J. Yagyu

Localized tungsten deposition in divertor region in JT-60U

Tungsten local transport in the diverter region emitted from the outer divertor was studied in JT-60U. Tungsten-coated CFC tiles were installed in the outer divertor position, on which outer strike points did not normally sit. A neutron activation method to measure tungsten surface density was used for the first time. On the outer wing tile, W existed only in the top surface, while on the inner divertor tile W was embedded in the thick carbon deposition layer. Poloidal distribution in the same toroidal section as the tungsten tiles showed dense W deposition near the inner strike points, the dome top and outer wing of the dome. Carbon ions by puffing 13CH4 at the same toroidal position of the W-tiles were also deposited near the inner strike points. Toroidal distribution on the outer wing showed significant localization of W deposition near the W tile array. In addition, toroidal asymmetry of W deposition on the outer wing probably due to the plasma parallel flow was observed.

Tritium Distribution on First Wall Carbon Tiles in JT-60U

Abstract T retention and its depth profile in the graphite tiles used for first wall of JT-60U have been measured by a tritium imaging plate technique and a full combustion method. T was found only limited depth beneath the plasma facing surface and little in both the surface region shallow than 1 μm and in bulk more than 1mm in depth. Although most of T produced by DD reactions are thermalized and neutralized in plasma and impinge on the plasma facing surface and penetrate into the inner surface, they are isotopically replaced by subsequently incoming D. Only some of high energy T escaping from plasma are directly implanted beneath the surface and retained escaping from the isotopic replacement until attainment of a saturation concentration.

Wall conditioning and experience of the carbon-based first wall in JT-60U

JT-60U has a variety of wall-conditioning methods such as baking of the vacuum vessel, helium Taylor discharge cleaning, helium glow discharge cleaning, tokamak discharge cleaning, and boronization. Using these wall-conditioning methods, the high-power operation of JT-60U has been successfully carried out with the carbon-based first wall. The material behavior of the carbon-based first wall has been investigated, and important knowledge was obtained on mechanical engineering and plasma surface interactions. In order to understand the tritium behavior in JT-60U, tritium retention in the first wall and tritium exhausted through the pumping system were measured. These results yield useful information on the tritium behavior in a future DT fusion machine.

Hydrogen Isotopes Retention in Gaps at the JT-60U First Wall Tiles

Retentions of Hydrogen (H) and deuterium (D) in the side surfaces (gaps between tiles) of the carbon tiles used as first wall in JT-60U were measured by the thermal desorption spectroscopy. In the gaps, the H and D retention were dominated in carbon deposited layer. The gap retention was less than that of the eroded plasma facing surface, where the retention was saturated, and linearly increased with exposure time. Overall retention rate in the gaps of the first wall tiles was determined to be 4.0 × 1019 H+D/s, and was comparable or larger than those in the re-deposited layers on the plasma facing surfaces and in the shadowed areas in the divertor region.

Tritium Retention in First Wall Tile Gaps of JT-60U

Abstract Tritium (T) retentions in tile gaps (side surfaces) of the first wall of JT-60U were measured by a tritium imaging plate technique (TIPT). For all first wall tiles measured here, the T retention decreased from the front (entrance) to the bottom of the side surfaces showing superposing two exponential decays, which were already observed in the divertor region. Heavier erosion on the plasma-facing surface resulted in higher T retention in the front-side surfaces in the vicinity of the plasma-facing surface. In addition, wider gap width also resulted in higher T retention in the bottom side surfaces. Using the TIPT results, overall T retention in the side surfaces of the whole first wall was estimated to be ~6 × 1017 T atoms, which was only one-tenth of total T retention in the plasma-facing surface of the first wall in JT-60U.