N. Miya

Tritium retention of plasma facing components in tokamaks

Abstract The areal distribution of tritium retention in tiles from TEXTOR, TFTR, JT-60U and JET has been measured via the imaging plate technique and the results are discussed from the perspective of carbon–hydrogen chemistry. It is found that the observed tritium distribution clearly shows asymmetries in poloidal and toroidal directions and also reflects the local temperature history of the analyzed tiles. We show the first clear evidence of the loss of high energy tritons by toroidal magnetic field ripple. We distinguish three different contributions to tritium retention in tokamaks with carbon plasma facing components: high energy tritons escaping from the core plasma, low energy ions and neutrals from the edge plasma, and molecular tritium from gas fueling. These components are retained at different depths and with different concentrations. Tritium from the edge plasma dominates the retained inventory but could be reduced if the surface temperature was higher. We propose tokamak operation with plasma facing components above 1000 K as a possible way to reduce the tritium inventory.

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.

Energy confinement of beam-heated divertor and limiter discharges in Doublet III

Observation of the intensity of the recycling particle flux at the main plasma edge for various limiter and divertor discharges indicates that the gross energy confinement of beam-heated discharges is closely related to the intensity of the edge particle flux. In limiter discharges, the global particle confinement time and the energy confinement time τE show many similarities: 1) linear Ip dependence at Ip < 600 kA, 2) no BT dependence, and 3) deterioration against injection power. Improvement of τE by increasing Ip, for example, is associated with high temperatures at the plasma edge region accompanied by reduced particle recycling. – Divertor discharges with low particle recycling around the main plasma show better energy confinement than limiter discharges at high plasma densities. The improvement of τE is primarily originated in the reduction of heat transport at the main plasma edge region, which is associated with the reduction of recycling particle flux at the main plasma edge. Under certain operation condition, for example, excessive cold-gas puffing, the discharge shows relatively high scrape-off plasma density and strong particle recycling between the main plasma and the limiter. The energy confinement time of these discharges degrades somewhat or reduces completely to that of the limiter discharge. – In low-recycling divertor discharges, the central electron and ion temperature is proportional to the injection power, and the plasma stored energy is proportional to ePabs (scales as INTOR scaling). With ≈ 4 MW beam injection, high-temperature and high-density plasmas were obtained (stored energy up to 280 kJ, Te(0) ≈ Ti(0) ≈ 2.5–3.0 keV at e ≈ (6–7) × 1013 cm−3, τE* ≈ 70 ms).

JT-60 power supplies

The paper gives a description of the electric power supplies of the JT-60 tokamak system, required to energize the magnetic field coils for plasma excitation and confinement.

Objectives and design of the JT-60 superconducting tokamak

A fully superconducting tokamak named JT-60SC is designed for the modification programme of JT-60 to enhance economical and environmental attractiveness in tokamak fusion reactors. JT-60SC aims at realizing high-β steady-state operation in the use of low radio-activation ferritic steel in a low ν* and ρ* regime relevant to the reactor plasmas. Objectives, research issues, plasma control schemes and a conceptual design for JT-60SC are presented.

Observation of very dense and cold divertor plasma in the beam-heated doublet III Tokamak with single-null poloidal divertor

A Langmuir probe array in the divertor plate has been used to investigate the dense, cold divertor plasma associated with remote radiative cooling in neutral-beam-heated, single-null open-divertor discharges in Doublet-Ill. With injected powers of up to 1.2 MW, the divertor plasma becomes denser and colder as the main plasma line-averaged density e increases, reaching ned= 2.8 X 1014 cm−3. Since the electron temperature drops to Ted = 3.5 eV under these conditions, this cold, dense plasma can provide a solution to the problem of wall erosion.

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.

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.

H-mode experiments with outer and lower divertors in JT-60

In JT-60, H-mode experiments with outer and lower divertors have been performed. In the outer divertor discharge, an H-mode similar to the modes observed in the lower/upper divertor discharges is obtained. Its threshold absorbed power and electron density are 16 MW and 1.8 × 1019 m−3. In the two schemes of combined heating with NB + ICRF and NB + LHRF, H-mode discharges are also obtained. Moreover, in the new configuration with the JT-60 lower divertor, H-mode phases with and without edge localized modes are obtained. The improvement in the energy confinement time in both divertor configurations is limited to values within 10%. The paper mainly presents the H-mode results of the outer divertor discharges. Also, typical results of the lower divertor discharges are shown for a comparison of the H-mode characteristics of the two configurations.