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Applying multiplicity spectrometry and using 238 U as the reference sample, the coincidence multiplicity spectra and the radiative capture cross-sections of 232 Th were measured for the energy range 21.5–215 eV. The gamma rays originating from neutron capture were detected by 16 sections of liquid scintillator, where the coincidence multiplicity ranged from 1 to 16. It seems that the γ-multiplicity spectra of the s resonances of 232 Th in the above energy region are almost identical. The radiative capture cross-sections were determined for each s resonance and for the three energy groups 21.5–46.5, 46.5–100 and 100–215 eV. Our group average cross-sections agree well with the values calculated with the resonance parameters given in the evaluated data libraries ENDF/B-VI, JENDL-3 and BROND-2.

Electrostatic biasing experiments using the Advanced Limiter Test (ALT-II) pump limiter in the TEXTOR tokamak have been carried out with the dual goals of: (a) improving the core plasma confinement in the tokamak and (b) enhancing the performance of the pump limiter. The fully toroidal belt limiter has been biased during both ohmic and neutral beam heated discharges. Both polarities of bias have been applied up to a maximum of +or-500 V with no evidence of impurity accumulation in the central plasma, although applying either polarity of bias to the limiter increases recycling from both the limiter face and the vacuum vessel liner. This in turn results in an increase of the central density. The application of a negative bias to the limiter produces a barrier to radial particle transport in the region between the limiter and the wall. This barrier is not observed in either the no bias or the positive bias case. Neither polarity of limiter bias affects the central plasma energy confinement, apparently because the electric field structure producing the radial barrier is outside the limiter tangency radius. The enhanced recycling, coupled with high edge density, increases the radiated power from the plasma edge and may lower the power flux to the plasma facing surface of the limiter blade. In the case of positive limiter biasing, the pressure in the pumped plasma collection scoops of the limiter increases by approximately 20%, corresponding to a similar increase in the particle removal rare of the pump limiter. The increase in the particle removal rate appears to result from a lower edge electron temperature. This is consistent with the observation of an increase in edge radiated power

Thermographic measurements at the ALT-II pump limiter on TEXTOR have been performed with the following results. (1) The design of the toroidal pump limiter provides a rather uniform temperature distribution and prevents an overheating of the graphite cover tiles. (2) The local measurements of the convective energy flux to the limiter (IR scanner) and of the radiated power (bolometer) show normally a good power balance; this indicates that the power is leaving the plasma toroidally symmetrically. (3) From the analysis of the temperature distribution in poloidal and toroidal direction along the ALT-II tiles a power decay length of 5–7 mm is derived. (4) The ion temperature at the last closed flux surface is obtained from the energy flux along the magnetic field lines. At low plasma densities, the ion temperature is up to a factor of 2 higher than the electron temperature and at high densities both temperatures converge to the same value.

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.

Rapidly changing heat fluxes deposited on the limiter blades are observed during disruptions by infrared (IR) scanners. These scanners are a suitable tool for the analysis of these heat fluxes because they provide both spatial and temporal information with sufficient resolution. Several new features of the power flux to the plasma facing surfaces, during a disruption have been found. The disruptive heat flux occurs on three different time-scales. The fastest ones are for heat bursts with a duration of ≤ 0.1 ms; several of these bursts form a thermal quench of about one millisecond duration, and some of these thermal quenches are found to occur during the current decay phase. Power flux densities of the order of 50 MW/m2 have been observed during a burst. The spatial extent of the area on which this power is deposited during a burst is larger than or equal to the size of half an ALT-II blade, i.e. about 1 m in the toroidal direction. Simultaneous measurements with two cameras show that the correlation length of a single burst is smaller than half the toroidal circumference, probably of the order of half a blade or a full blade length. This is consistent with plasma islands of low mode number. The typical heat deposition patterns at the limiter blades for normal discharges are preserved during a disruption. The magnetic structure near the plasma surface can therefore not be destroyed completely during the thermal quench. The power flux follows the field lines. However, the power e-folding length is about a factor of two to three times larger than under normal discharge conditions