Accident-tolerant fuels (ATFs) are designed to increase coping time following an accident scenario while preserving or improving current steady-state reactor operational performance. A potential ATF concept is using silicon carbide (SiC)-SiC-composite claddings. Fuel-performance simulations were conducted on a SiC-SiC-based cladding concept utilizing a multilayered approach for improved performance. This cladding concept is referred to in this paper as "the duplex concept" as it is a duplex structure composed of a monolithic SiC layer placed on the outside of an inner SiC-SiC composite layer. The monolithic SiC layer is used to provide gas tightness to the rod and protect the SiC-SiC-composite layer from exposure to the coolant. A liquid metal is added to the fuelcladding gap for improved thermal transport between the fuel and the cladding. In this work, the BISON fuelperformance code was used to conduct fuel-performance simulations on the cladding concept. Comparisons are made with a current prototypic fuel-rod design consisting of uranium dioxide (UO2) fuel enclosed in Zircaloy4 cladding under four relevant conditions. For condition I (normal operations) two representative steady-state cases were considered, one with a constant rod average heat rate, and one with an initially higher heat rate. For condition II events, a pellet-cladding interaction (PCI) ramp case was simulated to analyze potential anticipated operational occurrences. Condition III/IV transient responses during a loss of coolant accident (LOCA) and a reactivity-initiated accident (RIA) were also simulated. This computational study demonstrated that for normal operating conditions, the SiC concept cladding performed as well as the baseline for the standard-power cases evaluated. The ramping evaluations indicate potential for earlier fracturing of the SiC-SiC composite cladding compared to the Zircaloy-4 cladding due to the temperature gradient and the subsequent differential thermal conductivity degradation and swelling across the composite thickness. In condition III/IV events the SiC-SiC duplex concept remains intact after a 700 J/g RIA similar to Zircaloy-4 fuel systems. Under LOCA conditions, the duplex concept showed significantly improved performance, remaining intact in contrast to Zircaloy-4 which ballons and bursts.

This work evaluates the microstructural evolution of cold-worked, stress-relieved Zircaloy-4 cladding from pristine to uniaxial and biaxial deformed states. Differential aperture Laue diffraction and electron backscatter diffraction techniques are used to characterize intragranular strains, strain gradients, and grain fragmentation as metrics of deformation microstructure. The effects of mechanical anisotropy on deformation microstructure are investigated by comparing characterization results of samples subjected to different applied loads, including biaxial internal pressure and uniaxial tension along the rolled direction at 400 degrees C. Quantitative comparisons are made between the pristine microstructure and deformation-induced microstructure under both loading modes. Viscoplastic self-consistent simulations are performed to further investigate the microstructural evolution. Results indicate that biaxial loading from internal pressurization increases the deformation microstructure more than uniaxial loading along the rolled direction due to the relationship between loading and texture symmetry. Additionally, characterization results and simulations show distinct deformation-induced micro-textures: axial loading promotes a prismatic {10.0} fiber texture in the rolled direction, which strengthens the micro-texture inherited from pilgering, whereas pressure loading results in a {21.0} texture fiber, weakening the original micro-texture inherited from pilgering.

Zircaloy-4 cladding tubes were hydrogen-charged using a static pressure hydrogen charging method to experimentally produce a zirconium-hydride rim structure similar to that observed in Zircaloy-4 cladding tubes after operation in commercial light-water nuclear reactors. Detailed characterization was performed using electron energy loss spectroscopy (EELS) and four-dimensional scanning transmission electron microscopy (4D-STEM). By mapping the position of the plasmon peak in the low-loss EELS spectrum, it was determined that the zirconium hydrides formed were predominantly solid delta-phase zirconium hydrides. 4D-STEM was used to generate phase and strain maps at nanoscale. The interactions between two zirconium hydride platelets less than 200 nm apart leads to the localized lattice rotations and possible phase change, revealing a potential hydride growth mechanism.

This work evaluates the stress relaxation behavior of textured Zircaloy-4 cladding to understand how mechanical anisotropy influences pellet-cladding interactions. Uniaxial and biaxial stress relaxation tests are performed using full-tube axial tension and internal pressurization, respectively, aiming to achieve 0.25 %, 1 %, and 2 % equivalent strains in the cladding samples at a temperature of 300 degrees C. Internal pressure relaxation test results display enhanced stress relaxation compared to axial testing results, particularly for samples loaded beyond yield. Results of electron backscatter diffraction indicate increased deformation microstructure during loading and increased strain homogenization and recovery during relaxation for samples loaded via internal pressurization. Analysis indicates that the increased production and activity of basal (a) dislocations play a significant role in the enhanced relaxation measured in samples subjected to internal pressurization.

This work evaluates deformation activity in textured zircaloy-4 cladding with the aim of elucidating the mechanisms that govern mechanical behavior across a temperature range anticipated in reactor service. In particular, the dependence of mechanical anisotropy on temperature and stress state (ratio of applied stresses) is examined via uniaxial and biaxial tensile experiments at room temperature and 400 °C. The mechanical behavior is interpreted based on micro-texture and dominant slip system activity. Thermal activation of basal systems is found as a key mechanism affecting mechanical behavior across the temperature range. The biaxial stress states produce resolved shear stresses contrasting to uniaxial stress states. Mechanical testing and implications for reactor pellet-cladding mechanical interaction stress conditions are discussed.

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.

This paper presents the results of High-burnup Experiments for Reactivity-initiated Accident (HERA) Modeling & Simulation (M&S) exercise. The HERA project under the Nuclear Energy Agency (NEA) Second Framework for Irradiation Experiments (FIDES-II) program is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity-Initiated Accident (RIA) conditions. The Part I M&S cases are based on a series of tests in the Transient Reactor Test (TREAT) facility in the United States and the Nuclear Safety Research Reactor (NSRR) in Japan. The purpose of this work is to evaluate the test design to accomplish its goals in establishing clearer understanding of the effects of power pulse width during RIA conditions. The blind predictions using various computational tools have been performed and compared amongst to interpret the behaviors of high burnup fuels during RIA. While many international participants evaluate the thermal-mechanical behavior of fuel rod under different conditions, a considerable scatter of outputs comes out for the cases due to the disparity between codes in predicting mechanical behaviors. In general, however, the results of thermal-mechanical analysis elaborate that nominal design conditions the shorter pulse width tests in NSRR should cause cladding failures while the TREAT tests appear to have more split prediction of failure or not. Furthermore, the sensitivity analysis varying key testing parameters reveals the considerable effect of power pulse width and total energy deposition on prediction of fuel rod failure.

I-Loop is an irradiation facility that is currently being installed at the Advanced Test Reactor. It is a two-loop test facility capable of performing Light Water Reactor (LWR) irradiations in prototypic coolant conditions. The two loops are being installed to be capable of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) pressure, temperature, and chemistry environments. Each loop is nominally dedicated as a BWR or PWR for simplicity of operations. In-reactor water loop testing that an I-Loop provides is key to the deployment of new accident tolerant fuel technologies and other advanced LWR fuel concepts. Currently, pressurized water loops are the only testing facilities available to test BWR fuel concepts. Their test environments are non-prototypic at higher pressure/temperature and at single-phase fluid flow conditions. This void in the LWR test bed capabilities is one that the I-Loop is uniquely situated to provide. This report discusses the mechanical design, thermal hydraulic calculations, and neutronic calculations of a proposed standard experiment of accident tolerant BWR fuel concepts. Mechanical design examines the geometry and features of the main components. Thermal hydraulic calculations examine the modeling and results of the two-phase flow options available. Lastly, neutronic calculations examine the Monte Carlo analysis of enrichment, heat rates, and flux spectrum.