This paper presents a new feature of the Finite Element Method (FEM) code FENNECS (Finite ElemeNt NEutroniCS) for modelling reactivity control systems. The interface between materials within a finite element (usually referred to as a mixed element) is modelled using a flux-weighting method. While the method has demonstrated its accuracy in modelling the vertical movement of control rods in traditional Light Water and Fast Reactors (LWR and FR), it has limitations in modelling the rotation of control drums.The projection-based cusping treatment is another method that defines an effective homogenized cross-section for the mixed element. Unlike the flux-weighting method, this method does not involve any approximations. Instead, it exactly solves the weak form of the neutron diffusion equation.In order to illustrate the appropriate implementation of the method in the code, three exercises were solved. A comparison with the legacy flux-weighting model was performed, highlighting the benefits of the projection-based de-cusping method.In every case, if it is not completely removed, the cusping effect is mitigated, enabling the production of a solution compatible with nuclear safety analysis. Furthermore, it has been demonstrated that the projection-based method clearly outperforms the flux and volume weighting method in terms of accuracy.In terms of runtime, the projection-based method has demonstrated an average reduction of 40% for control rod exercises, while control drum exercises have shown a reduction of 15%.
The present paper deals with scaling in nuclear -system thermal -hydraulics (SYS TH), including the connection with Nuclear Reactor Safety Technology (NST). The paper is entirely derived from the S -SOAR document issued by CSNI of NEA, NEA/CSNI/R(2016)14, 2016 (Bestion et al., 2016). Scaling has constituted 'an issue' since the beginning of the exploitation of nuclear energy for civil purposes, with main reference to the generation of electricity. A Nuclear Power Plant (NPP) constitutes a technologically complex industrial system and it is characterized by the impossibility of, or the large difficulty in, characterizing the system's performance under the conditions of the design. So, models were designed, constructed, and operated under downscaled ranges of values for one or more of selected parameters (e.g. power, volume, height, pressure, etc). These features lay at the origin of the scaling issue, i.e. the difficulty in demonstrating that a model behaves like the prototype. Integrated definitions of the widely adopted terms, 'scaling', 'scaling issue', and 'addressing the scaling issue' are part of the present document. The related application domain includes the NST, and the licensing for water-cooled nuclear reactors under operation, under construction, or under an advanced
This paper contains a comparison of five modelling approaches for a simplified nuclear waste repository in a domal salt formation. It is the result of a four-year collaboration between five international teams on Task F of the DECOVALEX-2023 project on performance assessment modelling. The primary objectives of Task F are to build confidence in the models, methods, and software used for performance assessment (PA) of deep geologic nuclear waste repositories, and/or to bring to the fore additional research and development needed to improve PA methodologies. This work demonstrates how these objectives are accomplished through staged development and comparison of the models and methods used by participating teams in their PA frameworks.Participating teams made a wide range of model assumptions, ranging from compartmentalized networks to full 3D models of the salt formation and repository. Despite differences in the modelling strategies, all models indicate that salt compaction and diffusion of radionuclides in brine are key processes in the repository. For the isothermal spent nuclear fuel and vitrified waste scenario with multiple early failures considered, all models indicate little of the disposed radionuclides will migrate beyond the repository seal over the 100,000-year simulations. In general, the model output quantities have the largest differences over the short term and near the waste. Disparities between the models are believed to be due to differing simplifications from the conceptual model.
A sealed container for the geological disposal of spent nuclear fuel and vitrified high-level waste is the only component of a deep geological repository that provides complete containment of radionuclides. As such, attention is focused on its lifetime. The lifetime of the container is influenced by material degradation processes during disposal and is typically of the order of several millennia and, for some container materials, up to one million years. Designing, manufacturing, and predicting the performance of containers over such long periods requires an in-depth understanding of their material properties, fabrication processes, and degradation mechanisms. Scientific and technological progress can improve both the performance of containers and the robustness of lifetime predictions. Optimization of these aspects is of primary importance for many national radioactive waste disposal programs. In this article, the state of the art of complex coupled degradation processes, as well as the optimization potential of novel container materials, is presented. Furthermore, the existing tools allowing the prediction of long-term barrier integrity are discussed.
