Phase transformation of the corrosion layers on nuclear fuel cladding. Raman spectroscopy study.

The goal of this work was the phase analysis of corrosion layers on zirconium alloys. In the environment of nuclear reactors, zirconium alloys are covered with a protective layer of zirconium oxide, which occurs in two crystalline modifications - monoclinic and tetragonal. The distribution of these phases in the corrosion layer can affect the overall corrosion rate. Raman spectroscopy was used to determine the composition of the corrosion layers. The use of this method is advantageous because the monoclinic and tetragonal phases can be easily distinguished in the spectra of the corrosion layers. In total, samples of two alloys were measured. The samples were pre-exposed at 360 °C in Li+ containing water (70 mg/l Li as LiOH) . Exposure times were between 21 d and 231 d, so the series contained both pre- and post- transition samples. The relative proportion of the tetragonal phase decreases significantly after the transient. It has also been found that the corrosion layers are highly heterogeneous in terms of the distribution of crystalline modifications.

Computational Geometry-Based 3D Yarn Path Modeling of Wound SiCf/SiC-Cladding Tubes and Its Application to Meso-Scale Finite Element Model

The filament winding process is a competitive performing technology for nuclear fuel cladding tubes due to its high automation. The study of the yarn path on the mandrel surface is vital to design and produce the cladding tube with the desired mechanical properties, reducing manufacturing time and costs. The geodesic and semi-geodesic trajectories are used to create a 3D yarn path in this paper. A 3D yarn path optimization method based on the principle of minimum potential energy is proposed to simulate the overlap effect in accord with the real winding process. The finite element (FE) mesh based on the 3D yarn path has been used for the mechanical analysis of the cladding tube. The embedded region constraint is applied to define the interaction between the matrix mesh and the yarn mesh to model the meso-structure of the cladding tube. Based on the meso-scale FE model, the mechanical behavior of the wound SiCf/SiC nuclear fuel cladding tube is studied in detail. The results show that due to the neglect of the overlap effect, the conventional laminate model overestimates the cladding tube strength. The proposed meso-scale FE model can accurately predict the failure of the cladding tube. The results also confirm that the creation of a 3D yarn path and the derived meso-scale FE model, representing an accurate wound structure, are of importance to the prediction of the performance of the cladding tube.

On the Use of Chromium Coating for Inner-Side Fuel Cladding Protection: Thickness Identification Based on Fission Fragments Implantation and Damage Profile

Inner-side coatings have been proposed as a complementary solution within the accident tolerant fuel (ATF) framework, to provide enhanced protection for the nuclear fuel cladding. Unlike external surface, the degradation of irradiated internal cladding surface has not been studied extensively. Fission fragments produced during the fission of nuclear fuel is one of the key players in this degradation. This study aimed to estimate the minimum thickness of the thin chromium film, required to protect the inner side of the nuclear fuel cladding. The approach used is based on a set of calculations, of Ion ranges and damage profiles, for a group fission fragments, using the TRIM code. The calculation results were verified by comparison with the experimental data associated with the phenomena of the inner cladding degradation of thermo-releasing elements. The recommended minimum thickness for such a film was found to be 9 microns. Calculations also showed that chromium metal has a greater stopping power compared to the zirconium-based alloy E110, which indicates an increased ability of chromium to withstand exposure to energetic fission fragments during reactor operation.

Multiscale Modeling of SiCf/SiC Nuclear Fuel Cladding Based on FE-Simulation of Braiding Process

A generalized multiscale (micro-macro) finite element (FE) model for SiC-fiber reinforced SiC-matrix ceramic (SiCf/SiC) nuclear fuel claddings is established. In the macro level, the solid mesh of braided preform, which can be tailored by machine settings (braid angle, yarn width, and so on), is generated based on the braiding process simulation using the dynamic FE-solver, hiring the contact constraints. The matrix mesh and the yarn mesh are integrated by the embedded region constraint, with which the meshing difficulties can be avoided. In the micro-UD model, the progressive damage of the ceramic matrix is modeled using the phase field method (PFM) and the fracture is captured by Mohr–Coulombs criterion, which are stable and efficient in the description of the brittle crack initiation, coalition, and branching. Based on this multiscale model, the mechanical behavior of the braided SiCf/SiC nuclear fuel cladding tube is studied in detail. The superiorities over the homogenized tube model are demonstrated, too.