The tritium balance in DEMO reactor is a key factor for the success of the production of energy from the thermonuclear fusion reactor. Three of the four breeder blankets (BBs) concepts candidates for DEMO used the eutectic Pb-16Li enriched at 90% in 6Li as breeder, the water-cooled lithium-lead, dual-coolant lithium-lead, and helium-cooled lithium-lead (HCLL) BBs; therefore, the design and characterization of the tritium extraction and removal system (TER) from PbLi with high efficiency are a critical issue in the European Roadmap. In ITER research reactor, a PbLi BB concept will be qualified, the HCLL BB. An HCLL test blanket module, PbLi loop, instrumentations, and auxiliary systems will be characterized with the support of European infrastructures. However, the tritium extraction unit from PbLi (TEU) selected and designed for ITER is based on the gas-liquid contactor technology, a reliable technology but with less than 40% efficiency. Instead, the candidates TER technologies for the lead-lithium loops of DEMO BBs, the permeator against vacuum (PAV) and vacuum sieve tray (VST), will not qualified in ITER, because these systems will not be fully mature by the start of the reactor. PAV and VST can theoretically achieve efficiency above 80%. The present works aim to analyze the technologies candidates for ITER and DEMO reactors, describe and compare TEU and TER design for each concept of BB and the integration of TER in DEMO tokamak building taking into account two design requirements: self-sufficient sustainable of fusion nuclear reactor and safety requirements.

Tritium Extraction from HCLL/WCLL/DCLL PbLi BBs of DEMO and HCLL TBS of ITER

Utili M.;Tincani A.;Martelli D.;
2019

Abstract

The tritium balance in DEMO reactor is a key factor for the success of the production of energy from the thermonuclear fusion reactor. Three of the four breeder blankets (BBs) concepts candidates for DEMO used the eutectic Pb-16Li enriched at 90% in 6Li as breeder, the water-cooled lithium-lead, dual-coolant lithium-lead, and helium-cooled lithium-lead (HCLL) BBs; therefore, the design and characterization of the tritium extraction and removal system (TER) from PbLi with high efficiency are a critical issue in the European Roadmap. In ITER research reactor, a PbLi BB concept will be qualified, the HCLL BB. An HCLL test blanket module, PbLi loop, instrumentations, and auxiliary systems will be characterized with the support of European infrastructures. However, the tritium extraction unit from PbLi (TEU) selected and designed for ITER is based on the gas-liquid contactor technology, a reliable technology but with less than 40% efficiency. Instead, the candidates TER technologies for the lead-lithium loops of DEMO BBs, the permeator against vacuum (PAV) and vacuum sieve tray (VST), will not qualified in ITER, because these systems will not be fully mature by the start of the reactor. PAV and VST can theoretically achieve efficiency above 80%. The present works aim to analyze the technologies candidates for ITER and DEMO reactors, describe and compare TEU and TER design for each concept of BB and the integration of TER in DEMO tokamak building taking into account two design requirements: self-sufficient sustainable of fusion nuclear reactor and safety requirements.
Breeding blankets; DEMO; dual-coolant lithium-lead (DCLL); helium-cooled lithium-lead (HCLL); ITER; tritium extraction (TRIEX); water-cooled lithium-lead (WCLL)
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Utilizza questo identificativo per citare o creare un link a questo documento: http://hdl.handle.net/20.500.12079/52792
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