DEVELOPMENT OF MULTIPHYSICS, SEMI-AUTOMATIC, AND MULTISCALE TOOLS FOR BREEDING BLANKET DESIGN

Nuclear fusion reactors could be a long-term solution to the world's energy supply problem. However, this technology has several critical points that have not yet enabled its industrial deployment, such as fuel supply and the machine operating conditions from the point of view of the physical phenomena involved, either in terms of neutronics or thermohydraulics and thermomechanics.One of the key components of a nuclear fusion reactor that will have the fuel production function for its self-sufficiency is the breeding blanket. Given the complexity of this component and the multi-physical nature of the loading conditions under which it will operate, it is easy to conclude that its design studies require a complex approach that typically consists of several phases. The first phase involves the study of the component's nuclear response that provides the thermal power data deposited by neutrons and photons in order to proceed, in the second phase, to the study from the thermohydraulics point of view. This allows to evaluate a temperature spatial distribution that is arisen in the component to be provided as input, together with the other mechanical loads, for the evaluation of the proposed design solution's structural performance. Lastly, the results obtained are compared with the reference standards to determine whether the component studied is capable of operating under the expected loading scenarios without incurring incipient structural crisis conditions or, in any case, without violating any of the imposed design requirements. This modus operandi is typically characterized by two major bottlenecks which are the use of different computational software and the need to start over with all the calculation when one is faced with the need to make geometric changes, thus involving a considerable computational and time burden. Hence, the objective of the following PhD thesis is to define two methodologies to study the interconnected phenomenologies involved during the deployment of the breeding blanket to arrive at a design definition and verification.The first chapter will give a brief description of the fusion reaction, the breeding blanket, and the roadmap defined by the EUROfusion consortium for the coming years for the development of this technology. Both the study of neutron transport and the thermomechanical problem will also be explored from a theoretical point of view.The second chapter, on the other hand, presents a multi-scale and semi-automatic methodology coupling neutronics and thermal analysis that allows, starting from a radial power density profile and a neutron source from which it is derived, to obtain a tubular and/or plate design of a breeding blanket portion located anywhere in the reactor. The procedure, at first, is explained in detail in each of its steps, starting from the generation of a first attempt geometry through a simplified approach developed in MATLAB and then verified with ANSYS parametric design language, to its neutron analysis, using the Monte Carlo N-Particle code with use of unstructured mesh. Once the definition of the procedure and its validation are completed, it is applied to a particular breeding blanket architecture, namely the Water-cooled Lead Ceramic Breeder. The results validated the methodology and made it possible to exploit one of the generated geometries for a parametric study on the influence of the blanket material composition with respect to the tritium breeding ratio. A total of nineteen models were studied with the same neutron boundary conditions by varying the composition first of the first wall then of the breeding zone allowing noteworthy results to be obtained. It has been shown how, the use of materials with moderating qualities, can lead to a greater exploitation of tritium breeding reaction from lithium 6 in the first 40 cm of breeding zone and to a greater overall tritium breeding ratio of the local model, allowing a possible reduction of the radial breeding blanket thickness.The third chapter describes another multi-scale and semi-automatic methodology for the study of a particular stellarator-type nuclear fusion reactor, which by nature has a very high geometrical complexity. In particular, the procedure follows a back-and-forth flow in which, starting from local thermal analyses, thermal fields are derived, which are then averaged to obtain average temperatures to be associated with the components of a less detailed global model in order to study it at the structural level while maintaining the actual boundary conditions of the system. Once the results of this analysis are derived in terms of displacements, they are then applied as boundary conditions for the local model's mechanical analysis. Lastly, it is validated from a mechanical point of view through the application of RCC-MRx code rules. The methodology was applied to a DEMO Dual-Coolant Lithium-Lead breeding blanket architecture adapted to the HELIAS 5B geometry, which was examined for this study. The results, which are not relevant from a design point of view since that it wasn't the main objective, showed how the procedure is highly versatile and functional, making it a viable alternative for future studies in the field of stellarator-type reactors.Both activities were developed through the use of codes written with different programming languages, Python being the main one, and some of the most important have been listed and explained in the appendix.