K. Masaki

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.

Role of divertor geometry on detachment and core plasma performance in JT60U

Experimental results related to the divertor geometry such as divertor plasma detachment, neutral transport and plasma energy confinement, were compared in the open and W-shaped divertors. The ion flux near the outer strike point was larger than in the open divertor, and the electron temperature at the target, T e div , was reduced. Divertor detachment and x-point MARFEs occurred at n e 10-20% lower than that for the open divertor. Although the leakage of neutrals from the divertor to the main chamber decreased, a neutral source in the main chamber due to an interaction of the outer scrape-off layer (SOL) plasma to the baffle plates became dominant above the baffle. Degradation in the enhancement factor of the energy confinement was observed similarly in the open and W-shaped divertors. The neutral density inside the separatrix was estimated to be a factor of 2-3 smaller, which did not affect the energy confinement.

Deuterium retention of DIII-D DiMES sample

Abstract The deuterium retention property of B4C converted graphite and isotropic graphite exposed to DIII-D deuterium plasma was examined by using a technique of thermal desorption spectroscopy. Major outgassing species were HD, D2 and CD4 in both the graphite and the B4C. In the case of the graphite, the ratios of deuterium desorbed in the forms of HD, D2 and CD4 to the total desorption amount of deuterium were 40%, 27% and 33%, respectively. In the case of the B4C, which was covered by carbon due to redeposition, these ratios were similar to those of the graphite. In a thermal desorption spectrum of deuterium, three desorption peaks appeared in both the graphite and the B4C covered by the redeposition layer. At low temperature region, the desorption rate of deuterium for the B4C covered by the redeposition layer was larger than that of the graphite. From two dimensional distribution of deuterium retention, it was seen that the retained amount at the electron drift side was quite large. The amount at the ion drift side and the edge of inward major radius was also observed to be large. The average retained amount of the graphite was almost the same as that of the B4C covered by the redeposition layer.

Imaging plate technique for determination of tritium distribution on graphite tiles of JT-60U

Abstract The tritium imaging plate technique was applied to determine surface tritium distributions on graphite tiles used as the first wall and W-shaped divertor in JT-60U, in which tritium produced by the D–D nuclear reaction in the plasma was implanted and/or deposited depending on the incident energy. Measured samples were isotropic graphite (IG-430U) and CFC graphite (CX-2002U), used as divertor tiles and/or baffle plates just outside the divertor. Tritium areal distributions on graphite divertor tiles, dome units and baffle plates of JT-60U were successfully measured for the first time. Tritium distributions observed in JT-60U tiles can be explained by homogeneous implantation of high energy tritium which is influenced by redeposited layers and redistributed by the temperature increase due to the plasma heat load. The tritium retention in graphite heated above 800 K was significantly small.

Hydrogen isotope behavior in in-vessel components used for DD plasma operation of JT-60U by SIMS and XPS technique

Abstract The behavior of hydrogen and deuterium retained in the graphite tiles placed in the dome top unit, the outer divertor unit and the outer baffle plate of JT-60U with W-shaped divertor were investigated by secondary ion mass spectroscopy and X-ray photoelectron spectroscopy (XPS). It was found that deuterium on the surface of the tiles was replaced by hydrogen due to hydrogen discharges at the final stage. The amount of deuterium on the dome top unit largely depends on the location of the dome unit. Especially, the amount of that on the inner side and the center of the dome top units were high compared with that on the outer side. The binding energies of Cxa01s and Bxa01s peaks on the surface of the inner side of dome top unit were largely shifted to higher energy by XPS. These results indicate that hydrocarbons and boron oxide might be formed on the surface. It is concluded that the behavior of hydrogen and deuterium and the chemical state of graphite were influenced by the temperature of the surface, the position of the strike point and the flux of the particles during the operation.

Ablative removal of codeposits on JT-60 carbon tiles by an excimer laser

The codeposits on JT-60 tiles experienced hydrogen plasma burning were irradiated by focused beams of an excimer laser. The removal rate of the JT-60 codeposits was low when the laser energy density was smaller than the ablation threshold (1.0 J/cm 2 ), but reached to 1.1 μm/pulse at the laser energy density of 7.6 J/cm 2 . The effective absorption coefficient k in the JT-60 codeposits at ArF excimer laser wavelength was determined to be 1.9 μm -1 , which is almost one order smaller than the optical absorption coefficient at the same wavelength in graphite (16.4 μm -1 ). In the process of ablative removal of the codeposits, hydrogen was released predominantly in the form of hydrogen molecule and water formation could be ruled out. The temperature rise on the surface was measured on the basis of Plancks law of radiation, and the temperature during the irradiation at the laser energy density of 0.5 J/cm 2 decreased from 3570 K at the beginning of the irradiation to 2550 K at 1000th pulse of the irradiation.

Tritium retention in graphite inner wall of JT-60U

Abstract In JT-60U, D-D experiments have been performed since July 1991. Although tritium, which is produced by the D-D reaction, is programmed to be exhausted through pumping systems, a part of the tritium is trapped in the graphite inner wall. To evaluate the tritium retention in the graphite wall of JT-60U, several sample tiles were removed from JT-60U in November 1993. The retained tritium distribution in the graphite tiles was measured by means of burning sample pieces, converting the tritium into HTO and scintillation counting the resultant water. The result of this measurement showed that tritium of 1.6 × 1010 Bq still remained in the graphite first wall. This value corresponds to approximately 50% of the tritium generated from July 1991 to October 1993 in JT-60U. The tritium remaining in the divertor tiles was in higher concentration than that in the first wall. In particular, the tritium concentration peaked in the tiles between the strike points. Tritium at the strike points, where an electron or ion strikes directly, was in low concentration.

First wall issues related with energetic particle deposition in a tokamak fusion power reactor

Abstract Energetic particle deposition to the wall due to toroidal magnetic field (TF) ripple was assessed for a 2 GW fusion power reactor. When the present allowance for the loss is applied, the alpha particle flux to the wall can be as high as 2×10 18 m −2 s −1 in the reactor, eroding tungsten by ∼20 μm per year. The peak particle fluence over a 2-year operation cycle can reach 10 26 m −2 , probably being larger than a critical fluence for blister formation. The result suggests that, for the steady-state tokamak fusion reactor, we should introduce a new design methodology of determining an acceptable level of TF ripple on the basis of particle fluence to the wall, instead of the present one based on a tolerable heat flux.

High heat load test of CFC divertor target plate with screw tube for JT-60 superconducting modification

Abstract A flat carbon fiber composite (CFC) tile mock-up with screw tubes, which have helical fins like a nut, was fabricated aiming at further improvement of the heat removal performance of the cost-effectively manufactured divert or target for JT-60SC (modified JT-60 as a superconducting coil tokamak). The heat removal performance of the mock-up was successfully demonstrated on the JAERI Electron Beam Irradiation Stand. The estimated heat transfer coefficient of the screw tube at the non-boiling region was roughly three times higher than that of the smooth tube. This corresponds to 1.5 times that of the swirl tube. A heat cycle test of 10 MW/m2 showed that the mock-up with the screw tubes could withstand for 1400-cycles. These results indicate that the divertor target plate with the flat CFC tile and the screw tube can be a promising candidate for the JT-60SC divertor target.