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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Milestone 1.2.9: Radiolytic Gas Generation Measurements from Helium Backfilled Samples of AA1100 and AA6061 Coupons

Nearly 18 metric tons of aluminum-clad spent nuclear fuel (ASNF) is safely managed by the U.S. Department of Energy (DOE). These assemblies are currently in interim storage, with the intention of extended storage (>50 years) until final disposal. Strategies for the continued safe storage of this material are under evaluation, of which a key criterion is the extent of molecular hydrogen gas (H2) formation from the radiolysis of hydrated (oxy)hydroxide aluminum corrosion layers arising from in-reactor and wet storage conditions. Radiation-induced H2 formation has the potential to compromise cladding and storage canister integrity, in addition to promoting the formation of unfavorable gaseous environments. Consequently, understanding this radiation-induced phenomenon is essential for the development of predictive modeling capabilities to support technical considerations and the identification of radiation related challenges for the extended storage of ASNF. Here, we report radiolytic H2 yields (G-values, G(H2)) from the gamma irradiation of ‘pristine’ and pre-corroded aluminum coupons in helium (He) environments as a function of alloy composition (AA1100 and AA6061), relative humidity, and absorbed gamma dose. Measured yields were lower than corresponding values reported for argon environments, a positive result for proposed extended dry storage strategies that would employ helium as a backfill gas. Interestingly, the presented G(H2)He values are comparable to those previously measured in nitrogen environments, suggesting a He mediated H2 inhibition process, attributed here to Penning ionization.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
NASA NTRS1992

LWIR scene simulator developed for end-to-end performance evaluation of focal planes

The development of a long-wave infrared optical simulator facilitates evaluation of the end-to-end performance of long wavelength infrared (LWIR) focal plane arrays (FPAs) in a system-like environment. This simulator provides selectable structured scene inputs to a focal plane module or array. Background irradiances as low as 10 exp 10 photons/sq cm s are achievable when the simulator is cooled with liquid helium. The optical simulator can generate single or multiple targets of controllable intensities, and uniform or structured background irradiances. The infrared scenes can be viewed in a stationary mode or dynamically scanned across the focal plane.

Thompson, Niels A.↗
NASA NTRS2002

Spatial distribution and yield of DNA double-strand breaks induced by 3-7 MeV helium ions in human fibroblasts

Accelerated helium ions with mean energies at the target location of 3-7 MeV were used to simulate alpha-particle radiation from radon daughters. The experimental setup and calibration procedure allowed determination of the helium-ion energy distribution and dose in the nuclei of irradiated cells. Using this system, the induction of DNA double-strand breaks and their spatial distributions along DNA were studied in irradiated human fibroblasts. It was found that the apparent number of double-strand breaks as measured by a standard pulsed-field gel assay (FAR assay) decreased with increasing LET in the range 67-120 keV/microm (corresponding to the energy of 7-3 MeV). On the other hand, the generation of small and intermediate-size DNA fragments (0.1-100 kbp) increased with LET, indicating an increased intratrack long-range clustering of breaks. The fragment size distribution was measured in several size classes down to the smallest class of 0.1-2 kbp. When the clustering was taken into account, the actual number of DNA double-strand breaks (separated by at least 0.1 kbp) could be calculated and was found to be in the range 0.010-0.012 breaks/Mbp Gy(-1). This is two- to threefold higher than the apparent yield obtained by the FAR assay. The measured yield of double-strand breaks as a function of LET is compared with theoretical Monte Carlo calculations that simulate the track structure of energy depositions from helium ions as they interact with the 30-nm chromatin fiber. When the calculation is performed to include fragments larger than 0.1 kbp (to correspond to the experimental measurements), there is good agreement between experiment and theory.

NASA Program Biomedical Research and Countermeasur↗
DOE OSTI2026

FGMS-poster

Idaho National Laboratory (INL) performs post irradiation examination (PIE) of tri-structural isotropic (TRISO)-coated particle fuel to help qualify it for high temperature gas cooled reactors. TRISO fuel compacts are re-irradiated in the Neutron Radiography Reactor (NRAD) to generate the short lived fission products needed for fission product release testing. The Fuel Accident Condition Simulator (FACS) furnace and newly added Screen Neutron Irradiated Fuel for Failure (SNIFF) furnace heat the compacts in helium to temperatures of up to 2,000°C, prompting fission product release—predominantly gaseous xenon and krypton isotopes and condensable products such as cesium—from failed particles. These released isotopes are transported to a fission gas monitoring system (FGMS 1 or FGMS 3), where they accumulate in cryogenic cold traps and are quantified using high-purity germanium (HPGe) detectors. The addition of SNIFF and FGMS 3 increases throughput by enabling simultaneous testing of multiple compacts. Furthermore, automated INL developed software provides continuous, near-real time monitoring of fission product inventories and manages the liquid nitrogen cooling of the traps. These system enhancements improve the efficiency, data quality, and testing capacity of TRISO fuel performance evaluations.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2022

Design and demonstration of an economical SRF structure for Continuous Wave (CW), high-energy, Megawatt-class beams

Due to its very high RF-to-beam efficiency, superconducting radiofrequency (SRF) cavities have become the technology of choice for the construction of large linear accelerators (linacs) for basic science applications (examples, ESS and PIP-II). The technology is also attractive for high volume industrial applications such as electron irradiation processing of wastewater and flue gas. However, SRF’s current reliance on liquid helium for cryogenic cooling introduces infrastructural and safety challenges for use in industrial settings. Cryocooler conduction-cooling is a recently developed novel technique for operating SRF cavities without requiring liquid helium. This project aims for technical design and economic assessment of a medium energy, high average power e-beam linac based on the conduction-cooled SRF technique for high volume irradiation treatment of wastewater. The goals of this project are (1) to design a medium energy (10 MeV), high average power (1 MW) electron beam accelerator around the cryocooler conduction-cooled SRF cavity technology, (2) demonstrate the required accelerating voltage on a prototype cavity, and (3) build a cost-model to evaluate capital and operating expense of the accelerator.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2025

The effect of heat treatment on the defect evolution in LPBF 316H stainless steel

This work investigates the effect of processing and heat treatment on defect evolution and irradiation response of laser powder bed fusion (LPBF) 316H stainless steel, with comparisons to LPBF 316L and wrought 316L/316H. Using in-situ and ex-situ ion irradiations across a wide parameter space—temperature (300–675 °C), dose (0.2–25 dpa), dose rate (10 -3 –10 -5 dpa/s), and helium co-implantation (20–2500 appm)—we correlated void swelling, dislocation loop evolution, and segregation behavior with pre-irradiation microstructures (as-built, stress-relieved, solution-annealed, and cold-worked). Results show that swelling is strongly controlled by dislocation density: intermediate densities maximize swelling, while solution annealing or cold working reduce susceptibility to levels comparable to wrought alloys. Low dose rates and helium both promote cavity nucleation and lower the incubation barrier, with helium suppressing the role of dislocation density and driving swelling behavior toward wrought-like response. Loop evolution in SA LPBF 316H resembles wrought 316L but shows localized denuded zones near low-angle grain boundaries. STEM-EDS mapping further revealed Ni segregation at voids and sparse Al-rich oxides, without evidence of Ni–Si precipitates. Collectively, these findings identify dislocation density, helium content, and dose rate as the factors governing swelling in LPBF 316H and provide mechanistic datasets for model validation, supporting LAIN and the qualification of AM austenitic steels for nuclear service.

316 stainless steel↗
NASA NTRS1987

Morphometrics of cellular damage in mice testis receiving X-ray and high-energy particle irradiation

Murine tests were exposed to single, low doses of either X-ray, helium, or argon radiation. Animals were sacrificed seventy-two hours later. Testes were fixed for transmission electron microscopy (TEM) and sectioned at either 60 nm for TEM observation or at 2 micron for counting using routine light microscope methods. Counts of the total population of surviving spermatogonia, including all type A cells, intermediate, and type B cells, were taken from tubule cross sections identified as Stage 6 and Stage 1 according to spermatogonial configuration. The surviving fraction of spermatogonia as compared to control, S/S sub o, was calculated for each dose. For both ions and X-rays, there was a rapid decline in survival at dose levels of .10 to .15 Gy in Stage 6 tubules. This was followed by a more gradual decrease in population. At higher doses, 0.30 Gy for argon and 0.80 Gy for helium and X-rays, the cell survival rates declined rapidly. Pre-leptotene spermatocytes in Stage 1 tubules exhibited a different survival curve indicating the extreme radio-sensitivity of type B spermatogonia. Data verify that the seminiferous tubules are composed of a heterogeneous population of cells with different radio-sensitivities and that these differences are manifested even at very low doses.

Sapp, Walter J.↗
NASA NTRS2018

Computational Model Prediction and Biological Validation Using Simplified Mixed Field Exposures for the Development of a GCR Reference Field

The yield of chromosomal aberrations has been shown to increase in the lymphocytes of astronauts after long-duration missions of several months in space. Chromosome exchanges, especially translocations, are positively correlated with many cancers and are therefore a potential biomarker of cancer risk associated with radiation exposure. Although extensive studies have been carried out on the induction of chromosomal aberrations by low- and high-LET radiation in human lymphocytes, fibroblasts, and epithelial cells exposed in vitro, there is a lack of data on chromosome aberrations induced by low dose-rate chronic exposure and mixed field beams such as those expected in space. Chromosome aberration studies at NSRL will provide the biological validation needed to extend the computational models over a broader range of experimental conditions (more complicated mixed fields leading up to the galactic cosmic rays (GCR) simulator), helping to reduce uncertainties in radiation quality effects and dose-rate dependence in cancer risk models. These models can then be used to answer some of the open questions regarding requirements for a full GCR reference field, including particle type and number, energy, dose rate, and delivery order. In this study, we designed a simplified mixed field beam with a combination of proton, helium, oxygen, and iron ions with shielding or proton, helium, oxygen, and titanium without shielding. Human fibroblasts cells were irradiated with these mixed field beam as well as each single beam with acute and chronic dose rate, and chromosome aberrations (CA) were measured with 3-color fluorescent in situ hybridization (FISH) chromosome painting methods. Frequency and type of CA induced with acute dose rate and chronic dose rates with single and mixed field beam will be discussed. A computational chromosome and radiation-induced DNA damage model, BDSTRACKS (Biological Damage by Stochastic Tracks), was updated to simulate various types of CA induced by acute exposures of the mixed field beams used for the experiments. The chromosomes were simulated by a polymer random walk algorithm with restrictions to their respective domains in the nucleus [1]. The stochastic dose to the nucleus was calculated with the code RITRACKS [2]. Irradiation of a target volume by a mixed field of ions was implemented within RITRACKs, and the fields of ions can be delivered over specific periods of time, allowing the simulation of dose-rate effects. Similarly, particles of various types and energies extracted from a pre-calculated spectra of galactic cosmic rays (GCR) can be used in RITRACKS. The number and spatial location of DSBs (DNA double-strand breaks) were calculated in BDSTRACKS using the simulated chromosomes and local (voxel) dose. Assuming that DSBs led to chromosome breaks, and simulating the rejoining of damaged chromosomes occurring during repair, BDSTRACKS produces the yield of various types of chromosome aberrations as a function of time (only final yields are presented). A comparison between experimental and simulation results will be shown.

Hada, M.↗
DOE OSTI2022

AGR-5/6/7 Final Release-to-Birth Ratio Data Analysis

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel for use in the high-temperature gas cooled reactor (HTGR). AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product (FP) transport models and codes. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. Release-to-birth ratios (R/B) for fission-gas isotopes released from each capsule are calculated from release rates, measured by germanium detectors in the Fission Product Monitoring System (FPMS) installed downstream from each capsule, and birth rates calculated using numerical models of FG generation. The R/Bs are a critical measure of the ability of the fuel kernel, the particle coating layers, and the compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, and the impact of initially defective particles and/or particle-coating failures that occur during irradiation. For fission-gas isotopes, particle failure is defined as failure of all coating layers, allowing gaseous fission atoms to escape from a particle. During the first five cycles (162B ? 165A), R/Bs were stable in the 10-8?10-6 range, and no in-pile particle failures were observed, based on the gross gamma counts. The maximum R/B value of around 2 ? 10-6 for Kr-85m resulted from the presence of as-fabricated exposed kernels (based on the high exposed kernel fraction), the dispersed uranium, and high fuel particle temperatures in Capsule 1. Comparison of capsule-measured R/Bs from these early cycles to predictions using the previously developed AGR R/B model demonstrated FG release from the AGR-5/6/7 TRISO fuel was comparable to that of previous experiments. In addition, the Kr-85m R/B per-exposed-kernel values are comparable to R/B values obtained in AGR-3/4 irradiation experiment and four irradiation experiments performed during 1980s: (1) HRB-17/18, (2) COMEDIE-BD1, (3) HFR B1, and (4) HRB-21. In contrast, all measured R/B values are lower than predictions by the commonly used Richards and German models, which are intentionally conservative. A large number of in-pile particle failures occurred in Capsule 1 by the end of Cycle 166A. During the final four cycles (166A ? 168A), apparent damage to the Capsule 1 gas line appeared to cause FG leakage from that capsule into the other four capsules, resulting in an increase in fission gas (FG) detected in the effluent for all capsules. Isolation of the Capsule 1 gas line during the last three cycles also prevented measurement of its FG release. Thus, R/Bs in all capsules after Cycle 166A are highly uncertain because of undefined amount of leakage from Capsule 1, especially for long-lived isotopes. A few hundred in-pile particle failures were estimated for Capsule 1 before the end of Cycle 166A, but the total number of failures is unknown due to the lack of FG release data in the later cycles. Based primarily on evidence from the gross gamma counts during Cycle 168A, approximately 15 particles failed in Capsule 3 and four particles failed in Capsule 2. In-pile failures in Capsule 3 were anticipated because this capsule was designed to operate beyond the HTGR normal operating temperature range. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR model, developed in INL/EXT-14-32970, for R/B of the short-lived isotopes (Kr-89 and Xe-137) with minimal leakage from Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Benchmark Exercise for the Control Rod Swelling Evaluation

The VTR core has six reactivity control assemblies and three safety assemblies. The control assemblies or primary control rods are adjusted during the normal operation to balance the core reactivity and to control the reactor power. A typical control assembly radial layout is presented in Figure 1. The figure shows the swelled absorber (B 4 C) rod. Initially, helium gas fills the gap between the pin and the cladding before irradiation swelling takes place. For VTR, HT9 steel was selected as the cladding and duct material. The main neutron absorbing material used in the VTR is B 4 C. When residing in the core, the neutronics, thermophysical, and mechanical properties of the materials used in a control assembly will degrade due to accumulated neutron damage. Material degradation limits how long a control assembly can reside in the core. Many phenomena affect the control assembly lifetime, such as the loss of reactivity worth due to B 4 C depletion, the mechanical interaction of the absorber rod and the cladding due to B 4 C swelling, the helium gas buildup in the pin due to B-10 capture, etc. B 4 C swelling, which causes closure of the gap between the absorber rod and the cladding, is usually considered as the main limiting factor from past experience. An initial study was conducted at PNNL to evaluate the irradiation behavior of a VTR control assembly. The evaluation was performed using the CNRD2 code that was initially developed for the FFTF. The study also included an assessment of the VTR control assembly and focused on a 61-pin control assembly design, which is different from that used (37-pin design) in the core design study. The study conducted by PNNL was reviewed independently by ANL. A Python script referred to as the Control Assembly Evaluation Script (CAES) was developed for the independent review and additional assessment of 37-pin control assembly design. The script has focused on the assessment of the absorber rod swelling for its importance in determining the control assembly lifetime. CAES uses geometry, neutronics, materials data as input to predict the swelling of the absorber rod during its residence in the reactor core. The results from CAES showed some non-negligible differences against the PNNL results. Some of the differences can be attributed to the different interpretation of the control rod assembly dimensions. To resolve this issue, a benchmark exercise was proposed. The benchmark specification was developed by PNNL. The benchmark exercise was performed independently at PNNL and ANL using different codes/scripts (CRND2 and CAES). This memo documents the results calculated using the different codes. However, this report is limited to presenting the results obtained. Further investigation of the cause of the observed difference will be performed as part of future activities, pending continuation of the VTR program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Nanostructured Alumina Forming Austenitic Alloy (NAFA) Production Using Mechanical Alloying and High-Temperature Consolidation

Alumina-forming austenitic (AFA) alloys are a promising class of nuclear materials because of their high-temperature oxidation/corrosion resistance and mechanical properties. Unfortunately, these alloys are limited in use for core material applications, and they are specifically limited for use as nuclear fuel cladding because of their high Ni-transmutation and helium generation rate in-service. Improving these alloys through a fine dispersion of oxide precipitates and thus increasing the effective irradiation sink strength of the alloy system may mitigate many of the degradation phenomena expected during alloy deployment. These phenomena include high-temperature helium embrittlement and cavity swelling for lead-cooled fast reactor applications. This work effort uses combination of conventional and advanced manufacturing approaches are being used to fabricate nanostructured AFA (NAFA) materials. As the first objective of this initiative, a conventional AFA was modified using mechanical alloying and extrusion to alter the precipitation characteristics to include a fine dispersion of nanoscale oxides intended to serve as traps for irradiation-induced point defects and transmuted He within the lattice. This analysis compared the efficacy of the conventional mechanical alloying and extrusion approach with the unalloyed AFA consolidated approach using hot isostatic pressing (HIP). It was found that, although the mechanical alloying approach is successful in producing a fine distribution of oxides within the first nanostructured AFA (NAFA-1), the distribution is heterogeneous because of the mild milling parameters used to prevent cold welding of powder to the spherical milling media. The additional dispersion of oxides, coupled with a higher volume fraction of other secondary phases in the NAFA-1, produces higher alloy strengths that range up to 600°C in comparison to the unalloyed HIP AFA, thus exceeding the operating temperature of lead-cooled fast reactors. However, the strength of the NAFA-1 is lower than that of the HIP AFA at 800°C, which is presumed to be a function of increased secondary phases from the nonoptimized AFA chemistry. Future work is planned on a newly procured NAFA-2 chemistry that is more suitable for advanced reactor applications exploring new manufacturing routes.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Fusion Materials Semiannual Progress Report for the Period Ending December 31, 2020

Three ferritic/martensitic alloys; F82H, Fe8Cr2W and CNA3 were studied for the effect of H and He on swelling. Single ion beam (Fe 2+ ), dual ion beam (Fe 2+ +He 2+ ), and triple ion beam (Fe 2+ +He 2+ +H + ) irradiations were conducted at 500°C to a damage level of 50 dpa and dual and triple beam irradiation experiments were conducted at 450°C to 50 dpa. Cavity distribution and swelling were comparable in dual ion beam and triple ion beam irradiated F82H at 500°C, indicating that the role of hydrogen in swelling in triple beam irradiated F82H appeared to be insignificant compared to the role of helium. Swelling in Fe8Cr2W and CNA3 at 500°C was approximately 3-4X greater than that at 450°C. In F82H, the average cavity size and swelling were minimally larger at 500°C than at 450°C. The significant role of hydrogen in swelling has yet to be confirmed. Future efforts will focus on separate effects experiments to understand the role of hydrogen in cavity swelling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

A Multiscale Investigation of the Mechanisms Controlling Materials Degradation in the Fusion Energy Environment (DE-SC0006661: Final Report)

Realizing the promise of fusion as a commercially attractive energy source requires the development of advanced structural materials capable of sustained operation in an incredibly hostile environment. The fusion environment presents many challenges associated with high cyclic thermal-mechanical loadings, welding and joining disparate materials, and achieving chemical compatibility with coolants and tritium breeders. Yet, the overarching concern is the degradation of physical and mechanical properties, resulting from a neutron energy spectrum peaked at 14 MeV. The high-energy fusion neutron irradiation produces both displacement damage and high levels of hydrogen and helium through transmutation reactions. Advanced materials development for use in such a hostile environment is predicated on understanding the underlying mechanisms responsible for physical and material property degradation. This project has closely combined computational, theoretical and experimental techniques within a multiscale materials science paradigm to determine the mechanisms controlling material degradation in the fusion environment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1986

First results from Spacelab 2

Preliminary results of physical experiments carried out during the Spacelab 2 Shuttle mission are summarized. Attention is given to experiments in the fields of plasma dynamics; solar physics; high-energy astrophysics; and astronomy. Plasma experiments included an ejectable plasma diagnosics package and measurements of the passive charging of the Shuttle vehicle in the surrounding space plasma. The solar physics instrument package consisted of a solar spectral irradiance monitor; a solar optical universal polarimeter (SOUP); and a solar helium abundance high-resolution telescope and spectrograph (HRTS). Astronomical observations were performed using a scanning infrared telescope (IRT) which consisted of a highly baffled herschelian telescope and 10 detectors covering wavelengths from 2 to 120 microns. Cosmic-ray nuclei were detected and analyzed using gas Cerenkov counters and a transition radiation detector. Addition experiments included a thin film fluid dynamics payload and analysis of blood samples taken from the mission specialists. Complete data records from the experiments have now been distributed for an analysis period which will take at least a year. A table listing the Spacelab 2 experiments and their principal investigators is provided.

Urban, E. W.↗
DOE OSTI2020

Thermal Model Heat Rate Predictions of the AGR-5/6/7 Experiment

Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train. AGR-5/6/7 will test low-enriched uranium oxycarbide (UCO – a heterogeneous mixture of uranium oxide and uranium carbide) TRISO fuel. The AGR-5/6 portion of the experiment will provide data to support qualification of the selected reference fuel design, while the AGR-7 portion will serve as a margin test, irradiating the fuel beyond normal operating conditions. During irradiation, the temperature in each capsule is controlled by varying the helium / neon gas mixture flowing through the capsule until the desired thermocouple temperature is reached. Further adjustments can be made by changing the lobe power and the neutron filter. Three neutron filters are available to use as the experiment burns up the fissionable fuel. A thermal finite element model has been created for the five capsules comprising the AGR-5/6/7 experiment. Heat rates from a physics analysis are imported into the model along with fast neutron fluence. Graphite shrinkage due to the fast neutron fluence is incorporated into the model. Gas gaps change as a function of fast neutron fluence. Radiation heat transfer is a major contributor in this model. This is a large model with more than 1 million finite element brick elements. More than 150 different parts are modeled in the finite element model and communicate with each other from a heat transfer sense. More than 50 thermocouples are used in the experiment and are compared to actual measurements. The experiment is composed of five separate stainless steel capsules all welded together. This paper presents a method used to predict the gas mixture in the event that all of the thermocouples fail in a particular capsule. Curve fitting of heat rates from previous cycles and projecting them into the future is the basis of the method.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Anisotropy in Thermal Conductivity of Ga2O3 under Irradiation: Ab-initio Investigations

Ga2O3 is a wide-bandgap semiconducting material known for its high breakdown electric field and thermal stability. Its various crystalline forms have shown great potential in power electronics, optoelectronics, and high-temperature applications. Specifically, ß-Ga2O3 has been recognized for its suitability in neutron and gamma radiation detection, making it ideal for sensor and detector applications [1]. Notably, ß-Ga2O3 exhibits anisotropic thermal conductivity, meaning that the magnitude of thermal conductivity varies with the crystallographic direction [2]. Under neutron irradiation, Ga2O3 undergoes a significant phenomenon, resulting in the formation of Helium bubbles within its crystalline matrix. The presence of Helium in Ga2O3 has been found to influence the material's anisotropic thermal conductivity. In this study, the objectives were twofold: firstly, to employ Density Functional Theory (DFT) to determine the most probable sites of Helium occupation in ß-Ga2O3, and secondly, to investigate the anisotropic variation in thermal conductivity when Helium is present in the crystalline matrix of ß-Ga2O3. These ab-initio DFT studies provide crucial insights into understanding the behavior of Helium in Ga2O3 and its impact on thermal conductivity, which can contribute to further advancements in sensor and detector applications based on this wide-bandgap semiconductor.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Microstructural characterization of cold-worked 316 stainless steel flux thimble tubes irradiated up to 100 dpa in a commercial Pressurized Water Reactor

Two flux thimble tubes (FTT) made of 15% cold-worked 316 stainless steel (SS) were harvested from Ringhals Pressurized Water Reactor (PWR) Unit 2, with peak damages of 76 and 100 displacements per atom (dpa) after 29 and 34 years’ service, respectively. Specimens sectioned from parent tubes were comprehensively characterized with nominal damage levels of ~0, ~41, ~74, 76, and 100 dpa at a nominal temperature range of 285-323 °C. Both FTTs contained helium and hydrogen gases as transmutation products. The helium follows a production rate of ~9.8 appm/dpa, while environmental factors complicate hydrogen production obscuring an exact H/dpa ratio. Irradiation-induced dislocation loops, nano-cavities, solute clusters, and microsegregation were all observed. The dislocation loops and nano-cavities indicated saturation at 41 dpa. The solute clusters continued to evolve with Ni-Si clusters formed at 41 dpa, and Ni-Si-Mn-P clusters formed at 74 and 100 dpa, but neither clusters exhibited distinct diffraction patterns at any damage levels. Solute clusters were observed to frequently be co-located with dislocation loops, but fully decorated loops were rarely detected. Significant radiation-induced segregation (RIS) was observed around grain boundaries at all damage levels. The modified inverse Kirkendall (MIK) model captured the RIS behavior of major elements. Large cavities within or around an Mn-S rich region were observed for the first time. Through all the damage levels, void swelling is always below 0.05%, making significant dimensional change unlikely in core internals when used at similar conditions. Meanwhile, the role of overwhelming nanocavities, presumably helium bubbles, should be considered in other potential degradation mechanisms, including irradiation-assisted stress corrosion cracking, embrittlement, and loss of fracture toughness, which remain the concerns for extended operation of nuclear power plants.

Post Irradiation Examination, Isotope Dilution Mas↗
DOE OSTI2023

Milestone 1.2.14: Surface Characterization of Irradiated Surrogate Non-Native Corrosion Plumes on Aluminum 6061-T6 Specimens

The United States Department of Energy is seeking to evaluate the feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled cannisters. A significant research effort has been devoted to determining the amount of radiation-induced molecular hydrogen generation from corrosion layers that would be present on the fuel assembly surfaces. However, to date, no evidence of radiation-induced changes in the microstructure of the corroded aluminum surfaces have been reported. This report provides surface characterization, in terms of corrosion layer composition and morphology, for aluminum alloy 6061-T6 “plume” samples subjected to one of three potential ASNF drying techniques—“vacuum only”, “vacuum + 4 hours at 100 oC”, and “vacuum + 4 hours at 220 oC”—and gamma irradiated (= 53 MGy) in the presence of helium gas with ~0% added relative humidity. At these high absorbed gamma doses, for the first time ever, radiation-induced circular defects were found on the sample surface, regardless of the drying regime employed. Additionally, large cracks that penetrated through the corrosion layer to expose bare aluminum metal were observed in irradiated “plume” specimens subjected to drying conditions of “vacuum + 4 hours at 220 oC”. The nature and implications of these defects on the extended dry storage of ASNF is unclear, requiring further study.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗