Thermo-mechanical analysis of iron-chromium-aluminum (FeCrAl) alloy cladding for light water reactor fuel elements

Alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys, in order to improve the accident tolerance of light water reactor (LWR) fuel. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys that exhibit much slower oxidation kinetics in high-temperature steam than Zr-alloys. This behavior should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely.This dissertation documents efforts to develop fuel performance capabilities to assess the behavior of FeCrAl cladding during normal and transient reactor operating scenarios. Within this work, simulations were performed for FeCrAl cladding using constitutive models and representative reactor operating conditions implemented into the finite-element fuel performance code BISON.Simulations were performed targeting the cladding behavior during normal operation of a boiling water reactor using boundary conditions derived from neutronics data. These simulations indicate that the fuel compliance plays a much larger role in the evolution of the cladding stress state after gap closure for the FeCrAl cladding than for Zircaloy. Individual sensitivity analyses of the fuel and cladding creep responses were then performed, which indicated the influence of compliance for each material, separately, on the stress state of the fuel cladding.To improve calculations of the fuel expansion and compliance, an additional investigation was performed to assess the role of creep, relocation, and explicit fracture in the fuel. Fuel rods using each of these models are simulated under representative conditions and compared to test rod measurements. This analysis provides a start toward the development and incorporation of explicit fracture in fuel performance analysis.Additionally, performance and stability under transient conditions must also be demonstrated for FeCrAl cladding. This analysis focused on modeling the integral thermo-mechanical performance of FeCrAl clad uranium dioxide fuel during transient reactor operation. Results from this simple analysis show similar bursting time and temperature between both FeCrAl and Zircaloy cladding, however, beyond cladding burst in these conditions, the superior high temperature oxidation kinetics of the FeCrAl cladding significantly reduce hydrogen gas production and provide longer fuel integrity

Numerical Analysis of Structural Behavior Inside a Pressurized Water Reactor (PWR)

The mechanisms of creep for zircaloy-4 have been studied for many years. The creep data published in the recent 35 years were collected and analyzed to identify different creep mechanisms (dislocation-glide creep, power-law and power-law breakdown creep and diffusional creep), based on the forms of the relationships between stress, temperature and strain rate. This identification allowed the activation energies and other associated creep parameters to be derived for each mechanism. The creep parameters were used to construct a deformation-mechanism map for zircaloy-4 that shows the conditions under which different mechanisms are dominant. This multi-mechanism creep model was implemented into the ABAQUS CREEP subroutine to study the effect of creep on the structure deformation and stress evolution. This subroutine allowed the finite-element analysis to select the most proper creep mechanism naturally based on the local stress and temperature, which improved the accuracy for structures with a complex geometry and stress distribution. The multiple-mechanism creep model was coupled with a wear model to study the in-pile relaxation of the contact force between the spacer grid and cladding, and the evolution of the wear profile before formation of a gap. These two processes occur at different time scales. Therefore, an effective-cycle technique was developed to couple the two mechanisms in a fashion that combined an acceptable level of both efficiency and accuracy. The simulations indicated that two stages exist during the relaxation of the contact force: partial slip and full slip. During partial slip, wear damage occurs at the edges of the contact region. Creep is the dominant relaxation mechanism during the partial-slip, and allows the wear scar to propagate across the entire contact, which causes a transformation from the partial slip to full slip. Once full-slip occurs, the contact forces are relatively low, and the creation of the wear scar becomes the dominant relaxation mechanism. In this regime, reducing the wear coefficient and the amplitude of excitation force delays the formation of a gap between the grid and cladding. The wear profile developed during full-slip occurs homogeneously. For a given initial interference, there is a master curve for the wear scar, which does not depend on the friction coefficient, the amplitude of the excitation pressure, or the wear coefficient. In addition to vibration, the coolant can also cause corrosion to the cladding. The reaction between water and zirconium on the cladding surface produces hydrogen in addition to the oxide. Hydrogen diffuses into the cladding and reacts with zirconium below the oxide layer to produce δ-hydride (ZrH1.66). δ-hydride formation is associated with a volumetric expansion of about 17%, which causes both misfit stress and geometrical deformation. Therefore, a multi-scale framework was developed to simulate hydrogen diffusion and hydride formation in the cladding. The hydride formation and growth were simulated by others using a phase-field model at the mesoscale. Then, the results of the phase-field model were used in a continuum-level finite-element analysis to study the effect of hydride on the structural behavior. The multiscale framework was first used to simulate an experiment for validation. Then, the framework was used to study the hydride formation in the cladding. Hydride forms a rim on the cladding outer surface with a maximum thickness of 0.12 mm. The hydride volume fraction distribution is plotted at different times.PHDMechanical EngineeringUniversity of Michigan, Horace H. Rackham School of Graduate Studieshttps://deepblue.lib.umich.edu/bitstream/2027.42/138624/1/wanghai_1.pd

Hydrogen contribution to the thermal expansion of hydrided Zircaloy-4 cladding tubes

International audienceThis study is focused on the hydrogen-induced dimensional change or "growth" of zirconium alloys. Dilatometric experiments were performed on samples taken from a unirradiated Zircaloy-4 (Zy-4) fuel cladding loaded up to 940 wppm hydrogen. Samples were taken in the axial direction of the tube or at 45° to the axial and transverse directions. The results indicate that hydrogen-induced expansion is anisotropic. Theoretical expansion calculations were carried out considering the partition of hydrogen in solid solution and hydrides together with the material crystallographic texture. Hydride-induced expansion was calculated using two different assumptions reported in the literature, namely "Pure Lattice Transformation Strains" (PLTS) and "Pure Shear Transformation Strains" (PSTS). Calculations based on the PSTS hypothesis satisfactorily predicted the anisotropy observed in the dilatometric curve. Under this assumption, the contribution of hydrides to the axial growth of high-burnup Zy-4 cladding is limited to 12%. This study shows it is important to consider the respective contribution of hydrogen in both states, together with the material crystallographic texture, to understand the dilatometric behavior of hydrided zirconium alloys

26th International QUENCH Workshop : 6-9 December 2021, virtual event organized by Karlsruhe Institute of Technology Karlsruhe, Germany

The 26th International QUENCH Workshop was organized by KIT as a virtual event during 6-9 December 2021. The main topics are related to nuclear safety during reactor accidents and long-term dry intermediate storage of used fuel elements. One focus was on accident tolerant fuel (ATF) cladding behavior during design-basis and severe accidents. 80 participants from 16 countries participated in the event with 30 scientific presentations and fruitful discussions

Development of M5 Cladding Material Correlations in the TRANSURANUS Code: Revision 1

The technical report is based on an earlier research on material properties of the M5 structural material. Complementing this research with new M5 data found in open literature, a set of correlations has been developed for the implementation to the TRANSURANUS code. This includes thermal, mechanical, and chemical (corrosion) properties of M5. As an example, thermal capacity or burst stress correlations have been proposed using the available experimental data. The open literature provides a wide range of experimental data on M5, but for some quantities they are not complete enough to be suitable for the implementation to the TRANSURANUS code. A balanced consideration of similarity of M5 characteristics to those of Zircaloy-4 (Zry-4) or E110 have therefore led to the recommendation to use some of these data selectively also for M5. As such, creep anisotropy coefficients of E110 are recommended to be used also for M5.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

Evolution of microstructure and nanoscale chemistry of Zircaloy-2-type alloys during nuclear reactor operation

Zirconium alloys are used as fuel cladding tubes in nuclear reactors. During reactor operation, these alloys are degraded by corrosion, hydrogen pickup (HPU), and radiation-induced growth, processes influenced by the alloying elements. The alloy Zircaloy-2, which contains Sn, Fe, Cr, Ni, and O as alloying elements, is commonly used in boiling water reactors (BWRs). This thesis deals with atom probe tomography (APT) investigations of Zircaloy-2 and a similar model alloy, Alloy 2, before and after up to nine years of BWR operation. Alloy 2 contains more Fe and Cr and exhibits lower corrosion and HPU. Less than 10 wt ppm each of Fe, Cr, and Ni was observed in the matrix of as-produced Zircaloy-2 and Alloy 2 of commercial heat treatment, a consequence of very low solubility and formation of second phase particles (SPPs). After reactor exposure, these elements were found in nanoscale clusters that were located at radiation-induced 〈a〉-type dislocation loops. The amount of Fe, Cr, and Ni in clusters increased with increasing fluence. There were two main types of clusters, spheroidal Fe–Cr clusters and disc-shaped Fe–Ni clusters. On average there were no large differences in clusters before and after acceleration in degradation, only small increases in cluster number density, cluster size, and cluster Cr content. 〈c〉-component loops decorated with Sn, Fe, and Ni were observed after but not before acceleration in degradation. Sn formed a network-like structure. No differences in cluster and matrix chemistry between Zircaloy-2 and Alloy 2 were observed after reactor exposure, indicating that the improved properties of Alloy 2 are related to additional Fe and Cr being located in SPPs.It was possible to analyse the materials using voltage-pulsed APT. Voltage pulsing was needed to reliably determine Fe–Ni cluster composition and shape. Fe–Cr clusters were observed also using laser-pulsed APT. Focused-ion-beam (FIB) preparation of APT specimens at room temperature resulted in phase transformation from α-Zr to γ-hydride, whereas cryo-FIB preparation did not. The average number of ions detected before specimen fracture was higher for γ-hydride specimens. There were no significant differences in clustering of Fe, Cr, and Ni between α-Zr and γ-hydride specimens