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Fabrizio Franza

Tritium transport analysis in HCPB DEMO blanket with the FUS-TPC Code

(KIT Scientific Reports ; 7642)

AutorFranza, Fabrizio

VerlagKIT Scientific Publishing, Karlsruhe


UmfangX, 74 S.






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In thermonuclear fusion reactors, the fuel is an high temperature deuterium-tritium plasma, in which tritium is bred by lithium isotopes present inside solid ceramic breeder (e.g. Li-Orthosilicate) or inside liquid eutectic alloys (e.g. Pb-16Li alloy). In the breeding areas a significant fraction of the tritium produced is extracted out from the Breeding Zone by the He gas purging the breeding ceramic in the Helium Cooled Pebble Bed (HCPB) blanket concept or transported in solution by the owing alloy in the Helium Cooled Lead Lithium (HCLL) blanket concept.
Tritium produced in the breeding blanket by neutrons interacting with lithium nuclei can enter the metal structures, and can be lost by permeation to the environment. Tritium in metallic components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents and the need for maintenance.
In this study the problem of tritium transport in HCPB DEMO blanket from the generation inside the solid breeder to the release into the environment has been studied and analyzed by means of the computational code FUS-TPC (Fusion Devoted-Tritium Permeation Code). The code has been originally developed to study the tritium transport in HCLL blanket and it is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. The code has the main goal to estimate the total tritium losses into the environment and the tritium inventories inside the breeder, inside the multiplier, inside the purge gas and the main coolant loops and inside the structural materials.
Different simulations of the code were performed by adopting the configuration of the European HCPB blanket for DEMO.
Total tritium losses from a generic fusion power plant, is often considered a key parameter to evaluate the tritium containment capabilities (added to tritium inventories) of a certain nuclear plant. Without any tritium control techniques, permeation can be quite significant, thus some tritium transport mitigation devices are required. The code is able to model and compute different tritium fluxes exchanged in the overall tritium system. A sensitivity study for the tritium losses and inventories is performed in this work.