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Tritium production is of critical importance to prospective DT fusion power plants. Lithium ceramic and beryllium based solid-type breeder blankets are an option for supplying the tritium required to sustain the DT plasma. This research investigates the time-varying tritium production in solid breeder blankets with different compositions. The breeder fraction was varied in conjunction with the Li 6 enrichment. The parameter study considered 198 different blanket compositions for three blanket thicknesses. The cheapest configuration capable of meeting the tritium requirements were found. The cost of Li 4 SiO 4 (including Li 6 enrichment) and Be 12 Ti were considered. The time-varying tritium production of each blanket configuration was simulated using the interface code, FATI, that couples the radiation transport code MCNP 6 with the inventory code FISPACT-II. Economical blanket configurations capable of self-sufficiency were found. The cost of producing excess tritium for start-up inventories was found to be between $18,000 and $27,000 per g. Fitting functions to predict the time-averaged tritium breeding fraction and the tritium inventory at five years, were obtained for inclusion in the PROCESS systems code. PROCESS is now able to consider different breeding blanket compositions and thicknesses when assessing the engineering, physics and economic feasibility of reactor designs. [ABSTRACT FROM AUTHOR]

PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system. PROCESS allows the user to choose which constraints to impose and which to ignore, so when evaluating the results it is vital to study the list of constraints used. New algorithms submitted by collaborators can be incorporated – for example safety, first wall erosion, and fatigue life will be crucial and are not yet taken into account. This paper describes algorithms relating to the engineering aspects of the plant. The toroidal field (TF) coils and the central solenoid are assumed by default to be wound from niobium-tin superconductor with the same properties as the ITER conductors. The winding temperature and induced voltage during a quench provide a limit on the current density in the TF coils. Upper limits are placed on the stresses in the structural materials of the TF coil, using a simple two-layer model of the inboard leg of the coil. The thermal efficiency of the plant can be estimated using the maximum coolant temperature, and the capacity factor is derived from estimates of the planned and unplanned downtime, and the duty cycle if the reactor is pulsed. An example of a pulsed power plant is given. The need for a large central solenoid to induce most of the plasma current, and physics assumptions that are conservative compared to some other studies, result in a large machine, with a cryostat 36 m in diameter. Multiple constraints, working together, restrict the parameter space of the optimised model. For example, even when the ratio of operating current to critical current in the TF coils is increased by a factor of five, the total coil cross-section decreases only a little, because of the need for copper stabiliser, insulation, and structural support. The result is that the plasma major radius hardly changes. It is these surprising results that justify the development of systems codes. [ABSTRACT FROM AUTHOR]

Highlights: [•] A three-stage compound cryopump could be used in fusion reactors. [•] Helium “ash” is adsorbed at 4.5K, and deuterium and tritium are adsorbed at 15–22K. [•] We used a free Direct Simulation Monte Carlo code (DS2 and DS2V). [•] 90% of the deuterium and tritium are successfully trapped at 15K. [•] Fuel is returned directly to the plasma, without isotopic separation or storage. [ABSTRACT FROM AUTHOR]

The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing. [ABSTRACT FROM AUTHOR]

High energy neutrons produced in future fusion reactors will cause significant transmutation reactions in the breeder blanket, including the important tritium breeding reactions. The reaction rate for a given reaction type depends on the neutron spectrum and material properties. The inventory consequences for tritium production of a solid-type breeder blanket are discussed in this paper. A DEMO fusion reactor with 19 homogeneous breeder blanket modules made of Eurofer, helium, Be 12 Ti and Li 4 SiO 4 , with enriched Li 6 content. Each blanket module was segmented radially in order to analyse the time-dependent reaction rate as a function of depth. The resolution of radial segmentation was varied and 5 radial divisions were found to be sufficient to accurately model the tritium inventory. Time-dependent tritium production was simulated with the use of the interface code, FATI, which couples radiation transport code MCNP 6 with the inventory code FISPACT-II. The simulated results show how the tritium production varies over the expected lifetime of the blanket. The overall tritium production of the solid-type blanket decreases as the Li 6 and Li 7 are burnt up. The effect of Li 6 burn-up in the breeder zones nearest to the plasma is identified as the main contributing factor to the decreasing tritium production of the whole breeder blanket. [ABSTRACT FROM AUTHOR]

In normal operation the JET neutral beam injectors have the operating gas supplied to the ion source and the neutraliser. For tritium operation the gas is supplied to both the ion source and neutraliser at a point close to the earth grid (“grid gas”) due to the difficulty in producing a gas line with a secondary containment and a ceramic break for high voltage standoff. In preparation for the next JET tritium campaign the JET EP2 PINIs have been characterised with grid gas flow. This paper reports measurements of arc efficiency, species and divergence in both normal and grid gas operation with hydrogen and deuterium. The data is used to predict the performance in tritium operation. [ABSTRACT FROM AUTHOR]