A.A. Haasz

Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in todays tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

In-vessel tritium retention and removal in ITER

Abstract Tritium 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 – and more generically in any other next-step experimental fusion facility fuelled with tritium – 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.

Isotopic effects in hydrocarbon formation due to low-energy H+/D+ impact on graphite

Recent developments with gaseous divertors in tokamaks have led to prospects of less energetic ion bombardment (10s of eV) of material surfaces in the divertor region. The present experiments were undertaken with the objective of studying hydrocarbon formation due to H+/D+ impact on graphite for energies extending down to the 10 eV range. Mass spectrometry in the residual gas was used to measure the hydrocarbon formation rates as a function of pyrolytic graphite temperature (300–1000 K) and ion energy (10–200 eV/H or D) for mass-analysed H3+ and D2+ beams (1018 H+(D+)/m2 s) providing a unique opportunity to also investigate isotopic effects. The results indicate that, as the ion impact energy is reduced, a reduction in the maximum chemical yield (Ym) is observed and the broadening of the temperature dependence profile for hydrocarbon formation leads to significant erosion for low-energy impact at room temperature. The room-temperature methane and total chemical yields display maxima at about 50 eV and decrease as the ion energy is further reduced. Where kinetic effects are expected to affect the erosion process, viz., low energy ( 50 eV)/near Tm, an isotopic yield ratio, YD/YH of about 1.5–2 was measured. Under all other conditions studied, this ratio was near unity. Furthermore, it is evident that the dominating erosion mechanisms at low energies (<100 eV) differ from those occurring for higher energy impact.

Deuterium retention in tungsten for fusion use

The retention of deuterium in polycrystalline W foils has been measured as a function of ion fluence and implantation temperature. At temperatures in the range 350-550 K, retention levels were found to be above the 300 K value. This retention enhancement is attributed to an increase in the D diffusion coefficient, which allows a greater diffusion depth. Furthermore, while retention at room temperature saturates as a function of fluence, at 500 K no saturation is observed. D( 3 He, 4 He)p nuclear reaction analysis measurements show that at room temperature most of the D is trapped in the near-surface region of the specimen, but significantly beyond the implantation range. At higher temperatures, much lower levels are observed in the near-surface, and diffusion through to the back of thin specimens is observed: the front and back surface D concentrations are similar. While the D retention in a proposed W alloy for ITER applications (W-1%La 2 O 3 ) is similar to that measured in the pure W foils over most of the range of the experiments, two important differences are noted: a trend to saturation of the amount retained is observed at fluences >10 24 D/m 2 and 500 K, and the D concentration at the back of the W-1%La 2 O 3 alloy is about 1% of that at the implanted surface.

Recent advances on hydrogen retention in ITER's plasma-facing materials: Beryllium, carbon and tungsten

Abstract Management of tritium inventory remains one of the grand challenges in the development of fusion energy, and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER’s plasma-facing materials – Be, C, and W – and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this paper together with recommendations for ITER. Basic parameters of diffusivity, solubility, and trapping in Be, C, and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping, but long-term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be- and C-containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak, and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.

Two‐region model for hydrogen trapping in and release from graphite

A new model has been developed for hydrogen retention and trapping in and release from graphite. Two different regions in the graphite with different hydrogen transport and trapping behaviors are distinguished, the bulk region within, and the surface region on graphite crystallites. The model incorporates new experimental results related to atom diffusion and recombination on inner surfaces. Recombination is explained from a fundamental viewpoint by linking it to diffusion using a classical expression. The model is applied to a number of reemission and thermal desorption experiments, in particular, the reemission of hydrogen atoms during irradiation with energetic hydrogen ions and the formation of HD during irradiation with H+ and D+ or during thermal desorption of graphite that was preimplanted with H+ and D+ ions with different energies.

Hydrocarbon formation due to combined H + ion and H0 atom impact on pyrolytic graphite

Graphite, which is a favoured material for first wall use in fusion devices, is subject to chemical erosion under hydrogenic impact. Until recently, most published results on hydrogen-induced chemical erosion of graphite have concentrated only on the production of methane. For incident H+ energies ≳ 300 eV/H+, methane is in fact the dominant reaction product. For low energy (< 300 eV) H+ ion impact on pyrolytic graphite, however, the relative formation rates of heavier hydrocarbons (C2Hx, C3Hx) become more important. For thermal H0 atom (sub-eV) impact, the formation of heavier hydrocarbons dominates the total carbon erosion process. For combined H+ ion and H0 atom impact, as would occur in a fusion reactor with a ratio [H+]/[H0] ~ 0.1 to 1, the total C-erosion yield is 2–5 times larger than the methane yield, depending on the ion energy and the relative fluxes. The absolute level of the maximum erosion yield under all conditions observed remains ≲ 0.1 C/H.

Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions show a difference of several orders of magnitude in both inventory and permeation rate to the coolant.

Model for the chemical erosion of graphite due to low-energy H+ and D+ impact

A methane erosion yield model has been developed using the principal atomistic reactions outlined by Kuppers and co-workers [eg., A. Horn, A. Schenk, J. Biener, B. Winter, C. Lutterloh, M. Wittmann, and J. Kuppers, Chem. Phys. Lett. 231, 193 (1994)] with additional terms to account for the energy of the incident particles, namely, kinetic ejection and damage deposition. Furthermore, modifications were made to the previous models by using distributed activation energies for methyl and hydrogen release as well as an activated Eley-Rideal abstraction process. Fitting of this model to experimentally measured methane yield data shows excellent agreement, except for low energy (⩽25 eV) impact at temperatures above ∼800 K. We have provided a sound physical basis for the behavior of the free fitting parameters and conclude that most of the processes associated with low-energy impact on pyrolytic graphite leading to methane production have been incorporated. Possible extensions of the model to include heavy hydroc...

Impurity release from low-Z materials under light particle bombardment

Abstract Recent progress in the understanding of the erosion of low-Z materials under bombardment conditions characteristic of magnetic fusion experiments is reviewed. The role of physical sputtering, chemical sputtering and radiation-enhanced sublimation in tokamaks is considered, and observations are related to laboratory measurements. The role of physical sputtering is largely understood, and tokamak measurements, under conditions where physical sputtering is expected to dominate, can be well predicted, except in the energy range near the sputtering threshold. Chemical erosion and radiation-enhanced sublimation are less-well understood, and predictions of erosion yields under tokamak conditions require assumptions (primarily related to energy and flux density dependence) which do not have a solid experimental basis. Also, only a few quantitative results from tokamaks are available to confirm predictions, and those which are available are not always consistent.