In-vessel tritium retention and removal in ITER

Abstract

Invited Review Paper -R2 CEI SUN 3 0 O S T ITritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER.We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation.We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER, the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in-situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

Effect of ion flux on helium retention in helium-irradiated tungsten

“…This is due to a number of features that it presents: low sputtering yield, low-activation, high melting point, high thermal conductivity and low thermal expansion [1][2][3]. In the case of inertial confinement fusion by laser (laser fusion) with direct drive targets (as it is the case of the European project HiPER) W is proposed as an armor material [4,5] with the function of protecting the underlying structural materials against the intense ion fluxes stemming from the target explosions.…”

High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

“…Additionally, hydrogen neutrals (CX) from charge exchange reactions will be present with energies far above the sputtering threshold [26]. The fluxes and energies of the particles expected in ITER [8,27] together with the consideration of off-normal events (see below), led to the present choice of PFMs in ITER, which foresees the use of Be in the main chamber, W at the divertor entrance and CFC at the divertor strike zones [1]. In a reactor, the most promising material solutions for the first wall armour seem to be tungsten as a coating on low activation steel, or low activation steel alone [28], since the erosion of Be, as planed in ITER, will be also too large.…”

Removal of the deposition on JT-60 tile by nano-sec pulsed-laser irradiation

“…Particularly, codeposition in plasma-shadowed area is one of the critical concerns because it's difficult to remove it by general plasma-discharge cleaning methods, meaning this can cause continuous tritium accumulation in the vacuum vessel. From the safety reason, periodic tritium removal will be required before in-vessel tritium inventory reaches the administrative goal of ITER (700 g) [2].…”