A. Kaminaga

Tritium distribution in JT-60U W-shaped divertor

Abstract Detailed tritium profiles on the JAERI Tokamak-60U (JT-60U) W-shaped divertor and the first wall tiles were examined by a tritium imaging plate technique (TIPT) and full combustion method. The highest tritium level (60 kBq/cm 2 ) was observed at the dome top tiles. The tritium level of the divertor target was lower (2 kBq/cm 2 ). The result of the triton deposition simulation using orbit following Monte-Carlo code was consistent with the tritium distribution obtained by TIPT and full combustion method. These results indicate that the tritium distribution of the JT-60U W-shaped divertor reflects mainly the distribution of the energetic triton impinging on the wall. According to the simulation, the tritium atoms produced by D–D nuclear reaction in JT-60U are not loosing completely their initial energy of 1 MeV and around 1/3 of them are implanted into the wall.

Analyses of erosion and re-deposition layers on graphite tiles used in the W-shaped divertor region of JT-60U

Erosion and re-deposition profiles were studied on graphite tiles used in the W-shaped divertor of JT-60U in June 1997-October 1998 periods, operated with all-carbon walls with boronizations and inner-private flux pumping. Continuous re-deposition layers were found neither on the dome top nor on the outer wing, while re-deposition layers of around 20 pm thickness were found on the inner wing, in the region close to the dome top. On the outer divertor target, erosion was found to be dominant: maximum erosion depth of around 20 μm was measured, while on the inner target, re-deposition was dominant: columnar structure layers of maximum thickness at around 30 pm on the inner zone while laminar/columnar-layered structures of maximum thickness around 60 μm were found on the outer zone. Poloidal distributions of the erosion depth/re-deposition layer thickness were well correlated with the frequency histograms of strike point position, which were weighted with total power of neutral beam injection, on both the outer and inner targets. Through X-ray photoelectron spectroscopy, composition of the re-deposition layers at a mid zone on the inner target were 3-4 at.% B and <0.6 at.% O, Fe, Cr, and Ni with remaining C. Boron atoms are mostly bound to C atoms but some may precipitated as boron.

Evaluation of radiation shielding, nuclear heating and dose rate for JT-60 superconducting modification

Abstract The radiation shielding, nuclear heating and dose rate have been evaluated for JT-60 superconducting modification. The double-walled structure of a vacuum vessel was adopted so that the maximum nuclear heating at the inboard side of the superconducting TF coil could be suppressed to be lower than 2.5 mW/cm3. The 316 stainless steel (SS316L) boards outside the vacuum vessel are installed for effective gamma-ray shielding. The dose rate with a reduced activation ferritic steel pedestal for the first wall in the vacuum vessel was estimated. This pedestal reduced the in-vessel dose rate by 40% compared to SS316L pedestal. The dose rate outside the boron carbide doped concrete cryostat was estimated at shutdown after 10 years operation.

Performance of B4C-converted carbon fiber composites in high power neutral beam heated divertor discharges in JT-60U

B 4 C conversion-coated carbon fiber composites (B 4 C-converted CFC) have been installed on the outboard strike point of the divertor plate in JT-60U to form a complete toroidal ring. The plasma-surface interactions of the B 4 C-converted CFC have been investigated in high power divertor operation with neutral beam (NB) heating power of - 30 MW for 2 s. It is found that the carbon impurity is reduced without a significant increase of the boron impurity using a 100 μm thick B 4 C layer with a B /C ratio of 2.7. The oxygen impurity is reduced significantly by the evaporation of a 300 μm thick B 4 C layer with a B /C ratio of 3.8. A reduction in the detrimental effects on plasma-surface interactions with carbon-based materials has been demonstrated for more than 1000 shots by the use of surface-boronized carbon tiles in JT-60U

Investigation of plasma facing components in JT-60U operation

Abstract The mechanical fracture of three carbon fiber composite (CFC) first wall tiles was observed. This damage was probably caused by the electromagnetic force due to halo current during disruption. The required current to break the CFC tile is estimated to be 25 kA. The broken tile was rotated poloidally around the plasma with a speed of about 10 m/s during the following discharge. A possible driving force of this rotation might be the electromagnetic force due to the scrape-off layer (SOL) current. The required current to rotate the piece of the broken tile is 1 kA. These results indicate that electromagnetic interaction between SOL plasma and the plasma facing component is important in the research on the plasma wall interactions in fusion devices.

Quality evaluation of graphites and carbon/carbon composites during production of JT-60U plasma facing materials

Abstract The variations of physical and mechanical properties have been investigated for three grades of isotropic graphites and four grades of carbon/carbon (C/C) composites on the basis of the sample inspection data which have been obtained during production of the first wall and divertor plate materials for JAERI Tokamak-60 Upgrade (JT-60U). The evaluated properties are density, electrical resistivity, coefficient of thermal expansion (CTE), thermal conductivity, bending, tensile and compressive strengths. It is found that the maximum standard deviations normalized by the mean values are 22.7% for the C/C composites and 9.2% for the isotropic graphites. This scatter of the material quality should be considered in the design of the isotropic graphite and C/C composite armor tiles. Correlations between these properties are also observed for several materials.

Measurement of gaseous impurities in JT-60U

Using quadrupole mass analyzers, measurements were made of gaseous impurity pressures at the inner and outer divertor regions and at the mid-plane during neutral beam heated discharges in JT-60U. The hydrocarbon partial pressures in both divertor regions increase with the deuterium pressure, as a common function of deuterium pressure, (i.e., P CD4 = 3 x 10 −2 P D2 0.7 for CD 4 ), irrespective of different discharge conditions of neutral beam he power, direction of ion grad-B drift and gas puffing. On the other hand, the pressures of CO and CO 2 are rather insensitive to the deuterium pressure. The sum of the gaseous impurity partial pressures is dominated by CO and CO 2 in the low recycling discharges, and by hydrocarbons in the high recycling discharges. The total gaseous impurity pressure, as a fraction of the deuterium pressure, is on the order of 50% and 1-10% in the respective recycling regimes

Clean-up of graphite first walls in JT-60

During the venting period from April to May 1987 about a half of TiC-coated molybdenum and Inconel 625 first walls were removed and graphite first walls were installed. The overall surface area and the weight of the graphite tiles are 200 m2 and 3500 kg, respectively. Two kinds of isotropic graphite (Ibiden ETP-10 and Hitachi-Chemical HCB-18S) were used. Special attention was paid to the contamination of the graphite first walls during the courses of fabrication and installation. Main discharge were tried after 7 d conditioning, which included 3 d bakeout at the temperature of 300°C, 18 h glow discharge cleaning and 36 h Taylor-type discharge cleaning keeping the vessel temperature around 280°C. A 0.5 MA stable plasma was successfully attained at the 6th shot and a very rapid start-up of the discharge conditioning with plasma current up to 1 MA was achieved with 14 shots. An ultimate pressure of 6.4 × 10−7 Pa was obtained with the vessel temperature at RT and the residual gases were water, carbon oxide and carbon hydride.

Engineering design study of JT-60 superconducting modification

The modification of JT-60 is planned as a fully superconducting tokamak (JT-60SC). The mission of JT-60SC program is to establish scientific and technological bases for an advanced operation in an economically and environmentally attractive DEMO reactor and ITER. The research objectives are to accomplish high performance steady state operation with high beta and non-inductive full current drive, with high bootstrap current fraction, and demonstrate the plasma applicability of reduced-activation material for a plasma of break-even class relevant to the reactor plasma. Basic design of JT-60SC has been completed and the detailed design is under way. The engineering design for main components of JT-60SC is described.

OPERATION EXPERIENCES WITH JT-60U PLASMA FACING COMPONENTS AND EVALUATION TESTS OF B4C-OVERLAID CFC/GRAPHITES

Erosion of carbon fiber composite divertor tiles of JT-60U has been reduced significantly by the precise alignment and insitu taper-shaping of tile edges. The divertor tiles are coated with redeposited carbon films. None of graphite first wall tiles has been broken. Evaluation tests of B4C-converted and -coated CFC/graphite have been performed from viewpoints of high heat fluxdurability, thermal shock, deuterium retention, and erosion yields. JT-60U in-pile test has also been carried out. The results exhibit satisfactory performance for the divertor plate and first wall of JT-60U.