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At least 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A Practical Irradiance Model for Bifacial PV Modules

A model, suitable for a row or multiple rows of photovoltaic (PV) modules, is presented for estimating the backside irradiance for bifacial PV modules. The model, which includes the effects of shading by the PV rows, is based on the use of configuration factors to determine the fraction of a source of irradiance that is received by the backside of the PV module. Backside irradiances are modeled along the sloped height of the PV module, but assumed not to vary along the length of the PV row. The backside irradiances are corrected for angle-of-incidence losses and may be added to the front side irradiance to determine the total irradiance resource for the PV cell. Model results are compared with the measured backside irradiances for NREL and Sandia PV systems, and with results when using ray tracing software.

14 SOLAR ENERGY↗
DOE OSTI2019

Establishing the World Reference for Measuring the Atmospheric Longwave Irradiance with Traceability to the International System of Units

Advancing climate change research requires accurate and traceable measurement of atmospheric longwave irradiance. Measurement capabilities are limited to an estimated uncertainty of larger than +/- 6 W/m2 using the interim World Infrared Standard Group (WISG). Two independently designed and calibrated absolute radiometers measuring downwelling longwave irradiance were compared during five outdoor comparisons in 2013, 2015, and 2017 at the Physikalisch Meteorologisches Observatorium Davos World Radiation Center (PMOD/WRC) and the U.S. Department of Energy (DOE) Atmospheric Radiation Measurement program (ARM) in the Southern Great Planes (SGP). Two Absolute Cavity Pyrgeometers (ACPs) developed by the National Renewable Energy Laboratory (NREL) and four Integrating Sphere Infrared Radiometers (IRISs) developed by PMOD/WRC took part in these intercomparisons.From the five comparisons, the difference between the irradiance measured by ACPs and IRISs varied from 0.2 W/m2 to 2.5 W/m2 based on atmospheric conditions, which is within the combined stated uncertainties of +/- 3 W/m2. The irradiance measured by the WISG is lower than the average irradiance measured by ACPs and IRISs. The magnitude of difference varied from 0.2 W/m2 to 6.6 W/m2 depending on the integrated water vapor. A concerted effort to establish a world reference for measuring the atmospheric longwave irradiance with lower uncertainty and with traceability to the International System of Units (SI) by using the ACPs and IRISs as the reference started at the World Meteorological Organization (WMO). The Commission for Instruments and Methods of Observation Commission (CIMO) Task Team on Radiation References (TT) meeting was held in 2017 at the National Physical Laboratory (NPL), in Teddington, United Kingdom to discuss specific recommendation on the traceability of the atmospheric longwave irradiance to the International System of units (SI).

14 SOLAR ENERGY↗
DOE OSTI2018

UNCERTAINTY QUANTIFICATION OF CALCULATED FUEL TEMPERATURE FOR AGR-3/4 IRRADIATION EXPERIMENT

The third in a series of Advanced Gas Reactor (AGR) experiments conducted at the Idaho National Laboratory, AGR-3/4, was completed in April 2014, resulting in irradiation of the tristructural isotropic fuel for 369 effective full power days. A number of designed-to-fail fuel particles were embedded along the central line of each irradiated compact to provide a known source of fission products for subsequent transport through the compact matrix and structural graphite materials. The fuel temperature in each capsule during irradiation, a significant factor in the release and transport of fission products, was maintained within a predefined range representative of a high temperature gas reactor. Fuel temperatures were not measured directly because contact between a thermocouple and the fuel could lead to unwanted particle failures. An ABAQUS-based finite element heat transfer code, calibrated based on temperatures measured in the surrounding graphite rings, was used to predict fuel temperatures during irradiation. To quantify the uncertainty of calculated fuel temperatures, thermal model parameters with high sensitivity and/or large uncertainty were identified. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of calculated temperatures. Quantification of input uncertainties over the extended irradiation period of the AGR-3/4 experiment is a challenging task due to changes in capsule thermal properties and geometry caused by exposure to neutron irradiation at high temperatures. In the absence of direct measurements or complete knowledge about these changes, inputs to thermal model and associated uncertainties are estimated based on modeling assumptions and expert judgment. Unplanned events that occurred during irradiation contribute to input uncertainties as well. For example, shrinkage of fuel compact and graphite rings led to significant increase in the gap size uncertainties. This paper focuses on uncertainty quantification of calculated fuel temperature for AGR-3/4 capsule thermal models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Yttrium Hydride Post-Irradiation Examination Plan

Micro nuclear reactors are expected to use metal hydride moderators to ensure thier compact design for the flexible operational purposes. Department of Energy’s Microreactor program therefore aims to provide infrasture and qualification bed for the development of micro nuclear reactors. Because micro nuclear reactors must be compact to offer a flexible operational purposes, metal hydrides are expected to be used as neutron moderator materials due to their high hydrogen number density. Los Alamos National Laboratory (LANL) directed by the DOE to provide a qualified fabrication techniques of metals hydride, especially the yttrium hydride (YHx) which provides advantages over other metal hydride candiates. Qualification of YHx involves fabrication, property, and performance pillars, which must be met for the deployment of microreactors. In the support of the qualification process metal hydride, LANL has developed fabrication techniques based in direct hydriding and powder compacting. For both manufacturing techniques, fresh material properties of metal hydrides have been reported in the literature. To establish the performance limits and the qualification parameters, an irradiation campaign has been initated at Idaho National Laboratory (INL). The irradation campaign aims to investigate the critical irradiated YHx properties as a function of neutron fluence and temperaute. The first phase of the irradiation campaign includes the effect of the operating temperature under neutron irradiation which provides initial data on the metal hydrides and informs the manufacturing processes. So that, drop-in irradations in Advanced Test Reactor (ATR) has been designed. This document depicts the post-irradation examination (PIE) plan of the ATR irradiated YHx. Objective of the PIE plan is to serve as a qualification basis for YHx as a neutron moderator material by understanding the impacts irradiation has on the physical and thermophysical properties; to establish a link between fabrication and the irradiated YHx properties; to institute a PIE strategy for other solid-state neutron moderator candidates, and; to inform micro nuclear reactor fuel performance and safety codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A preliminary study on irradiation growth behavior modelling of N36 zirconium alloy tube

As a new kind of Zr-Sn-Nb series alloy, N36 has not been explored in irradiation growth behavior before. In this paper, four empirical correlations including power, logarithmic, exponential and hyperbolic tangent law were used to build irradiation growth model of N36 zirconium alloy preliminarily. Before that, lengths of the fuel rods with N36 claddings from pool-side measurement were analyzed to get irradiation growth strains, which were then used to fit the four empirical correlations mentioned above. Distribution of fast fluence along axial direction of tubes was considered in all irradiation growth models. The fitting results showed that the hyperbolic tangent law fit irradiation growth strains the best, while power law did the worst. In order to improve the precision of the hyperbolic tangent model, several improvements were done. Verification and validation of the improved hyperbolic tangent irradiation growth model suggested that the size of second phase may have an influence on irradiation growth behavior of N36 tubes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Microstructural Characterization of Irradiated Fuels and Materials for the United States High Performance Research Reactor Fuel Development Program and the Fuel Cycle Research and Development Program

In an effort to contribute to the development of advanced nuclear fuel for research reactors, monolithic and dispersion plate fuels will be examined to determine how irradiation parameters such as fission density, fission rate, and irradiation temperature, as well as mechanical parameters such as composition and fabrication, affect the final microstructure of the fuel, post irradiation. Electron Backscatter Detection (EBSD) analysis capabilities of irradiated materials will be developed as well as improvements on irradiated sample preparation in order to provide higher quality samples for SEM, FIB, and TEM. In addition, image analysis techniques developed at Idaho National Laboratory (INL) will be employed to examine microstructural features such as fission product precipitates and fission gas bubbles. The main goals are to further develop EBSD capabilities on irradiated material; characterize microstructure of irradiated fuel to support qualification; develop a consistent, effective process for image analysis using INL developed software; and provide repeatable, statistically-confident, critically assessed data to further modeling efforts. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Additive Manufacturing of Research Reactor Control Elements and Subsequent Neutron Irradiation

Additive manufacturing techniques have made major strides in recent years and are now capable of creating complex components in near-final form. The nuclear industry has many applications with complicated and even unique components that are time consuming, expensive, and more difficult to make using traditional fabrication routes. Modern fabrication techniques possess the potential to reduce fabrication lead-time and costs while increasing quality and efficiency. However, these methods have not been adopted due to the need for strict validation for nuclear applications. Unlike structural components that largely consist of a uniform material, for nuclear reactor core components incorporation of fissile or neutron poison materials inside a cladding becomes necessary. This in turn requires a more complex and flexible manufacturing method to produce these hybrid components. In this study, demonstration of ultrasonic additive manufacturing (UAM) process to produce components resembling research reactor control elements was carried out]. Specifically, the control elements for ORNL's High Flux Isotope Reactor (HFIR) were targeted. These control elements (consisting of an inner cylinder and four outer plates) are placed on the periphery of the circular HFIR core. The control elements consist of Al6061-T6 alloy with a 13-cm-long grey and a 56-cm-long black region, corresponding to 40 wt% of Ta and 30 wt% Eu{sub 2}O{sub 3} (europia) are incorporated along their length. In this study various fabrication parameters for production of aluminum blocks with and without embedded Eu{sub 2}O{sub 3} and Ta bearing compacts were explored. Select optimized specimens we then irradiated inside HFIR for one reactor cycle. The UAM process exploits ultrasonic energy to facilitate solid-state welding of metal foils in contact under pressure. This additive process is carried out in successive layers to build up the material and when coupled with CNC milling, it yields complex 3D components. The milling and subsequent additive steps may be used to embed foreign material into the material undergoing buildup. In this manner fissile or neutron poison elements may be incorporated into the complex structure, though in an inhomogeneous fashion. Multiple Al-6061 blocks were produced by UAM using 25 mm wide by 0.15 mm thick Al-6061-H18 tapes. The mechanical properties of these blocks in the as-fabricated condition was characterized in detail and it was shown that a great degree of anisotropy in the mechanical behavior exists between the direction in the tape plane and the direction perpendicular to the tapes (buildup direction) The latter exhibited reduced modulus and significantly smaller strength. Post production heat treatment appeared to greatly enhance the strength in the buildup direction. Multiple tensile specimens machined per SS-J3 geometry were prepared for irradiation testing. The tensile specimen matrix was set up to enable post irradiation examination of mechanical properties in various directions as well as after heat treatment. A small block with embedded Ta- and Eu{sub 2}O{sub 3}-bearing compacts, three of each kind, was also produced, as demonstrated previously. These compacts were 5.7 mm in diameter and 1.6 mm in height and contained 60 vol Ta or Eu{sub 2}O{sub 3}. The poison-bearing Al-6061 block alongside the tensile bars was loaded into a perforated rabbit (irradiation capsule) for irradiation in HFTK. The perforated rabbit allows penetration of HFIR water coolant into the Interior of the capsule in direct contact with the test specimens. The rabbit was loaded into the flux trap region at the uppermost position in the core. Irradiation temperature was 68 deg. C and fast neutron (E>0.1 MeV) fluence was 10{sup 25} neutrons/m{sup 2}. Currently PIE activities are ongoing that are focused on mechanical testing of irradiated tensile bars and metallographic examination of neutron poison bearing block.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Microstructure Investigations of U{sub 3}Si{sub 2} Irradiated by Heavy Ions at LWR Temperatures

Searching for a nuclear fuel material with enhanced accident tolerance compared to the conventional UO{sub 2} fuel is one of the key objectives of the accident tolerant fuel (ATF) campaign. Having both high thermal conductivity and high uranium load, U{sub 3}Si{sub 2} is expected to provide lower operating temperature/stored energy and advantageous neutron economy as a light water reactor (LWR) fuel material. Therefore, U{sub 3}Si{sub 2} has been regarded as a promising candidate fuel material by the ATF community. To evaluate the qualification of U{sub 3}Si{sub 2} as an ATF fuel in LWRs, its fuel performance during both steady-state operation and power transients must be well understood. As a typical nuclear fuel material applied in research reactors, the fuel performance of U{sub 3}Si{sub 2} at low temperatures (<300 deg. C) has been extensively studied through both experimental and computational approaches. However, under LWR conditions, the temperature of U{sub 3}Si{sub 2} can go beyond 600 deg. C. In that case, the references collected in U{sub 3}Si{sub 2}'s applications in research reactors are inadequate for the validation of U{sub 3}Si{sub 2} as an LWR ATF. Experimental efforts that are dedicated to the LWR conditions must be made to help examine U{sub 3}Si{sub 2}'s qualifications as well as provide valuable references for the development related models in advanced fuel performance code. Gaseous fission products accumulate in nuclear fuel during operation and form fission gas bubbles. The evolution of those fission gas bubbles results in increase of volume (swelling), degradation of thermal conductivity, and eventually fission gas release, compromising the integrity and performance of fuel elements in multiple aspects. Hence, fission gas behavior is one of the most important components of fuel performance. In-pile irradiation and corresponding post-irradiation examinations (PIEs) undoubtedly provide most reliable information on the actual performance of a nuclear fuel. However, the high cost and long lead time involved in in-pile irradiation limit its wide applications. On the other hand, heavy ion irradiation is capable of simulating the neutron effects on materials, especially the behavior of energetic fission fragments in nuclear fuel materials, to a considerable extent. Therefore, heavy ion irradiation was used in this study to investigate the microstructural evolution of U{sub 3}Si{sub 2} under LWR-temperature irradiation, focusing on the difference from low temperature irradiation and fission gas behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Prioritizing Nuclear Materials for SAM-3 Neutron Irradiation Campaign: Structural and Cladding Materials Candidates

This report outlines a framework for selecting structural and cladding materials for the Nuclear Science User Facilities (NSUF) SAM-3 neutron irradiation campaign to support the advancement of nuclear energy technologies. The document begins with an introduction that provides background context, highlights the motivations for launching a new irradiation campaign, and defines the overall objectives. The core of the report describes the design considerations for the irradiation campaign, including capsule configurations, irradiation temperature ranges, and target dose levels (defined by displacements per atom, or dpa). The material recommendation was guided by the Specimen Identification and Prioritization (SIP) Working Group, a multidisciplinary team of experts representing national laboratories, academia, industry, federal government and agency. This group played a central role in identifying candidate materials, evaluating technical justifications, and ensuring alignment with boarder programmatic goals. A detailed set of criteria for material prioritization is then presented, taking into account reactor relevance, performance gaps, advanced manufacturing methods, and emerging material classes. Based on the input of SIP working group, specific materials were selected and justified for inclusion in the irradiation campaign by the NSUF leadership and its U.S. Department of Energy (DOE)-Office of Nuclear Energy (NE) management. The final section provides recommended capsule designs, summarizing critical parameters such as material type, fabrication method, sample geometry, irradiation conditions, and specimen quantities. This report serves as a foundation for executing a focused and high-impact neutron irradiation campaign aimed at addressing key materials challenges for both existing and advanced nuclear reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Dynamic Scaling Analysis of Accelerated Irradiation Testing on Additive Manufacturing Materials by Positron Annihilation

The timely applications of Additive Manufacturing (AM) materials in nuclear environments require accelerated irradiation tests, mainly ion irradiation to enable rapid prototyping. Low dose ion irradiation would cause sub-nanostructure changes by generation of lattice defects, vacancies, vacancy clusters and voids and void swelling caused by cellular dislocations. Positron Annihilation Lifetime (PAL), a novel technology, sensitive towards sub-nanostructure morphology with high accuracy (about 10-7 vacancy per atom), supported by Transition Electron Microscope (TEM) would be applied to identify the type and total size of the defects. The subsequent PAL measurements and TEM surface studies would be followed by PAL analysis that includes sophisticated trapping model. The PAS results would become an input to dynamic scaling analysis (that predicts radiation effects from low dose studies for high dose effects), which incorporate mean-field theory model. The final effect is an in-depth understanding of the microstructure evolution of AM materials under ion irradiation which can be extrapolated to the studies of neutron irradiation, since ion-irradiation takes less time and do not cause the irradiation hazard. The working hypothesis is that PAL technology, that have excellent sensitivity to low-defect concentration would help to identify ion-induced material damage on the atomic and nano-scale level, which then could be extrapolated to understand the neutron damage better.

accelerated irradiation testing↗
DOE OSTI2021

Mechanical Performance and Mesostructure Analysis of Proton-Irradiated Fused Filament Fabrication Acrylonitrile Butadiene Styrene Material

The use of fused filament fabrication (FFF) acrylonitrile butadiene styrene (ABS) and other thermoplastics in radiation environments is beginning to be studied as possible replacements for traditionally manufactured parts and tools. Interlayer adhesion within the mesostructure has been shown in published literature to be an integral component in the strength of FFF ABS. Research of irradiated 3D printed polymers has primarily focused on the influence of gamma irradiation on the mechanical properties of FFF ABS samples, without evaluating its impact on the mesostructure of the samples. The purpose of this paper is to understand the mechanical damage caused by proton radiation on FFF ABS samples through the evaluation of the mesostructure of the ABS samples. To achieve this objective, proton radiation at 40 MeV was applied to FFF ABS samples at radiation doses up to 1.0 MGy at a dose rate of 1 MGy/hr. Following the irradiation, tensile testing was performed on the samples. Here, the fractured surfaces of the tested samples were subsequently observed using a scanning electron microscope. The mechanical test results show an increase in the ultimate tensile strength (UTS) and a decrease in the ductility of the irradiated samples. Statistical analysis on the results shows that there is a statistically significant difference in the UTS and the ductility of unirradiated and 1.0 MGy irradiated samples and between irradiated samples and 1.0 MGy samples. In addition, the percentage elongation at break has statistically significant differences in the means between irradiated samples and 1.0 MGy samples. The UTS has statistically significant differences in the means between 0.1 MGy and 1.0 MGy. The difference in the means between the unirradiated and the 1.0 MGy samples is the most significant for both the UTS and elongation at break. The scanning electron microscopy (SEM) results indicated that interlayer adhesion improved as a function of radiation dose corresponding to the increase in tensile strength. The SEM results also showed that crazing and plastic deformation were reduced; aligning with the loss in ductility observed in the tensile tests results. The proton radiation is causing these mechanical and physical changes through two mechanisms: (1) dose rate effects on ionizing radiation-induced oxidative degradation; and (2) radiation heating effects of high energy (>1 MeV) charged particles.

electron microscopy↗
DOE OSTI2018

Loop and void damage during heavy ion irradiation on nanocrystalline and coarse grained tungsten: Microstructure, effect of dpa rate, temperature, and grain size

Displacement damage, through heavy ion irradiation was studied on two tungsten grades (coarse grained tungsten (CGW) and nanocrystalline and ultrafine grained tungsten (NCW)) using different displacement per atom rates and different irradiation temperatures (RT and 1050 K). Percentage of <111> and <100> type loops at the irradiation conditions was determined. Irradiation damage in the microstructure was quantified using average loop areas and densities (method A) and loop areal fraction in the grain matrices under 2-beam diffraction conditions (method B). Average values of <111> and <100> loops were calculated from method A. Loop coalescence was shown to occur for CGW at 0.25 dpa. Using both methods of quantifying microstructural damage, no effect of dpa rate was observed and damage in CGW was shown to be the same at RT and 1050 K. Swelling from voids observed at 1050 K was quantified. The loop damage in NCW was compared to CGW at the same diffraction and imaging conditions. NCW was shown to possess enhanced irradiation resistance at RT regarding loop damage and higher swelling resistance at 1050 K compared to CGW. For irradiation at 1050 K, the NCW was shown to have a similar defect densities to the CGW which is attributed to higher surface effects in the CGW, vacancy loop growth to voids and a better sink efficiency in the CGW deduced from the vacancy distribution profiles from Kinetic Monte Carlo simulations. Loop density and swelling was shown to have similar values in grains sizes that range from 80-600 nm. No loop or void denuded zones occurred at any of the irradiation conditions. This work has a collection of experiments and conclusions that are of vital importance to materials and nuclear communities.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Nano-scale microstructure damage by neutron irradiations in a novel Boron-11 enriched TiB 2 ultra-high temperature ceramic

Ultra-high temperature transition-metal ceramics are potential candidates for fusion reactor structural and plasma-facing components. We reveal the irradiation damage microstructural phenomena in Boron-11 enriched titanium diboride (TiB 2 ) using mixed-spectrum neutron irradiations, combined with state-of-art characterization using transmission electron microscopy (TEM) and high resolution TEM. Irradiations were performed using High Flux Isotope Reactor at ~220 and 620 °C up to 2.4 x 10 25 n.m -2 (E > 0.1 MeV). The calculated dose including contribution from residual Boron-10 ( 10 B) transmutation recoils, was ~4.2 displacements per atom. TiB 2 is susceptible to irradiation damage in terms of dislocation loop formation, cavities and anisotropic swelling induced micro-cracking. At both 220 and 620 °C, TEM revealed dislocation loops on basal and prism planes, with nearly two orders of magnitude higher number density of prism-plane loops. HRTEM, electron diffraction and relrod imaging revealed additional defects on {10$ \overline{1}\ $0} planes, identified as faulted prism-plane dislocation loops. High defect cluster density on prism planes will induce a-lattice parameter swelling of TiB 2 , as reported in literature. This anisotropic lattice parameter swelling induced grain boundary micro-cracking, the extent of which decreased with increasing irradiation temperature. The dominance of irradiation defect clusters on prism planes is different than typical hexagonal ceramics where dislocation loops predominantly form on basal planes, causing c-lattice parameter swelling. Helium generation and temperature rise from the transmutation of residual 10 B resulted in matrix and grain boundary cavities for the irradiation at 620 °C. In conclusion, the study additionally signifies isotopic enrichment as a viable approach to produce transition-metal diborides for potential nuclear structural applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A combined theoretical-experimental investigation of thermal transport in low-dose irradiated thorium dioxide

During reactor operation, nuclear fuels are subject to extreme temperature and irradiation conditions which can significantly degrade the fuel's thermal transport properties. The reduction in thermal conductivity of the fuel as a result of irradiation-induced lattice defects is arguably the most important fuel performance metric in regard to reactor efficiency and safety. Because thorium dioxide (ThO2) is suitable as a model system for more complex materials such as UO2 and its mixed oxides, we present a theoretical investigation of thermal conductivity reduction seen in defect-bearing thorium dioxide and compare directly to experimental measurements. Phonon-mediated thermal transport of the fuel is modeled by a solution to the Boltzmann transport equation (BTE) for phonons. A cluster dynamics (CD) model for lattice defect evolution during irradiation predicts defect densities which are used as input to the BTE for modeling phonon-defect scatterings. Phonon scatterings by lattice defects include those from point defects and vacancy clusters and interstitial clusters of various sizes. The CD model is benchmarked against structural defect characterization of irradiated thorium dioxide using electron microscopy. Thermal conductivity predicted by the BTE model is compared to measured values for irradiated thorium dioxide specimens below room temperature to isolate effects of phonon-defect scattering from intrinsic 3-phonon processes, which dominate at higher temperatures. The computed conductivity values are in partial agreement at temperatures close to room temperature while slight deviations are observed at the lowest measured temperatures, suggesting that implemented phonon-defect scattering cross-section expressions may not be adequate for low temperatures. The presented work provides a necessary investigation of the influence of irradiation induced defects on fuel performance and represents a first step toward a full characterization of phonon mediated thermal transport in irradiated materials with complex defect microstructure.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Response of 11 B enriched ZrB 2 ultra-high temperature ceramic to neutron irradiation at elevated temperatures

ZrB 2 , an ultra-high temperature ceramic (UHTC) is being considered for use in fusion reactor first-wall structures, yet its response to irradiation remains poorly understood. This study employed scanning/transmission electron microscopy (S/TEM), synchrotron X-ray diffraction (XRD), finite element calculations, and thermal property measurements to thoroughly investigate the neutron-irradiation effects on 11 B-enriched ZrB 2 . Neutron irradiations were conducted at 220 °C and 620 °C, with a neutron fluence of 2.2 × 10 25 neutron/m 2 (energy > 0.1 MeV), resulting in 3.9 dpa and 4200 appm He. The study revealed the unusual prevalence of prism loops and a > c anisotropic lattice swelling, likely linked to the low c/a ratio of ZrB 2 , leading to grain boundary microcracking. Reducing the grain sizes was effective in reducing intergranular cracking and macroscopic swelling. The observation of cavities in ZrB 2 irradiated at 620 °C, as opposed to 220 °C, prompts questions about the temperature at which vacancies in ZrB 2 become mobile, and the role of neutron absorption by 10 B in elevating irradiation temperatures. Isotopic enrichment in 11 B proves to be a viable strategy for mitigating helium production in transition-metal diborides, which is a critical consideration for nuclear applications. Irradiation-induced defects reduce the thermal diffusivity and conductivity of ZrB 2 by a factor of 4–9, which has important implications for its role as a plasma-facing material in fusion reactors that drive high heat fluxes through first-wall materials. Here, this comprehensive study lays the foundation for understanding ZrB 2 behavior under neutron irradiation and highlights important phenomena to consider for various material applications.

36 MATERIALS SCIENCE↗
DOE OSTI2018

In-situ TEM study of the ion irradiation behavior of U 3 Si 2 and U 3 Si 5

We report that U 3 Si 2 and U 3 Si 5 are two important uranium silicide phases currently under extensive investigation as potential fuel forms or components for light water reactors (LWRs) to enhance accident tolerance. In this paper, their irradiation behaviors are studied by ion beam irradiations with various ion mass and energies, and their microstructure evolution is investigated by in-situ transmission electron microscopy (TEM). U 3 Si 2 can easily be amorphized by ion beam irradiations (by 1MeV Ar 2+ or Kr 2+ ) at room temperature with the critical amorphization dose less than 1 dpa. The critical amorphization temperatures of U 3 Si 2 irradiated by 1MeV Kr 2+ and 1MeV Ar 2+ ion are determined as 580±10K and 540±5K, respectively. In contrast, U 3 Si 5 remains crystalline up to 8 dpa at room temperature and is stable against ion irradiation-induced amorphization up to ~50 dpa by either 1MeV Kr 2+ or 150KeV Kr + at 623K. In conclusion, these results provide valuable experimental data to guide future irradiation experiments, support the relevant post irradiation examination, and serve as the experimental basis for the validation of advanced fuel performance models

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Pair distribution function analysis of neutron-irradiated silicon carbide

We have employed x-ray total scattering to investigate the structure of polycrystalline 3C-silicon carbide following neutron irradiation. Here, the structure as a function of irradiation temperature and dose was quantified by analyzing pair distribution functions. Although the SiC matrix retains its crystal structure after irradiation, a significant increase in the diffuse scattering component is observable indicating that neutron irradiation leads to changes in both the short- and medium-range order. These changes include both an irradiation dose- and temperature-dependent increase in the vacancy concentration leading to an increase in the Si and C atomic displacement parameters. A dose-dependent decrease in the size of defect free material is also quantified from the structural refinements due to an increase in the number of defects and defect clusters. Evidence of additional correlations in the short-range order (up to ~4 Å) from differential pair distribution function analysis indicate that combinations of atomistic defects including anti-site defects, vacancies and defect clusters are present after these irradiation conditions. Such structural information will be valuable for direct comparison of experimental and simulated atomic structures of irradiated silicon carbide.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Helium bubble formation in nickel under in-situ krypton and helium ions dual-beam irradiation

Helium has profound effects on the microstructure and mechanical property of materials used in nuclear power systems. Here to understand the interaction between helium and irradiation-induced defects and its effect on bubble formation, single-beam and dual-beam irradiations on nickel were performed with 16 keV helium ions and 1 MeV krypton ions under in-situ TEM observation. For 1 MeV krypton single-beam irradiation, voids were observed at 600ºC but not at 500ºC and 700ºC. For helium single-beam irradiation at 500ºC, helium moved freely and formed bubble-loop complexes. For dual-beam irradiation at 500ºC, helium was trapped by vacancies and formed bubbles homogeneously. The structure of dislocation loops was also affected by the presence of helium. For krypton single-beam irradiation at 500ºC, the microstructure was {100} periodic wall of faulted and perfect dislocation loops. For dual-beam irradiation at 500ºC, periodic wall was not observed, and dislocation loops were predominantly perfect. For single-beam helium at 500ºC, dislocation loops were produced by athermal process of SIAs ejection or loop punching mechanisms. The interplay between helium and vacancies played an important role on the evolution of helium bubbles and defect microstructures.

36 MATERIALS SCIENCE↗