THIRD INTERNATIONAL WORKSHOP
ON TRITIUM-MATERIAL INTERACTIONS
(Satellite Meeting of the 7th ISFNT)

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Preface

 The third "International Workshop on Tritium-Material Interactions" was held on May 20-21st, 2005, at Wel Sunpia Tateyama, locating in a resort area by Tateyama mountains in Toyama, as a satellite meeting of the 7th International Symposium on Fusion Nuclear Technology (May 23-27th, 2005, Tokyo, Japan). The basic topics of tritium-material interactions were taken up as scope of this workshop, because it is of great importance in the future fusion devices as well as ITER to establish the safety handling technologies of a large amount of tritium. Site selection of ITER has been finished on the last summer. After this the ITER organization will be built up, and construction of the experimental reactor will start. From this viewpoint, basic data with respect to adsorption, dissolution, diffusion, permeation, trapping/detrapping of tritium in the fusion materials will play an important role in the state-of-the-art design. Seven papers were presented by invited speakers from overseas and Japan. The foreign invited speakers were from Sweden and Germany. The Japanese speakers were from the Japan Atomic Energy Research Institute, Hokkaido University, Shizuoka University, and Toyama University. The meeting was in an appropriate size of about 40 attendants (including three attendees from overseas), giving a good opportunity to keep close contact for exchanging views and to deepen acquaintance each other.
  By focusing attention to the tritium-material interactions, up-to-date information were presented from both basic and technological viewpoints as evaluation of the retention of tritium in plasma-facing materials, absorption of hydrogen isotopes in Pd-alloys, processing of tritiated water, and effects of radiation damage in ceramic breeders.
  Each paper commanded strong interest of the attendants and aroused intense discussions. The discussions were liveliest in a very frank atmosphere. The workshop confirmed the importance of further studies on fundamental phenomena and elemental techniques related to the safety handling and confinement of tritium in a fusion reactor. There is a good reason to believe that the tritium management and safety control techniques can be developed extensively through worldwide systematic studies and cooperation.
  Finally, thanks should be extended to all of the participants who made the workshop enjoyable and fruitful. Special thanks are due to Drs. M. Rubel, C.H. Wu and I. Cristescu visiting Toyama from overseas for presenting invaluable papers.

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[workshop 2005_01]

　Components of JET divertors have been examined to in order to assess the fuel retention in areas shadowed from the direct plasma impact: castellation groves in beryllium blocks, gaps separating tiles of Mk-I, and the interior of the septum module in the Mk-II Gas Box (GB) structure. The results show that: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) in these gaps the fuel inventory exceeds that on plasma facing surfaces by up to a factor of 2; (iii) in the grooves of castellation the fuel content is less than 3.0 % of that found on top surfaces; (iv) fuel retention in the septum of Mk-IIGB is insignificant in comparison to that detected in the shadowed region in the inner divertor corner. Implications of these results for a next-step device are addressed.
Keywords: fuel inventory, material migration, carbon, beryllium, JET

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[workshop 2005_02]

　The retention properties and depth profiles of deuterium and/or hydrogen in the graphite tiles located in the divertor region used in the W-shaped divertor of JT-60U with inner side pumping and JT-60 were investigated by thermal desorption spectroscopy (TDS) and secondary ion mass spectroscopy (SIMS). Total amount of retained hydrogen in the samples used in JT-60 and hydrogen and deuterium in JT-60U was proportional to the thickness of the re-deposited layers. Assuming homogeneous hydrogen distribution in the re-deposited layers and carbon density of 0.9 g/cm³, H/C in the re-deposited layers was determined to be approximately 0.03 from this slope of the linear relationships. This value is almost the same as that observed in the tiles of JT-60U and is much smaller than those observed in JET and other devices. The re-deposited layers are very likely subjected to high temperatures because of their poor thermal contact to the substrate owing to their porous structure.
Keywords: hydrogen, deuterium, retention, re-deposited layer, depth profile, JT-60, JT-60U

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[workshop 2005_03]

　In the next step D/T fusion device, the accurate predictions of tritium retention in Plasma Facing Component is important both from the point of view of the safety and the physics performance.
  However, for reliable prediction of T-retention, a detailed knowledge and understanding on the mechanisms, which are responsible for T-retention is essential. The mechanisms are quite complex, for example, in fusion devices; the PFC is in contact with plasma and PFC surface exposed to large flux of ion and neutral particles. The large flux of ions may be retained in the PFC materials by implantation in the depth of ion range, the particles will diffuse to the bulk of materials and eventually, the particles are trapped. In addition, the ion and neutral particles cause the sputtering of surface. The sputtered particles are ionized in plasma and redeposited somewhere on the surface within the fusion devices. The sputtering and deposition lead to T-retention via co-deposition.
  After a decade experimental investigation and theoretical study on D/T plasma wall interaction, D/T transport in materials as well as neutron effects, the progress on understanding of the complex of T-retention is significant, so that the prediction of T-retention can be performed more accurate.
  In this paper, a detailed analysis on the mechanisms of tritium retention has been performed, based on:
1. Tritium retention via implantation,
2. Tritium retention via bulk diffusion,
3. Tritium retention via neutron transmutation,
4. Tritium retention via co-deposition,
5. Tritium retention via neutron damage induced trapping sites
6. Tritium retention in dust and flakes,
Keywords: tritium, plasma facing materials, implantation, co-deposition, neutron effects

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[workshop 2005_04]

　　Tungsten plates annealed at 1473 K were exposed to deuterium plasmas with incident energies ranging from 7 to 98 eV and a fixed flux of 10²² D/m²/s, and the blistering and retention in the near-surface region were investigated with a variety of techniques, such as scanning and transmission electron microscopy (SEM and TEM), thermal desorption spectroscopy (TDS), nuclear reaction analysis (NRA), elastic recoil detection (ERD), secondary ion mass spectroscopy (SIMS) and X-ray diffraction (XRD),. Blisters with the maximum diameter of about 2 microns (comparable to grain size) were formed on the tungsten surfaces after plasma exposure, and small blisters with a diameter of around 30 nm and microcracks were formed in the near-surface region before the formation of larger blisters. Within the experimental error there was a zero change in the lattice parameter after the plasma exposure, implying that deuterium does not exist in the lattice interstitial sites, but instead forms a deuterium-vacancy complex and then clusters and further bubbles (predominantly in the form of molecules in vacancy clusters and voids) in the near-surface region. After deuterium plasma exposure, deuterium was retained in the depth up to a few microns from the surface, and the maximum atomic ratio of deuterium against tungsten reached as high as 1-2% in the near-surface region. These evidences suggest that crystal defects like vacancies should be generated due to lowering of the formation energy of vacancies by the intrusion of a great number of hydrogen isotope atoms into the near-surface region of tungsten.
Keywords: tungsten, retention, blistering, hydrogen isotopes

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[workshop 2005_05]

　To find out the most efficient column material for a newly developed gas chromatographic hydrogen isotope separation system, changes in the heat of hydrogen absorption and in the thermodynamic isotope effect with alloy composition were investigated for Pd-based alloys, Pd_(1-x)AE_x, where AE=Co, Ni, Cu, Rh, Pd, Ag, Pt, or Au. It was experimentally observed that the heat of absorption decreased with increasing AE content for Co, Ni, Cu, Rh, Pt, and Au, although the extent of the reduction differed from each others. On the other hand, alloying of Ag caused the increase in the heat of absorption. With respect to the isotope effect, variations of the alloying element and the extent of alloying showed no noticeable effect on the isotope effect among the alloying investigated including pure Pd.
  Those features were analyzed from electronic structures of the alloys by use of Gaussian 03 and DVXα packages, where the caluculations were carried out for small clusters as Pd_(8-y)AE_y for the former and Pd_(14-y)AE_y for the latter. The ab initio caluculations showed that the energy of the highest occupied molecular orbital of the clusters changes almost linearly with alloy composition, suggesting that the Fermi energy changes with alloy composition. By assuming the Fermi energies of alloys to be arithmetic means of the Fermi energies of pure Pd and AE, a linear relation was found between the observed heat of absorption and the Fermi energy of the alloys. Vibrational analysis of Pd_(8-y)AE_y-H systems showed that the force constant for the bond between a H and a host metal atoms dose not change much irrespective of different alloying element, causing the isotope effect to be almost invariant with alloying element and alloy composition.
Keyword: hydrogen absorption, Pd-alloys, heat of absorption, isotope effect, alloying effect

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[workshop 2005_06]

　The Combined Electrolysis Catalytic Exchange (CECE) method in combination with Cryogenic Distillation (CD) was chosen for tritium recovery from tritiated water which will be produced during ITER operation. One of the key components with impact to both the tritium inventory and safety is the electrolyser. The solid polymer electrolyser type is proposed but the main concern is the life time in tritium environment. On overview of main activities devoted to the life time of a SPM and carried out at Mound facility-US, TPL JAERI and TL Karlsruhe are presented in this paper.

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[workshop 2005_07]