The Large Helical Device (LHD) now under construction is a heliotron/torsatron device with a closed divertor system. The edge LHD magnetic structure has been studied in detail. A peculiar feature of the configuration is the existence of edge surface layers, a complicated three dimensional magnetic structure which does not, however, seem to hamper the expected divertor functions. Two divertor operational modes are being considered for the LHD experimenthigh density, cold radiative divertor operation as a safe heat removal scheme and high temperature divertor plasma operation. In the latter operation, a divertor plasma with a temperature of a few keV, generated by efficient pumping, is expected to lead to a significant improvement in core plasma confinement. Conceptual designs of the LHD divertor components are under way.
The first encouraging experiments demonstrating direct, explicit control of the He 2+ density in a tokamak plasma have been performed in the TEXTOR tokamak with the Advanced Limiter Test-II pump limiter. Helium is injected in a short gas puff from the outside of the plasma, is observed to reach the plasma core, and then is readily removed from the plasma. An exhaust efficiency of -8% is obtained. Active charge-exchange spectroscopy is used to study the exhaust and transport of He 2 * within the plasma, and the density evolution is modeled with a diffusive-convective transport code.PACS numbers: 52.25.Fi, 34.70.+e, 52.55.Fa, 52.70.Kz In future burning fusion devices, helium (He) ash must be continuously removed from the core to prevent dilution of the deuterium-tritium (D-T) fuel and concomitant quenching of the burn. Thus, He ash removal is fundamental to the operation of any fusion reactor, since the rate at which the a-particle by-products of the fusion reaction are purged from the core plasma will determine the pulse length available before the burn is quenched. For proposed steady-state tokamaks, such as the International Thermonuclear Experimental Reactor (ITER), continuous purging of the He ash is essential. Recent estimates 1 show that newly created He ions must be removed within 7 to 15 energy confinement times to maintain continuous reactor operation.The recycling of injected He from the wall and the lack of direct diagnostic capability to measure He concentrations have complicated previous efforts to determine the He removal rate. 2 " 4 Estimating this rate requires knowledge of the recycling coefficient, which can only be determined indirectly and depends on the wall conditioning status and history of the device. The experiments reported here are the first in which direct, explicit removal of injected He has been demonstrated. This is made possible by the Advanced Limiter Test-II (ALT-II) system, a toroidal belt limiter 5-7 that uses turbomolecular pumps (TMPs). Most existing particle exhaust schemes use gettering materials or cryopumping systems, which do not pump He. We simulate the presence of recycled He ash in a tokamak by puffing concentrations of 3%-5% He (relative to n e ) into the TEXTOR plasma just before or during neutral-beam injection (NBI). The transport of the He into the plasma core and its subsequent pump-out phase using the ALT-II system are followed with spectroscopic techniques by observing the He in three locations: the plasma core, the plasma edge at the ALT-II limiter, and the ALT-II pumping duct. By combining the results from these measurements, the exhaust efficiency for the He found in the plasma core is obtained during ALT-II pumping.In the plasma core, the He density is measured by charge-exchange excitation (CXE) spectroscopy, in combination with NBI. Measuring spatially and temporally resolved ion temperatures and absolute densities using CXE line intensities is a well-established technique on many tokamaks. 8 " 11 We use CXE spectroscopy to obtain the local He 2+ densi...
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