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The Japanese Government designated JAEA as the Domestic Agency of Japan (JADA) for the ITER Project just after the entry into force of the ITER Agreement on 24th Oct. 2007. In-kind procurements allotted to JADA amount to about 16% of the total according to the "Common Understandings on Procurement Allocation ". The first procurement arrangement for 25% of the TF superconductors was concluded in November 2007 between the ITER Organization (IO) and the JADA. After the bidding, venders for fabricating 15% of the TF conductors were selected in March 2008. The Quality Assurance Program of JADA, that is indispensable for implementing procurement, was approved by IO, and the Project Management of JADA subsequently got underway. Preparation for the procurement continues for the instruments and equipment that JA has the responsibility to procure. JADA also has performed tasks requested by IO on the PF conductor and the accelerator of the neutral beam injector system, etc.

Argon, krypton and xenon were puffed with and without simultaneous hydrogen gas puffing into Ohmically heated plasmas of the JT-60U tokamak with low plasma currents in order to study the capability of disruption mitigation. It was found that krypton gas puffing can provide a plasma termination with smaller amounts of runaway electrons in comparison to argon and xenon gas puffing.

Characteristics and underlying mechanisms for plasma current spikes, which have been frequently observed during the thermal quench of JT-60U disruptions, were investigated through tokamak simulation code simulations including the passive shell effects of the vacuum vessel. Positive shear and reversed shear (PS and RS) plasmas were shown to have various current spike features in the experiments, e.g. an impulsive increase in the plasma current (positive spike) in the majority of thermal quenches, and a sudden decrease (negative spike), that has been excluded from past consideration, as an exception. It was first clarified that the shell effects, which become significant especially at a strong pressure drop due to the thermal quench of high βp plasmas, play an important role in the current spike in accordance with the initial relation of the radial location between the plasma equilibria and the vacuum vessel. As a consequence, a negative current spike may appear at thermal quench when the plasma is positioned further out from the geometric centre of the vacuum vessel. It was also pointed out that a further lowering in the internal inductance, in contradiction to previous interpretation in the past, is a plausible candidate for the mechanism for positive current spikes observed even in RS plasmas. The new interpretation enables us to reason out the whole character of current spikes of JT-60U disruptions.

The time dependence of the current decay during the current quench phase of disruptions, which can significantly influence the electro-magnetic force on the in-vessel components due to the induced eddy currents, is investigated using data obtained in JT-60U experiments in order to derive a relevant physics guideline for the predictive simulations of disruptions in ITER. It is shown that an exponential decay can fit the time dependence of current quench for discharges with large quench rate (fast current quench). On the other hand, for discharges with smaller quench rate (slow current quench), a linear decay can fit the time dependence of current quench better than exponential.

Intensive studies on the physics of disruptions and developments of avoidance/mitigation methods of disruption-related phenomena have being carried out in JT-60U. The characteristics of the disruption sequence were well understood from the observation of the relationship between the heat pulse onto divertor plates during thermal quench and the impurity influx into the plasma, which determined the speed of the following current quench. A fast shutdown was first demonstrated by injecting impurity ice pellets to the plasma and intensively reducing the heat flux on first wall. The halo current and its toroidal asymmetry were precisely measured, and the halo current database was made for ITER in a wide parameter range. It was found that TPF x I{sub h}/I{sub p0} was 0.52 at the maximum in a large tokamak like the JT-60U, whereas the higher factor of 0.75 had been observed in medium-sized tokamaks such as Alcator C-Mod and ASDEX-Upgrade. The vertical displacement event (VDE) at the start of the current quench was carefully investigated, and the neutral point where the VDE hardly occurs was discovered. MHD simulations clarified the onset mechanisms of the VDE, in which the eddy current effect of the up-down asymmetric resistive shell was essential. The real-time Z{submore » j} measurement was improved for avoiding VDEs during slow current quench, and plasma-wall interaction was avoided by a well-optimized plasma equilibrium control. Magnetic fluctuations that were spontaneously generated at the disruption and/or enhanced by the externally applied helical field have been shown to avoid the generation of runaway electrons. Numerical analysis clarified an adequate rate of collisionless loss of runaway electrons in turbulent magnetic fields, which was consistent with the avoidance of runaway electron generation by magnetic fluctuations observed in JT-60U. Once generated, runaway electrons were suppressed when the safety factor at the plasma surface was reduced to 3 or 2.« less

Progress in the understanding of magnetohydrodynamic (MHD) stability is summarized on JT-60U tokamak discharges with improved confinement such as the (hot-ion) H-mode, high-βp mode, high-βp H-mode,...

Plasma current ramp-up assisted by outer vertical field coils with non-inductively driven current is studied during the ignition access phase in a high aspect ratio tokamak such as ITER-FEAT class tokamak reactor. In a tokamak without the Ohmic transformer, the pl1asma current ramp-up to 8 MA could be achieved by the vertical field with the heating/current drive power of 100 MW. The current ramp-up time can be chosen arbitrarily, as shorter than that expected in the usual, non-inductive current drive operation analyses, and it depends on the fusion power ramp-up time. This makes the reactor operation flexible. Experimental evidence showing the current drive by a vertical field in JT-60U has also been presented with plasma circuit theory.

Fast plasma shutdown without runaway electron generation by gas puffing is investigated in the JT-60U tokamak. Argon-only injection enables a fast shutdown; however, it induces runaway electron generation. Hydrogen-only injection generates much less runaway electrons; however, the shutdown time is considerably longer. Mixed injection of hydrogen and argon achieves a fast plasma shutdown without runaway generation. Argon atoms contribute to radiate the energy of plasma leading to a fast plasma shutdown, whilst hydrogen atoms contribute to increase the electron density for avoiding runaway electron generation.

The possibility of using massive quantities of injected deuterium with and without additional admixture of lesser quantities of higher-Z impurities to effect fast thermal shutdown of a tokamak discharge without causing runaway electron avalanching is systematically studied. It is found that various combinations of deuterium alone and deuterium with impurities (helium, nitrogen, argon, and krypton) can provide acceptable runaway-free thermal shutdown. The admixture of impurities cited reduces the quantity of deuterium needed and also reduces the radiative cooling time needed for the plasma to reach final thermal equilibrium, where radiation losses balance ohmic input. In contrast, the addition of neon does not appreciably reduce deuterium quantity or promote faster cooling. This difference relative to the other impurity studies can be understood from the radiation versus plasma temperature characteristics for neon.

The characteristics of internal transport barrier (ITB) structures are studied and active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of ITBs with steep Ti gradients is limited to the minimum safety factor location ρqmin. However, ITBs with reduced Ti gradients can move to the outside of ρqmin. The lower boundary of the ITB width is proportional to the ion poloidal gyroradius at the ITB centre. Furthermore, active control of the ITB strength based on modification of the radial electric field shear profile is successfully demonstrated by toroidal momentum injection in different directions or an increase of heating power by neutral beams.

An advanced diagnostic technique to measure the ion temperature and fast ions in open magnetic field systems is described. The method for the determination of deuterium to tritium ratio and the hel...

Simulations have demonstrated that magnetic islands having the widths expected in major disruptions cause collisionless loss of relativistic electrons and that the resultant loss rate is high enough to avoid or to suppress runaway generation. This is because, for the magnetic fluctuations in the disruptions, the loss of electron confinement due to the breakdown of toroidal momentum conservation overwhelms the runaway electron confinement due to the phase averaging effect of relativistic electrons. The simulation results provide strong support for the JT-60U experiments on fast plasma shutdown avoiding runaway generation.

A period-one unstable periodic orbit (UPO) is identified in the inter--edge-localized-mode (ELM) intervals of a high-performance quasisteady state discharge of JT-60U using UPO identification based on their recurrence properties and the so-called fixed point transform. This result is tested by use of surrogate data sets. This diagnostics invariant interpretation of the ELM confinement state gives detailed and applicable quantitative information since the UPOs in general form the skeleton of dynamical invariants and have in numerous systems been shown to be controllable by weak parameter perturbations.

The operation conditions to avoid runaway electron generation at the major disruption have been investigated in JT-60U tokamak plasmas. It has been found that runaway electrons are not observed for low Bt of ? 2.2?T or low plasma current quench rates (I? ? -(dIp/dt)/Ip) of

Tokamak startup characteristics with a low-loop-resistance vacuum vessel were studied in the HT-2, which is a tokamak with a major radius of 41 cm, a minor radius of 11 cm, and a plasma current of I p < 55 kA. Precise poloidal magnetic field control is possible using independently and multivariably controlled poloidal field coils. The vacuum vessel loop resistance Ω v was originally high (14 mΩ), but it was modified to be able to operate with a low value (0.3 mΩ). The latter is approximately one-tenth of the plasma resistance (2.2 mΩ) at breakdown (T e = 10 eV). With Ω V = 0.3 mΩ, the magnetic field induced by the eddy current is large, and it disturbs the breakdown. However, careful compensation of the poloidal field makes a well-controlled ohmic startup possible. Other results are as follows: very little difference in the consumed ohmically heated flux was observed between discharges with low and high loop resistances; well-controlled startup was obtained with a very low loop voltage of 2.5 V, which corresponded to the 1 V/m electric field; the breakdown condition is well described by the electron avalanche model. It is concluded that low Ω V is applicable to a tokamak design, as long as the poloidal magnetic field is well controlled even during the breakdown phase.

A review on the plasma position and current control system for the JT-60 upgrade is presented. In the control system, advanced digital technologies such as the versa module Europe bus, the computer automated measurement and control technology, and Ethernet are applied to fast equilibrium control of the vertically elongated divertor plasmas. The delay of the overall feedback loop with a sampling period of 0.25 ms is

Isolation of the core with improved confinement (inside the internal transport barrier) from high recycling and strongly radiative boundary plasmas was demonstrated for the first time in the reversed shear plasmas of the JT-60U tokamak. The internal transport barrier remained intact, while edge radiation and recycling was enhanced with neon and hydrogen gas puffing. The internal transport barrier reduced the particle transport by a factor 10{endash}15, while shrinking in the time scale of resistive diffusion towards the center. {copyright} {ital 1997} {ital The American Physical Society}

A killer pellet is an impurity pellet that is injected into a tokamak plasma in order to terminate a discharge without causing serious damage to the tokamak machine. In JT-60U neon ice pellets have been injected into OH and NB heated plasmas and fast plasma shutdowns have been demonstrated without large vertical displacement. The heat pulse on the divertor plate has been greatly reduced by killer pellet injection (KPI), but a low-power heat flux tail with a long time duration is observed. The total energy on the divertor plate increases with longer heat flux tail, so it has been reduced by shortening the tail. Runaway electron (RE) generation has been observed just after KPI and/or in the later phase of the plasma current quench. However, RE generation has been avoided when large magnetic perturbations are excited. These experimental results clearly show that KPI is a credible fast shutdown method avoiding large vertical displacement, reducing heat flux on the divertor plate, and avoiding (or minimizing) RE generation.

Plasma breakdown for an electric field of (loop voltage of 1.7 V) was demonstrated in JT-60U using helium prefilling gas and LHRF heating. Runaway electrons were not generated even for the low prefilling pressure of required for plasma breakdown at low loop voltage. Preionization was not obtained by LHRF heating, but the time taken to pass the radiation barrier was significantly shortened and a subsequent plasma-current ramp-up at a plasma current of < 30 kA was successfully obtained. As a result, a plasma-current start-up of with a loop voltage of was demonstrated when the plasma current was 35 kA to 1 MA.

A high toroidal eddy current induced in a vacuum vessel during plasma-current quench, Ip quench, results in errors in determining the vertical position of the plasma-current center, ZJ, calculated ...

Experimental analyses are presented of fluctuations observed in a plasma-current-center, I p -center, location and the inward clearance between the outermost flux surface and the first wall at plasma perturbations (e.g., β p collapse and disruption) for JT-60U divertor plasmas. The neutral point for vertical and horizontal movement of the I p center at plasma disturbances is confirmed, and attachment of the plasma to the inward first wall is avoided because of the increase in l i at fast plasma-current quench. These experimental observations will support the design of fusion reactors like the International Thermonuclear Experimental Reactor (ITER).

Extremely fast vertical displacement events (VDEs) induced by a strong collapse were found in a vertically elongated , high tokamak with a resistive shell through computer simulations using the tokamak simulation code. Although the plasma current quench which has been shown to be the prime cause of VDEs in a relatively low tokamak (Nakamura Y et al 1996 Nucl. Fusion 36 643), was not observed during the VDE evolution, the observed growth rate of VDEs was almost five times faster than the growth rate of the usual positional instability . The essential mechanism of the -collapse-induced VDE was clarified to be the intense enhancement of positional instability due to a large and sudden degradation of the magnetic field decay n-index in addition to the significant destabilization due to a reduction in the stability index . The radial shift of the magnetic axis caused by the collapse induces eddy currents on the resistive shell, and these eddy currents produce a large degradation of the n-index. It is pointed out that the shell geometry characterizes the VDE dynamics, and that the VDE rate depends strongly both on the magnitude of the collapse and the n-index of the equilibria just before the collapse occurs. The JT-60U vacuum vessel is shown to possess the capability of preventing -collapse-induced VDEs.

Vertical displacement events (VDEs), which are frequently observed in disruptive discharges of elongated tokamaks, are investigated using the Tokamak Simulation Code. We show that disruption events such as a sudden plasma pressure drop (β p collapse) and the subsequent plasma current quench (I p quench) can accelerate VDEs due to the adverse destabilizing effect of the resistive shell, which has previously been thought to stabilize VDEs. In a tokamak with a surrounding shell which is asymmetric with respect to the geometric midplane, the I p quench also causes an additional VDE acceleration due to the vertical imbalance of the attractive force. While the shell-geometry characterizes the VDE dynamics, the growth rate of VDEs depends strongly on the magnitude of the β p collapse, the speed of the I p quench and the n-index of the plasma equilibrium just before the disruption. An amelioration of I p quench-induced VDEs was experimentally established in the JT-60U tokamak by optimizing the vertical location of the plasma just prior to the disruption. The JT-60U vacuum vessel is shown to be suitable for preventing the β p collapse-induced VDE.

A poloidal halo current due to a vertical displacement event (VDE) is observed in experimentally simulated VDE discharges and density limit disruptions in the JT-60U tokamak. In the case of a clockwise I{sub p} and B{sub T} discharge, the halo current flows into the vacuum vessel from the inside separatrix and goes back to the plasma from the outside separatrix. A maximum halo current is produced by a change in the poloidal flux generated by plasma current decay. A toroidal asymmetry factor of 2.5 is estimated from the requirements of the fracture of the carbon-fiber composite tiles. The toroidal asymmetry is caused by the poloidal field (PF) that is produced by the toroidal field (TF) ripple, the deformation of the vacuum vessel, the setting error between the vacuum vessel and the TF and PF coils, the low-n mode during current quench, etc. To consider this asymmetry, in JT-60U, one must estimate the total halo current as nearly 26% of the plasma current just before a current quench. 25 refs., 10 figs.

Effects of the shallow pellet injection on the ELMy H-mode plasma have been investigated in JT-60U for single null open-divertor plasmas. The penetration depth of the pellet was 20–30% of the plasma minor radius from the outermost flux surface with NB heating power of 15–25 MW and the plasma density of 1–2 × 10 19 m −3 . (1) A particle stream was generated transiently by the shallow pellet injection from the main plasma to the divertor region without the degradation of the energy confinement, and was accelerated by the enhanced ELM activity. (2) The total power on to the divertor plate (mainly to the inboard divertor region) increased just after the pellet injection by ∼ 100% with ∼ 60% rise of the maximum heat flux density. (3) The screening of the helium influx by the shallow pellet injection has been confirmed from the delayed rise of the helium density in the main plasma, which has been obtained by the helium gas puffing just after the pellet injection.

Reduction of the plasma-surface interaction at the energy quench is essential for avoiding or softening the subsequent plasma-current quench. This reduction suppresses vertical instability and generation of runaway electrons during the plasma-current quench. Resultant slow plasma-current quench improves controllability of plasma movement and reduces halo currents. This amelioration of the major disruption will drastically relax the design requirements of tokamak fusion reactors.

Change-over experiments using hydrogen and helium were carried out after deuterium operations in JT-60U. We measured the behavior of charge-exchanged deuterium flux with a neutral particle energy (CX) analyzer and the amount of evacuated tritium with a tritium monitor system. By analyzing the time evolution and shot sequence of the deuterium flux, effective plasma operations for rapid depletion of hydrogen isotopes are obtained. The deuterium flux behavior in the range of days was correlated with the evacuated tritium.

The mechanical fracture of three carbon fiber composite (CFC) first wall tiles was observed. This damage was probably caused by the electromagnetic force due to halo current during disruption. The required current to break the CFC tile is estimated to be 25 kA. The broken tile was rotated poloidally around the plasma with a speed of about 10 m/s during the following discharge. A possible driving force of this rotation might be the electromagnetic force due to the scrape-off layer (SOL) current. The required current to rotate the piece of the broken tile is 1 kA. These results indicate that electromagnetic interaction between SOL plasma and the plasma facing component is important in the research on the plasma wall interactions in fusion devices.

A multivariable non-interacting control technique has been applied to the configuration control of single-null divertor plasmas in JT-60U. Strong interactions between the controls of the plasma current and the plasma horizontal position—which inevitably occur during their independent control, owing to large inductive coupling between the ohmic heating coil and vertical magnetic field coil—have been drastically suppressed by adopting the non-interacting control. The multivariable non-interacting control technique is a general method and can be applied to any tokamak machine, including tokamak fusion reactors such as ITER. Poloidal field coil currents and voltages required for configuration control in ITER-CDA have been investigated using this control algorithm.

By modification of current and heat deposition profiles, the integrated core plasma performance was improved in ELMy H-mode. In a quasi steady state, beta p approximately 2.5-3, beta N approximately 2.5-3.1 and H-factor approximately 1.8-2.1 were sustained simultaneously under the full current drive condition (bootstrap current 55%, beam driven current 45% at Ip=0.5MA). The achievable beta p value and the energy confinement performance increases systematically with safety factor at the edge qeff and a high qeff operation (qeff>6) is required to obtain favorable integrated performance. To obtain these improved high beta ELMy plasmas, it is essential to control MHD activities; suppression of beta p-collapse and control of ELM activity.

Spontaneous formation of internal transport barrier was observed in JT-60U high- beta p discharges, where q=3 surface is the most likely explanation for its radial location. Further improved confinement was brought about by the subsequent transition to the edge transport barrier.

Spontaneous formation of an internal transport barrier was observed associated with improved confinement in the high-${\mathrm{\ensuremath{\beta}}}_{\mathit{p}}$ discharges in the JT-60U tokamak. The radial location of the transport barrier was found to be on the q=3 surface. A fast magnetohydrodynamic event localized at the transport barrier triggered the subsequent formation of an edge transport barrier that resulted in the further confinement improvement. In these discharges, a high poloidal rotation velocity that significantly exceeded the prediction of the present neoclassical theory was also observed at r/a=0.8.

The toroidally uniform vessel current in the JT-60U tokamak is estimated by a magnetic fitting code for the plasma initiation and current ramp-up phases, the vertically unstable discharge and the rapid major disruption. The estimated vessel current agrees well with the measured current, and the poloidal vessel current distribution profile is consistent with the plasma displacements.

Divertor phenomena prior to density limit disruptions have been studied in ohmic and neutral beam heated discharges with power of 7.5–8.5 MW in JT-60U, The MARFE occurs when the outer divertor electron temperature falls down to 10 eV or less for both ohmic and neutral beam heated discharges. The MARFE forms in the vicinity of the X-point, and grows toward the inside of the main plasma accompanied with an increase in particle recycling. The divertor radiation loss during the MARFE stays at a saturation level as a whole though the radiating region changes. After the detachment, the MARFE moves up from the divertor region and repeats the up-down movement in a period of several milliseconds, leading to disruptions.

The effects of wall conditioning techniques used in JT-60U are described. Helium glow discharge cleaning is effective in reducing oxygen and particle recycling. Both carbon and oxygen impurities are reduced by a factor of 2 by reducing the wall temperature from 300 to 150°C. The dilution ratio H/(H+D) decreased by D 2 -TDC, ohmic discharges and He-GDC. At the initial phase of wall conditioning, the dilution ratio increased by He-TDC, and a sequence of tokamak discharges with neutral beam injection, suggesting that hydrogen particles were sputtered from inside the graphite wall. After sufficient wall conditioning had reduced the dilution ratio below 0.25, the dilution rationdid not deteriorate by neutral beam injection or by a sequence of shots.

This paper reports on the technique of ripple injection that has been proposed for refueling in tokamak reactors. The usefulness of ripple-assisted fueling has been investigated by using an orbit-following Monte Carlo code. The penetration depth strongly depends on the beam energy. The ripple-enhanced outward flow of ripple-detrapped fast ions is not a serious problem. If E{sub b}/T{sub e0} {le} 4 is chosen, the fuel efficiency becomes {gt}80%. There is an optimum toroidal angle of the injection beamline to enhance the penetration depth of fast ions, and the range of angles that are effective for fueling is rather wide. The loss of alpha particles incident to the fueling has also been investigated by using the same code. By regulating the shape of the ripple-well region, the total alpha-particle loss can be reduced to {lt} 5%. Ripple-assisted fueling in the International Thermonuclear Experimental Reactor (ITER) has also been investigated. Because of the small aspect ratio, the field ripple is strongly decayed in the plasma. Consequently, central fueling presents some difficulties in ITER. However, fueling near one-half of the plasma minor radius is possible with an {approximately} 6% alpha-particle power loss.

Temperature measurements and inverse thermal analyses of the heat flux on the divertor plates of JT-60 have been performed in both outside X-point and lower X-point operations. These investigations showed that heat load behavior during normal operation is quite different depending on parameters such as plasma density, the shape of the magnetic surface near the divertor plate, polarity of the plate (ion-side vs. electron-side) and the plate materials (TiC-coated Mo vs. graphite). Due to contracting magnetic fluxes near the divertor plate in the outside X-point operation, the peak heat flux on the outside divertor plate was higher than that on the lower divertor plate. Large heat concentration was produced at the edge of the plate due to a small gradient of the magnetic field line incident on the plate, which resulted in serious damage to the divertor plates. Heat deposition due to high energy runaway electrons produced during disruption was observed on the lower divertor plate. Based on these experimental results, heat load and design issues of the JT-60U divertor plate are discussed.

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.

Studies on plasma confinement, steady-state tokamak operations and boundary plasmas are main subjects in tokamaks relating to fusion plasmas. Emphases of recent JT-60 experiments from 1987 to the shutdown in October, 1989 were put on these studies. Investigations of L-mode improved modes such as pellet injection, high Ti mode, giant sawtooth by 3rd harmonics of ICRF etc., and H-mode were carried out as confinement studies. Current drive by LHCD made a large progress by using a new launcher. Bootstrap current reached 80% of plasma current, which has brought a new idea of steady-state tokamaks. Results on particle confinement, particle and heat control studies with divertor are also presented

Ammonia has been shown to function as a morphogen at multiple steps during the development of the cellular slime mold Dictyostelium discoideum ; however, it is largely unknown how intracellular ammonia levels are controlled. In the Dictyostelium genome, there are five genes that encode putative ammonium transporters: amtA , amtB , amtC , rhgA , and rhgB . Here, we show that AmtA regulates ammonia homeostasis during growth and development. We found that cells lacking amtA had increased levels of ammonia/ammonium, whereas their extracellular ammonia/ammonium levels were highly decreased. These results suggest that AmtA mediates the excretion of ammonium. In support of a role for AmtA in ammonia homeostasis, AmtA mRNA is expressed throughout the life cycle, and its expression level increases during development. Importantly, AmtA-mediated ammonia homeostasis is critical for many developmental processes. amtA − cells are more sensitive to NH 4 Cl than wild-type cells in inhibition of chemotaxis toward cyclic AMP and of formation of multicellular aggregates. Furthermore, even in the absence of exogenously added ammonia, we found that amtA − cells produced many small fruiting bodies and that the viability and germination of amtA − spores were dramatically compromised. Taken together, our data clearly demonstrate that AmtA regulates ammonia homeostasis and plays important roles in multiple developmental processes in Dictyostelium .

We isolated and characterized a gene from Dictyostelium discoideum , which encodes a protein of 279 amino acids (30.6 kDa) containing six transmembrane domains with two highly conserved motifs of asparagine–proline–alanine (NPA) found in the aquaporin family of water-channel proteins, although the second motif of the protein has been modified into NPV (asparagine–proline–valine). The deduced amino acid sequence of the gene, which we have named aqpA , is 39% identical to D. discoideum WacA, 26% identical to human Aqp5, 26% identical to Oryza sativa PIP2a, 25% identical to yeast Aqy1 and 24% identical to E.coli AqpZ. Southern analyses indicated that aqpA is present as a single copy in the genome. Northern blot analysis showed that the developmentally regulated 1 kb mRNA transcript first appears at the tight mound stage (12 h), and is abundant in fingers (16 h) and late culminants (20 h). In-situ hybridization of slugs revealed that aqpA mRNA accumulated in cells of the prespore region but not in those of the prestalk region. Disruption of aqpA by homologous recombination did not significantly affect growth or developmental morphogenesis. Although mutant spores were viable, when assayed soon after encapsulation, they became permeable to propidium iodide and lost viability after a week on the top of a fruiting body. Thus, AqpA is essential to maintain spore dormancy perhaps through the regulation of water flow.

Four hydrogen pellets, two of 3 mm■×3 mml pellets and two of 4 mmF/×4 mml pellets, with the velocity of ∼2.3 km/s, were injected to OH and NB heated plasmas by a four‐barrel pneumatic injector on the JT‐60 tokamak. The pellet ablation profile was estimated by the measurement of the Hα emissions originating from the ablation cloud using a fiber optics array combined with the Hα ‐filtered multichannel photodiode. The fiber optics array consisting of ten channel bundles views the pellet pass in the vacuum vessel. The optical fiber was developed to tolerate baking up to 300 °C. The measured ablation profiles were compared with the calculated one by the neutral gas shielding model with a self‐limiting effect. A good agreement was obtained between measured and calculated ablation profiles in OH plasmas; however, the fast‐ion‐induced ablation needed to be taken into account in the calculation for NB‐heated plasmas.

Behavior of a spindle-cusp plasma produced at its central null-field point from a thin wire target by laser pulse is experimentally studied, mainly in a low plasma-beta regime, by means of many different plasma diagnostics. As the results, somewhat queer confinement properties have been found, and some considerations are given for the observed results.

Improved energy confinement for additionally heated JT-60 limiter and lower x-point discharges has been obtained with hydrogen pellet injection. The energy confinement time is enhanced by up to 40% over that of usual gas fuelled discharges at a medium NB or NB + ICRF heating power level of < 15 MW, for which a strongly peaked electron density profile with ne(0)/ne < 5 and ne(0) < 2.0 × 1020 m−3 is sustained within 0.5 s after a series of pellet injections. The achieved values of the fusion product ne(0)τE(a)Te(0) (

An m=3/n=l locked mode has been observed in JT-60 at safety factors slightly below three during the current ramp-up phase of high plasma current discharges. The locked mode grows suddenly and there is no Mirnov oscillation just before the onset of this mode. The maximum plasma current of Ohmic discharges is restricted to approximately 2.4 — 2.45 MA by disruptions, which are preceded by a large m=3/n=l locked mode. The locked mode is stabilized by high power neutral beam injection. The stabilizing effect depends on neutral beam heating power, plasma density, plasma current ramp-up rate and divertor/limiter configuration.