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Zechstein (Upper Permian) dolomitic grainstones saturated with high pressure natural gas (including N2, CH4 and CO2) are present in the roof of workings in the Rudna copper mine in SW Poland. Natural gas commonly occurs in Zechstein reefal carbonates but this is the first occurrence of gas in basinal facies of the Zechstein Limestone in Poland. The reservoir dolomites are thin (ca. 1m), with porosity of 2-15% and permeability of less than a few mD. Pore size and pore distribution was analysed using microscope observations, X-ray computed microtomography, mercury porosimetry and argon adsorption to investigate the dolomitic reservoir interval. The dolomites have variable reservoir properties, and stratigraphic traps with dimensions of several metres are present where high poro-perm dolo-grainstones are sealed by tighter dolomites. Geochemical data indicate that the methane and carbon dioxide present are derived from source rock intervals in the Kupferschiefer (Late Permian) and the Zechstein Limestone. The high content of nitrogen in the gas may be related to the thermal decomposition of Carboniferous organic matter. However, a biogenic origin for the nitrogen is also possible. [ABSTRACT FROM AUTHOR]

The spherical tokamak (ST) is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the US actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3-6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1-1.5 s to 5-8 s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up as needed for an ST-based FNSF. In boundary physics, NSTX measures an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavourably impact next-step devices. Recently, NSTX has successfully demonstrated substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade is described. [ABSTRACT FROM AUTHOR]

TFTR long pulse ion sources have been operated with gas fed only into the neutralizer. Gas for the plasma generator entered through the accelerator rather than directly into the arc chamber. This modification has been proposed for tritium beam operation to locate control electronics at ground potential and to simplify tritium plumbing. Source operation with this configuration and with the nominal gas system that feeds gas into both the ion source and the center of the neutralizer are compared. Comparison is based upon accelerator grid currents, beam composition, and neutral power delivered to the calorimeter. Charge exchange in the accelerator can be a significant loss mechanism in both systems at high throughput. A suitable operating point with the proposed system was found that requires 30% less gas than used presently. The extracted D+, D+2, and D+3 fractions of the beam were found to be a function of the gas throughput; at similar throughputs, the two gas feed systems produced similar extracted ion fractions. Operation at the proposed gas efficient point results in a small reduction (relative to the old high throughput mode) in the extracted D+ fraction of the beam from 77% to 71%, with concomitant changes in the D+2 fraction from 18% to 26%, and 6% to 3% for D+3. The injected power is unchanged, ∼2.2 MW at 95 kV. [ABSTRACT FROM AUTHOR]

Energy flow within TFTR neutral beamlines is measured with a waterflow calorimetry system capable of simultaneously measuring the energy deposited within four heating beamlines (three ion sources each), or of measuring the energy deposited in a separate neutral beam test stand. Of the energy extracted from the ion source on the well-instrumented test stand, 99.5±3.5% can be accounted for. When the ion deflection magnet is energized, however, 6.5% of the extracted energy is lost. This loss is attributed to a spray of devious particles onto unmonitored surfaces. A 30% discrepancy is also observed between energy measurements on the internal beamline calorimeter and energy measurements on a calorimeter located in the test stand target chamber. Particle reflection from the flat plate calorimeter in the target chamber, which the incident beam strikes at a near-grazing angle of 12°, is the primary loss of this energy. A slight improvement in energy accountability is observed as the beam pulse length is increased. This improvement is attributed to systematic error in the sensitivity of the energy measurement to small fluctuations in the supply water temperature. An overall accuracy of 15% is estimated for the total power injected into TFTR. Contributions to this error are uncertainties in the beam neutralization efficiency, reionization and beam scrape-off in the drift duct, and fluctuations in the temperature of the supply water. [ABSTRACT FROM AUTHOR]

An array of 64 silicon surface-barrier diodes on a circular arc view Tokamak Fusion Test Reactor (TFTR) plasmas through a slot aperture to provide poloidal imaging of x-ray emission in the 200 eV-15 keV range. Information is inferred on magnetohydrodynamic (MHD) instabilities, disruptions, radiation, impurity transport, electron temperature., and electron thermal conductivity. Spatial resolution is 2.5 cm. Movable absorber foil arrays provide energy selection. Preamplifier-amplifier pairs have gains of 0.05-100 V/µA. Two outputs are provided with (1) 40-, 80-, and 300-Hz and (2) 40-, 80-, and 600-kHz filtering. The signals are digitized at rates up to 500 kHz and stored in 128 K (total system) memory. Foils, gains, and filters are selectable from the control room by a computer. [ABSTRACT FROM AUTHOR]

The x-ray pulse-height analysis (PHA) diagnostic uses a liquid-nitrogen cooled array of 5 Si(Li) and one HpGe detectors to do time-resolved (5-100 ms) x-ray spectroscopy of the central horizontal chord of the Tokamak Fusion Test Reactor (TFTR) plasmas in the 1-50-keV range. Central electron temperature T[sub e] and concentration of medium and low-Z impurities are derived from the spectra. Remotely selectable absorber-foil arrays provide selection of the energy range. Fixed and movable aperture arrays allow approximate equalization of count rates in different energy bands and extend dynamic range. Amplifier pulse shapes are approximately triangular. Main amplifier peaking time is 4 µs yielding 230-eV FWHM at 5.9 keV. Pileup inspection times are selectable at 0.13, 0.4, or 0.9 µs. Throughput is up to 42 kHz. The PHA has been used to study temperature and impurities over a wide range of TFTR operational parameters. Dramatic variations in metal impurities with density, plasma current, and major radius have been observed. [ABSTRACT FROM AUTHOR]