O. V. Ogorodnikova

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

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 around the strike points on the divertor plates, results both from the attempt to reduce the tritium inventory and to optimize the lifetime of the plasma-facing components.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.

Deuterium retention in tungsten exposed to low-energy pure and helium-seeded deuterium plasmas

Influence of helium (He) on the deuterium (D) retention in tungsten (W) under simultaneous He-D plasma exposure was investigated. Bulk polycrystalline tungsten and two W coatings on carbon substrate, namely, plasma-sprayed tungsten and combined magnetron-sputtered and ion implanted tungsten (CMSII-W) were exposed to pure and He-seeded D plasmas generated by electron-cyclotron-resonance plasma source. The D retention in each sample was subsequently analyzed by various methods such as nuclear reaction analysis for the D depth profiling up to 6 μm and thermal desorption spectroscopy for the determination of total amount of D retention. It is shown that seeding of helium into D plasma with helium ion flux fraction of 10% reduces the deuterium retention for all tungsten grades but more significant reduction was observed for polycrystalline W and less significant effect was found for W coatings. From the thermal desorption spectroscopy measurements, we conclude that the presence of He modifies the density of ex...

Deuterium retention in different tungsten grades

Deuterium retention in different tungsten (W) grades was investigated for various incident ion energies ranging from 3 to 200?eV per deuterium atom and fluences ranging from 1?1023?m?2 to 2?1025?m?2. Irradiation temperatures were 320 and 500?K. The retained amount was determined by ion beam analysis and thermal desorption spectroscopy. For irradiation at 320?K, deuterium retention in plasma-sprayed tungsten (PSW) is higher compared to polycrystalline W for all investigated energies. For irradiation at 500?K, the deuterium retention for polycrystalline W increases strongly with fluence in contrast to PSW. The highest deuterium inventory was found for ITER grade W irradiated with 200?eV per deuterium at 500?K.

Surface effects on plasma-driven tritium permeation through metals

The model for the plasma-driven tritium permeation proposed by Doyle and Brice [J. Vac. Sci. Technol. A 5 (4) (1987) 2311] and by Waelbroeck et al. [Forschungazentrum, Julich, Germany, Jul-1996, 1984] has been developed. An improved analytical expression for the plasma-driven tritium permeation in the steady state is presented. Using the presented analytical expression, the influence of surface conditions on the front and the back sides of a metal has been considered for the plasma-driven permeation of tritium through both endothermic and exothermic metals. The presented pictures allow us to define the steady-state tritium permeation in a wide range of the surface conditions: as for an absolutely clean surface and a bare (presence of small amounts of impurities) surface as for contaminated with different impurities surface. This result can be useful for choosing the plasma-facing material for the future steady-state fusion device.

Effect of nitrogen seeding into deuterium plasma on deuterium retention in tungsten

An interaction of deuterium plasma with seeding species with a material is, in particular, an issue for medium- and high-Z plasma-facing materials for fusion devices to reduce the power load in front of material surfaces. In this paper, we investigated the influence of seeding of nitrogen (N) into deuterium plasma on the accumulation of deuterium (D) in tungsten (W). Tungsten samples were exposed to pure and N-seeded deuterium plasmas generated by an electron-cyclotron resonance plasma source in the PLAQ (Plasma Quelle) experiment. D and N retention in each sample was subsequently analysed by nuclear reaction analysis for depth profiling up to 6 μm. It was found that the amount of N in W is (7–9)×1019 N m−2 in the temperature range 300–650 K and slightly decreases down to 6×1019 N m−2 at 800 K. This means that the nitrogen-containing layer formed upon exposure of W to N-seeded D plasma is thermally stable and does not decompose at least up to 800 K. It is shown that the seeding of nitrogen into D plasma does not prevent blister formation and even results in an increase of the size of blisters in some cases. Depth profile measurements show that there is an enhancement of D diffusion into the bulk and, consequently, an increase of D retention in W in the presence of N seeding in D plasma compared to pure D plasma. The influence of N seeding on D retention depends strongly on the applied bias and fluence in our plasma conditions. The mechanisms of deuterium retention in W in the presence of nitrogen seeding are discussed.

In situ nuclear reaction analysis of D retention in undamaged and self-damaged tungsten under atomic D exposure

In situ deuterium depth profile measurements by nuclear reaction analysis (NRA) were performed on undamaged and neutron-like damaged W during D atom exposure at several sample temperatures. Damaged W was produced by high energy W-ion irradiation. In situ results of D retention after exposure termination are in good agreement with ex situ measurements for damaged W. In contrast, in situ NRA during D exposure on undamaged W showed significantly higher D retention than ex situ NRA measurement. It is also shown that this increased retention is lower at the location non-irradiated by the probing 3He beam. These observations indicate that the D retention is influenced by secondary damage produced by the probing 3He beam, however the dynamic retention could also partially contribute to the observed retention increase.

Tritium retention in reduced-activation ferritic/martensitic steels

Abstract The effects of displacement damage, plasma exposure and heat loads on T retention in reduced-activation ferritic/martensitic (RAFM) steels were investigated by exposing the steels to DT gas at 473 K. Despite enormous change in surface morphology, T retention in the heat-loaded specimen was comparable with that in the unloaded specimen. The exposure to plasma resulted in a drastic increase in T retention at the surface and/or sub surface. However, the T trapped at the surface/subsurface was easily removed by maintaining the specimens in air at ~300 K. Formation of radiation-induced defects led to a significant increase in T retention, and T trapped in the defects was not removed at ~300 K. These observations suggest that displacement damages have the largest effects on T retention at ~473 K.