Tritium inventory in ITER plasma-facing materials and tritium removal procedures

Abstract

Abstract:Interactions between the plasma and the vessel walls constitute a major engineering problem for next step fusion devices, such as ITER, determining the choice of the plasma-facing materials. A prominent issue in this choice is the tritium inventory build-up in the vessel, which must be limited for safety reasons. The initial material selection, i.e. beryllium (Be) on the main vessel walls, tungsten (W) on the divertor upper baffle and dome, and carbon fibre composite (CFC) around the strike points on the divertor plates, results both from the attempt to reduce the tritium inventory and to optimise the lifetime of the plasma facing components (PFCs).In the framework of the EU Task Force on Plasma-Wall Interaction (PWI TF), the many physics aspects governing the tritium inventory are brought together. Together with supporting information from international experts represented by the ITPA SOL/DIV section, this paper describes the present status of knowledge of the in-vessel tritium inventory build-up. Firstly, the main results from present fusion devices in this field are briefly reviewed. Then, the processes involved are discussed: implantation, trapping and diffusion in plasma facing materials are considered as well as surface erosion and co-deposition of tritium with eroded material. The intermixing of the different materials and its influence on hydrogen retention and co-deposition is a major source of uncertainty on present estimates and is also addressed.Based on the previous considerations, estimates for the tritium inventory build up are given for the initial choice of ITER materials, as well as for alternative options. Present estimates indicate a build-up of the tritium inventory to the administrative limit within a few hundred nominal full power D:T discharges, co-deposition with carbon being the dominant process. Therefore, tritium removal methods are also an active area of research within the EU PWI TF, and are discussed. An integrated operational scheme to slow the rate of tritium accumulation is presented, which includes plasma operation as well as conditioning procedures. † Current address: Fusion for Energy,

Fuel retention in carbon materials under ITER-relevant mixed species plasma conditions

“…However, to describe the retention in a CFC material correctly, the dependences on the incident fluence, flux, ion energy and surface temperature have to be taken into account. As an example, the retention in CFC estimated in [20] has been based on the measurements with the low flux ion beams. Not taking into account the dependence of the flux has resulted in a significant overestimation of the in-bulk CFC retention.…”

Erosion, formation of deposited layers and fuel retention for beryllium under the influence of plasma impurities

“…If ITER will forgo the deployment of carbon-based plasma-facing materials, the co-deposition with beryllium is believed to be the dominant fuel retention mechanism [16]. Varying the Be deposition flux, the incident ion energy and the substrate temperature, a scaling for the D/Be ratio in co-deposited layers was obtained in PISCES-B [17,18].…”

Fuel re-absorption by thermally treated co-deposited carbon layers

“…It was determined by NRA using a 2 MeV 3 He + beam and by detecting high-energy protons emerging from the reaction D( 3 He, p) 4 He. The information depth with the 2 MeV 3 He + beam in graphite is 6 µm (ρ = 1.9 g cm −3 ) and 7 µm in the deposit (ρ was assumed to be 1.6 g cm −3 ).…”

“…This issue becomes even more important if carbon plasma-facing components (PFCs) are used [2][3][4]. Two basic schemes for fuel removal are currently considered: (i) desorption of hydrogen-containing species and (ii) removal of the entire fuel-rich co-deposit.…”