F. Franza

Tritium Transport Issues for Helium-Cooled Breeding Blankets

Tritium mobility through breeding blanket (BB) and steam generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason, it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modeling and parametric issues. In this paper, the main permeation issues are identified and possible solutions are discussed to address the modeling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled BBs: 1) helium-cooled pebble beds and 2) helium-cooled lithium-lead. For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC to define proper and feasible tritium mitigation techniques, which are needed to keep the tritium losses lower than the allowable environmental release (i.e., 20 Ci/d).

Sensitivity study for Tritium Permeation in Helium-Cooled Lead-Lithium DEMO Blanket with the FUS-TPC Code

Abstract Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of hydrogen and its isotopes in materials. We address the problem of tritium transport in Helium Cooled Lead-Lithium (HCLL) DEMO blanket from lead-lithium breeder through different heat transfer surfaces to the environment by developing a computational code (FUS-TPC). The main features of the code are briefly described and a parametric study is performed in order to identify the most influencing parameters in terms of tritium releases into the environment and of tritium inventories. The results showed that the results are strongly affected by the tritium Sievert’s constant in Lead-Lithium and the efficiency of permeation barriers.

Tritium permeation issues for helium-cooled breeding blankets

Tritium permeation through Breeding Blanket and Steam Generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration in order to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modelling and parametric issues. In this study the main permeation issues are identified and possible solutions are discussed to face the modelling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled breeding blankets, i.e.: 1) Helium-Cooled Pebble Beds (HCPB) and 2) Helium-Cooled Lithium-Lead (HCLL). For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC in order to define proper and feasible tritium mitigation techniques which are needed to keep the tritium losses lower than the allowable environmental release (i.e. 20 Ci/d).

Design integrated system for power plant development

In the KIT activities for the development of Tokamak fusion reactors are performed since more than one decade. Recently an approach has been launched for the creation of a design integrated system. This system will support the analysis work towards an optimized Tokamak DEMO reactor configuration. A system analysis tool aiming to describe main plasma physics and engineering features will be combined with semiautomatic systems for creation of reactor configurations and sets of boundary conditions Thus, data is generated and provided to different engineering platforms dedicated to each reactor system respectively. These platforms are the basis for detailed optimization and improvement of conceptual designs. This paper focuses of the semi-automatic computational tools in between the system code and the engineering platforms which will be used to create and provide analysis data and boundary conditions.

A High Field Tokamak Neutron Source Facility

Fusion creates more neutrons per energy released than fission or spallation, therefore DT fusion facilities have the potential to become the most intense sources of neutrons for material testing. An Ignitor-like device, that is a compact, high field, high density machine could be envisaged for this purpose making full use of the intense neutron flux that it can generate, without reaching ignition. The main features of this High Field Neutron Source Facility, which would have about 50% more volume than Ignitor, are illustrated and the R&D required in order to achieve relevant dpa quantities in the tested materials are discussed, in particular the adoption of superconducting magnet coils. Radiation damage evaluations have been performed by means of the ACAB code, showing the potential of high field, neutron-rich devices for fusion material testing. Few full-power months of operation are sufficient to obtain significant radiation damage values (in terms of dpa) of large size samples (~m3). The setup of a duty cycle for the device in order to obtain such operation times is discussed. The problem of radiation damage to the insulator of the Toroidal Field Coils has been explored. Two strategies for mitigating damage to the TF coil insulators have been demonstrated, and it is likely that both will need to be implemented to ensure the survival of the insulating material for the lifetime of the tokamak.