Table of contents

Hydrogen, the first, the lightest and the smallest element of the Periodic Table is important in every corner of our Universe and our terrestrial micro-world. Isotopes and compounds of hydrogen have paramount importance and countless applications in our daily life. They have also a great potential to be globally used for hydrogen energy systems: fuel cells, metal hydrides and thermonuclear fusion. In all these cases, the progress highly depends on deep knowledge and understanding of processes underlying the interaction of hydrogen with matter.

Hydrogen isotopes are essential as thermonuclear fuel and the consequences of their interaction with fusion reactor materials belong to the most important and challenging issues when the deuterium-tritium operation of a reactor-class device is considered. The major point is related to processes leading to a long term retention of hydrogen isotopes in plasma facing components, the so-called tritium inventory. Therefore, diffusion, chemical reactivity of the isotopes with materials and, in particular, problems connected with the radioactivity of tritium trigger world-wide studies aiming at the understanding of transport and re-deposition of eroded wall material. The ultimate goal is a proper assessment of the inventory and the development of methods for efficient removal of tritium retained in the reactor wall. These topics are addressed at all major conferences on controlled thermonuclear fusion research and at topical meetings focused on a review of the most recent ideas and progress made in this field. It is no wonder that there were two series of workshops dedicated to this appealing subject. An international Workshop on "Tritium Effects in Plasma Facing Components" was organised every second year as a satellite symposium to the International Conference on Plasma–Surface Interactions in Controlled Fusion Devices (PSI conferences). Beginning in 1992 it was held in Livermore, Nagoya, Ispra, Santa Fe and in Stockholm (in 2000 organised as an International Workshop on "Hydrogen Isotopes in Solids"). The history of the second series, International Workshop on "Hydrogen Isotope Recycling at Plasma Facing Materials in Fusion Reactors", also began in 1992 on the initiative of scientists from the University of Tokyo and from the Bonch-Bruevich University in St. Petersburg. Originally, the symposium was organised under the name "Interactions of Fuel Particles with Fusion Materials". The meetings were held annually either in Japan or in Russia until 1999 and then, in 2000 and 2001, they were organised at the Argonne National Laboratory, Illinois, USA. In 2002, in our world of ever expanding number of conferences, a rare event occurred: the committees of the two series decided to merge having a combined workshop which would both continue the tradition of the two series and would also create a forum for more deep and concentrated discussion. This time it was in Tokyo in the charming scenery of the Hongo Campus at the University of Tokyo. The Workshop organised prior to the 15th PSI (Gifu n/b Nagoya) gathered 70 participants from twelve countries.

Authors of invited talks, oral and poster presentations gave an overview and detailed description of processes related to the retention, migration and recycling of hydrogen isotopes on various materials being of interest for a fusion reactor: carbon, beryllium and high-Z metals. Experimental and theoretical approaches have been presented. These were results from major fusion experiments (JET, JT-60, TFTR, ASDEX, TEXTOR) and also obtained under laboratory conditions. Carbon is a major first wall component that has been selected to be used in the bottom of the ITER divertor. Chemical erosion of carbon by hydrogen followed by the re-deposition of hydrocarbons on wall structures is the driving force for a build-up of carbonaceous films (called "co-deposits") accumulating significant quantities of fuel. Over many years the community has examined the morphology and properties of co-deposits. Despite extensive trial of different methods (e.g., based on plasmo-chemical and isotope exchange processes) little progress has been achieved in the past towards the efficient in-situ decomposition of co-deposits accompanied by a significant tritium removal rate. We are glad that breakthrough results obtained with laser-assisted ablation have been presented for TFTR tiles during this Workshop. The successful proof of principle paves the way for testing the laser cleaning method on plasma facing components of JET. Other important issues regarding the recent studies of co-deposits are those related to: (a) in-situ determination of their growth rate by means of quartz microbalance devices and (b) mapping of co-deposits distribution, thickness and the content of retained fuel on the entire walls of various machines. New imaging techniques allow very precise mapping and measurements of tritium produced in devices fuelled with deuterium (D–D reaction). As a result of these studies we are able to make more and more accurate assessments of a global material transport in controlled fusion devices. Significant progress has also been reported in studies of hydrogen interaction with high-Z metals.

We want to express our deep gratitude to the sponsor of the Workshop: The Atomic Energy Society of Japan. We are also grateful to the staff members of the University of Tokyo and Nagoya University for their support in the organisation and all technical matters associated with the preparation of conference materials. The sponsorship and the great assistance in organisation helped both to keep the registration fee low and to make the meeting fruitful and effective. It is our pleasure to publish the proceedings of this symposium which brought so many new and interesting results. We thank all the participants for their contributions and we thankthe referees of submitted papers. Thank you all for your hard work and kind co-operation. We are looking forward to the next meeting planned for June 2003 in Helsinki.

Superabundant vacancies of metal atoms, of concentrations as high as 10 ~ 30 at %, can be formed in the presence of interstitial hydrogen as a consequence of reduction of the formation energy by trapping H atoms. The equilibrium concentration and mobility of Vac-H clusters were determined by in situ XRD and resistivity measurements, and their sources were identified. The binding energies of trapped H atoms were determined by thermal desorption spectroscopy. Some of these experimental results are described, with particular reference to Pd, Ni and Cr.

Depth profiles of deuterium and hydrogen retained in plasma facing graphite tiles placed in the divertor region of JT-60U, and chemical bonding states of carbon on the surface of the tiles, were respectively investigated by SIMS and XPS. The deuterium in the near surface of all graphite tiles was replaced by hydrogen due to exposure to hydrogen plasma in the clean-up stage after the D-D operation. The depth profiles of the H/¹²C and D/¹²C signal intensity ratios varied with the location of the tile. The integrated (H + D)/¹²C intensity ratio, Σ(H + D)/¹²C, in the outer divertor tile was the lowest at the strike point, and increased with the distance from the strike point. At the inner divertor tile with a thick re-deposition layer, Σ(H + D)/¹²C was very low (values in the range of 0.01 ~ 0.07). After the removal of the re-deposition layer, the integrated ratios Σ(H + D)/¹²C and ΣD/¹²C increased by factors of 4 and 70, respectively. In the cases of the baffle plates and the dome top tiles, Σ(H + D)/¹²C was higher. The intensity ratios, D/¹²C and H/¹²C, were strongly influenced by the temperature of the tile, the existence of redeposition layers or eroded regions, and the fluxes of hydrogen isotope ions and charge exchange particles during the D-D operation.

Recent results regarding the formation of co-deposits, fuel accumulation and overall material transport at the TEXTOR tokamak are described. Two categories of brittle flaking co-deposits were identified: (i) smooth stratified layers of a thickness of up to 50 µm and a fuel content of up to 16 at.% , (ii) granular and columnar structures reaching 1 mm in thickness and con-taining around 0.5 at.% of fuel species. They were formed on the blades of the toroidal belt pump limiter (~15000 s of plasma operation) and on the neutral-iser plates of this limiter (~90000 s), respectively. A comparison is made to the fuel inventory measured in other controlled fusion devices with carbon walls.

Underneath the ASDEX Upgrade divertor two different types of amorphous hydrocarbon layers were found. Transparent soft layers with a typical ratio of D/C ≈ 1 and brownish hard layers with a ratio D/C ≈ 0.5. The amount of carbon deposited at the inner divertor is a factor ≈ 3 larger than at the outer divertor. Typical 0.3% of the total deuterium input during one experimental campaign are found in these layers in the divertor and only about 8 · 10^-5 of the D input in the pumping ducts. From the deposition pattern it is concluded, that the layers are mainly build up by species with high sticking probability (0.1 < β < 0.9). Neutral collisions and a parasitic plasma below the divertor are involved in the layer formation. Discharge resolved measurements of the layer deposition reveal a proportionality of the layer growth on the neutral density below the divertor.

Surface induced dissociation of diatomic and tri-atomic hydrogen and deuterium ions upon impact on a hydrogen covered graphite surface has been investigated in the collision energy range from about 20 up to 100 eV. In this typical energy regime for fusion edge plasma conditions relative abundances of H⁺ and D⁺ fragments as a function of collision energy were measured, no di-atomic secondary product ions could be observed. As target surfaces graphite tiles from the Tokamak experiment TORE SUPRA in CEA-Cadarache/France have been used.

High heat flux interactions with plasma facing components have been studied at microscopic scales. The beam from a continuous wave neodymium laser was scanned at high speed over the surface of graphite and carbon fiber composite tiles that had been retrieved from TFTR and JET after DT plasma operations. The tiles have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma operations and laser scanning has released more than 80% of the codeposited tritium. The temperature rise of the codeposit was much higher than that of the manufactured material and showed an extended time history. The peak temperature varied dramatically (e.g. 1436°C compared to >2300°C) indicating strong variations in the thermal conductivity to the substrate. A digital microscope imaged the codeposit before, during and after the interaction with the laser and revealed 100-micron scale hot spots during the interaction. The heat flux produced a pattern of beads on a mixed Be/C deposit. Heat pulse durations of order 100 ms resulted in brittle destruction and material loss from the surface, whilst a duration of ≈ 10 ms showed minimal changes to the codeposit. These results show that reliable predictions for the response of deposition areas to off-normal events such as ELMs and disruptions in next step devices need to be based on experiments with tokamak generated codeposits.

For fine-grain graphites with different final heat treatment, the influences of the porosity, degree of graphitization, and dopant (TiC, VC, WC, and ZrC) on the fluence dependence of the retention of 1 keV deuterium were investigated using thermal desorption spectroscopy. A strong decrease of the D retention for fluences higher than 10²¹ D/m² was observed for the undoped graphites graphitized at temperatures above 2000 K compared to material only calcined at 1270 K. Due to the identical manufacturing processes for the carbide-doped graphites used in this study, the structure is comparable for all of them. The choice of dopant as well as the ratio of open to closed porosity show no influence on the D retention. Therefore, these properties of the graphites can be neglected for hydrogen retention estimations.

Interaction of hydrogen atoms with highly oriented pyrolytic graphite (0001) surface at near room temperature was studied by means of thermal programmed desorption (TPD), atomic force (AFM) and scanning tunneling (STM) microscope. Bubbled surfaces were observed after irradiation of graphite with atomic hydrogen. The average height of bubbles was about 3–5 nm and the lateral size up to 100 nm. AFM and STM tips are not only used for surface imaging but also for destruction of bubbles. The possibility of bubble destruction with STM and AFM tips without noticeable traces of remaining substance of the bubbles on the surface, and a flat bottom of the produced craters may be regarded as a further support in favor of our previous hypothesis of hydrogen intercalation between upper graphite layers.

Carbon films co-deposited in divertor regions are known to represent an important problem in future fusion devices due to their strong tritium content. This has also been identified as a problem in the design of the ITER-FEAT divertor, and many methods have been proposed to remove the films, most of them implying the exposure of the inner walls to oxidizing agents. On the other hand, experiments at Ciemat have recently shown the potential of using ion scavengers, such as nitrogen, to completely suppress the film deposition from hydrocarbon/hydrogen mixtures at relative concentrations typical of chemical sputtering scenarios in divertor plasmas. In the present work, Thermal Desorption Spectroscopy and He Glow Discharge Desorption have been used to characterize the hydrogen and hydrocarbon content of the films created under the presence of nitrogen. Preliminary experiments on the use of other ion and radical scavengers (NO, NH₃) are also presented. Results for the deposition from glow discharge plasmas of hydrogen-diluted methane are presented. The implications of the results for the control of T co-deposition in divertors are also addressed.

On the tokamak TEXTOR hydrogen atom velocities and penetration depths have been determined both by active Lyman- and passive Balmer-line spectroscopy of atomic hydrogen in combination with molecular spectroscopy for a variety of surface temperatures and gas blow rates. These variations influence both the molecule to atom ratio and the local boundary plasma parameters, which in turn determine the Franck–Condon energies of the dissociated atoms. Most of the released atoms have thermal energies and reflect the temperature of the emitting surface. Molecules from these surfaces are selectively vibrationally excited and may dissociate with small energies because e.g. excitation can now take place into lower energies of the repulsive state 2p ³Σ_u. On the other hand this state can also give rise to high Franck–Condon energies for ground state molecules in a cold plasma.

Tritium Imaging Plate Technique (TIPT) is used to measure the tritium areal distribution on plasma facing graphite tiles of JT-60U. Measurement of the W-shaped divertor region showed that the surface tritium distribution of tiles used under the "both-side pumping" system was similar to that of tiles used under the "inner-side pumping" system. This result suggests that tritium retained in those tiles is not affected by the type of exhaust system because this tritium was implanted with high energy. Thus, tritons produced by D-D reaction were implanted in the plasma facing wall without completely losing their initial energy. The tritium level beneath the redeposited layer on the inner divertor target tile was about three times as high as that of the surface of the redeposited layer. The toroidal surface tritium distribution suggested ripple loss of high energy tritons. These results also indicate that long-term tritium retention in JT-60U was caused by implantation of high-energy tritons.

Tungsten and tantalum were examined in the TEXTOR tokamak as test limiters in order to compare their performance under plasma operation and to recognise fuel recycling on endothermic (W) and exothermic (Ta) hydrogen absorbers. Differences have been noticed in the distribution and microstructure of co-deposits, in the fuel inventory in the bulk of metals and, in the deuterium release mechanism (ratio of molecules to atoms). As a result of poor thermal conductivity, the surface temperature of Ta during the power deposition was higher than that of W and it increased shot-by-shot because of the degradation of thermal properties due to surface modification. Results on thermal response, fuel recycling and inventory show that, as a candidate material for plasma facing components, tungsten is superior to tantalum.

In this paper, we describe the formation of HD molecules in front of a graphite surface in the boundary layer of TEXTOR. The surface release of HD molecules was observed under different experimental conditions: in neutral beam experiments, in surface saturation experiments and in gas injection experiments. Molecular line emission of HD could be measured for the first time in a high temperature boundary plasma by means of Fulcher-α-band spectroscopy. The Q-branches of the first four diagonal bands were identified and analysed with respect to their rotational and vibrational population. No differences between isotopomers could be ascertained in the vibrational ground state population. Calibration spectra, which describe the isotope differences, as well as first experimental results with regard to the isotope exchange are shown. The main result is that HD molecules have to be included into the determination of the particle flux of recycled hydrogen isotopomers. By means of a model constructed on the chemical law of mass action, the release of the isotopomer H₂, HD, and D₂ can be described in the form of concentration ratios.

The D retention in W (100) single crystal irradiated with 6 keV deuterium ions up to a fluence of 1 × 10²² D/m² at temperatures of 300, 430 and 480 K has been studied by TDS. Moreover, the structural evaluation of D irradiated W with depth has been investigated by using X-ray diffraction technique. The obtained results show that the retention curve of implanted deuterium deviates from the line of 100% trapping at RT. Almost all (about 99%) trapped deuterium was accumulated in the near-surface layer with a thickness less than 0.25 µm at an average D concentration of about 5 at.%. No changes (within ±10%) of retained D for a long holding time (more than 600 h) between the irradiation and TDS experiments was observed. Increasing the temperature of the irradiated target from 300 to 430 K leads to a decrease of the amount of trapped D by a factor of about 30. The retention of D at 480 K and the fluence of 1 × 10²² D/m² was below the detection limit of 2.5 × 10¹⁸ D/m². The structural analysis has shown that during the irradiation at 300 K mainly vacancy type defects (vacancy clusters and cavities) in the near surface layer (~0.5 µm) and interstitial type defects (dislocation loops) within a layer of a ~2 µm were formed.

The depth profiles of deuterium trapped as D atoms and D₂ molecules in WO₃ films both exposed to D plasma at temperatures in the range from 300 to 650 K and implanted with 10 keV D ions at 300 K and 650 K respectively have been determined by means of secondary ion mass spectrometry and residual gas analysis. After plasma exposure at 350–550 K and ion implantation at 300 K, most of the deuterium is accumulated as D₂ molecules. The appearance of molecular deuterium in the D implanted WO₃ film is considered to be related to the formation of cavities filled with deuterium. D atoms seem to be chemically bonded to O atoms forming deuterium tungsten bronze and captured by grain boundaries. Further irradiation with 10 keV D ions at both 300 K and 650 K leads to the reduction of WO₃ to metallic W within the ion stopping zone.

Bulk hydrogen retention and hydrogen diffusion in graphite and carbon materials have been studied to estimate hydrogen recycling and tritium inventory in a fusion reactor environment. Hydrogen may permeate into a filler grain in the form of a hydrogen molecule, diffuse through crystallite boundaries and finally be trapped as hydrogen atoms at the edge surface of a crystallite. In the estimation of hydrogen transport, the activation energy of hydrogen diffusion can be determined from absorption experiments, and the activation energy of detrapping can be obtained from desorption experiments. The activation energies of hydrogen trapping are 2.6 eV in ordinary graphite and 4.4 eV for irradiated or mechanically milled graphite samples.

The retention and re-emission behaviors of hydrogen isotopes in SiC have been studied by means of the elastic recoil detection (ERD) technique. Protium ions or deuterium ions were implanted into SiC sample at room temperature up to almost saturation. It was found that the concentrations of protium and deuterium in SiC reach values of 0.70 and 0.75, respectively. The ratio of the saturation concentration of H to D is calculated to be 0.92 by the solution of the mass balance equations, which is consistent with the experimental value of 0.93. Isochronal annealing for 10 min from room temperature to 1123 K was applied to deuterium implanted SiC sample to analyze the thermal release behavior. It was found that the deuterium re-emission takes place in three stages. The major second and third stages are ascribed to the re-emission as deuterium molecule of deuterium atoms bound to Si and C in SiC, respectively. The first stage might be the re-emission of deuterium-silicon compounds such as SiD₂ or SiD₄. Isothermal annealing at 573 K and 773 K were applied to evaluate the amounts of deuterium re-emitted in the first and second stages. The mass balance equations were also used to evaluate the effective molecular recombination rate constant, which was determined to be 7.3 × 10^-5.

A computer code incorporating real surface reliefs for calculations of integral and differential parameters of particle reflections from solids is briefly described and validated by comparing with experiments and known analytical solutions for ideally smooth surfaces.

Calculations of the angular distributions of deuterons reflected from rolled SS surfaces and grinded W samples show differences depending on the angle of irradiation relative to the direction of the relief structure. A comparison of calculations with experimental data on 100 eV deuteron reflection from grinded W shows reasonable agreement for the experimentally investigated interval of angles. Some new features of angular distributions in the plane of incidence and in the perpendicular plane are found. So, the possibility of more accurate accounte of reflected particles contributing to hydrogen recycling near real PFC of fusion devices is demonstrated.

Simulation of sputtering experiments were carried out in steady state plasma of a mirror Penning discharge under low and high hydrogen recycling regimes. The samples for studies were Pd (99.98 purity) foils with and without hydrogen saturation of concentrations up to H/Pd = 0.65. Irradiation doses were 10¹⁸–10¹⁹ ions/cm². No significant difference in Pd erosion behavior was found under hydrogen, helium or nitrogen plasma impact for the different recycling regimes. The absolute values of the erosion coefficient in helium and nitrogen plasmas are in sufficient agreement with ion sputtering data in literature. However, in the case of hydrogen plasma disagreement of more than an order of magnitude was observed. Possible reasons are suggested and discussed to explain such erosion behavior.

In this paper we consider the possibility of using a hydride-forming Zr-V alloy as a first-wall material for fusion devices. Theoretical and experimental studies of the Zr-V hydride irradiation with flows of high-energy plasma particles have been performed. It is shown that the use of these materials allows the reduction of erosion of the first wall and also a better control of the hydrogen isotope recycling.

Diffusion of tritium and helium-3 in 21-6-9 stainless steel vessels, during the storage of tritium at the pressure of 6.13 MPa for about 4 and 6 years at room temperature and later exposed to air for about 3 and 1 years respectively, was investigated by determination of the concentration distribution through etching the surface with acid solution and analyzing the released gases, and by calculation based on diffusion and decay theories. Results show that the natural oxide layers of stainless steel and some micro crannies have taken effects on the diffusion behavior of tritium, which made a "valley" but smooth T₂ in the concentration distribution curves of tritium. Helium-3 atoms are mainly distributed near the surface and have some very high local concentrations due to the strong tendency to be captured by defects. The calculation by numerical methods shows that the depths which tritium and helium-3 can reach in the two vessels are both about 500 µm. And due to the complex disturbances by many factors, the measured results differ much from the calculation ones in absolute values.

In this study, the kinetics of desorption of adsorbed hydrogen from an ideal metallic surface is modelled in Cellular Automaton (CA). The modelling is achieved by downgrading the surface to one dimension. The model consists of two parts that are surface migration and desorption. The former is attained by randomly sorting the particles at each time, the latter is realised by modelling the thermally-activated process. For the verification of this model, thermal desorption is simulated then the comparison with the chemical kinetics is carried out. Excellent agreement is observed from the result. The results show that this model is reasonable to express the recombinative desorption of two chemisorbed adatoms. Though, the application of this model is limited to the second-order reaction case. But it can be believed that the groundwork of modelling the transport dynamics of hydrogen through the surface under complex conditions is established.

Permeation through a sandwich structure "martensitic steel-liquid LiPb alloy" has been modeled by numerical calculations. Diffusion-limited, absorption-limited, and recombination-limited regimes have been observed in various conditions. Features of permeation dynamics, steady state permeation rate, concentration profile, and the net accumulation have been analyzed in these regimes. A comparison of permeation from different sides of the sandwich has been made.

The effects of hydrogen ion (H₃⁺) bombardment on the optical properties of polycrystal stainless steel (SS) mirrors were investigated. Ellipsometry, profilometry, scanning electron microscopy, angle distribution of scattered light, and spectral reflectance were used to characterize the surfaces. Results for the bombardment of SS mirrors with ions of fixed energy (0.3, 0.65 and 1.5 keV/H), with ion flux density (0.5–2) · 10²⁰ H/m²s up to fluences of (1.1–4.3) · 10²⁴ H/m² are presented.

The data show that the surface roughness rises with both increasing ion energy and ion fluence, and that surface roughening leads to an increase of the scattered light with a corresponding decrease of reflectance at normal incidence. The thickness of the apparent layer, obtained by ellipsometry, was found to rise with increasing ion energy at fixed ion fluence and with increasing ion fluence at fixed ion energy.

Possible mechanisms for the ion energy effect on the degradation rate of stainless steel mirrors and the origin of the apparent layer are discussed.

A systematic study was carried out on the relation between the surface recombination rate constant of deuterium at a Nb surface and the oxygen concentration in bulk. Oxygen was introduced into a Nb specimen up to concentrations from 0.12 to 4.3 at.%. Then, deuterium was introduced into the specimen, and the surface recombination rate constant was measured by thermal desorption in a temperature range from 623 to 913 K. The surface recombination rate constant of deuterium decreased with increasing oxygen concentration in the solid solution phase. This reduction in the recombination rate constant was attributed to the increase of oxygen coverage at the surface with the bulk concentration. Irrespective of the bulk oxygen concentration, no significant increase was observed in the height of the potential barrier against desorption processes.

An apparatus equipped with X-ray Photoelectron Spectroscopy (XPS) and Thermal Desorption Spectroscopy (TDS) was constructed to study interactions of energetic hydrogen isotopes with plasma facing materials. It is a remarkable feature of the apparatus that energetic ion implantation is carried out at around 150 K to study reactions of energetic ions with matrix by suppressing the reactions of thermalized ions. Using this apparatus, TDS experiments for pyrolytic graphite implanted with energetic D₂⁺ ions at 173 and 373 K were carried out. The experimental results suggest that the deuterium implanted was released through a four-step release processes, involving three D₂ and one CD_x (x = 2, 3 and 4) desorption processes. Two deuterium and CD_x desorption processes were observed in the temperature range from 700 to 1200 K. In addition, a new deuterium desorption process was observed for the deuterium-implanted sample at 173 K. This has never been observed for deuterium-implanted graphite implanted at temperatures higher than room temperature (R.T.).

Hydrogen trapping in tantalum was experimentally studied using an in-situ observation technique. One side of a tantalum membrane was continuously exposed to a deuterium plasma and was bombarded with 0.8 MeV ³He ions to introduce traps. After the bombardment, the deuterium depth profiles near the surface were observed by nuclear reaction analysis with the reaction of D(³He, p)⁴He. The result indicated that many deuterium atoms were present in traps which were associated with radiation damages. The temperature dependence of the deuterium concentration showed that there were two kind of traps, called trap 1 and trap 2 here. The difference in the peak shape of the depth profiles and the difference in the trapping energy suggested that traps 1 and 2 were caused by vacancy clusters and monovacancy sites, respectively. The concentration of the trapped deuterium was found to be of the same order of magnitude as those in molybdenum and nickel under the same experimental conditions.

The photon or electron stimulated desorption (ESD or PSD) of D₂O and H₂O adsorbed on iron surface covered with thin iron oxide film was studied with QMS. The photon energy was varied from 3.1 to 10.8 eV using a deuterium lamp and a Hg-Xe lamp. The electron energy was 30 eV. During photon or electron irradiation, water or hydrogen molecules were desorbed. During photon irradiation with energy < 4 eV, water desorption was initiated by thermal effect. During photon or electron irradiation with energy < 4 eV, water desorption was initiated by thermal effect and electronic excitation. During photon or electron irradiation with energy > 10 eV, hydrogen molecule desorption was initiated by electronic excitation. During photon irradiation with energy < 10 eV, hydrogen molecule desorption was not observed. Water desorption efficiency by these methods was also discussed.

A list of participants and their email addresses from the International Workshop on Hydrogen Isotopes in Fusion Reactor Materials, held at the University of Tokyo, Japan on 22–24 May 2002.