C. C. Klepper

Divertor heat flux reduction by D2 injection in DIII-D

D{sub 2} gas injected into ELMing H-mode discharges in DIII-D reduced total integrated heat flux to the divertor by {approximately}2{times} and peak heat flux by {approximately}5{times}, with only modest degradation to plasma stored energy. Steady gas injection without particle pumping results in eventual degradation in stored energy. The initial reduction in peak heat flux at the divertor tiles may be primarily due to the increase in radiated power from the X-point/divertor region. The eventual formation of a high density region near the X-point appears to play a role in momentum (and energy) transfer from the flux surfaces near the outboard strike point to flux surfaces farther out into the scrapeoff. This may also contribute to further reduction in peak heat flux.

Spectroscopic characterization of the DIII-D divertor

Radiative losses along a fixed view into the divertor chamber of the DIII-D tokamak [Plasma Physics Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol I, p. 159] have been characterized for attached and partially detached discharges by analyzing line-integrated vacuum ultraviolet (VUV) signals. Essentially all the emission can be ascribed to carbon and deuterium. Because the majority of the most intense lines, which lie at wavelengths above 1100 A, are not accessible to the present instrumentation, extensive use has been made of collisional-radiative (CR) calculations for level populations of the important ions in order to relate the total radiated power to shorter wavelength transitions. In beam-heated plasmas, the fraction of radiation detected from carbon along the VUV spectrometer view is usually between 50% and 80% of the total. Carbon densities are estimated from a simplified approach to modelling the emission using a one-dimensional transport code. For p...

Control of wall particle inventory with divertor pumping on DIII-D

Reduction of the net wall particle inventory has been achieved in the DIII-D tokamak by the use of a divertor cryopump for the particle exhaust. A sequence of 12 discharges was conducted without helium glow discharge cleaning (HeGDC) and without active cryopump exhaust, resulting in a net wall loading of 1250 torr.L (8.8*1022 atoms) by the end of the sequence. The cryopump was activated for the subsequent set of 10 discharges. At the end of this set of discharges, the net wall loading was reduced to its initial value or less, i.e. the wall loading state for the reference discharges with HeGDC. Plasma performance did not suffer without the HeGDC in that the stored energy during the ELMy phase was restored to the reference discharge level during the discharges with the active cryopump. Hence, with continuous particle exhaust (provided, for example, by a pumped divertor), next generation, long pulse devices, such as ITER, will not require interdischarge HeGDC for particle control, which would require turning off superconducting coils between discharges

Deuterium-tritium concentration measurements in the divertor of a tokamak via a modified Penning gauge

Abstract The measurement of the relative concentrations of hydrogen, deuterium, tritium and helium is an important task in the nuclear fusion research area. Control of the deuterium-tritium (D-T) isotropic ratio and limiting the helium ash content in a fusion plasma are the key to optimizing the fuel burn in a fusion reactor such as ITER. A diagnostic technique has been developed to measure the D-T isotopic ratio in the divertor of a tokamak with a Penning vacuum gauge. The Penning discharge provides a source of electrons to excite the neutral deuterium and tritium in the pumping duct. Subsequently, the visible light from the hydrogen isotopes is collected in an optical fibre bundle, transferred away from the tokamak into a low radiation background area and detected in a high resolution Czerny-Turner spectrometer, equipped with a fast CCD (charge-coupled device) camera for optical detection. The intensity of the observed line emission (D α , 6561.03 A; T α , 6560.44 A) is directly proportional to the partial pressure of each gas found in the divertor. The line intensity of each isotope is calibrated as a function of pressure. The ratio of the line intensities thus provides a direct measurement of the D-T isotopic ratio. The lower limit for the determination of the D-T isotopic ratio is about 0.5%. This system is applicable for the pressure range from 10 −5 mbar to a few times 10 −2 mbar.

Experiments on steady state particle control in Tore Supra and DIII-D

Particle control is playing an increasingly important role in tokamak plasma performance. The present paper discusses particle control of hydrogen/deuterium by wall pumping on graphite or carbonized surfaces, as well as by external exhaust with pumped limiters and pumped divertors. Wall pumping is ultimately a transient effect and by itself not suitable for steady state particle exhaust. Therefore, external exhaust techniques with pumped divertors and limiters are being developed. How wall pumping phenomena interact and correlate with these inherently steady state, external exhaust techniques, is not well known to date. In the present paper, the processes involved in wall pumping and in external pumping are investigated in an attempt to evaluate the effect of external exhaust on wall pumping. Some of the key elements of this analysis are: (1) charge-exchange fluxes to the wall play a crucial role in the core-wall particle dynamics, (2) the recycling fluxes of thermal molecules have a high probability of ionization in the scrape-off layer, (3) thermal particles originating from the wall, which are ionized within the scrape-off layer, can be directly exhausted, thus providing a direct path between wall and exhaust which can be used to control the wall inventory. This way, the wall can be kept in a continuous pumping state in the sense that it continuously absorbs energetic particles and releases thermal molecules which are then removed by the external exhaust mechanism. While most of the ingredients of this analysis have been observed individually before, the present evaluation is an attempt to correlate effects of wall recycling and external exhaust.

Application of a species-selective Penning gauge to the measurement of neon and hydrogen-isotope partial pressures in the plasma boundary

A species-selective Penning gauge, previously applied to He partial pressures, has been applied to the detection of small concentrations of Ne in a D{sub 2} gas. This is important for the study of Ne in the boundary region of magnetic fusion devices, where this impurity is deliberately injected to enhance the radiated power in that region. The application of the technique to the detection of the partial pressure of a minority hydrogen isotope is also examined. In this latter application, the detection system and the data analysis are more complex, because of the proximity of the spectral lines from the isotopes. In both applications, it is found that use of a proper detection scheme permits reliable measurements of concentrations as low as 0.5% of the minority neutral species, without requiring changes to the standard commercial Penning gauge setup.

On the challenge of plasma heating with the JET metallic wall

The major aspects linked to the use of the JET auxiliary heating systems: NBI, ICRF and LHCD, in the new JET ITER-like wall are presented. We show that although there were issues related to the operation of each system, efficient and safe plasma heating was obtained with room for higher power. For the NBI up to 25.7 MW was safely injected; issues that had to be tackled were mainly the beam shine-through and beam re-ionization before its entrance into the plasma. For the ICRF system, 5 MW were coupled in L-mode and 4 MW in H-mode; the main areas of concern were RF sheaths related heat loads and impurities production. For the LH, 2.5 MW were delivered without problems; arcing and generation of fast electron beams in front of the launcher that can lead to high heat loads were the keys issues. For each system, an overview will be given of: the main modifications implemented for safe use, their compatibility with the new metallic wall, the differences in behaviour compared with the previous carbon wall, with emphasis on heat loads and impurity content in the plasma.

Effect of low density H-mode operation on edge and divertor plasma parameters

We present a study of the impact of H-mode operation at low density on divertor plasma parameters on the DIII-D tokamak. The line-average density in H-mode was scanned by variation of the particle exhaust rate, using the recently installed divertor cryo-condensation pump. The maximum decrease (50%) in line-average electron density was accompanied by a factor of 2 increase in the edge electron temperature, and 10% and 20% reductions in the measured core and divertor radiated power, respectively. The measured total power to the inboard divertor target increased by a factor of 3, with the major contribution coming from a factor of 5 increase in the peak heat flux very close to the inner strike point. The measured increase in power at the inboard divertor target was approximately equal to the measured decrease in core and divertor radiation.

Helium exhaust studies with the ALT-II pump limiter in TEXTOR

Abstract In TEXTOR helium removal experiments with the pump-limiter ALT-II have started. To simulate the presence of helium ash in the plasma, helium is injected into the discharge (e.g. at t = 0.7s) as a short pulse of Δt = 20ms. It is found that the He is removed from the discharge in an e-folding time of about half a second for neutral beam heated plasmas and in an e-folding time of about 1.5 s in an OH plasma. The exhaust efficiency of helium amounts to about 8% and is close to the one for deuterium. The fuelling efficiency for the injected helium is found to be in the range of 50–100%; the remaining part seems to be stored in the TEXTOR walls. An estimate of the surface density leads to a value of several times 1013 cm−2. This helium can easily be liberated in succeeding discharges and can be removed efficiently when ALT is pumping.

Helium transport in enhanced confinement regimes on the TEXTOR and DIII-D tokamaks

Comparisons of helium (He) transport and exhaust in L-mode and in an enhanced confinement regime (H-mode), which is induced by a polarizing electrode, have been made for the TEXTOR tokamak. The results show an increased tendency for He accumulation when bulk plasma energy and particle confinement are improved during the polarization induced H-mode. Since these results imply that a high He pumping efficiency may be necessary for H-mode burning plasmas, we have begun exploring He transport in a divertor H-mode, similar to that proposed for International Thermonuclear Experimental Reactor (ITER). A collaborative program has been initiated to measure He transport and scaling on DIII-D during L-mode, H-mode, and ELMing H-mode plasma conditions. To simulate the presence of He ash in DIII-D, a 25 ms He puff is injected into a DIII-D plasma resulting in a He concentration of {approx}5%. The time dependence of the He{sup 2+} density profiles in the plasma core is measured by charge-exchange recombination spectroscopy at 11 radial locations.