Analysis of different factors on the features of thermal destruction of a fuel pin

The presented work studies the influence of various factors that affect the specific features of fuel pins melting. For this purpose, fuel pins with different geometries and energy release are considered. Numerical simulation of melting is carried out using a program module for calculating the destruction of fuel rods. Comparison with theoretical calculations is made. The analysis of the convergence of calculations with respect to the time step value and the number of calculated cells along the radius and height is carried out. As a result of work with the use of numerical methods, the characteristic times of destruction of fuel elements during an accident with a loss of coolant flow rate (an accident of the ULOF type) and the dependence of weight loss on time are obtained under various conditions.

Structural integrity investigations of spent nuclear fuel with finite element modeling

Radioactive waste in Switzerland will be disposed of in a deep geological repository (DGR). Responsible for the planning and preparation of realization of this task is National Cooperative for the Disposal of Radioactive Waste (Nagra). Spent fuel assemblies (SFA) constitute the main high-level waste (HLW) stream that will be disposed in the DGR. Prior to final disposal they will be transferred or transported to an encapsulation plant, where they will be loaded into final disposal canisters. To ensure that the structural integrity of SFAs is not compromised during handling and transportation, it is desirable to characterize the expected mechanical parameters of SFAs after long-term interim storage. Experimental research activities performed at the JRC Karlsruhe include safety aspects of radioactive waste management, encompassing also spent fuel storage and spent fuel/HLW disposal activities. Nagra and JRC have established a collaboration to jointly study relevant properties and behaviours of spent fuel rods, with the support of the Gösgen nuclear power plant and of Framatome, and in collaboration with other partners in Europe and internationally. As part of this collaboration, 3-point bending and impact tests were performed at the hot-cell facilities of JRC Karlsruhe, to determine the mechanical response of spent fuel rodlets under quasi-static and dynamic loads. The structural integrity of fuel rods was also evaluated under different handling scenarios using finite element (FE) analysis. Starting with the construction of a static 3D FE model of a Pressurized Water Reactor (PWR) nuclear fuel rodlet in ANSYS Mechanical, Nagra has developed a series of FE models over the years. Mechanical properties of the original rodlet model were derived through an extensive validation process, using experimental data from the 3-point bending tests. To evaluate the mechanical response of an SFA in different loading scenarios, this model was expanded using 1D beam modeling approach. The development of the simplified 1D models is shown in this presentation. In particular, the effect of the contact formulation between the spacer grid and the rods is discussed. Finally, preliminary results of the bending response of a 15×15 PWR SFA sub-model are presented.

Diamond Coating Reduces Nuclear Fuel Rod Corrosion at Accidental Temperatures: The Role of Surface Electrochemistry and Semiconductivity

If we want to decrease the probability of accidents in nuclear reactors, we must control the surface corrosion of the fuel rods. In this work we used a diamond coating containing <60% diamond and >40% sp2 “soft” carbon phase to protect Zr alloy fuel rods (ZIRLO ®) against corrosion in steam at temperatures from 850 °C to 1000 °C. A diamond coating was grown in a pulse microwave plasma chemical vapor deposition apparatus and made a strong barrier against hydrogen uptake into ZIRLO® (ZIRLO) under all tested conditions. The coating also reduced ZIRLO corrosion in hot steam at 850 °C (for 60 min) and at 900 °C (for 30 min). However, the protective ability of the diamond coating decreased after 20 min in 1000 °C hot steam. The main goal of this work was to explain how diamond and sp2 “soft” carbon affect the ZIRLO fuel rod surface electrochemistry and semi conductivity and how these parameters influence the hot steam ZIRLO corrosion process. To achieve this goal, theoretical and experimental methods (scanning electron microscopy, Raman spectroscopy, electrochemical impedance spectroscopy, carrier gas hot extraction, oxidation kinetics, ab initio calculations) were applied. Deep understanding of ZIRLO surface processes and states enable us to reduce accidental temperature corrosion in nuclear reactors.

Idea of an experimental technique for quantitative passive gamma emission tomography on irradiated nuclear fuel assemblies

An idea is presented in which passive gamma emission tomography of irradiated nuclear fuel is developed to enable quantitative information of the spatial activity distribution of selected isotopes within the fuel rods of the assembly. The idea is based on using well-known calibration sources mounted in the measurement device during measurement. The image reconstruction would include the sources, thereby enable quantification of the activity distribution. Should the idea be proven viable, the outcome would be valuable to the global community dealing with characterisation of nuclear fuel in terms of safety, security, safeguards and fuel development.