Integrated multiscale experiment and model analysis of radially resolved microstructure and thermal conductivity in mixed oxide fuel
Not Available
Fuel performance modeling↗
Engineering topics
Cappia, Fabiola
Publications and source records attributed to Cappia, Fabiola.
Design and Evaluation of a Ring Tension Test Grip for Remote Mechanical Testing of Irradiated Tubular Specimens
Here, the ring tension test (RTT) is a mechanical testing method for determining bulk mechanical behavior in the circumferential or hoop direction for tubular materials. The test is especially useful for testing materials with anisotropic mechanical properties, such as zirconium alloys, which are commonly used as nuclear fuel cladding. Anisotropy requires direction-specific testing to determine the hoop strength. Historically, several RTT methods and grips have been used, each method has its strengths and weaknesses, and, in all cases, the measured strength is subject to uncertainty due to variations of the testing geometry and experimental tolerances. Recent analysis has shown that grips with a hemicylindrical mandrel configuration are recommended as the most robust configuration. The two strictest aspects to be controlled are the ability to determine gage region orientation and closely matching the size of the mandrels to the test specimen. This last requirement is particularly challenging when the dimensions of the specimen vary because of environmental effects such as dimensional changes due to irradiation. This paper presents a new RTT grip designed to incorporate this mandrel shape, hold the gage at the desired orientation, be suitable for remote operation in a hot-cell environment, and be adaptable for different sizes or variations in the specimen size. The general description and the unique design features of the test specimen and grips are given in detail. The performance of the grips in mechanical testing, including in a remote hot-cell environment, is also provided.
42 ENGINEERING↗
Length scale effects of micro- and meso-scale tensile tests of unirradiated and irradiated Zircaloy-4 cladding
Zircaloy-4 is an essential material for cladding structures within fission-based reactors. To explore the changes in properties measured on differing length scales, FIB-machined micro-scale tensile tests were performed on both irradiated and control groups of Zircaloy-4. This was correlated with tensile testing on femtosecond laser-machined meso-scale specimens. Pronounced size effects were found when varying specimen geometry. Increases in tensile geometry size were associated with a reduction in measured yield stress for both irradiated and unirradiated samples. Here, meso-scale testing found strength and strain values similar to that of bulk-scale testing.
36 MATERIALS SCIENCE↗
FUEL PERFORMANCE STUDIES AT IDAHO NATIONAL LABORATORY MAKING USE OF THE BYRON FUEL SHIPMENT
In December of 2023 a shipment of commercially irradiated fuel rods from the Byron Generating Station in Illinois was successfully shipped to the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The make-up of the rods includes a mix of cladding types from traditional ZirloTM, advanced zirconium alloys, and chrome coated ZirloTM. Burnups range from regular end of life values to over 70 GWd/MTU rod average. The R&D plan for the rods involves multiple projects from developing licensing data for new claddings to integral transient tests to support burnup extension efforts in the United States. The R&D began in early 2024 with the nondestructive examinations of the rods after which they will be sectioned for microscopy, mechanical testing, and analytical chemistry. Additionally, many rod segments will be refabricated into new test pins and inserted into a static water capsule for integral Reactivity Initiated Accident (RIA) and Loss of Coolant Accident (LOCA) testing at the TREAT reactor.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
THE U.S. ACCIDENT TOLERANT FUELS PROGRAM: RECENT ACCOMPLISHMENTS SUPPORTING IMPLEMENTATION
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Capps, Nathan↗
Status Report on Characterization of High Burnup Fuel with Advanced Nondestructive Pulsed Neutron PIE
Characterizing irradiated or spent nuclear fuels with pulsed neutron techniques provides microstructural data such as phase fractions as well as crystallographic data, e.g. lattice parameters, from diffraction analysis. Diffraction characterization is complemented by spatially resolved mapping of isotope densities from energy-resolved neutron imaging, in particular neutron absorption resonance imaging, and overall bulk isotope assay with better sensitivity for minority isotopes from neutron absorption resonance spectroscopy without spatial resolution. Furthermore, after characterization at ambient condition, heating of irradiated or spent fuel will allow to characterize differences of e.g. lattice thermal expansion or phase transition temperature and kinetics compared to fresh fuel as well as enable the study of disappearance of irradiation defects. This data enables benchmarking of predictions of properties of irradiated fuels for which otherwise experimental data is sparse. The effort described here strives to characterize a section cut from a high-burnup fuel. Volumes smaller than entire fuel pellets or rodlets as proposed here, e.g. sections cut from a fuel pellet, to pave the way to characterize entire pellets or rodlets in the future.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Enabling Digital Image Correlation for Deployment in the Hot Fuels Examination Facility
Digital image correlation (DIC) is a non-contact image-based technique for displacement and strain measurement. It is also a full-field method that can provide excellent spatial resolution. DIC has been widely used for measurements in many styles of mechanical tests and is desirable in part because of this flexibility to measure nearly any material across a range of time and length scales. Its ability to measure localized strain for tests and in environments not accessible to other strain measurement techniques makes it an ideal option for several mechanical test cases that exist in hot cell facilities such as the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). However, DIC has not been demonstrated at HFEF, and several barriers to implementation in the hot cell environment exist. This report outlines the possible paths for deploying DIC at HFEF, weighing the technical advantages and risks with each path. It also discusses the steps which must be taken to add this capability to the suite of post-irradiation examination (PIE) techniques available at HFEF. Finally, it concludes with recommendations for the best method to perform DIC at the facility.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Integrated Approach to Post-Irradiation Examination of Nuclear Materials at Idaho National Laboratory
Idaho national Laboratory (INL) is the U.S. lead national laboratory for the Department of Energy’s Office of Nuclear Energy (DOE-NE), providing much of the nuclear research, development and demonstration capability needed to move nuclear innovation forward to deployment. INL’s Materials and Fuels Complex hosts a unique combination of personnel, facilities and infrastructure and offers the ability to perform post-irradiation examinations (PIE) of nuclear materials spanning multiple length scales. The ability to combine engineering-scale analysis and sub-microscopic characterization provides valuable insights into fuel and structural material behavior and degradation mechanisms. The holistic approach is used to accelerate these materials demonstration and deployment. Selected studies will be presented, highlighting the impact of these techniques on improving fuel reliability, safety, and efficiency, thereby advancing the development of sustainable and advanced nuclear energy technologies.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗