irradiation-induced structural damage on the physical properties of these complex materials have remained largely unexplored. To this end, changes in the hardness and elastic moduli of a high-entropy TM diboride (Hf 0.2 Nb 0.2 Ta 0.2 Ti 0.2 Zr 0.2 )B 2 and three of its quaternary subsets following irradiation with 10 MeV gold (Au) ions to fluences of up to 6 × 10 15 Au cm -2 are investigated at the micrometer and sub-micrometer length-scales via the dispersion of laser-generated surface acoustic waves (SAW) and nanoindentation, respectively. The nanoindentation measurements show that the TM diborides exhibit an initial increase in hardness following irradiation with energetic Au ions, with a subsequent decrease in hardness following further irradiation. One quaternary composition, (Hf 1/3 Ta 1/3 Ti 1/3 )B 2 , exhibits a notable exception to the trend and continues to exhibit an increase in hardness with ion irradiation fluence. Although differences in the absolute values of the effective elastic moduli obtained from the measured SAW dispersion and nanoindentation are observed (and attributed to microstructural variations at the measurement length-scale), both techniques yield similar trends in the form of an initial reduction and subsequent saturation in the elastic modulus with increasing ion irradiation fluence. The quaternary TM diboride (Hf 1/3 Ta 1/3 Ti 1/3 )B 2 again exhibits a departure from this trend. The high-entropy TM diboride (Hf 0.2 Nb 0.2 Ta 0.2 Ti 0.2 Zr 0.2 )B 2 exhibits the greatest recovery in hardness and modulus when irradiated to high ion fluences following initial changes at low fluence, indicating superior resistance to radiation-induced damage over its quaternary counterparts. Opportunities for designing HECs with superior hardness and modulus for enhanced radiation resistance (compared to their single constituent counterparts) by tailoring chemical disorder and bond character in the lattice are discussed.

36 MATERIALS SCIENCE↗

Ion irradiation induced strain and structural changes in LiTaO 3 perovskite

LiTaO 3 crystals irradiated with 3 MeV and 1.162 GeV Au ions were studied by single crystal x-ray diffraction and Raman scattering measurements. The maximum lattice strains after 3 MeV Au ion irradiation to a fluence of 1.2 × 10 13 cm -2 were 1.2% and 0.6% along the c- and a-/b-axes, respectively. Two effects were observed in 1.162 GeV Au ion irradiated samples: (i) the (0006) and (1120) Bragg peaks were split into doublets, which suggested a subtle structural change due to slight modification of chemical composition; and (ii) the pre-damaged 1.2% lattice strain along the c-axis was relaxed to 0.9% after subsequent irradiation with 1.162 GeV Au ions, while relaxation along the a- or b-axis was not obvious. A distinct change in the Raman spectrum of the <0001> oriented LiTaO 3 crystals was observed after 1.162 GeV Au ion irradiation, but no obvious change was observed in the <1120> oriented samples or in 3 MeV Au ion irradiated samples. Strain and structural changes in crystalline LiTaO3, with or without pre-existing defects, upon ion irradiation are delineated in its responding to inelastic ionization and elastic nuclear collisions.

36 MATERIALS SCIENCE↗

Ion irradiation of iron chalcogenide superconducting films

Iron-chalcogenide FeSe x Te 1-x (Fe(Se,Te)) superconductors have attracted a great deal of interests in both fundamental physics and potential applications. While Majorana zero-modes hosted by the Fe(Se,Te) superconductors hold great promise for topological quantum computing, low-anisotropy and very high upper critical fields make Fe(Se,Te) an excellent candidate for high magnetic field and energy applications. Recently, we demonstrated a route for creating structural defects by ion irradiation in the FST films, leading to strong flux pinning. Ion irradiation into Fe(Se,Te) films significantly improved the critical current density J c over a wide field range and for all orientations with respect to magnetic field. In particular, a robust enhancement of critical temperature T c and J c was realized simultaneously in the Fe(Se,Te) film irradiated with proton. Here, we provide a review of the ion irradiated Fe(Se,Te) films, including the film growth and characteristics before and after irradiations, design of irradiation process for different ion energy and species with irradiation simulations and experimental setups that are scalable for long length Fe(Se,Te) coated conductors.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Post-Irradiation Evaluation of Eurofer97 Fracture Toughness Using Miniature Multinotch Bend Bar Specimens

In this study, we performed fracture toughness characterization of ten neutron-irradiated Eurofer97 variants using precracked miniature multi-notch bend bar (M4CVN) specimens based on the Master Curve method in the ASTM E1921 standard. The neutron irradiation was performed in the flux trap position of the High Flux Isotope Reactor (HFIR) of the Oak Ridge National Laboratory (ORNL) with the nominal irradiation temperature of 300°C and irradiation dose of 2.5 displacements per atom (dpa). Depending on the irradiation temperature and materials, we observed different degrees of irradiation hardening and embrittlement for ten Eurofer97 variants. The upper shift in the Master Curve reference temperature T0Q vs. the increase in Vickers microhardness values showed a liner relationship for only a few materials indicating different irradiation responses of the Eurofer97 variants.

Chen, Xiang↗

Aggressive Mammary Cancers Lacking Lymphocytic Infiltration Arise in Irradiated Mice and Can Be Prevented by Dietary Intervention

Because the incidence of breast cancer increases decades after ionizing radiation exposure, aging has been implicated in the evolution of the tumor microenvironment and tumor progression. In this paper, we investigated radiation-induced carcinogenesis using a model in which the mammary glands of 10-month-old BALB/c mice were transplanted with Trp53-null mammary tissue 3 days after exposure to low doses of sparsely ionizing γ-radiation or densely ionizing particle radiation. Mammary transplants in aged, irradiated hosts gave rise to significantly more tumors that grew more rapidly than those in sham-irradiated mice, with the most pronounced effects seen in mice irradiated with densely ionizing particle radiation. Tumor transcriptomes identified a characteristic immune signature of these aggressive cancers. Consistent with this, fast-growing tumors exhibited an immunosuppressive tumor microenvironment with few infiltrating lymphocytes, abundant immunosuppressive myeloid cells, and high COX-2 and TGFβ. Only irradiated hosts gave rise to tumors lacking cytotoxic CD8 + lymphocytes (defined here as immune desert), which also occurred in younger irradiated hosts. These data suggest that host irradiation may promote immunosuppression. To test this, young chimera mice were fed chow containing a honeybee-derived compound with anti-inflammatory and immunomodulatory properties, caffeic acid phenethyl ester (CAPE). CAPE prevented the detrimental effects of host irradiation on tumor growth rate, immune signature, and immunosuppression. These data indicated that low-dose radiation, particularly densely ionizing exposure of aged mice, promoted more aggressive cancers by suppressing antitumor immunity. Dietary intervention with a nontoxic immunomodulatory agent could prevent systemic effects of radiation that fuel carcinogenesis, supporting the potential of this strategy for cancer prevention.

60 APPLIED LIFE SCIENCES↗

Design of an Out-Of-Pile Experimental Facility to Demonstrate the Feasibility of In Situ Thermal Conductivity Measurements of Nuclear Fuels Under Irradiation

There is substantial merit in quantifying nuclear fuel performance under irradiation. At Oak Ridge National Laboratory (ORNL), the MiniFuel irradiation platform has become the primary test vehicle for conducting separate-effects fuel performance irradiation experiments. The MiniFuel experiment is a passively controlled capsule design deployed in the High Flux Isotope Reactor (HFIR) through which fuel performance data is collected post-irradiation. Separate effects fuels irradiation capabilities are being expanded at ORNL by developing instrumented capsule designs that aim to capture fuel performance phenomena in-situ. One such capsule will specifically target fuel specimen thermal conductivity changes as a function of fuel burnup. Due to the complexity of making this measurement on nuclear fuel in-pile, this paper describes the necessary out-of-pile testing conducted on the thermal conductivity capsule (TCC) design. The measurement is ascertained via a thermopile system with heat transferred unidirectionally through a surrogate fuel specimen sandwiched between two conductive materials. The capsules investigated in this study are representative of the in-pile design, with the primary departure from irradiation conditions being the distribution of heat generation within the capsule. In the out-of-pile experiment, an external heater was used to drive heat through the conductive slug materials and into the specimen. This paper expounds the design of the out-of-pile experimental system and the thermal conductivity measurement technique. Predictive models used to determine the sensitivity of the measurement to variables governing thermal contact conductance between the specimen and slug materials and to predict experimental results are also described. Data from the out-of-pile experiment will be used to validate the readiness of the design for insertion into HFIR for irradiation.

Parker, Trevor [ORNL]↗

Response of GaN-Based Semiconductor Devices to Ion and Gamma Irradiation

GaN has electronic properties that make it an excellent material for the next generation of power electronics; however, its radiation hardening still needs further understanding before it is used in radiation environments. In this work we explored the response of commercial InGaN LEDs to two different radiation environments: ion and gamma irradiations. For ion irradiations we performed two types of irradiations at the Ion Beam Lab (IBL) at Sandia National Laboratories (SNL): high energy and end of range (EOR) irradiations. For gamma irradiations we fielded devices at the gamma irradiation facility (GIF) at SNL. The response of the LEDs to radiation was investigated by IV, light output and light output vs frequency measurements. We found that dose levels up to 500 krads do not degrade the electrical properties of the devices and that devices exposed to ion irradiations exhibit a linear and non- linear dependence with fluence for two different ranges of fluence levels. We also performed current injection annealing studies to explore the annealing properties of InGaN LEDs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Irradiation-Induced Defect Evolution in Nuclear Graphite

Graphite has historically been used as a moderator material in nuclear reactor designs dating back to the first man-made nuclear reactor to achieve criticality (Chicago Pile 1) in 1942. Additionally, graphite is a candidate material for use in the future envisioned next-generation nuclear reactors (Gen IV); specifically, the molten-salt-cooled (MSR) and very-high-temperature reactor (VHTR) concepts. Gen IV reactor concepts will introduce material challenges as temperature regimes and reactor lifetimes are anticipated to far exceed those of earlier reactors. Irradiation-induced defect evolution is a fundamental response in nuclear graphite subjected to irradiation. These defects directly influence the many property changes of nuclear graphite subjected to displacing radiation; however, a comprehensive explanation for irradiation-induced dimensional change remains elusive. The objectives of this project were focused on the characterization of irradiation-induced defect evolution in nuclear graphite via transmission electron microscopy (TEM). With the use of novel TEM specimen preparation techniques, high-temperature electron-irradiation and characterization of high-temperature neutron-irradiated nuclear graphite, novel fullerene-like defects are shown to be a dominant defect type, especially at higher temperatures. These results contradict the historical models of defect evolution and provide valuable insight into the macroscopically observed property changes in irradiated nuclear graphite.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Rapid Sample Insertion/Extraction Systems for Gamma Irradiation

The Rapid Sample Insertion/Extraction System for Gamma Irradiation, otherwise known as the "rabbit" system, was a four-week long project which included many different aspects; from coding an Arduino to building PVC piping to 3-D printing the "rabbit" capsules. The "rabbit" system is a system of PVC piping that allows a quick and efficient transfer of materials into/out of one of the irradiation chambers in the Gamma Irradiation Facility (GIF) with the use of a 3-D printed "rabbit." This "rabbit" encapsulates material to be irradiated and carries it from a position outside of the irradiation chamber to the basket inside of the chamber. The main purpose of this system is to save time and provide more exact data without any delays that normally occur when a person has to enter the chamber, retrieve data, and then analyze the data. This system should take measurements and retrieve the data instantaneously. The way in which the "rabbit" is sent through the PVC piping is with an advanced bi-directional, high-throughput pneumatic system, or a shop vacuum cleaner. When the vacuum is set to blow or suck then the "rabbit" will be pulled through the PVC piping to its intended destination and will hit sensors along the sides of the tubing when it reaches the end of the piping. These sensors tell the Arduino that the "rabbit" is finished moving throughout the tubing and stops a timer. Another timer is used to see how long the "rabbit" is being irradiated so when the "rabbit" reaches the sensors in the basket in the irradiation chamber another timer is started and it ends when the sensors no longer detect the "rabbit," which means that it has begun its journey back to the starting point.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Cracking Behavior of Irradiated Ex-Plant Materials

The cracking behavior of reactor core internal materials is a key factor for the long-term safety and availability of light water reactors. Subject to high-temperature coolant, thermal and mechanical loading, and neutron irradiation, reactor core internal materials are susceptible to several degradation mechanisms during power operation. To understand the long-term effects of neutron irradiation, irradiated materials harvested from a decommissioned reactor were studied for their cracking susceptibility and fracture resistance. Small compact-tension specimens were machined from different locations on the baffle plates with irradiation doses ranging from 0.06 to 48.5 dpa. Crack growth rate and fracture toughness J-resistance (J-R) curve tests were performed on these specimens in a simulated light water reactor environment, and the fracture morphologies of the tested samples were examined with a scanning electron microscope. All samples behaved similarly under cyclic loading, and no deteriorated corrosion-fatigue behavior was observed in the test environments. Under constant load, no elevated cracking susceptibility was observed either in these samples at low stress intensity factors. However, a rapid cracking behavior featuring very high crack growth rates was observed in several high-dose samples. This rapid crack growth behavior was activated when the applied stress intensity factor, K, was sufficiently high. Once activated, the rapid crack growth remained steady even when the K was lowered to the pre-activation level and stopped only when the K was reduced drastically. Yet, the rapid crack growth could be re-activated repeatedly in these samples by raising the K above the initial activation level. This rapid cracking response may be related to the severe irradiation embrittlement experienced by these samples. The J-R curve tests performed with these samples showed a significant decline in fracture toughness with increasing neutron irradiation. Local cracking at the crack tip facilitated by the severe irradiation embrittlement of this material may be responsible for the rapid crack growth behavior demonstrated by this decommissioned material.

36 MATERIALS SCIENCE↗

AGR-5/6/7 Irradiation Disassembly and Metrology First Look

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of high-temperature gas-cooled reactors (HTGRs), which are graphite-moderated nuclear reactors cooled with helium. This work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing, deployment, and operation of HTGRs in the United States. To achieve these goals, the program includes fuel fabrication, irradiations of TRISO fuels and high-temperature materials (e.g., graphite), safety testing and post-irradiation examination (PIE), fuel performance modeling, and fission product transport and source term determination. The ART AGR program has conducted four distinct fuel irradiation experiments in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The first of these irradiation tests, designated AGR-1, began in ATR in December of 2006 and ended in November 2009. This experiment was primarily to act as a shakedown test of the multi-capsule test train design and to provide early data on fuel performance that would be used in fuel fabrication process development. AGR-1 fuel kernels were produced on an engineering scale, but the TRISO coatings and cylindrical fuel compacts were fabricated on a laboratory scale. The AGR-1 PIE was completed and the final report was published in 2015. The second irradiation test, AGR-2, started in ATR in June 2010 and ended in October 2013. The AGR-2 irradiation test was designed to provide fuel performance data for coated particles fabricated on an engineering-scale pilot line using a coater with an internal chamber diameter of 150 mm (6 in.). The final PIE report was published in 2021. AGR-3/4, a single irradiation that combined what were originally conceived as the third and fourth tests, was to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport. PIE of the AGR-3/4 experiment is still in progress as of this writing. The subject of this report is AGR-5/6/7, the final qualification test of AGR TRISO fuel made entirely at the engineering scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Accelerated Irradiation and Qualification of Ceramic Nuclear Fuels

Accelerated irradiation efforts are an important topic in accelerated qualification of new nuclear fuel concepts for light water reactor (LWRs), microreactors, and other special purpose reactors. Accelerated fuel qualification combines both experimental and modeling work to expedite the total qualification time from approximately 20-25 years to 5-10 years timeframe. The scope covered under the current work package focuses on the experimental contributions for the neutron irradiations. Accelerated irradiation aims to reduce the total time needed for neutron irradiation to achieve targeted burnup, which is timely using conventional irradiation profiles. The purpose of this report is to detail the samples preparation for such irradiations, which contain standard, medium, and large grain undoped UO 2 , as well as large grain, Cr 2 O 3 -doped UO 2 . The data that will be generated from the irradiation of these samples will yield important information regarding fission gas retention as a function of grain size and the effect that Cr dopants have on diffusivity of fission gases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Final Report on Characterization of Irradiated Sensors and Coupling Adhesive Bonds

This report summarizes characterization via scanning/transmission electron microscopy of the microstructures of the unirradiated and irradiated piezoelectric ultrasonic sensor/aluminum substrate assemblies using four commercially available inorganic coupling adhesives to bond two types of piezoelectric crystals to the substrates. The sample assemblies were irradiated in the PULSTAR reactor at NC State University and ultrasonic data was collected in-situ. ORNL LAMDA Laboratory capabilities were utilized to perform pre- and post-irradiation examination of the sensor assemblies. This document summarizes the PIE performed at the ORNL LAMDA laboratory. The results of the PIE described in this report are consistent with the ultrasonic data collected during irradiation – in particular, high temperature epoxy adhesive seemed to provide the best coupling as compared to the refractory ceramic adhesives. It was determined that the quality of the sensor-adhesive-substrates governed the ultrasonic performance of the sensors. It was also apparent that irradiation did not significantly affect bond quality, which is also supported by the ultrasonic data collected during irradiation. A more comprehensive DOE NSUF Final Report including details of materials selection, sample fabrication, initial ultrasonic testing, irradiation and in-situ ultrasonic testing, positron annihilation lifetime spectroscopy and doppler broadening spectroscopy performed at EPRI and NC State University will be published at the conclusion of the project.

36 MATERIALS SCIENCE↗

ROADRUNNER MiniFuel Experiment: Irradiation Target Design and Sample Characterization

High-density uranium nitride (UN) is a fuel candidate for several advanced nuclear reactor designs currently under development. Because there are limited UN performance data relative to fuel fabrication impurity and density variation, an irradiation campaign has been developed as part of a collaborative effort among the University of Texas at San Antonio (UTSA), Westinghouse Electric Company, Oak Ridge National Laboratory (ORNL), and Los Alamos National Laboratory (LANL) under the Nuclear Science User Facilities program. This project, entitled ROADRUNNER, or Research On ADvancing the peRformance of UraNium Nitrides in Extreme enviRonments, aimsto support UN fuel qualification for advanced reactors by investigating the impact of density and impurity variations on UN performance as a function of irradiation temperature and burnup. The MiniFuel experiment vehicle developed by ORNL, which leverages the High Flux Isotope Reactor, was selected to perform this accelerated separate-effects irradiation testing. The experiment test matrix consists of six MiniFuel targets containing miniature UN fuel disks, and targets three distinct burnup levels (37.5, 60, and 75 MWd/kg U) and three distinct temperatures (600, 900, and 1200°C). Neutronics and thermal analyses were performed to determine the experimental parameters needed to meet the desired irradiation conditions and to predict the experiment components temperatures. UN pellets were fabricated at LANL with tightly controlled parameters to produce specimens with three distinct densities and three levels of carbon content. The pellets were then thinned down by UTSA to the experiment-required thickness. The pre-characterization of the specimens includes density measurements, carbon and oxygen contents, microstructure analysis, and x-ray computed tomography. The selected specimens will be assembled into the MiniFuel experiment, and the first ROADRUNNER MiniFuel targets are intended for HFIR insertion during the Fall of 2024. After irradiation, the targets will be shipped to ORNL’s hot cell facility for disassembly. The post-irradiation examination on the fuel specimens includes fission gas release measurements, visual inspection, fuel swelling measurements, gamma spectroscopy, and microstructure analysis. The data collected post-irradiation will be used to develop fuel performance models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Ion irradiation testing of 14YWT ODS and HT-9

Ion irradiations were performed to compare a new generation of 14YWT, NFA-2, to the prior generation NFA-1. Irradiations were performed at 450 °C and to a peak dose of 500 dpa. The irradiation condition was chosen as a relevant temperature for scoping for an upcoming JOYO neutron irradiation, and the high dose was chosen to evaluate the exceptional microstructural stability to irradiation induced changes of 14YWT. For comparison, HT-9 was irradiated as the benchmark. Irradiation induced changes to the microstructures were limited for the 14YWT samples and extensive for the HT-9 benchmark.

36 MATERIALS SCIENCE↗

AMMT Round Robin 316H HFIR Irradiation Test Matrix and Readiness for Insertion

The Round Robin 316H HFIR-2 irradiation campaign, conducted under the Advanced Materials and Manufacturing Technologies (AMMT) Program, is designed to evaluate the variability in irradiation response of laser powder bed fusion (LPBF) stainless steel (SS)-316H across multiple national laboratories, powder heats, and processing conditions, with comparison to wrought counterparts. A total of 11 materials are intended for irradiation in the High Flux Isotope Reactor (HFIR), targeting two irradiation temperatures (400°C and 600°C) and one level of irradiation damage (2 dpa). This irradiation campaign utilizes the standardized general tensile (GENTEN) capsule design to accommodate subsized tensile specimens of these materials. Thermal analyses were performed to ensure appropriate temperature control by design and uniformity within the capsules. All the capsules for this irradiation except one have been successfully assembled, welded, and tested and are ready for HFIR insertion. The last capsule requires re-build following a failed weld. All eight capsules are intended to be inserted in HFIR cycle 517, which is scheduled to start September 8, 2026. This report summarizes the design, material selection, and readiness for insertion of the capsules.

36 MATERIALS SCIENCE↗