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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Understanding irradiation growth through atomistic simulations: Defect diffusion and clustering in α-zirconium and the influence of alloying elements - 2016-0093

Irradiation-induced structural changes of α-zirconium alloys and in particular the effect of iron were investigated by molecular dynamics simulations using embedded atom potentials derived from first-principles calculations. The simulations revealed that at temperatures between 500 and 600 K self-interstitial atoms (SIAs) diffuse rapidly in a cooperative movement, preferably parallel to basal planes (a directions; ), forming nano-clusters with an extension in and . Vacancies diffuse more slowly than SIAs and remain isolated for a longer period of time. Nano-clusters associated with SIAs cause a pronounced overall expansion in a directions, as well as local strains. Under compressive strain in the c direction, vacancy diffusivity increases in the c direction. In contrast, the diffusivity of SIAs increases in the c direction under a tensile strain in the c direction. SIA nano-clusters are highly mobile within basal planes. Vacancy clusters grow by merging, leading to a contraction in the a direction, compensating for the expansion caused by SIA nano-clusters and possibly contributing to the plateau in growth after the initial rapid expansion. At the onset of breakaway growth, possibly due to stress buildup, the vacancy nano-clusters can condense into c loops, thereby diminishing the compensation effect. The alloying elements iron, nickel, chromium, and niobium liberated from secondary phase particles under irradiation or already in solution are attracted to vacancies and SIAs and are found inside vacancy and SIA loops. The interaction of alloying elements with defect clusters is discussed, with a particular focus on iron. Iron has been found to promote cluster formation in zirconium, and the structures of zirconium-iron clusters have been analyzed. Tin is repelled by SIA clusters and only weakly attracted by vacancies. Niobium impedes the diffusion of SIAs (and therefore may increase annihilation rates with nearby vacancies) and does not destabilize vacancy or SIA clusters. Ab initio calculations of the dimensional and elastic coefficients of the intermetallic phases occurring in secondary phase particles, such as Zr{sub 2}Fe and Zr{sub 3}Fe, are presented, allowing an assessment of local strains in a zirconium matrix. Thus, novel results from extended molecular dynamics simulations provide new insights and contribute to a deeper understanding of the complex mechanisms causing irradiation-induced dimensional changes and the breakaway growth of zirconium alloys. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Equivalent radiation damage in zirconium irradiated in various reactors - 2016-0101

The physical properties and in-reactor performance of core components made from zirconium alloys are a function of the radiation-induced atomic displacements that occur during service. The vacancy and interstitial point defects and their clusters created by the atomic displacements modify the microstructure, and therefore physical properties such as strength and ductility, by forming dislocation loops. The rate of production of point defects also affects time-dependent properties such as creep and diffusion (mass transport). To simulate the effect of irradiation on core components, researchers have used materials test reactors for many years to irradiate and test materials. Such reactor tests can target an accumulated displacement damage dose to simulate the state of a material at a given stage in reactor life but at an accelerated rate. Each nuclear reactor has a specific neutron energy spectrum that also depends on the in-reactor location. Therefore, to use test reactor data to determine the effect of radiation damage on power reactor materials, one must convert the accumulated dose into a unit that is common in its effect on the material properties. For many property changes in nuclear reactor cores, this unit is displacements per atom. This paper compares the damage production for zirconium irradiated under differing neutron energy spectra in various power and test reactors. The damage rates as a function of different measures of neutron flux were used to relate atomic displacements in various materials test reactors to CANDU and light-water reactor conditions. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2023

Artificial intelligence models, photos, and data associated with the manuscript “Quantifying Streambed Grain Size, Uncertainty, and Hydrobiogeochemical Parameters Using Machine Learning Model YOLO” (v2)

This data package is associated with the manuscript “Quantifying Streambed Grain Size, Uncertainty, and Hydrobiogeochemical Parameters Using Machine Learning Model YOLO” published in Water Resources Research (Chen et al., 2024). This data package includes the training, validation, testing, and prediction data used by the artificial intelligence (AI) model for automated grain size and hydro-biogeochemistry quantification using streambed photos. The grain size data are extracted for each photo using You Look Only Once (YOLO), a pre-trained object detection model. This data package was originally published in October 2023. It was updated August 2025 (v2; new and modified files). File and folder names were not revised to indicate changes. See the change history section in the readme for more details. Please see flmd.csv for a list of all files contained in this data package and descriptions for each. Please see dd.csv for a data dictionary that defines the column headers of .csv files in the data package. This dataset is comprised of one data folder containing (1) file-level metadata; (2) data dictionary; (3) readme; and (4) six subfolders. Subfolders 1 to 4 include the training, validation, testing, and prediction data. Subfolder 5_Summary includes the summary results of different combinations of training, validation, testing, and prediction data. Subfolder 6_SupplementalData includes additional data downloaded from public sources (Kaufman et al., 2023a; Kaufman et al., 2023b; Garefalakis et al., 2023; Mair et al., 2024; https://github.com/river-corridors-sfa/Geospatial_variables). In total, the data package includes 110 folders and 44,283 files. These files include 9,047 .jpg photos, 1 .png photo, 3 .tif photos; 26,639 photo labels and individual grain sizes and probability from AI (.txt); 8,447 grain size distribution data (.dat); and 126 CSV files for results summary, and 14 required metadata files (.xlsx). The summary CSV files contain 68 columns and approximately 2,200 rows that represent photo names, site locations, recording time, GPS coordinates, grains sizes (D10, D50, D60, and D84), number of grains, and additional hydro-biogeochemical data such as water depth, flow velocity, Manning’s coefficient, friction factor, hydraulic conductivity, permeability, streambed interstitial velocity magnitude, mass transfer rate, and nitrate uptake velocity. The photos were obtained from 75 sites in the Yakima River Basin and the Columbia River shorelines, and other associated data from samples and sensors obtained when the photos were taken are publicly available (Fulton et al. 2022; Grieger et al. 2023). All files are .csv, .txt, .dat, .jpg, or .pdf. We acknowledge the Yakama Nation as owners and caretakers of the lands where we collected some of these data. We thank the Confederated Tribes and Bands of the Yakama Nation Tribal Council and Yakama Nation Fisheries for working with us to facilitate sample collection and optimization of data usage according to their values and worldview.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Review of Calculations on Point Defect Properties in Delta-Pu

The results of theoretical predictions of properties for vacancies and self-interstitial atoms (SIA) in δ-Pu are presented and reviewed. Three methods have been used for these predictions, namely the modified embedded atom method (MEAM), density functional theory (DFT) with and without spin polarization, and continuum mechanics (CM) models adapted to plutonium. The properties considered are formation and migration energies, and relaxation volumes of vacancies and SIA. Predicted values vary considerably. Nevertheless, all three methods predict that the activation energy for self-diffusion by vacancies is of similar magnitude as the SIA formation energy. Furthermore, the absolute magnitudes of relaxation volumes for vacancies and SIA are also similar, indicating that there exist no large bias for radiation-induced void swelling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Report on simulation of fission gas and fission product diffusion in UO 2

In UO 2 nuclear fuel, the retention and release of fission gas atoms such as xenon (Xe) are important for nuclear fuel performance by, for example, reducing the fuel thermal conductivity, causing fuel swelling that leads to mechanical interaction with the clad, increasing the plenum pressure and reducing the fuel–clad gap thermal conductivity. We use multi-­scale simulations to determine fission gas diffusion mechanisms as well as the corresponding rates in UO 2 under both intrinsic and irradiation conditions. In addition to Xe and Kr, the fission products Zr, Ru, Ce, Y, La, Sr and Ba have been investigated. Density functional theory (DFT) calculations are used to study formation, binding and migration energies of small clusters of Xe atoms and vacancies. Empirical potential calculations enable us to determine the corresponding entropies and attempt frequencies for migration as well as investigate the properties of large clusters or small fission gas bubbles. A continuum reaction-­diffusion model is developed for Xe and point defects based on the mechanisms and rates obtained from atomistic simulations. Effective fission gas diffusivities are then obtained by solving this set of equations for different chemical and irradiation conditions using the MARMOT phase field code. The predictions are compared to available experimental data. The importance of the large Xe U3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and high binding energy. We find that the Xe U3O cluster gives Xe diffusion coefficients that are higher for intrinsic conditions than under irradiation over a wide range of temperatures. Under irradiation the fast-­moving Xe U3O cluster recombines quickly with irradiation-induced interstitial U ions, while this mechanism is less important for intrinsic conditions. The net result is higher concentration of the Xe U3O cluster for intrinsic conditions than under irradiation. We speculate that differences in the irradiation conditions and their impact on the Xe U3O cluster can explain the wide range of diffusivities reported in experimental studies. However, all vacancy-­mediated mechanisms underestimate the Xe diffusivity compared to the empirical radiation-­enhanced rate used in most fission gas release models. We investigate the possibility that diffusion of small fission gas bubbles or extended Xe-­vacancy clusters may give rise to the observed radiation-­enhanced diffusion coefficient. These studies highlight the importance of U divacancies and an octahedron coordination of uranium vacancies encompassing a Xe fission gas atom. The latter cluster can migrate via a multistep mechanism with a rather low effective barrier, which together with irradiation-induced clusters of uranium vacancies, gives rise to the irradiation-enhanced diffusion coefficient observed in experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to the process and analysis can be made, and neutron diffraction can become a viable and efficient technique for gamma/alpha phase composition determination for nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

TEX-Hf: Integral Experiment Execution of Thermal/Epithermal eXperiments using Highly Enriched Uranium with Polyethylene and Hafnium (IER-532 CED-3b Report)

This report documents the experimental configurations and measurements for IER-532, Thermal/Epithermal eXperiments (TEX) with highly enriched uranium (HEU) fuel and interstitial hafnium (Hf), moderated and reflected by polyethylene. TEX-Hf is a variation of and based on the TEX-HEU (IER-297) design, with the inclusion of hafnium. These configurations provide integral experiments for validation of hafnium in the thermal, intermediate, and fast neutron energy regimes by maximizing the sensitivity in k eff to the hafnium isotope cross sections. The experiment campaign was completed over seven weeks during the end of FY22 and beginning of FY23 at the National Critical Experiments Research Center at the Nevada National Security Site. The campaign produced seven experimental configurations, four reproducibility measurements, and many additional dimensional measurements that will be of use to the future benchmark evaluation of this experiment and other experiments using the same HEU fuel. Table 1 summarizes the TEX-Hf experimental configurations, including their physical parameters, calculated fission fractions, and estimated excess reactivities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Simulation of fission gas diffusion in UO{sub 2} by vacancy clusters

The low solubility of fission gas in nuclear fuel leads to inter- and intra-granular bubbles. These bubbles affect performance of the fuel rod and increase the fuel temperature through the increase of porosity in the fuel pellets and decreased thermal conductivity across the pellet-clad gap as gas is released into the plenum. The nucleation, growth, and inter-connectivity of the fission gas bubbles ultimately depend on the diffusivity of gas atoms through the fuel matrix. Using mechanisms and data previously established from density functional theory (DFT) and empirical potential calculations, diffusion and reaction models have been derived for xenon atoms, uranium atoms, and vacancies in the substitutional and interstitial locations. Although similar to previous models, the diffusion and reaction rates have been re-derived in a more robust manner. Previous studies have under-predicted the experimentally observed diffusivity of xenon atoms in UO{sub 2}, however it has been previously shown that by including xenon atoms bound to two vacancies, the calculated diffusivity approaches the measured data. The derived model for species diffusion and inter-species reactions has been implemented in the MARMOT phase-field code in a modular structure so as to capture the enhanced diffusion of xenon due to bound gas + vacancy-cluster movement. The MARMOT code allows for the diffusion, reaction, and source rates to be solved simultaneously, resulting in steady state or transient concentrations of each species, and ultimately the diffusion behavior of xenon atoms. By including the diffusion of clusters of different combinations of xenon and vacancies, the calculated diffusivity more closely captures the measured diffusivity of xenon when using preliminary parameters. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Optimization of salt-cake removal flowsheet at SRS through incorporation of testing and in-tank waste experience - 15263

Salt-cake removal at SRS may be performed for several reasons: to provide space for evaporator operation (i.e., to precipitate more salt in the drop tank), to provide feed for salt processing (i.e. immobilize the waste), or to remove the salt for tank closure. Many different salt dissolution techniques have been employed in the 40 years that SRS has been performing salt removal, from a basic 'Add, Sit, Remove' method (water is added on top of the salt-cake and time is allowed for diffusion), to performing interstitial liquid removal, or using mixing devices to promote contact with the liquid. Lessons learned from previous salt-cake removal campaigns, in addition to testing and modeling, have led to opportunities for improvements to the overall salt-cake removal process. This includes better understanding of salt properties and behavior during dissolution; the primary concerns for salt dissolution are the release of radiolytic hydrogen and criticality prevention (post-dissolution). Recent developments in salt dissolution include the reuse of dilute supernate and a semi-continuous dissolution (SCD) process, where low volume mixing eductors are used to deliver water near the surface of the salt-cake at the same rate as the salt solution is removed and transferred to a receipt tank. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Investigation of reactivity differences in UF{sub 6} cylinder arrays with different masses at various mist conditions

In nuclear criticality safety, ensuring that all systems stay subcritical is of utmost importance to the safe operation of the facility as well as not imposing any undue risk to the public or the environment. This can be relatively straightforward for simple and isolated systems as the factors affecting criticality are well understood and proven based on past safety experience in the industry. However, array configurations can be challenging and present hidden complications not seen in simpler systems. While the reactivity of arrays depends on the same factors as simple/isolated systems, arrays are also influenced by interaction/spacing (including moderation effects) and reflection, which can introduce competing effects in the array that become harder to attribute to any one given factor. These effects can lead to some undesired effects and unanticipated trends in array systems, making it more important to understand those effects and in particular in UF{sub 6} cylinder arrays. Interstitial moderation/reflection is one of the effects that can lead to unanticipated trends in cylinder arrays The moderation/reflection effect of specific interest in this paper is mist or water vapor. Herein the reactivity differences in fixed pitched cylinder arrays with different cylinder masses are investigated as a function of mist density. The cylinder mass is based on varying the UF{sub 6} density to achieve the desired UF{sub 6} mass by completely filling the cylinder. Initial simulation results for different masses demonstrate that there could be some unique effects with mist densities in the range of approximately 0.1 and 0.3 g/cm{sup 3}. These results are then compared to the results of different cylinder masses using the nominal solid UF{sub 6} density with the same mass values. This comparison shows these results are unique to modeling the cylinder with a given mass using a density that allows the mass to completely fill the cylinder. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Focus on Different Binders for Concentrated Alkali and Magnesium Hydroxide Waste Conditioning - 18122

As part of a nuclear dismantling process, an effluent highly concentrated in alkali hydroxides and magnesium hydroxide could be produced and should be conditioned in a matrix capable of ensuring the adequacy between the hydraulic binder performances and the conditioning requirements. First, several matrices were chosen according to the compatibility of their interstitial pore water pH with our highly concentrated in alkali and magnesium hydroxide waste. Then, we observed their behavior from a macroscopic point of view when an increasing amount of alkali hydroxide is added in the mixing water. The compressive strength - an important requirement for the storage acceptance - along with some microstructural comprehension allowed us to discriminate the different binders. Only one was chose to continue the study with the integration of magnesium hydroxide in the formula. Finally, a validation of the early ages performance acquired without magnesium hydroxide was performed and a durability study is undergoing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Importance of Plutonium Turnings in Criticality Safety

In criticality safety, turnings are a concern due to the fact that they leave interstitial space in which a moderator can ingress and reduce the critical mass for a fissile material. As turnings are produced when a Plutonium (Pu) items are machined, they are commonly encountered in facilities that process Pu. The intent of this paper is to establish minimum subcritical mass data points for Pu turnings based on the minimum density achievable as measured in a turnings experiment performed at Los Alamos National Laboratory (LANL). Further discussion will cover the implications of these results with concern to criticality safety during normal operations and during a glovebox fire scenario. The results of the study show that a bare homogeneous Pu turnings (at 0.252 kg/L) and water mixture has a spherical critical mass of 3.80 kg, while the fully water reflected spherical critical mass of the same Pu turnings and water mixture has a critical mass of 1.65 kg. When mixed with an idealized hydrocarbon oil rather than water, the spherical critical mass is reduced. In this case, the bare critical mass of the homogeneous Pu turnings and oil mixture is 2.57 kg while the fully water reflected spherical critical mass of the Pu turnings and oil mixture is 1.17 kg. The results of this study show that up to approximately 1000 g of Pu turnings even if mixed with an ideal hydrocarbon and flooded would not be a criticality concern during fire fighting activities. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Multiscale Model for Simulating Fission-Induced Recrystallization in U-Mo Alloy

U-Mo alloy is one of the most promising fuels for the future high performance research and test reactors due to its high uranium loading density and stable irradiation behavior. However, a major concern to its qualification is the fuel swelling at high fission density due to the formation of large fission-gas bubbles. The fission-induced recrystallization further expedites the formation and growth of gas bubbles. So far, the exact mechanism of recrystallization is still unclear. One possible driving force for recrystallization is the production of point defects and interstitial loops due to irradiation. Following this, we developed a multi-scale model by integrating the rate-theory and phase-field models to study microstructural evolutions in irradiated U-Mo alloy, i.e., fission-induced gas bubble evolution and recrystallization. Density function theory (DFT) calculations are utilized to predict the defect formation energies, diffusivities of defects, the interfacial energy of gas bubble, and the elastic constants of U-Mo alloy, etc. The dislocation density evolution is predicted by the rate-theory developed by Jeff. The grain subdivision occurs once the dislocation density is above a critical value. The grain growth coupled with gas bubbles evolution is modeled using a multi-phase phase-field model. The effects of fission rate and initial grain microstructure on the recrystallization process are systematically investigated. The simulated dislocation density, sub-grain size distribution, and volume fraction of the recrystallization grains under different fission density are comparable with experimental data. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Topological Defects in Graphite under Irradiation using Molecular Dynamics

Graphite is the preferred material for moderating elements and structural components in a Very High Temperature Reactor (VHTR). It has a layered structure with basal planes constituted by carbon atoms arranged in a hexagonal lattice. Graphite accommodates a wide variety of defects such as vacancies, interstitials, and dislocation loops; a detailed description can be found in a review paper by Telling and Heggie. The layered configuration in graphite engenders energetic conditions that favor the formation of topological defects. Unlike the interlayer defects, which often are sp{sup 3} hybridized, topological defects maintain their sp{sup 2} bond connectivity. Several types of topological defects are known to arise in graphite; one of the most recognizable defects corresponds to a 5-7 Stone-Wales (SW) configuration that arises by a simple rotation of bonds. While the bond rotations result in a defected configuration, each carbon atom is still bonded to three other carbon atoms in the SW configuration, albeit with different bond lengths. Topological defects are expected to arise in irradiated graphite. A recent investigation with Raman and XPS spectroscopy have concluded that a large number of such defects are generated under neutron irradiation; an earlier ab-initio investigation has also arrived at the same conclusion. While several molecular dynamics (MD) simulations have been reported on irradiated graphite in the past, none have characterized the evolution of defects that are topological in nature. Using MD simulations, the current investigation quantifies the evolution and the distribution of topological defects under irradiation. Using MD simulations, we establish the presence of 5 and 7 membered topological defects in graphite following radiation. There is a greater probability for the formation of a 7 member ring relative to that of a 5 member ring. Since a heptagon causes a lower distortion to the 120 deg. bond angle of the sp{sup 2} hybridized orbital than a pentagon it is plausible to expect a larger number of 7 member rings following a radiation perturbation. Using a spatial distribution analysis, we have verified that 7 member rings tend to stay adjacent to 5 member rings as in Stone-Wales (SW) configuration. Since there are unequal number of 5 and 7 member rings, it follows that there are several isolated 7 member topological defects in graphite following radiation, possibly connected to dangling bonds or non-planar (sp{sup 3}) defects.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Radiation Effects in Innovative Structural Materials

The extreme radiation environments and demanding structural conditions of next-generation fission and proposed fusion reactors will require high performance materials with very high radiation resistance. Although conventional structural materials such as ferritic/ martensitic steels offer several attractive features, there is value in exploring alternative materials systems that may offer unprecedented radiation stability while also offering good thermomechanical properties. As noted elsewhere, options for designing radiation resistance include (a) the traditional approach of introducing a very high sink strength for recombination of radiation-induced vacancies and interstitials (e.g., Ti-modified stainless steels or 'nano-cluster-strengthened ferritic alloys', (b) utilizing materials with immobile vacancies at the operating temperature (e.g., silicon carbide), and (c) utilizing certain classes of materials that inherently exhibit superior resistance to radiation defect retention and accumulation (e.g., ferritic steels). Three relatively new but promising structural materials systems were selected for investigation, based on their favorable unirradiated mechanical properties and potential for good stability under irradiation: High entropy alloys (HEAs), bulk metallic glass (BMG), and MAX phase ceramics. The high configurational entropy in HEAs might alter point defect (solute and radiation defects) diffusion processes, thus potentially enhancing the recombination of radiation-produced damage. Therefore, HEAs offer the potential of a new class of materials for radiation resistance category (c) mentioned previously. A variety of BMGs such as Zr{sub 52.5}Cu{sub 17.9}Ni{sub 14.6}Al{sub 10}Ti{sub 5} offer good mechanical properties including fracture toughness values that may exceed 100 MPa-m{sup 1/2} and near net shape fabrication potential. The lack of crystallinity in BMGs prohibits the formation of conventional Frenkel pair defects under irradiation and therefore offer the possibility of another new material system for radiation resistance category (c). MAX phase ceramics (e.g., Ti{sub 3}SiC{sub 2}, Ti{sub 3}AlC{sub 2}, Ti{sub 2}AlC) have good mechanical properties and contain a nano-laminate structure that could provide a very high point defect sink strength. Due to their refractory nature, many of the MAX phase ceramics may exhibit negligible vacancy mobility up to high temperatures. Therefore, MAX phase ceramics offer the potential to provide radiation resistance in either categories (a) and (b), depending on the operating temperature. For this study, a novel 27%Fe-27%Mn-28%Ni-18%Cr single phase face centered cubic high entropy alloy was synthesized that exhibited good strength and ductility over a wide temperature range. Zr{sub 52.5}Cu{sub 17.9}Ni{sub 14.6}Al{sub 10}Ti{sub 5} (BAM-11) was selected as a prototypic advanced BMG. Several MAX phase ceramics (e.g., Ti{sub 3}SiC{sub 2}, Ti{sub 3}AlC{sub 2}, Ti{sub 2}AlC) that have good unirradiated thermal and mechanical properties were also investigated. The irradiation damage behaviors of these three novel materials systems have been studied after fission neutron at ∼75 deg. C up to 1 dpa and Ni ion irradiation at 25- 700 deg. C up to 30 dpa (lower temperatures and doses for the BMGs). All three sets of materials generally exhibited good radiation stability within certain temperature regimes. For example, the BMG exhibited only minor (<10%) change in strength that appeared to saturate after a dose of ∼1 dpa. A key shortcoming of current BMGs are that they can not be used at temperatures >>300 deg. C due to glass to crystalline transition at elevated temperature. The MAX phase ceramics irradiated at 400-700 deg. C appeared to be in the 'low temperature' regime where vacancies are immobile and therefore the radiation hardening saturated at relatively low doses. The Ti{sub 3}SiC{sub 2} MAX phase exhibited the best overall radiation stability over a wide temperature range. The HEA exhibited fine scale dislocation loop formation, no observable cavity swelling and reduced amounts of Cr depletion compared to traditional austenitic Fe-Cr-Ni alloys after ion irradiation, which suggests suppressed point defect and solute diffusion behavior. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2020

Critical Design Elements for an On-Site Mercury Concentration and Holding Facility - 20175

Administrative changes and funding issues have delayed the Department of Energy (DOE) from establishing and commissioning a National Repository for mercury as required under the Mercury Export Ban Act (MEBA) of 2008. In the interim, a few hazardous waste Treatment, Storage, and Disposal Facilities (TSDFs) have been authorized to accept mercury classified as a hazardous waste as under Subtitle C of the Resource Conservation and Recovery Act (RCRA). Waste generators are charged (either by weight or by volume) for the waste they manifest to these TSDFs. In many cases, mercury and mercury compounds have chemically sorbed to (or physically lodged into) the interstitial spaces of a substrate material. This almost always causes the amount of waste generated to be unnecessarily large, due to the extra volume (or weight) of the accompanying, non-mercury materials. Depending on a multitude of factors, it could be financially beneficial for waste generators to take greater control of their mercury affairs by separating these substrate materials from the mercury and mercury compounds that have become associated with them. Building and operating an on-site mercury Concentration and Holding Facility (CHF) could help relieve a waste generator's overall cost burden. The purpose of such CHFs is not to treat or dispose of mercury, but to recover and concentrate mercury from bulky mercury-laden materials to the extent that the host substrate to which the mercury may have previously been sorbed could pass a Toxicity Characteristic Leaching Procedure (TCLP) test and either be disposed of at a conventional landfill, or be disposed of at a hazardous waste landfill, but absent additional restrictions specifically for mercury. A CHF represents a way for waste generators to save money by separating mercury from common host materials, concentrating it, and then safely storing it until transportation to an authorized TSDF takes place This paper focuses on the critical design elements waste generators should consider when designing and operating a mercury CHF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Developing and Implementing New Waste Chemistry Controls for Hanford's Double-Shell Tanks - 20041

Waste chemistry controls for the Hanford double-shell tanks (DSTs) were established in the 1980's in response to tank failures caused by stress corrosion cracking (SCC) at the Savannah River Site (SRS). The controls were established based on a combination of SRS chemistry limits and corrosion testing results. Conditions in the Hanford DSTs have changed over time, with the temperature in the majority of the DSTs dropping below 50 deg. C. Testing indicated that the original chemistry control specification and SCC testing performed in 2010 did not protect against pitting corrosion mechanisms - what is believed to have caused the failure of tank 241-AY-102. New waste chemistry controls are necessary to protect the DSTs from both SCC and pitting corrosion. Savannah River National Laboratory (SRNL), with input from the Tank Integrity Expert Panel (TIEP) Corrosion Subgroup (CSG), performed a statistically based investigation of nitrate and halide ion induced pitting corrosion. The investigation was intended to develop a comprehensive waste chemistry envelope that minimizes the risk of both SCC and pitting caused by halide and nitrate ions. The waste chemistry envelope needed to be robust enough to address future waste retrieval and process changes which could significantly change the waste composition of the DSTs. The experimental design allowed the statistically significant ions and their effect on pitting corrosion to be determined. SRNL developed the 'pitting factor' which is an empirical relationship between the statistically significant inhibitive species (hydroxide, and nitrite), and pitting inducing species (nitrate, chloride, and fluoride). The coefficients are weighted factors that show the degree of influence the species have on the propensity for pitting. New waste chemistry control limits were recommended by SRNL, and concurred with by the TIEP CSG, that incorporated the pitting factor along with other limits. The controls developed to reduce the risk of pitting corrosion were evaluated to determine if they also protected against SCC. Hundreds of test results were reviewed and compared against the proposed limits. With the exception of one test result, instances where cracking was observed corresponded to chemistry that was outside of the proposed limits - indicating that the chemistry controls developed for pitting also adequately addressed SCC concerns. Existing tank composition estimates were evaluated against the new chemistry controls - three current tank chemistries are suspect, and may not meet the new chemistry control requirements. These included the interstitial liquid (liquid associated with solids) of DSTs 241-AN-102, 241-AN-106, 241-AN-107, and 241-AY-101. Measures are ongoing to evaluate the tanks with the suspect chemistry. These include varying combinations of: performing additional evaluations (history of use, modeling); DST specific laboratory testing; and, pursuing core samples of the DSTs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Statistical Study of Primary Knock-on Atoms under Neutron Irradiation in RMC Code

Materials in high flux neutron condition will suffer from an evolution of damage which ultimately restricts the lifetime of the fuel clad and reactor pressure vessel. Atoms in irradiated material transform to primary knock-on atoms (PKAs) by absorbing the kinetic energy transferred from neutrons, and the energy released from nuclear reactions. The secondary collision cascades which are induced by PKAs create point defects within the material, with residual vacancies and interstitials eventually causing swelling and embrittlement that degrades mechanical behavior. For a long time, Displacements per Atom (DPA) index has been identified as a measure of radiation damage in material. Nowadays, several advanced methods, such as Molecular Dynamics (MD) and Kinetic Monte Carlo (KMC) simulations, have been developed for more accurately simulating the evolution of the damage. However, all of these techniques need more comprehensive information of PKA, such as the type, energy, direction and spatial distribution. The limitations of DPA emerge and a method should be provided to simulate the generative process of PKAs. Recently, a deterministic analysis code SPECTRA-PKA has been developed by Gilbert to calculate the energy spectra of almost all types PKAs by solving the group-to-group cross-section matrices equation which includes neutron scattering, light charged particles, and recoils of the residual nuclei. But this deterministic analysis code could not give the direction, spatial and time distribution of PKAs. Based on the Geant4 program, Cui Zhenguo has simulated the irradiation by neutrons with 1 MeV energy in zirconium, while only the two adjacent elastic collisions between injected neutron and target atoms are taken into consideration. In this paper, Monte Carlo method to calculate PKAs information during neutron transport in RMC code is introduced. RMC is a continuous-energy Reactor Monte Carlo neutron and photon transport code being developed by Department of Engineering Physics at Tsinghua University, Beijing. PKAs in zirconium are calculated in fixed neutron source mode. Statistical information of PKAs in PWR fuel cell and SFR fuel cell is shown and compared. The results show that Monte Carlo method can provide more comprehensive information of PKAs. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗