R.P. Doerner

Chapter 4: Power and particle control

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137–2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma–materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral–neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma–materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

Key ITER plasma edge and plasma–material interaction issues

Abstract Some of the remaining crucial plasma edge physics and plasma–material interaction issues of the ITER tokamak are discussed in this paper, using either modelling or projections of experimental results from existing tokamak operation or relevant laboratory simulations. The paper covers the following subject areas at issue in the design of the ITER device: (1) plasma thermal loads during Type I ELMs and disruptions, ensuing erosion effects and prospects for mitigating measures, (2) control of co-deposited tritium inventory when carbon is used even on small areas in the divertor near the strike points, (3) efficiency of edge and core fuelling for expected pedestal densities in ITER, and (4) erosion and impurity transport with a full tungsten divertor. Directions and priorities of future research are proposed to narrow remaining uncertainties in the above areas.

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.

Sputtering yield measurements during low energy xenon plasma bombardment

The sputtering yields of molybdenum, titanium, beryllium, and carbon have been measured during xenon ion bombardment from a plasma in the energy range between 10 and 200 eV. The erosion rates of Mo, Be, and C are measured both spectroscopically in the plasma and using the standard weight loss technique. Spectroscopic measurements of Ti sputtering yields, where no atomic physics data is available, are normalized to the weight loss measurements. The erosion rates of the metals decrease with the reduced mass of the metal–xenon combination and decrease with the increasing metal’s binding energy, as expected. The erosion results for bombardment of graphite indicate that the sputtering rate of carbon, as atoms, from the surface is insufficient to explain the total carbon weight loss measured. The multiple mechanisms for carbon erosion during plasma bombardment are discussed and the sputter rates of carbon atoms and carbon dimers are presented.

Observations of suppressed retention and blistering for tungsten exposed to deuterium–helium mixture plasmas

Blister formation and D retention in W have been investigated for low energy (~55 ? 15?eV), high flux (~1022?m?2?s?1), high fluence (?4.5 ? 1026?m?2) ion bombardment at moderate temperature (~573?K) in mixed species D+He plasmas in the linear divertor plasma simulator PISCES-A. The amount of D retained in W is found to decrease significantly when compared with that in W exposed to pure D plasmas, as measured with high resolution thermal desorption spectroscopy. Scanning electron microscopy observations reveal the suppression of the blisters, a surface feature known to drive up retention, in the D + He mixture plasma exposed W samples. Reduced D retention is accompanied by the formation of nano-sized high density He bubbles in the near surface, observed with a transmission electron microscope (TEM). It is believed that the nano-bubbles act as a diffusion barrier to implanted D atoms and consequently reduce the amount of uptake in the W material. This newly observed effect implies that current predictions of D retention in W, in actual fusion devices, may be overestimated, since there will be He ash in fusion plasma. Toughness enhanced, fine-grained (grain size of ~1??m) W?TiC samples, exposed to pure D plasma conditions, also show little or no evidence of blistering. The measured D retention in the W?TiC samples was approximately 1 ? 1019?D?m?2 corresponding to about 2 ? 10?7 of the implanted D fluence, and is very low compared with the retention in pure stress relieved W, which exhibited surface blisters and had a D retention of about 1 ? 1021?D?m?2.

The influence of displacement damage on deuterium retention in tungsten exposed to plasma

Trapping of tritium at lattice damage from fusion neutron irradiation is expected to increase the tritium inventory in tungsten components in ITER. The magnitude of this increase depends on the concentration of traps that are produced, and on the depth to which the increased tritium retention extends into the material. Experiments to address these issues are described, in which displacement damage by ion irradiation was used as a surrogate for neutron damage. Irradiated samples were exposed to high flux deuterium plasma to simulate divertor conditions. The resulting deuterium content was measured by nuclear reaction analysis. Measurements were done at various damage levels up to those expected from the end-of-life neutron fluence in ITER. These experiments determine the number of traps produced by displacement damage and the rate at which they are filled during exposure to plasma. The role of defect annealing was explored through plasma exposures at various temperatures. In addition to trapping at damage, near-surface retention from internal precipitation was observed at lower temperatures. Addition of 5% helium to the deuterium plasma greatly reduced D retention by precipitation by localizing it closer to the surface. Results from these experiments indicate that the contribution to tritium inventory in ITER from trapping at neutron damage should be small.

Plasma?surface interaction, scrape-off layer and divertor physics: implications for ITER

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITERs use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Deuterium retention in liquid lithium

Measurements of deuterium retention in samples of lithium exposed in the liquid state to deuterium plasma are reported. Retention was measured as a function of plasma ion dose in the range 6?1019?4?1022 D atoms and exposure temperature between 523 and 673?K using thermal desorption spectrometry. The results are consistent with the full uptake of all deuterium ions incident on the liquid metal surface and are found to be independent of the temperature of the liquid lithium over the range explored. Full uptake, consistent with very low recycling, continues until the sample is volumetrically converted to lithium deuteride. This occurs for exposure temperatures where the gas pressure during exposure was both below and slightly above the corresponding decomposition pressure for LiD in Li.

Tungsten nano-tendril growth in the Alcator C-Mod divertor

Growth of tungsten nano-tendrils (?fuzz?) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The thickness of the individual nano-tendrils (50?100?nm) and the depth of the layer (600???150?nm) are consistent with observations from experiments on linear plasma devices. The observation of tungsten fuzz in a tokamak may have important implications for material erosion, dust formation, divertor lifetime and tokamak operations in next-step devices.

Behavior of tungsten exposed to high fluences of low energy hydrogen isotopes

Abstract Tungsten is a candidate plasma facing material under investigation in a Sandia National Laboratories project conducted at Los Alamos National Laboratory. Samples of 99.95% tungsten provided by Plansee Aktiengesellschaft were exposed to 100 eV deuterium and tritium ions at a range of fluxes from 2.3×1017 to 1.3×10 18 ions / cm 2 s for one hour at 623 K in the tritium plasma experiment. The samples were outgassed to determine the amount of retained hydrogen isotopes. The retention scaled at slightly greater than the square root of the fluence. The fractional retention was on the order of 10−5. The data from these experiments were combined with previous results to construct a comprehensive model of the migration and retention behavior for hydrogen in tungsten. A second set of experiments involved exposing 99.95% tungsten foils provided by AESAR to 100 eV deuterons at a flux of 6×10 17 D / cm 2 s for 30 min at 423 and 373 K. Scanning Electron Microscopy analysis was performed on the samples to determine the effects of the plasma exposure. Unannealed samples revealed extensive blistering with many blister caps removed. Samples annealed to 1473 K showed minor blistering, and samples annealed to 1273 K showed no blistering. The SEM analysis was used in conjunction with the retention results to understand the role of annealing and defects in trapping within the tungsten.