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The International Workshop on Carbon Materials, held in Stockholm on the 21st and 22nd of September 1995, was the seventh meeting in the series designed to highlight recent progress in studies and development of plasma facing materials for controlled fusion devices. The meeting was organized in a close cooperation between the groups from the Royal Institute of Technology (KTH) - Stockholm and Research Centre (KFA) in Juelich - Germany. It is worthwhile to mention that the workshop was for the first time organized outside Germany; all previous meetings of the series were held at the KFA - Juelich.

The workshop was sponsored by the Royal Swedish Academy of Sciences through its Nobel Institute for Physics, the Swedish Natural Science Research Council, the Royal Institute of Technology and the City of Stockholm.

The symposium gathered 48 registered participants from Germany, Japan, Sweden, France, the United States, Italy, the United Kingdom, Russia, Poland and India. The scientists were coming from the leading plasma physics and fusion-related materials research laboratories.

Presentations included 16 invited talks and several oral presentations given both by the researchers and representatives of industrial companies developing plasma facing materials and components. Moreover, the Toyo Tanso Ltd. (Japan) organized an exhibition of the Company products. Ample time was allowed for discussions and "hot" topic presentations. The workshop was preceded (September 20) by the EU Meeting on Tritium Removal in Controlled Fusion Devices.

The programme consisted of three major topics concerning: neutron irradiation effects in materials, hydrogen interaction-tritium inventory phenomena and comparison of carbon materials and beryllium for fusion applictions. The speakers presented both an overview and details of the field. For the first time, the results of the influence of radiation damage on the tritium inventory in the materials were presented by scientists from Europe, Japan and the United States. A few other talks dealt with a detailed description of the change in microstructure and thermo-mechanical properties in the neutron irradiated graphites, carbon-based composites and beryllium. Several presentations were concerned with erosion of the first wall materials in fusion devices. The issues of chemical erosion and formation of thick co-deposits (layers containing hydrogen isotopes together with plasma impurity species) and their impact on the operation of future devices were also deeply addressed and discussed. The present status of the design of the ITER (International Thermonuclear Experimental Reactor) first wall and blanket was presented by the members of the ITER Team. Studies and development of new carbon- based materials and plasma sprayed beryllium layers with improved thermo-mechanical properties were summarized by the members of several research laboratories and industrial companies.

On behalf of the Organizing and Programme Committees I would like to express my deep gratitude to all the sponsors and participants for making the symposium very fruitful and interesting.

Based on our previous works, we propose a new model for damage structure of graphite based on sp² to sp³ transition of carbon atoms that can explain the change of the materials' parameters observed both in damaging and annealing processes. In early stage of the irradiation, defects are produced in a basal plane and/or in-between the basal planes (in-plane or two dimensional defects) accompanying sp² to sp³ transition of some carbon atoms and losing their graphitic bonding. They play a critically important role on the reduction of thermal conductivity, increase of electrical resistivity, increase of hydrogen retention and so on. They also cause a dimensional change through the increase of lattice spacing between the basal planes. Such in-plane defects are rather easily annealed out and the materials' parameters for lightly irradiated graphite recover their initial values by annealing far below the graphitization temperature. However, there is no way to avoid their production under neutron irradiation, particularly at low temperatures. After high dose irradiation, the defects grow into three dimensional structures probably including the formation of sp³ clusters. Accordingly, the basal planes lose their ordering resulting in turbulence and bending, and remaining open spaces. They play an important role on the volume expansion, which cannot be caused totally by the lattice expansion. Once graphite loses its 2 dimensional structure, the recovery to the initial structure is very difficult.

Carbon–carbon composites are an attractive choice for fusion reactor plasma facing components because of their low atomic number, superior thermal shock resistance, and low neutron activation. Next generation tokamak reactors such as the International Thermonuclear Experimental Reactor (ITER), will require high thermal conductivity carbon–carbon composites and other materials, such as beryllium, to protect their plasma facing components from the anticipated high heat fluxes. Moreover, ignition machines such as ITER will produce a large neutron flux. Consequently, the influence of neutron damage on the structure and properties of carbon–carbon composite materials must be evaluated. Data from two irradiation experiments are reported and discussed here. Carbon–carbon composite materials were irradiated in target capsules in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). A peak damage dose of 4.7 displacements per atom (dpa) at an irradiation temperature of ~600°C was attained. The carbon materials irradiated here included uni-directional, two-directional, and three-directional carbon–carbon composites. Irradiation induced dimensional changes are reported for the materials and related to single crystal dimensional changes through fiber and composite structural models. Moreover, carbon–carbon composite material dimensional changes are discussed in terms of their architecture, fiber type, and graphitization temperature. Neutron irradiation induced reductions in the thermal conductivity of two, three-directional carbon–carbon composites are reported, and the recovery of thermal conductivity due to thermal annealing is demonstrated. Irradiation induced strength changes are reported for several carbon–carbon composite materials and are explained in terms of in-crystal and composite structural effects.

In the continuation of our study on neutron induced properties changes of carbon fiber composites at low irradiation temperature and low neutron doses, an experimental study on the changes of dimensional stability and thermal conductivity has been performed for irradiation temperature 400-1000°C and neutron doses 0.8-1.8 dpa. The investigated materials include: A 05 (2D felt-type CFC), CX 2002U (2D felt-type CFC), DMS 678 (2D woven CFC), N112 (3D CFC), N11 (3D CFC), FMI A27-130 (3D CFC) and MKC (1D CFC). The results show that dimensional changes in both parallel and perpendicular directions at irradiation temperature 400-600°C are negligible, whilst the thermal conductivity changes (K_i/K₀) are between 0.3-0.87, under those irradiation conditions. The results are critically analysed and discussed.

Tritium retention measurements have been performed for two carbon composites both before and after neutron irradiation. These two composites were FMI-222 manufactured by Fiber Materials Incorporated and MKC-1PH by Mitsubishi Kasei. The neutron irradiations were performed in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory at 473 K and spanned radiation levels from unirradiated up to 1.0 dpa. The high energy traps in the composites were saturated in tritium by heating the composites to 1673 K with a pure tritium gas pressure of 13.2 Pa. The samples were then outgassed in a separate system at a temperature of 1963 K where tritium analysis was performed using an ionization chamber and liquid scintillation counting. For all irradiation conditions, the retained tritium was less than that measured in earlier studies for different types of graphites. The results suggest carbon composites should be preferred over graphites for use in fusion reactors where both tritium and neutron irradiation exist.

In the present work, based on a thermodynamic analysis of hydrogen-beryllium interaction, a new approach for hydrogen behaviour in beryllium is presented. This approach is based on the assumption that "true" solubility of hydrogen in beryllium is extremely low and endothermic trapping in impurities and gas filled bubbles is responsible for hydrogen inventory in beryllium at temperatures higher than 700 K. Hydrogen implanted into beryllium at elevated temperatures is assumed to be retained as H-atoms trapped at impurities and as molecular H₂ inside pores and bubbles.

The Langmuir type adsorption and dangling sp²-bonds relaxation are proposed for the hydrogen-graphite interaction. Three kinds of traps are proposed: carbon interstitial loops with an adsorption enthalpy of −4.4eV/H₂ (Trap 1); graphite network edge atoms with an adsorption enthalpy of −2.3eV/H₂ (Trap 2); basal planes adsorption sites with enthalpy of +2.43eV/H₂ (Trap 3). The sorption capacity of every kind of graphite could be described with its own unique set of traps. The irradiation with neutrons or energetic carbon ions increases the number of traps. At damage level of ~1 dpa under room temperature irradiation the concentration of traps 1 and 2 reaches 1500 and 500 appm respectively.

Estimations of hydrogen inventory for ITER conditions for both beryllium and graphite are provided.

The effects of neutron-induced damage on the tritium retention have been studied experimentally for graphite and other carbon-based materials in the damage range ⩽4 dpa and for beryllium in the range ⩽40 dpa.

The results for graphite and other carbon-based materials are in qualitative agreement with previous studies what concerns the increase of tritium retention with neutron damage in the range below about 0.1 dpa. On the other hand the kinetics of tritium transport and adsorption seems to be faster and the estimated number of tritium traps of irradiated material in the asymptotic region above about 0.1 dpa is remarkably lower (~ 1000 appm) than in a previous study (~6000 appm).

The studies of tritium retention in neutron-irradiated beryllium are rendered difficult because of the huge amount of neutron-generated tritium, which has to be released before. This requires such high temperatures that probably any reversible neutron-induced traps are annealed. A gradual increase of tritium retention with increasing fast neutron fluence of about a factor ten is observed in the range <40 dpa, which is assumed to be due to irreversible changes in the microstructure of the samples.

In order to achieve a long burning time period in a fusion experimental reactor and a steady state operation in a Large Helical Device, LHD, the heat flux and the particle flow have to be adequately controlled, in addition to the suppressions of the plasma surface interactions such as erosions and recyclings. Recently, in Japan, high heat flux components which can remove the fluxes of 30 MW/m² and 10 MW/m², have been developed for ITER and LHD, respectively. In addition, the divertor concepts such as a Local Island Divertor and ripple limiters were progressed both for LHD and ITER. The data on erosion and hydrogen recycling were accumulated for low Z materials such as lithium, boron and boron carbide. It is shown that the conditioning for the boron-containing material is significantly easier compared with the case of graphite or CFC. Others activities associated with the plasma facing materials such as tritium and neutron damage are also briefly introduced.

The First Wall (FW) is one of the key components of the International Thermonuclear Experimental Reactor (ITER). It must be designed to remove the surface heat flux from the plasma, while, in combination with a structurally integral shield, reduce the nuclear effects in the vacuum vessel and protect the superconducting coils from excessive nuclear heating and radiation damage. As the principal plasma facing element it will be exposed to the 1500 MW fusion power at an average neutron wall loading of about 1 MW/m² and must be designed for a life of 0.3 MWa/m² during the Basic Performance Phase. In addition the First Wall must successfully withstand peaked thermal and electromagnetically loads resulting from plasma disruptions.

A conceptual design of the ITER First Wall has evolved as a result of the Engineering Design Activity. This design utilizes a beryllium plasma facing layer, bonded onto a copper layer which conducts the heat to the contained active cooling tubes. The copper is bonded to stainless steel for structural support and attachment to the water cooled stainless steel shield section. As discussed in this paper, the basic FW configuration developed for use at most locations, referred to as the primary wall, must be modified for different power requirements at two specific locations, the plasma limiter and the baffle. Carbon Fiber Composite (CFC) is being considered as an alternative material for the baffle and limiter FW because of its high thermomechanical performance and its high resistance to thermal shock.

An extensive test program has been set up to evaluate the performance of beryllium, carbon materials and tungsten as plasma facing materials for the divetor in thermonuclear fusion devices. Experiments described in this paper represent pre-irradiation testing being part of a comprehensive study on neutron damage of several plasma facing materials and components.

Thermal shock tests were carried out with several grades of beryllium, carbons and tungsten, at heat loads up to 5.3 MJ/m². In these experiments CFC materials and tungsten show less erosion than the beryllium grades. Best erosion behaviour of the beryllium grades is observed for S 65-C and for TGP-56.

Steady state heating tests with actively cooled Be/Cu samples and CFC/ metal samples, respectively were performed at loadings up to 5.8 MW/m². The main objective of these tests was the pre-selection of components to be included in a neutron irradiation program and to determine the heat removal efficiency before neutron irradiation. All samples investigated in steady state heating tests (Be/Cu and CFC modules) did not show any indications of failure, and are thus qualified to be used in the forthcoming neutron irradiation program. CFC modules however, show a strong variation of surface temperature due to the deviations in thermal conductivity of the CFC materials.

During irradiation of carbon materials with hydrogen ions hydrocarbon molecules are formed resulting in an enhancement of the erosion yield. At temperatures around 800 K hydrocarbon molecules are released in a thermal activated process, while at low temperatures and low ion energies physical sputtering of lightly bound hydrocarbon radicals enhances the erosion yield. Doping of carbon materials with B, Si and Ti results in a reduction of its chemical reactivity with hydrogen ions. While B reduces drastically the thermal activated process it does not alter the sputtering of hydrocarbons at low energies. For isotropic graphites doped with 10at% Si (LS10) and 10at% Ti(LT10) it is shown that preferential erosion of carbon leads to enrichment of the dopant at the surface. The thermal activated hydrocarbon emission is reduced already at low ion fluences for LS10 and LT10, while the low energy process is only reduced after high fluence irradiation and carbon surface depletion in the case of Ti doping. Depending on the microstructure of the material a very pronounced surface topography delevopes. Carbidic grains protect the underlying carbon material from erosion until a columnar structure evolves. Due to the high threshold for physical sputtering of Ti the total erosion yield for LT10 shows the predicted threshold behaviour for physical sputtering.

Mass spectroscopy and optical spectroscopy have been used to measure the formation of methane, higher hydrocarbons and of CO during the interaction of limiters with the boundary plasma and of special carbon targets with the scrape-off-layer plasma (SOL) of TEXTOR. Mass spectroscopic data are obtained by the Sniffer probe in the SOL under carbon, boronized and siliconized wall conditions. At target temperatures ⩽100 °C, methane yields range typically between 0.7 and 1.2%. They vary only little with changing plasma conditions. C₂-hydrocarbon formation dominates the overall carbon erosion under many conditions. Their yields increase with decreasing plasma temperature. Siliconization of the walls reduces the methane formation only little but suppresses the formation of higher hydrocarbons significantly. CO formation is dominated by the actual oxygen impurity fluxes and ranges between 0.2% up to 1.5% depending on the wall conditioning. Supporting data on hydrocarbon and CO formation are obtained from the outgassing after the discharge.

Optical spectroscopy has been used to determine methane formation yields from CH band emission in front of graphite test limiters positioned at the last closed flux surface. The yields are typically in the range between 1.5 and 5% and are generally a factor 2-3 higher compared to those from mass spectroscopy. The CH₄ formation is nearly constant between 200 °C up to 700 °C and decreases beyond 800-1000 °C. It decreases with increasing flux density. C₂ hydrocarbon emission from the limiters has not been observed by molecular band emission within the range of normal plasma conditions. They show up only for detached plasma conditions.

Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasmaspray process for depositing beryllium coatings on damaged beryllium surfaces. Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary H₂ gas additions to improve the melting of the beryllium powder and negative transferred-arc cleaning to improve the bonding between the beryllium coatings and the underlying surface. Information will also be presented on thermal cycle tests which were done on beryllium coated ISX-B beryllium limiter tiles using 10s cycle times with 60s cooldowns using a heat flux slightly in excess of 5 MW/m².

Although they have rather drastically different characteristics, currently beryllium and tungsten are both considered as candidate materials for plasma-facing components in the International Thermonuclear Experimental Reactor (ITER). The erosion behavior of these materials under steadystate deuterium plasma bombardment has recently been investigated using the PISCES-B Mod facility. Several effects of plasma impurities on the materials erosion behavior have been observed in these experiments. At elevated temperatures beryllium tends to be contaminated with carboncontaining plasma impurities, resulting in the deposition of carbonaceous films. These films shield beryllium from plasma bombardment and reduce erosion. In contrast, no sign of impurity deposition has been found with tungsten. However, the overall erosion behavior of tungsten is dictated by trace amounts of oxygen-containing impurities at energies below the sputtering threshold by deuterium. A universal materials balance model is presented to explain these impurity effects.