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Tritium recovery in fusion reactors is one of the main goals in R&D, as a limited inventory is available and its uneconomic production. That is the reason why efficient technologies are indispensable to be developed in order to achieve fast tritium recovery and its subsequent reuse in the system for increasing its self-sufficiency. In this work a flexible tritium recovery demonstrator prototype based on permeation against vacuum concept has been designed and manufactured, as well as all necessary equipment for a Pb15.7Li loop implementation in order to test and demonstrate that an in-pipe integrated solution is possible, and at the same time, to validate the manufacturing process. Thus, efficient rates for more optimized future models could be then extrapolated. The aim of this paper is to show the different testing results that have been carried out in this research project. These results include permeation properties of the material considered for the permeator, as long as it has been manufactured with a novel technique, Selective Laser Melting. They also include vacuum tests on the permeator to quantify possible leakages and to set up and analyze the capability to generate vacuum inside the permeator, and finally, permeation tests with the prototype, in a first stage with a gas mixture of hydrogen and argon inside the loop instead of Pb15.7Li.

The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

A fast and efficient recovery of bred tritium is a major milestone of tritium breeding technologies R&D for the demonstration of a fusion reactor tritium self-sufficiency. Permeator against vacuum (PAV) runs as a single-step process for tritium on-line recovery, acts as passive systems allowing to be thermally governed can be easily in-pipe integrated in LiPb loop systems and can be conceived with high compactness. An optimal PAV design is proposed with detailed design parameterization of tritium recovery efficiency at different velocity ranges from numerical simulation based on properly developed Openfoam® CFD code BelFoam® customized solver. Diverse structural design options are being considered to manufacture the PAV component that presents diverse manufacturing concerns. Fabrication options are exploring rolling of thin plate with internal armor in the vacuum gap up to the coating of a porous controlled thick wall. The aim of this paper is to emphasize on the advanced performance of the PAV proposed concept, to present and discuss the different technical solutions that have been studied as well as the fabrication tests carried out.

A preliminary neutronic assessment of the performances of a helium-cooled Li8PbO6 breeding blanket for the conceptual design of a DEMO fusion reactor is given. The study mainly focuses on TBR, power density responses and shielding factor optimization to estimate the feasibility of the design under the prescribed radiation deposition limits at TF-coils superconducting magnets. Computational analyses are based on three-dimensional 30° sector using the Monte Carlo code MCNPX 2.6. The scoping interest of helium-cooled Li8PbO6 blanket designs is based on a large potential minimization of the amount of Be required and the strong relaxation of 6Li enrichment requirements for this solution when compared to other solid breeder blanket options.

The fast and efficient recovery of bred tritium represents a major milestone of tritium breeding technologies R&D and is key for the demonstration of fusion reactor fuel self-sufficiency. For lead–lithium eutectic, diverse technologies are currently being investigated and qualified. Permeator Against Vacuum (PAV) solution represents a firm candidate because: (i) runs as a single-step process for tritium on-line recovery, (ii) works passively allowing to be thermally governed, (iii) can be easily in-pipe integrated in Pb15.7Li loop systems and (iv) can be conceived with high compactness. An optimal design of a PAV requires a detailed hydraulic design optimization for established operational ranges. An optimal PAV design is proposed and qualified by numerical simulation.

A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO_FUS based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils and material damage (dpa, gas production) to estimate the operational life of the steel-made structural components in the blanket and vacuum vessel (VV). In order to optimize the shielding of the coils different combinations of water and steel have been considered for the gap of the VV. The used neutron source is based on an axi-symmetric 2D fusion reaction profile for the given plasma equilibrium configuration. MCNPX has been used for transport calculations and ACAB has been used to handle gas production and damage energy cross sections.

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.