An experimental study of plasma fluctuations in the TCV and TEXTOR Tokamaks

International Nuclear Information System (INIS)

Mejeire de, C. A.

2013-01-01

The main body of this thesis reports on the commissioning and first measurements with a novel tangential phase-contrast imaging (TPCI) diagnostic, which had previously been installed in the TCV tokamak. The instrument measures fluctuations in line-integrated electron density along 9 parallel chords within a 6 cm diameter CO 2 laser beam. TPCI measurements reveal the first evidence in TCV of the geodesic acoustic mode (GAM), which is an oscillating zonal flow. Frequency, radial wavelength, radial extent and propagation are all in qualitative agreement with a gyro-kinetic simulation and recent theoretical work. The mode is found to have a modest, but measurable magnetic component, whose spatial structure is characterised for the first time in a toroidal plasma. For some experiments, clear evidence is found of the theoretically expected m/n = 2/0 mode structure, although in others the structure appears to be more complex. Electron energy confinement in X 2 heated TCV L-mode plasmas had previously been observed to increase on changing the triangularity (δ) of the poloidal plasma cross-section from δ = +0.4 to δ = −0.4. Measurements with the TPCI diagnostic reveal that this change coincides with a clear decrease in both the absolute level and the decorrelation time of broadband electron density fluctuations. This is in agreement with the conjecture that the increased confinement time is caused by a change in the turbulent state. The second part of the thesis reports on a fluctuation study in the TEXTOR tokamak. At sufficiently weak toroidal magnetic field, NBI heated, limited TEXTOR plasmas exhibit bursts of beam-ion driven ‘fishbone’ and Alfvén modes, which are characterised using the multi-antenna reflectometer and Mirnov coils. In H-mode the fishbone triggers ELMs and in L-mode it triggers previously unobserved bursts of particle recycling, resembling the ELMs. The reflectometer phase shows statistically significant bispectral coherence between the fishbone

Retention of neon in graphite after ion beam implantation or exposures to the scrape-off layer plasma in the TEXTOR tokamak

International Nuclear Information System (INIS)

Kim, Y.M.; Philipps, V.; Rubel, M.; Vietzke, E.; Pospieszczyk, A.; Unterberg, B.; Jaspers, R.

2002-01-01

The interaction of neon ions with graphite was investigated for targets either irradiated with ion beams (2-10 keV range) or exposed to the scrape-off layer plasma in the TEXTOR tokamak during discharges with neon edge cooling. The emphasis was on the influence of the target temperature (300-1200 K) and the implantation dose on the neon retention and reemission. The influence of deuterium impact on the retention of neon implanted into graphite has also been addressed. In ion beam experiments saturation is observed above a certain ion dose with a saturation level, which decreases with increasing target temperature. The temperature dependence of the thermal desorption corresponds to an apparent binding energy of about 2.06 eV. The retention of neon (C Ne /C C ) decreases with increasing ion energy with values from 0.55 to 0.15 following irradiation with 2 and 10 keV ions, respectively. The reemission yield during the irradiation increases with target temperature and above 1200 K all impinging ions are reemitted instantaneously. The retention densities measured using the sniffer probe at the TEXTOR tokamak are less than 1% of the total neon fluence and are over one order of magnitude smaller than those observed in ion beam experiments. The results are discussed in terms of different process decisive for ion deposition and release under the two experimental conditions

Multi scale study of carbon deposits collected in Tore-Supra and TEXTOR tokamaks; Etude multi echelle des depots carbones collectes dans les tokamaks Tore Supra et TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Richou, M

2007-06-15

Tokamaks are devices aimed at studying magnetic fusion. They operate with high temperature plasmas containing hydrogen, deuterium or tritium. One of the major issue is to control the plasma-wall interaction. The plasma facing components are most often in carbon. The major drawback of carbon is the existence of carbon deposits and dust, due to erosion. Dust is potentially reactive in case of an accidental opening of the device. These deposits also contain H, D or T and induce major safety problems when tritium is used, which will be the case in ITER. Therefore, the understanding of the deposit formation and structure has become a main issue for fusion researches. To clarify the role of the deposits in the retention phenomenon, we have done different complementary characterizations for deposits collected on similar places (neutralizers) in tokamaks Tore Supra (France) and TEXTOR (Germany). Accessible microporous volume and pore size distribution of deposits has been determined with the analysis of nitrogen and methane adsorption isotherms using the BET, Dubinin-Radushkevich and {alpha}{sub s} methods and the Density Functional Theory (DFT). To understand growth mechanisms, we have studied the deposit structure and morphology. We have shown using Transmission Electron Microscopy (TEM) and Raman micro-spectrometry that these deposits are non amorphous and disordered. We have also shown the presence of nano-particles (diameter between 4 and 70 nm) which are similar to carbon blacks: nano-particle growth occurs in homogeneous phase in the edge plasma. We have emphasised a dual growth process: a homogenous and a heterogeneous one. (author)

The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Tammen, H.F.

1995-01-10

One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics `Rijnhuizen`, was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL).

The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR

International Nuclear Information System (INIS)

Tammen, H.F.

1995-01-01

One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics 'Rijnhuizen', was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL)

Advanced limiter test (ALT-I) in the TEXTOR tokamak - concept and experimental design

International Nuclear Information System (INIS)

Conn, R.W.; Grotz, S.P.; Prinja, A.K.

1983-01-01

The concept and experimental design of a pump-limiter for the TEXTOR tokamak is described. The module is constructed of stainless steel with a compound curvature head designed to limit the maximum heat flux to 300 W/cm 2 . The head is made of TiC-coated graphite containing a variable aperture slot to admit plasma to a deflector plate for ballistic pumping action. The assembly is actively pumped using Zr-Al getters with an estimated hydrogen pumping speed of 2x10 4 1/s. The aspect ratio of the pump duct and the length of the plasma channel are both variable to permit study of plasma plugging, ballistic scattering, and enhanced gas conduction effects. The module can be moved radially by 10 cm to permit its operation either as the primary or secondary limiter. Major diagnostics include Langmuir and solid state probes, bolometers, infrared thermography, thermocouples, ion gauges, manometers, and a gas mass analyzer. (author)

Spectral measurements of runway electrons in the TEXTOR tokamak

International Nuclear Information System (INIS)

Kudyakov, Timur

2009-01-01

The generation of multi-MeV runaway electrons is a well known effect related to the plasma disruptions in tokamaks. The runaway electrons can substantially reduce the lifetime of the future tokamak ITER. In this thesis physical properties of runaway electrons and their possible negative effects on ITER have been studied in the TEXTOR tokamak. A new diagnostic, a scanning probe, has been developed to provide direct measurements of the absolute number of runaway electrons coming from the plasma, its energy distribution and the related energy load in the material during low density (runaway) discharges and during disruptions. The basic elements of the probe are YSO crystals which transform the energy of runaway electrons into visible light which is guided via optical fibres to photomultipliers. In order to obtain the energy distribution of runaways, the crystals are covered with layers of stainless steel (or tungsten in two earlier test versions) of different thicknesses. The final probe design has 9 crystals and can temporally and spectrally resolve electrons with energies between 4 MeV and 30 MeV. The probe is tested and absolutely calibrated at the linear electron accelerator ELBE in Rossendorf. The measurements are in good agreement with Monte Carlo simulations using the Geant4 code. The runaway transport in the presence of the internal and externally applied magnetic perturbations has been studied. The diffusion coefficient and the value of the magnetic fluctuation for runaways were derived as a function of B t . It was found that an increase of runaway losses from the plasma with the decreasing toroidal magnetic field is accompanied with a growth of the magnetic fluctuation in the plasma. The magnetic shielding picture could be confirmed which predicts that the runaway loss occurs predominantly for low energy runaways (few MeV) and considerably less for the high energy ones. In the case of the externally applied magnetic perturbations by means of the dynamic

Spectral measurements of runway electrons in the TEXTOR tokamak

Energy Technology Data Exchange (ETDEWEB)

Kudyakov, Timur

2009-07-22

The generation of multi-MeV runaway electrons is a well known effect related to the plasma disruptions in tokamaks. The runaway electrons can substantially reduce the lifetime of the future tokamak ITER. In this thesis physical properties of runaway electrons and their possible negative effects on ITER have been studied in the TEXTOR tokamak. A new diagnostic, a scanning probe, has been developed to provide direct measurements of the absolute number of runaway electrons coming from the plasma, its energy distribution and the related energy load in the material during low density (runaway) discharges and during disruptions. The basic elements of the probe are YSO crystals which transform the energy of runaway electrons into visible light which is guided via optical fibres to photomultipliers. In order to obtain the energy distribution of runaways, the crystals are covered with layers of stainless steel (or tungsten in two earlier test versions) of different thicknesses. The final probe design has 9 crystals and can temporally and spectrally resolve electrons with energies between 4 MeV and 30 MeV. The probe is tested and absolutely calibrated at the linear electron accelerator ELBE in Rossendorf. The measurements are in good agreement with Monte Carlo simulations using the Geant4 code. The runaway transport in the presence of the internal and externally applied magnetic perturbations has been studied. The diffusion coefficient and the value of the magnetic fluctuation for runaways were derived as a function of B{sub t}. It was found that an increase of runaway losses from the plasma with the decreasing toroidal magnetic field is accompanied with a growth of the magnetic fluctuation in the plasma. The magnetic shielding picture could be confirmed which predicts that the runaway loss occurs predominantly for low energy runaways (few MeV) and considerably less for the high energy ones. In the case of the externally applied magnetic perturbations by means of the dynamic

The influence of the dynamic ergodic divertor on the radial electric field at the Tokamak TEXTOR

International Nuclear Information System (INIS)

Coenen, Jan Willem

2009-01-01

In this work the influence of external Resonant Magnetic Perturbations (RMPs) on the radial electric field Er in magnetically confined plasmas is investigated by Charge Exchange Recombination Spectroscopy (CXRS) at the Tokamak TEXTOR. Here, the RMPs are produced with the Dynamic Ergodic Divertor (DED), a set of 16 helical perturbation coils located at the high field side of TEXTOR. Within this work, the base mode number of perturbations has been m/n=6/2. We have first investigated the influence of external torque from neutral heating beams on plasma rotation and E r . The ergodic zone causes an electron loss, and subsequently a vector j x vector B force driven by the compensating ion return current. In addition, the DED changes the global confinement properties. Depending on the edge safety factor (''field line twist'') q a , either increased or decreased particle confinement is observed. In case of the increased particle confinement (IPC) the increase in density (40%) and particle confinement time τ p (30%) is correlated to the connection of field lines at the q=5/2 surface to the DED target, locally changing the transport properties and the E r . Transport is reduced and the E r shear is increased locally at q=5/2 up to 1.5 . 10 5 s -1 , while the E r becomes more positive. (orig.)

Numerical modelling of pump limiter biasing on TEXTOR-94 and Tore Supra

International Nuclear Information System (INIS)

Gerhauser, H.; Claassen, H.A.; Mank, G.; Zagorski, R.; Loarer, T.; Gunn, J.; Boucher, C.

2002-01-01

The two-dimensional multifluid code TECXY has been used to model the biasing (with respect to the first wall) of the toroidal belt limiter ALT-II on the tokamak TEXTOR-94 and of the new toroidal pump limiter being installed on Tore Supra tokamak in the framework of the CIEL project. It is well known that the edge flow pattern can be influenced by the poloidal electric drifts from imposing radial electric fields. The modelling with TECXY introduces imprinted bias currents in the scrape-off layer (SOL) for the case of negative (limiter) biasing, and imprinted bias potentials for the case of positive biasing. This allowed us to simulate sufficiently well the experimental I-V characteristics for either biasing of ALT-II and also reproduced the essential features and trends of the observed plasma profiles in the SOL of TEXTOR-94. For negative biasing a moderate improvement of the pumping exhaust efficiency can be achieved in the case of TEXTOR. For Tore Supra, however, only a negligible improvement of the limiter performance with biasing can be predicted, which is explained by the relatively weak drift flows in Tore Supra. (author)

Preliminary design analysis of the ALT-II limiter for TEXTOR

International Nuclear Information System (INIS)

Koski, J.A.; Boyd, R.D.; Kempka, S.M.; Romig, A.D. Jr.; Smith, M.F.; Watson, R.D.; Whitley, J.B.; Conn, R.W.; Grotz, S.P.

1983-01-01

Installation of a large toroidal belt pump limiter, Advanced Limiter Test II (ALT-II), on the TEXTOR tokamak at Juelich, FRG is anticipated for early 1986. This paper discusses the preliminary mechanical design and materials considerations undertaken as part of the feasibility study phase for ALT-II

In situ measurements of fuel retention by laser induced desorption spectroscopy in TEXTOR

Science.gov (United States)

Zlobinski, M.; Philipps, V.; Schweer, B.; Huber, A.; Stoschus, H.; Brezinsek, S.; Samm, U.; TEXTOR Team

2011-12-01

In future fusion devices such as ITER tritium retention due to tritium co-deposition in mixed material layers can be a serious safety problem. Laser induced desorption spectroscopy (LIDS) can measure the hydrogen content of hydrogenic carbon layers locally on plasma-facing components, while hydrogen is used as a tritium substitute. For several years, this method has been applied in the TEXTOR tokamak in situ during plasma operation to monitor the hydrogen content in space and time. This work shows the LIDS signal reproducibility and studies the effects of different plasma conditions, desorption distances from the plasma and different laser energies using a dedicated sample with constant hydrogen amount. Also the LIDS signal evaluation procedure is described in detail and the detection limits for different conditions in the TEXTOR tokamak are estimated.

Parametric dependence of density limits in the Tokamak Experiment for Technology Oriented Research (TEXTOR): Comparison of thermal instability theory with experiment

International Nuclear Information System (INIS)

Kelly, F.A.; Stacey, W.M.; Rapp, J.

2001-01-01

The observed dependence of the TEXTOR [Tokamak Experiment for Technology Oriented Research: E. Hintz, P. Bogen, H. A. Claassen et al., Contributions to High Temperature Plasma Physics, edited by K. H. Spatschek and J. Uhlenbusch (Akademie Verlag, Berlin, 1994), p. 373] density limit on global parameters (I, B, P, etc.) and wall conditioning is compared with the predicted density limit parametric scaling of thermal instability theory. It is necessary first to relate the edge parameters of the thermal instability theory to n(bar sign) and the other global parameters. The observed parametric dependence of the density limit in TEXTOR is generally consistent with the predicted density limit scaling of thermal instability theory. The observed wall conditioning dependence of the density limit can be reconciled with the theory in terms of the radiative emissivity temperature dependence of different impurities in the plasma edge. The thermal instability theory also provides an explanation of why symmetric detachment precedes radiative collapse for most low power shots, while a multifaceted asymmetric radiation from the edge MARFE precedes detachment for most high power shots

The influence of the dynamic ergodic divertor on the radial electric field at the Tokamak TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Coenen, Jan Willem

2009-11-06

In this work the influence of external Resonant Magnetic Perturbations (RMPs) on the radial electric field Er in magnetically confined plasmas is investigated by Charge Exchange Recombination Spectroscopy (CXRS) at the Tokamak TEXTOR. Here, the RMPs are produced with the Dynamic Ergodic Divertor (DED), a set of 16 helical perturbation coils located at the high field side of TEXTOR. Within this work, the base mode number of perturbations has been m/n=6/2. We have first investigated the influence of external torque from neutral heating beams on plasma rotation and E{sub r}. The ergodic zone causes an electron loss, and subsequently a (vector)j x (vector)B force driven by the compensating ion return current. In addition, the DED changes the global confinement properties. Depending on the edge safety factor (''field line twist'') q{sub a}, either increased or decreased particle confinement is observed. In case of the increased particle confinement (IPC) the increase in density (40%) and particle confinement time {tau}{sub p} (30%) is correlated to the connection of field lines at the q=5/2 surface to the DED target, locally changing the transport properties and the E{sub r}. Transport is reduced and the E{sub r} shear is increased locally at q=5/2 up to 1.5 . 10{sup 5}s{sup -1}, while the E{sub r} becomes more positive. (orig.)

Combined interferometric and polarimetric diagnostics for TEXTOR

International Nuclear Information System (INIS)

Soltwisch, H.

1980-01-01

A method for combining Faraday rotation measurements with a phase modulated HCN interferometer is described. Extended to a multichannel system this technique should allow to investigate the distributions of electron density and poloidal magnetic field in a Tokamak plasma. We discuss the principle of operation and appropriate evaluation of measured data with regard to TEXTOR parameters and give an estimate for the expected experimental accuracy. (orig./GG) 891 GG/orig.- 892 KN

Density turbulence and disruption phenomena in TEXTOR

International Nuclear Information System (INIS)

Waidmann, G.; Kuang, G.; Jadoul, M.

1992-01-01

Disruptive processes are observed in tokamak plasmas not only at the operating limits (density limit or q-limit) but can be found under a variety of experimental conditions. Large forces are exerted then on vessel components and support structures. The sudden release of stored plasma energy presents a serious erosion problem for the first wall already in the next generation of large tokamak machines. Strong energy losses from the plasma and an influx of impurities are already present in minor plasma disruptions which do not immediately lead to a plasma current termination. The rapid loss of energy confinement was investigated within the framework of a systematic study on plasma disruption phenomena in TEXTOR. (author) 4 refs., 4 figs

Measurements of the runaway electron energy during disruptions in the tokamak TEXTOR

International Nuclear Information System (INIS)

Forster, M.; Finken, K. H.; Willi, O.; Lehnen, M.; Xu, Y.

2012-01-01

Calorimetric measurements of the total runaway electron energy are carried out using a reciprocating probe during induced TEXTOR disruptions. A comparison with the energy inferred from runaway energy spectra, which are measured with a scintillator probe, is used as an independent check of the results. A typical runaway current of 100 kA at TEXTOR contains 30 to 35 kJ of runaway energy. The dependencies of the runaway energy on the runaway current, the radial probe position, the toroidal magnetic field and the predisruptive plasma current are studied. The conversion efficiency of the magnetic plasma energy into runaway energy is calculated to be up to 26%.

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.

1984-05-01

Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.; California Univ., Los Angeles

1984-01-01

Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)

A constant heat flux plasma limiter for TEXTOR

International Nuclear Information System (INIS)

Mioduszewski, P.

1980-10-01

In future large tokamak machines heat removal from the plasma is going to play an important role. In TEXTOR the total plasma power is expected to be in the range of 0.5-2.5 MW. Typical fractions of about 50% of this power have to be removed from the plasma by limiters. The power flux from the limiter scrape-off layer to the limiter surface decays rapidly with distance into the scrape-off layer resulting in a highly space-dependent heat load on the limiter. Therefore, limiters are shaped in a way to smooth of the heat load, and the ideal limiter shape should produce a constant heat flux over the whole limiter surface. The ideally shaped limiter offers a better chance to handle the high heat loads with the preferred materials like stainless steel (or inconel 625 as in the case of TEXTOR). (orig./GG)

Evidence for reduction of the toroidal ITG instability in the transition from saturated to improved Ohmic confinement in the tokamak TEXTOR

International Nuclear Information System (INIS)

Kreter, A; Schweer, B; Tokar, M Z; Unterberg, B

2003-01-01

In high density Ohmically heated discharges in the tokamak TEXTOR a transition from the saturated Ohmic confinement (SOC) to the improved Ohmic confinement (IOC) was observed triggered by a sudden reduction of the external gas flow. The SOC-IOC transition was investigated regarding the influence of the toroidal ITG instability driven by the ion temperature gradient (ITG). The ion temperature profiles were measured with high radial resolution by means of charge-exchange recombination spectroscopy (CXRS) with a high-energetic diagnostic hydrogen beam recently installed at TEXTOR. On the basis of the measured ion temperature distributions the η i parameter (ratio of the density and ion temperature decay lengths) and the growth rate of the toroidal ITG instability were calculated. After the SOC-IOC transition η i drops and lies in a noticeably smaller radial region over the threshold for the toroidal ITG. In consequence of it, the IOC regime is characterized by a clear reduction of the ITG growth rate γ ITG which was calculated including finite Larmor radius effects. The steepening of the plasma density profile after the decrease of the external gas flow is the main reason for the reduction of the ITG growth rate and the subsequent confinement transition to the IOC regime

Disruptions and Their Mitigation in TEXTOR

International Nuclear Information System (INIS)

Finken, K.H.; Jaspers, R.; Kraemer-Flecken, A.; Savtchkov, A.; Lehnen, M.; Waidmann, G.

2005-01-01

Disruptions remain a major concern for tokamak devices, particularly for large machines. The critical issues are the induced (halo) currents and the resulting forces, the excessive heating of exposed surfaces by the instantaneous power release, and the possible occurrence of highly energetic runaway electrons. The key topics of the investigations on TEXTOR in the recent years concerned (a) the power deposition pattern recorded by a fast infrared scanner, (b) the runaway generation measured by synchrotron radiation in the infrared spectral region, (c) method development for 'healing' discharges that are going to disrupt, and (d) massive gas puffing for mitigating the adverse effects of disruptions

Heterodyne ECE diagnostic in the mode detection and disruption avoidance at TEXTOR

International Nuclear Information System (INIS)

Kraemer-Flecken, A.; Finken, K.H.; Larue, H.; Udintsev, V.S.; TEXTOR - team

2003-01-01

Disruptions cause major concerns for the operation of tokamaks. During disruption large forces act on the tokamak vessel and its interior parts. The huge amount of plasma energy deposited on the first wall components within one millisecond causes serious damage. Therefore disruptions should be avoided. One way to avoid disruptions is the operation of a tokamak in a regime which is easy to handle from the control point of view. However, the operation in the advanced scenarios or improved confinement modes is very complicated and even small deviation in one of the control parameters can cause a disruption. In this cases a method should be available to detect the disruption in advance and mitigate or even better avoid the energy quench by appropriate means. At TEXTOR we developed a method to detect the disruption precursor. The module is integrated in the plasma control system. The detection method was tested at TEXTOR for (i) combination with tangential neutral beam injection to increase the toroidal rotation profile and to tear apart the m = 2 disruption precursor by a steep rotation gradient across the island (ii) gas puff experiments with He used to mitigate the disruption effects specially to suppress the generation of the runaway electrons. The paper demonstrates the possibility to detect disruptions precursors and to avoid disruptions using two ECE-channels out of the standard electron temperature diagnostic. The system demonstrated its reliability during the last month of TEXTOR operation. The injection of co- as well as counter neutral beam to avoid the disruption was successful tested and a detailed analysis of the mode development is presented. The measured rotation profiles show the development of a step in the toroidal velocity in the vicinity of the q = 2 surface which prevents the plasma from a disruption. Furthermore detailed analysis of the frequency development of the m = 2 mode could explain the observed sudden increase in the mode frequency

Disruption generated secondary runaway electrons in present day tokamaks

International Nuclear Information System (INIS)

Pankratov, I.M.; Jaspers, R.

2000-01-01

An analysis of the runaway electron secondary generation during disruptions in present day tokamaks (JET, JT-60U, TEXTOR) was made. It was shown that even for tokamaks with the plasma current I approx 100 kA the secondary generation may dominate the runaway production during disruptions. In the same time in tokamaks with I approx 1 MA the runaway electron secondary generation during disruptions may be suppressed

The TEXTOR helium self-pumping experiment: Design, plans, and supporting ion-beam data on helium retention in nickel

International Nuclear Information System (INIS)

Brooks, J.N.; Krauss, A.; Mattas, R.F.; Smith, D.L.; Nygren, R.E.; Doyle, B.L.; McGrath, R.T.; Walsh, D.; Dippel, K.H.; Finken, K.H.

1990-01-01

A proof-of-principle experiment to demonstrate helium self-pumping in a tokamak is being undertaken in TEXTOR. The experiment will use a helium self-pumping module installed in a modified ALT-I limiter head. The module consists of two, ≅ 25x25 cm 2 heated nickel alloy trapping plates, a nickel deposition filament array, and associated diagnostics. Between plasma shots a coating of ≅ 50A nickel will be deposited on the two trapping plates. During a shot helium and hydrogen ions will impinge on the plates through a ≅ 3 cm wide entrance slot. The helium removal capability, due to trapping in the nickel, will be assessed for a variety of plasma conditions. In support of the tokamak experiment, the trapping of helium over a range of ion fluences and surface temperatures, and detrapping during subsequent exposure to hydrogen, were measured in ion beam experiments using evaporated nickel surfaces similar to that expected in TEXTOR. Also, the retention of H and He after exposure of a nickel surface to mixed He/H plasmas has been measured. The results appear favorable, showing high helium trapping (≅ 10-50% He/Ni) and little or no detrapping by hydrogen. The TEXTOR experiment is planned to begin in 1991. (orig.)

The TEXTOR helium self-pumping experiment: Design, plans, and supporting ion-beam data on helium retention in nickel

International Nuclear Information System (INIS)

Brooks, J.N.; Krauss, A.; Mattas, R.F.; Smith, D.L.; Nygren, R.E.; Doyle, B.L.; McGrath, R.T.; Walsh, D.; Dippel, K.H.; Finken, K.H.

1990-01-01

A proof-of-principle experiment to demonstrate helium self-pumping in a tokamak is being undertaken in TEXTOR. The experiment will use a helium self-pumping module installed in a modified ALT-I limiter head. The module consists of two, ∼25 x 25 cm 2 heated nickel alloy trapping plates, a nickel deposition filament array, and associated diagnostics. Between plasma shots a coating of ∼50 angstrom nickel will be deposited on the two trapping plates. During a shot helium and hydrogen ions will impinge on the plates through a ∼3 cm wide entrance slot. The helium removal capability, due to trapping in the nickel, will be assessed for a variety of plasma conditions. In support of the tokamak experiment, the trapping of helium over a range of ion fluences and surface temperatures, and detrapping during subsequent exposure to hydrogen, were measured in ion beam experiments using evaporated nickel surfaces similar to that expected in TEXTOR. Also, the retention of H and He after exposure of a nickel surface to mixed He/H plasmas has bee measured. The results appear favorable, showing high helium trapping (∼10--50% He/Ni) and little or no detrapping by hydrogen. The TEXTOR experiment is planned to begin in 1991. 12 refs., 2 figs., 2 tabs

Determination of central q and effective mass on textor based on discrete Alfven wave (DAW) spectrum measurements

International Nuclear Information System (INIS)

Descamps, P.; Wassenhove, G. van; Koch, R.; Messiaen, A.M.; Vandenplas, P.E.; Lister, J.B.; Marmillod, P.

1990-01-01

The use of the discrete Alfven wave spectrum to determine the current density profile and the effective mass density of the plasma in the TEXTOR tokamak is studied; the measurement, the validity of which is discussed, confirms independently the central q(r=0)<1 already obtained by polarimetry. (orig.)

Exposure of tungsten nano-structure to TEXTOR edge plasma

International Nuclear Information System (INIS)

Ueda, Y.; Miyata, K.; Ohtsuka, Y.; Lee, H.T.; Fukumoto, M.; Brezinsek, S.; Coenen, J.W.; Kreter, A.; Litnovsky, A.; Philipps, V.; Schweer, B.; Sergienko, G.; Hirai, T.; Taguchi, A.; Torikai, Y.; Sugiyama, K.; Tanabe, T.; Kajita, S.; Ohno, N.

2011-01-01

W nano-structures (fuzz), produced in the linear high plasma device, NAGDIS, were exposed to TEXTOR edge plasmas (ohmic He/D mixed plasma and pure D plasma) to study formation, erosion and C deposition on W fuzz in tokamak plasmas for the first time. Fuzz layers were either completely eroded or covered by C deposit. There was no clear indication of W fuzz growth under the present conditions. There was no significant difference of C deposition between 'thick' fuzz (500-600 nm in thickness) and 'thin' fuzz (300-400 nm) in the He/D plasma. On the W fuzz surface, C deposition was enhanced probably due to reduction of effective sputtering yield and effective reflection coefficient of carbon ions, similar to roughness effects. Formation and erosion of W fuzz in tokamak devices and role of impurities are discussed.

Transient electromagnetic analysis in tokamaks using TYPHOON code

International Nuclear Information System (INIS)

Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

1996-01-01

The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

ALT-II armor tile design for upgraded TEXTOR operation

International Nuclear Information System (INIS)

Newberry, B.L.; McGrath, R.T.; Watson, R.D.

1994-01-01

The upgrade of the TEXTOR tokamak at KFA Julich will be completed in the spring of 1994. The upgrade will extend the TEXTOR pulse length from 5 seconds to 10 seconds. The auxiliary heating systems are also scheduled to be upgraded so that eventually a total of 8.0 MW auxiliary heating will be available through a combination of neutral beam injection and radio frequency heating. Originally, the inertially cooled armor tiles on the full toroidal belt Advanced Limiter Test - II (ALT-II) were designed for 5-second operation with a total heating power of 6.0 MW. The upgrade of TEXTOR will increase the energy deposited per pulse onto ALT-II by more than 300%. Consequently, the graphite armor tiles for ALT-II had to be redesigned in order to increase their thermal inertia and, thereby, avoid excessively high graphite armor surface temperatures that would lead to unacceptable contamination of the plasma. The armor tile thermal inertia had been increase primarily by expanding the radial thickness of the tiles from 17 mm to 20 mm. This increase in radial tile dimension will reduce the overall pumping efficiency of the ALT-II pump limiter by about 30%. The final armor tile design was a compromise between increasing the power handling capability and reducing the particle exhaust efficiency of ALT-II. The reduction in exhaust efficiency is unfortunate, but could only be avoided by active cooling of the ALT-II armor tiles. The active cooling option was too complicated and expensive to be considered at this time

ALT-II armor tile design for upgraded TEXTOR operation

International Nuclear Information System (INIS)

Newberry, B.L.; McGrath, R.T.; Watson, R.D.; Kohlhaas, W.; Finken, K.H.

1994-01-01

The upgrade of the TEXTOR tokamak at KFA Juelich was recently completed. This upgrade extended the TEXTOR pulse length from 5 seconds to 10 seconds. The auxiliary heating was increased to a total of 8.0 MW through a combination of neutral beam injection and radio frequency heating. Originally, the inertially cooled armor tiles of the full toroidal belt Advanced Limiter Test -- II (ALT-II) were designed for a 5-second operation with total heating of 6.0 MW. The upgrade of TEXTOR will increase the energy deposited per pulse onto the ALT-II by about 300%. Consequently, the graphite armor tiles for the ALT-II had to be redesigned to avoid excessively high graphite armor surface temperatures that would lead to unacceptable contamination of the plasma. This redesign took the form of two major changes in the ALT-II armor tile geometry. The first design change was an increase of the armor tile thermal mass, primarily by increasing the radial thickness of each tile from 17 mm to 20 mm. This increase in the radial tile dimension reduces the overall pumping efficiency of the ALT-II pump limiter by about 30%. The reduction in exhaust efficiency is unfortunate, but could be avoided only by active cooling of the ALT-II armor tiles. The active cooling option was too complicated and expensive to be considered at this time. The second design change involved redefining the plasma facing surface of each armor tile in order to fully utilize the entire surface area. The incident charged particle heat flux was distributed uniformly over the armor tile surfaces by carefully matching the radial, poloidal and toroidal curvature of each tile to the plasma flow in the TEXTOR boundary layer. This geometry redefinition complicates the manufacturing of the armor tiles, but results in significant thermal performance gains. In addition to these geometry upgrades, several material options were analyzed and evaluated

Real-time evaluation of electron and current density profile parameters on TEXTOR

International Nuclear Information System (INIS)

Bruessau, W.D.; Soltwisch, H.

1985-08-01

The shapes of electron and current density profiles are monitored in real-time mode in order to get rapid qualitative information on the development of a TEXTOR tokamak plasma. The profiles are described by form parameters which relate to the signals of a 9-channel FIR-polari/interferometer in simple mathematical formulae. These profile parameters are obtained by real-time conversion of measured quantities for display on a storage oscilloscope or on a chart recorder. The application of the parameters is demonstrated in some examples. (orig.)

Charge exchange recombination in X-ray spectra of He-like argon measured at the tokamak TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Schlummer, Tobias

2014-06-16

Charge exchange recombination between ions and atomic hydrogen is an important atomic process in magnetically confined fusion plasmas. Besides radiative cooling of the plasma edge, charge exchange causes modifications of the ionization balance and the population densities of excited ion states. The central goal of this work is to investigate the influence of charge exchange on X-ray spectra measured at the tokamak TEXTOR. A new 2D X-ray spectrometer developed for future use at the stellarator W7-X was recently installed at TEXTOR. The spectrometer is optimized for measuring the K{sub α}-spectrum of He-like argon (1s2l - 1s{sup 2}) at wavelengths close to 4 Aa. K{sub α}-spectroscopy on He-like impurity ions is an established diagnostic for electron and ion temperature measurements in fusion plasmas. Still, up to now the observed intensity ratios of the K{sub α}-lines and their associated satellites are not fully understood. They show significant deviations from the predictions made by basic corona models. In the past charge exchange with the neutral particle background and radial impurity transport have been discussed as likely explanations. Yet a detailed description of the experimental spectra still has not been achieved. To reconstruct the 2D K{sub α}-spectra measured at TEXTOR the radial argon ion distribution is modeled using an impurity transport code. The model accounts for charge exchange and transport on basis of given radial profiles of the neutral particle density n{sub 0}(r) and the diffusion coefficient D {sub perpendicular} {sub to} (r). The theoretical spectrum is then constructed based on the processes relevant for line emission. Within an iterative procedure n{sub 0}(r) and D {sub perpendicular} {sub to} (r) are varied until consistency between the theoretical and the experimental spectra is achieved. It is shown that the 2D K{sub α}-spectra allow a clear distinction of charge exchange and transport effects, ensuring unique solutions for n

Applications of solid-state nuclear track detectors (SSNTDs) for fast ion and fusion reaction product measurements in TEXTOR experiments

Energy Technology Data Exchange (ETDEWEB)

Szydlowski, A.; Malinowski, K.; Malinowska, A. [Association EURTOM-IPPLM Warsaw, The Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland); Wassenhove, G. Van [EURATOM-Belgium State Association, LPP, ERM/KMS, Trilateral Euregio Cluster, B-1000 Brussels (Belgium); Schweer, B. [Association EURATOM-FZJ, Institutte of Plasma Physicx, Juelich (Germany)

2011-07-01

Full text of publication follows: The paper reports on measurements of fusion reaction protons which were performed on TEXTOR facility in January 2009. The basic experimental scheme was similar to that applied in the previous measurements [1, 2]. The main experimental tool equipment was a small ion pinhole camera which was equipped with a PM-355 detector sample and was attached to a water cooled manipulator. The camera was placed below the plasma ring in the direction of ion drifts, at a distance of 4.4 cm from LCFS. However, in the described experiment it was aligned at an angle to the mayor TEXTOR radius (contrary to previous experiments), so that the input pinhole was oriented first at {gamma} = 45 degrees (shots 108799 - 108818) and then {gamma} = 600 (shots 108832 - 108847). The discharges were executed with one neutral beam of the total power 0.6 - 1.0 MW. In the first series (Nos 108799 - 108818) the plasma was additionally heated by ICRH of frequency 38 MHz. The irradiated detector samples were subjected to the same interrupted etching procedure as the samples used in the CR-39/PM-355 detector calibration measurements [1, 2]. After that, track density distributions and track diameter histograms were measured under an optical microscope. By the use of the calibration curves, it was possible to distinguish craters produced by protons from other craters and to convert the obtained histograms into proton energy spectra. The craters induced by lower energy ions appeared to be concentrated in narrower areas, whereas higher energy ions were registered in a more diffused detector fields. The paper shows again that the CR-39/PM-355 detector is an useful diagnostic tool for tokamak experiments, for measurement of charged ions. References: [1] A. Szydlowski, A. Malinowska, M. Jaskola, A. Korman, M.J. Sadowski, G. Van Wassenhove, B. Schweer and the TEXTOR team, A. Galkowski, 'Application of Solid State Nuclear Track Detectors in TEXTOR Experiment for Measurements

Role of symmetry-breaking induced by Er × B shear flows on developing residual stresses and intrinsic rotation in the TEXTOR tokamak

International Nuclear Information System (INIS)

Xu, Y.; Shesterikov, I.; Berte, M.; Dumortier, P.; Van Schoor, M.; Vergote, M.; Hidalgo, C.; Krämer-Flecken, A.; Koslowski, R.

2013-01-01

Direct measurements of residual stress (force) have been executed at the edge of the TEXTOR tokamak using multitip Langmuir and Mach probes, together with counter-current NBI torque to balance the existing toroidal rotation. Substantial residual stress and force have been observed at the plasma boundary, confirming the existence of a finite residual stress as possible mechanisms to drive the intrinsic toroidal rotation. In low-density discharges, the residual stress displays a quasi-linear dependence on the local pressure gradient, consistent with theoretical predictions. At high-density shots the residual stress and torque are strongly suppressed. The results show close correlation between the residual stress and the E r × B flow shear rate, suggesting a minimum threshold of the E × B flow shear required for the k ∥ symmetry breaking. These findings provide the first experimental evidence of the role of E r × B sheared flows in the development of residual stresses and intrinsic rotation. (letter)

Particle balance studies in TEXTOR during experiments of pellet injection, helium injection, and ICR-heating

International Nuclear Information System (INIS)

Banno, T.; Finken, K.H.; Gray, D.S.; Winter, J.

1995-01-01

Analysis based on the particle conservation law has been carried out to observe the global fuelling process in tokamak discharges. The response of the net recycling flux from the first wall is investigated in the tokamak TEXTOR, using calibrated signals of the gas feed rate, the neutral gas pressure in the vessel, the total amount of electrons, and the particle removal rates by the ALT-II belt-pump limiter and by a main pump unit. Net absorption (pumping) of hydrogen by the wall is observed for almost all tokamak discharges since a new wall conditioning technique called siliconisation is employed. The net absorption or fuelling depending on the discharge condition influenced by injection of pellets, by helium gas injection combined with neutral beam injection, and by rf heating can be interpreted in terms of the particle-induced desorption effect with depth profile taken into consideration. ((orig.))

Dynamics of fast ions during sawtooth oscillations in the TEXTOR tokamak measured by collective Thomson scattering

DEFF Research Database (Denmark)

Nielsen, Stefan Kragh; Salewski, Mirko; Bindslev, Henrik

2011-01-01

Experimental investigations of sawteeth interaction with fast ions measured by collective Thomson scattering on TEXTOR are presented. Time-resolved measurements of localized 1D fast-ion distribution functions allow us to study fast-ion dynamics during several sawtooth cycles. Sawtooth oscillation...

Gyrosheath near the tokamak edge

International Nuclear Information System (INIS)

Hazeltine, R.D.; Xiao, H.; Valanju, P.M.

1993-03-01

A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results

ALT-I pump limiter results on TEXTOR

International Nuclear Information System (INIS)

Dippel, K.H.; Finken, K.H.; Guthrie, S.E.; Malinowski, M.E.; Pontau, A.E.; Campbell, G.A.; Goebel, D.M.; Conn, R.W.

1985-01-01

The ALT-I pump limiter is used to control hydrogen fluxes from the TEXTOR tokamak. The performance of two different modules, the open fixed geometry (FG) and the closed variable geometry (VG) is discussed. In unpumped scoop limiter operation, the pressure in the ALT-I chamber increases to 3x10 -4 torr(FG) and 2x10 -3 torr(VG). With pumping, the fraction of particles incident on the neutralizer plate that is removed is 25-50%(FG) and 50-80%(VG). These removed particles are estimated to be 2-4(8)%(FG) and 6-13%(VG) of the total plasma outflux (Nsub(e)/tausub(p)). The collection of helium from the plasma using the FG module is approximately half as effective as hydrogen collection. The higher particle removal efficiency for the VG module is attributed to lower neutral backstreaming. (author)

Control and RF-transmission in the ECW system on TEXTOR-94

International Nuclear Information System (INIS)

Dobbe, N.J.; Sterk, A.B.; Kruisbergen, R.P.J.J.M.; Kruyt, O.G.; Bestebreurtje, M.E.; Prins, P.R.; Hoekzema, J.A.; Grift, A.F. van der; Elzendoorn, B.S.Q.

2001-01-01

A real-time and multitasking control system has been developed for the new ECW system on the TEXTOR tokamak. It allows the system to be remotely controlled by client/server application. A quasi-optical transmission line has been installed which uses confocal mirrors and can be used for different frequencies (>100 GHz). It is suitable for transmission of up to two RF beams from different sources to the plasma. The launcher is mounted in a main horizontal port and injects a focused beam with a spot size of 2 cm (at 110 GHz) near the plasma axis. The launcher is steerable independently in the toroidal and poloidal directions

Control and RF-transmission in the ECW system on TEXTOR-94

Energy Technology Data Exchange (ETDEWEB)

Dobbe, N.J.; Sterk, A.B. E-mail: sterk@rijnh.nl; Kruisbergen, R.P.J.J.M.; Kruyt, O.G.; Bestebreurtje, M.E.; Prins, P.R.; Hoekzema, J.A.; Grift, A.F. van der; Elzendoorn, B.S.Q

2001-10-01

A real-time and multitasking control system has been developed for the new ECW system on the TEXTOR tokamak. It allows the system to be remotely controlled by client/server application. A quasi-optical transmission line has been installed which uses confocal mirrors and can be used for different frequencies (>100 GHz). It is suitable for transmission of up to two RF beams from different sources to the plasma. The launcher is mounted in a main horizontal port and injects a focused beam with a spot size of 2 cm (at 110 GHz) near the plasma axis. The launcher is steerable independently in the toroidal and poloidal directions.

Investigations of radial electric field and global circulation layer in limiter tokamaks

International Nuclear Information System (INIS)

Zagorski, R.; Gerhauser, H.; Lehnen, M.; Loarer, T.

2002-01-01

An updated version of the 2D multifluid code TECXY is used to study the radial electric field structure and the appearance of a global circulation layer (GCL) inside the separatrix of the limiter tokamaks TEXTOR-94 and Tore-Supra-CIEL. The dependence of the driving forces on device geometry, limiter position, magnetic field orientation, impurity content and other parameters is investigated. The centrifugal force in the vicinity of the limiter head always determines the direction of the poloidal velocity in the GCL. There is good agreement with experimentally measured profiles of the poloidal velocity at the TEXTOR low field side. (orig.)

Fast-ion redistribution due to sawtooth crash in the TEXTOR tokamak measured by collective Thomson scattering

DEFF Research Database (Denmark)

Nielsen, Stefan Kragh; Bindslev, Henrik; Salewski, Mirko

2010-01-01

Here we present collective Thomson scattering measurements of 1D fast-ion velocity distribution functions in neutral beam heated TEXTOR plasmas with sawtooth oscillations. Up to 50% of the fast ions in the centre are redistributed as a consequence of a sawtooth crash. We resolve various directions...

Runaway-ripple interaction in Tokamaks

International Nuclear Information System (INIS)

Laurent, L.; Rax, J.M.

1989-08-01

Two approaches of the interaction between runaway electrons and the ripple field, in tokamaks, are discussed. The first approach considers the resonance effect as an intense cyclotron heating of the electrons, by the ripple field, in the guiding center frame of the fast particles. In the second approach, an Hamiltonian formalism is used. A criterion for the onset of chaotic behavior and the results are given. A new universal instability of the runaway population in tokamak configuration is found. When combined with cyclotron losses one of its major consequence is to act as an effective slowing down mechanism preventing the free fall acceleration toward the synchrotron limit. This configuration allows the explanation of some experimental results of Tore Supra and Textor

2-D Imaging of Electron Temperature in Tokamak Plasmas

International Nuclear Information System (INIS)

Munsat, T.; Mazzucato, E.; Park, H.; Domier, C.W.; Johnson, M.; Luhmann, N.C. Jr.; Wang, J.; Xia, Z.; Classen, I.G.J.; Donne, A.J.H.; Pol, M.J. van de

2004-01-01

By taking advantage of recent developments in millimeter wave imaging technology, an Electron Cyclotron Emission Imaging (ECEI) instrument, capable of simultaneously measuring 128 channels of localized electron temperature over a 2-D map in the poloidal plane, has been developed for the TEXTOR tokamak. Data from the new instrument, detailing the MHD activity associated with a sawtooth crash, is presented

Langmuir probe measurements in the TEXTOR tokamak during ALT-I pump limiter experiments

International Nuclear Information System (INIS)

Goebel, D.M.; Campbell, G.A.; Conn, R.W.; Leung, W.K.; Dippel, K.H.; Finken, K.H.; Thomas, G.J.; Pontau, A.E.

1986-04-01

Langmuir probes have been used to characterize the edge plasma of the TEXTOR tokamak and measure the parameters of the plasma incident on the ALT-I pump limiter during ohmic and ICRH heating. Probes mounted directly on the ALT limiter, and a scanning probe located 90 0 toroidally from the limiter, provide data for the evaluation of pump limiter performance and its effect on the edge plasma. The edge plasma is characterized by density and flux e-folding lengths of about 1.8cm when ALT is the main limiter. These scrape-off lengths do not vary significantly as ALT is moved between the normal 42-46cm minor radii, but increase to over 2.2cm when ALT is inserted to 40cm. The flux to probes at a fixed position in the limiter shadow varies by less than 25% for core density changes of a factor of five. This suggests that the global particle confinement time tau/sub p/, scales as the core density. Estimates from the probes indicate that tau/sub p/ is on the order of the energy confinement time, tau/sub E/. The edge electron temperature, T/sub e/, typically decreases by a factor of two when the core density is raised from 1 to 4 x 10 13 cm -3 . The T/sub e/ profile is essentially flat in the limiter shadow, with values of 10-25 eV depending on the core plasma density and ICRH power. ICRH heating increases the electron temperature and flux in proportion to the coupled power. With ALT as the primary limiter and no direct shadowing, the ion side receives 2 to 3 times the flux of the electron side during both ohmic and ICRH heating. The edge plasma is not directly modified by pump limiter operation, but changes with the core plasma density as particle removal lowers the recycling of neutrals in the boundary

Wall reflection modeling for charge exchange recombination spectroscopy (CXRS) measurements on Textor and ITER

International Nuclear Information System (INIS)

Banerjee, Santanu; Vasu, P; Von Hellermann, M; Jaspers, R J E

2010-01-01

Contamination of optical signals by reflections from the tokamak vessel wall is a matter of great concern. For machines such as ITER and future reactors, where the vessel wall will be predominantly metallic, this is potentially a risk factor for quantitative optical emission spectroscopy. This is, in particular, the case when bremsstrahlung continuum radiation from the bulk plasma is used as a common reference light source for the cross-calibration of visible spectroscopy. In this paper the reflected contribution to the continuum level in Textor and ITER has been estimated for the detection channels meant for charge exchange recombination spectroscopy (CXRS). A model assuming diffuse reflection has been developed for the bremsstrahlung which is a much extended source. Based on this model, it is shown that in the case of ITER upper port 3, a wall with a moderate reflectivity of 20% leads to the wall reflected fraction being as high as 55-60% of the weak signals in the edge channels. In contrast, a complete bidirectional reflectance distribution function (BRDF) based model has been developed in order to estimate the reflections from more localized sources like the charge exchange (CX) emission from a neutral beam in tokamaks. The largest signal contamination of ∼15% is seen in the core CX channels, where the true CX signal level is much lower than that in the edge channels. Similar values are obtained for Textor also. These results indicate that the contributions from wall reflections may be large enough to significantly distort the overall spectral features of CX data, warranting an analysis at different wavelengths.

Wall reflection modeling for charge exchange recombination spectroscopy (CXRS) measurements on Textor and ITER

Energy Technology Data Exchange (ETDEWEB)

Banerjee, Santanu; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382 428, Gujarat (India); Von Hellermann, M [FOM Institute for Plasma Physics, Rijnhuizen (Netherlands); Jaspers, R J E, E-mail: sbanerje@ipr.res.i [Applied Physics Department, Eindhoven University of Technology, Eindhoven (Netherlands)

2010-12-15

Contamination of optical signals by reflections from the tokamak vessel wall is a matter of great concern. For machines such as ITER and future reactors, where the vessel wall will be predominantly metallic, this is potentially a risk factor for quantitative optical emission spectroscopy. This is, in particular, the case when bremsstrahlung continuum radiation from the bulk plasma is used as a common reference light source for the cross-calibration of visible spectroscopy. In this paper the reflected contribution to the continuum level in Textor and ITER has been estimated for the detection channels meant for charge exchange recombination spectroscopy (CXRS). A model assuming diffuse reflection has been developed for the bremsstrahlung which is a much extended source. Based on this model, it is shown that in the case of ITER upper port 3, a wall with a moderate reflectivity of 20% leads to the wall reflected fraction being as high as 55-60% of the weak signals in the edge channels. In contrast, a complete bidirectional reflectance distribution function (BRDF) based model has been developed in order to estimate the reflections from more localized sources like the charge exchange (CX) emission from a neutral beam in tokamaks. The largest signal contamination of {approx}15% is seen in the core CX channels, where the true CX signal level is much lower than that in the edge channels. Similar values are obtained for Textor also. These results indicate that the contributions from wall reflections may be large enough to significantly distort the overall spectral features of CX data, warranting an analysis at different wavelengths.

Non-axisymmetric SOL-transport study for tokamaks and stellarators

International Nuclear Information System (INIS)

Sardei, F.; Feng, Y.; Kisslinger, J.; Grigull, P.; Kobayashi, M.; Harting, D.; Reiter, D.; Federici, G.; Loarte, A.

2007-01-01

The paper addresses basic features of non-axisymmetric edge transport induced in tokamaks by local limiters or external magnetic perturbations and in low-shear stellarators by the presence of edge magnetic islands. 3D simulations and, if available for comparison, experimental results are presented and discussed for three devices, ITER during start-up operation, TEXTOR-DED and W7-AS, having edge topologies totally different from each other. The modeling is performed with the EMC3/EIRENE code, which treats self-consistently plasma, neutral and impurity transport in a general 3D scrape-off layer (SOL) with arbitrarily complex geometry of magnetic configuration and plasma-facing components. Shown are code predictions of the power load on the ITER start-up limiters as well as modeling results on the transport in the TEXTOR-DED stochastic edge and on the physics of stable detachment in W7-AS. Experimental observations confirming the code simulations are referenced for both TEXTOR-DED and W7-AS, a direct comparison between modeling and experimental results is shown for W7-AS

Effect of heating on the suppression of tearing modes in tokamaks.

Science.gov (United States)

Classen, I G J; Westerhof, E; Domier, C W; Donné, A J H; Jaspers, R J E; Luhmann, N C; Park, H K; van de Pol, M J; Spakman, G W; Jakubowski, M W

2007-01-19

The suppression of (neoclassical) tearing modes is of great importance for the success of future fusion reactors like ITER. Electron cyclotron waves can suppress islands, both by driving noninductive current in the island region and by heating the island, causing a perturbation to the Ohmic plasma current. This Letter reports on experiments on the TEXTOR tokamak, investigating the effect of heating, which is usually neglected. The unique set of tools available on TEXTOR, notably the dynamic ergodic divertor to create islands with a fully known driving term, and the electron cyclotron emission imaging diagnostic to provide detailed 2D electron temperature information, enables a detailed study of the suppression process and a comparison with theory.

An investigation of coupling of the internal kink mode to error field correction coils in tokamaks

International Nuclear Information System (INIS)

Lazarus, E.A.

2013-01-01

The coupling of the internal kink to an external m/n = 1/1 perturbation is studied for profiles that are known to result in a saturated internal kink in the limit of a cylindrical tokamak. It is found from three-dimensional equilibrium calculations that, for A ≈ 30 circular plasmas and A ≈ 3 elliptical shapes, this coupling of the boundary perturbation to the internal kink is strong; i.e., the amplitude of the m/n = 1/1 structure at q = 1 is large compared with the amplitude applied at the plasma boundary. Evidence suggests that this saturated internal kink, resulting from small field errors, is an explanation for the TEXTOR and JET measurements of q 0 remaining well below unity throughout the sawtooth cycle, as well as the distinction between sawtooth effects on the q-profile observed in TEXTOR and DIII-D. It is proposed that this excitation, which could readily be applied with error field correction coils, be explored as a mechanism for controlling sawtooth amplitudes in high-performance tokamak discharges. This result is then combined with other recent tokamak results to propose an L-mode approach to fusion in tokamaks. (paper)

Remote Participation tools at TEXTOR

International Nuclear Information System (INIS)

Kraemer-Flecken, A.; Krom, J.; Landgraf, B.; Lambertz, H.T.

2010-01-01

Remote Participation is a widely used term with different meanings. In the fusion community it has gained an increasing interest with the shut down of small experiments and participation of associations in larger experiments. Also at TEXTOR Remote Participation becomes more and more important with an increasing number of collaborations. At TEXTOR we differentiate between active and passive remote experiment participation. In addition potential users of TEXTOR like to be involved in the experiment preparation phase where the experiment schedule and the availability of diagnostic systems is discussed as well. After an experiment joint groups of users like to share the results and communicate with each other. The final step in publishing the results is also made more transparent for the users in a twofold process. Using a web based pinboard to spread the publication within the user community allows an extensive and early discussion of the results.

Turbulent transport reduction by E x B velocity shear during edge plasma biasing in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Van Oost, G. [Dept. of Applied Physics, Ghent Univ., Ghent (Belgium); Adamek, J.; Antoni, V.; Balan, P.; Boedo, J.A.; Devynck, P.; Duran, I.; Eliseev, L.; Gunn, J.P.; Hron, M.; Ionita, C.; Jachmich, S.; Kirnev, G.S.; Martines, E.; Melnikov, A.; Peleman, P.; Schrittwieser, R.; Silva, C.; Stoeckel, J.; Tendler, M.; Varandas, C.; Van Schoor, M.; Vershkov, V.; Weynants, R.R.

2004-07-01

Experiments in the tokamaks TEXTOR, CASTOR, T-10 and ISTTOK have provided new and complementary evidence on the physics of the universal mechanism of E x B velocity shear stabilization of turbulence, concomitant transport barrier formation and radial conductivity by using various edge biasing techniques. (orig.)

Relativistic runaway electrons in tokamak plasmas

International Nuclear Information System (INIS)

Jaspers, R.E.

1995-01-01

Runaway electrons are inherently present in a tokamak, in which an electric field is applied to drive a toroidal current. The experimental work is performed in the tokamak TEXTOR. Here runaway electrons can acquire energies of up to 30 MeV. The runaway electrons are studied by measuring their synchrotron radiation, which is emitted in the infrared wavelength range. The studies presented are unique in the sense that they are the first ones in tokamak research to employ this radiation. Hitherto, studies of runaway electrons revealed information about their loss in the edge of the discharge. The behaviour of confined runaways was still a terra incognita. The measurement of the synchrotron radiation allows a direct observation of the behaviour of runaway electrons in the hot core of the plasma. Information on the energy, the number and the momentum distribution of the runaway electrons is obtained. The production rate of the runaway electrons, their transport and the runaway interaction with plasma waves are studied. (orig./HP)

First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

DEFF Research Database (Denmark)

Bin, W.; Bruschi, A.; D'Arcangelo, O.

2015-01-01

Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin i...

Design and construction of the movable limiters for holding the plasma in position in the nuclear fusion experiment TEXTOR

International Nuclear Information System (INIS)

Butzek, D.A.; Derichs, K.

1983-11-01

The nuclear fusion experiment TEXTOR (Tokamak Experiment for Technology Oriented Research) has been constructed for investigation of plasma-wall-interaction. The plasma is generated inside a torus-shaped vacuum vessel. In addition to the magnetic fields mechanical limiters are provided to hold the plasma in position. The limiter scheme of textor consists of main limiters and reference limiters (measuring limiters). Main and reference limiters are mounted in different cross sections of the torus. The main limiters are movable during the plasma discharge while the reference limiters are kept fixed. They are adjustable. Thus, by moving the main limiters, the reference limiters can be exposed to different thermal loads during the discharge. Exposing the reference limiters to the plasma, first results have been obtained concerning the scrape off layer: thickness, fluxes of hydrogen and chromium through this layer. The limiter scheme, the final design and construction of the limiters and the first phase of operation are described in this report. (orig.) [de

Electron transport in the plasma edge with rotating resonant magnetic perturbations at the TEXTOR tokamak

International Nuclear Information System (INIS)

Stoschus, Henning

2011-01-01

Small three-dimensional (3D) magnetic perturbations can be used as a tool to control the edge plasma parameters in magnetically confined plasmas in high confinement mode (''H-mode'') to suppress edge instabilities inherent to this regime, the Edge Localized Modes (ELMs). In this work, the impact of rotating 3D resonant magnetic perturbation (RMP) fields on the edge plasma structure characterized by electron density and temperature fields is investigated. We study a low confinement (L-mode) edge plasma (r/a>0.9) with high resistivity (edge electron collisionality ν * e >4) at the TEXTOR tokamak. The plasma structure in the plasma edge is measured by a set of high resolution diagnostics: a fast CCD camera (Δt=20 μs) is set up in order to visualize the plasma structure in terms of electron density variations. A supersonic helium beam diagnostic is established as standard diagnostic at TEXTOR to measure electron density n e and temperature T e with high spatial (Δr=2 mm) and temporal resolution (Δt=20 μs). The measured plasma structure is compared to modeling results from the fluid plasma and kinetic neutral transport code EMC3-EIRENE. A sequence of five new observations is discussed: (1) Imaging of electron density variations in the plasma edge shows that a fast rotating RMP field imposes an edge plasma structure, which rotates with the external RMP rotation frequency of vertical stroke ν RMP vertical stroke =1 kHz. (2) Measurements of the electron density and temperature provide strong experimental evidence that in the far edge a rotating 3D scrape-off layer (SOL) exists with helical exhaust channels to the plasma wall components. (3) Radially inward, the plasma structure at the next rational flux surface is found to depend on the relative rotation between external RMP field and intrinsic plasma rotation. For low relative rotation the plasma structure is dominated by a particle and energy loss along open magnetic field lines to the wall components. For high

Electron transport in the plasma edge with rotating resonant magnetic perturbations at the TEXTOR tokamak

Energy Technology Data Exchange (ETDEWEB)

Stoschus, Henning

2011-10-13

Small three-dimensional (3D) magnetic perturbations can be used as a tool to control the edge plasma parameters in magnetically confined plasmas in high confinement mode (''H-mode'') to suppress edge instabilities inherent to this regime, the Edge Localized Modes (ELMs). In this work, the impact of rotating 3D resonant magnetic perturbation (RMP) fields on the edge plasma structure characterized by electron density and temperature fields is investigated. We study a low confinement (L-mode) edge plasma (r/a>0.9) with high resistivity (edge electron collisionality {nu}{sup *}{sub e}>4) at the TEXTOR tokamak. The plasma structure in the plasma edge is measured by a set of high resolution diagnostics: a fast CCD camera ({delta}t=20 {mu}s) is set up in order to visualize the plasma structure in terms of electron density variations. A supersonic helium beam diagnostic is established as standard diagnostic at TEXTOR to measure electron density n{sub e} and temperature T{sub e} with high spatial ({delta}r=2 mm) and temporal resolution ({delta}t=20 {mu}s). The measured plasma structure is compared to modeling results from the fluid plasma and kinetic neutral transport code EMC3-EIRENE. A sequence of five new observations is discussed: (1) Imaging of electron density variations in the plasma edge shows that a fast rotating RMP field imposes an edge plasma structure, which rotates with the external RMP rotation frequency of vertical stroke {nu}{sub RMP} vertical stroke =1 kHz. (2) Measurements of the electron density and temperature provide strong experimental evidence that in the far edge a rotating 3D scrape-off layer (SOL) exists with helical exhaust channels to the plasma wall components. (3) Radially inward, the plasma structure at the next rational flux surface is found to depend on the relative rotation between external RMP field and intrinsic plasma rotation. For low relative rotation the plasma structure is dominated by a particle and energy loss

Boronization in TEXTOR

International Nuclear Information System (INIS)

Winter, J.; Esser, H.G.; Koenen, L.; Reimer, H.; Seggern, J. v.; Schlueter, J.; Waelbroeck, F.; Wienhold, P.; Veprek, S.

1989-01-01

The liner and limiters of TEXTOR have been coated in situ with a boron containing carbon film using a RG discharge in a throughflow of 0.8 He + 0.1 B 2 H 6 + 0.1 CH 4 . The average film thickness was 30-50 nm, the ratio of boron and carbon in the layer was about 1:1 according to Auger Electron Spectroscopy. Subsequent tokamak discharges are characterized by a small fraction of radiated power ( eff lower than 1.2 are derived from conductivity measurements. The most prominent change in the impurity concentration compared to good conditions in a carbonized surrounding is measured for oxygen. The value OVI/anti n e of the OVI intensity normalized to the averaged plasma density anti n e decreases by more than a factor of four. The decrease in the oxygen content manifests itself also as a reduction of the CO and CO 2 partial pressures measured during and after the discharge with a sniffer probe. The carbon levels are reduced by a factor of about two as measured by the normalized intensity CII/anti n e of the CII line and via the ratio of the C fluxes and deuterium fluxed measured at the limiter (CI/D α ). The wall shows a pronounced sorption of hydrogen from the plasma, easing the density control and the establishment of low recycling conditions. The beneficial conditions did not show a significant deterioration during more than 200 discharges, including numerous shots at ICRH power levels >2 MW. (orig.)

Temporal evolution of confined fast-ion velocity distributions measured by collective Thomson scattering in TEXTOR

DEFF Research Database (Denmark)

Nielsen, Stefan Kragh; Bindslev, Henrik; Porte, L.

2008-01-01

reported [Bindslev , Phys. Rev. Lett. 97, 205005 2006]. Here we extend the discussion of these results which were obtained at the TEXTOR tokamak. The fast ions are generated by neutral-beam injection and ion-cyclotron resonance heating. The CTS system uses 100-150 kW of 110-GHz gyrotron probing radiation......Fast ions created in the fusion processes will provide up to 70% of the heating in ITER. To optimize heating and current drive in magnetically confined plasmas insight into fast-ion dynamics is important. First measurements of such dynamics by collective Thomson scattering (CTS) were recently...... of the velocity distribution after turnoff of the ion heating. These results are in close agreement with numerical simulations....

Preliminary design analysis of the ALT-II limiter for TEXTOR

International Nuclear Information System (INIS)

Koski, J.A.; Boyd, R.D.; Kempka, S.M.; Romig, A.D. Jr.; Smith, M.F.; Watson, R.D.; Whitley, J.B.; Conn, R.W.; Grotz, S.P.

1984-01-01

Installation of a large toroidal belt pump limiter, Advanced Limiter Test II (ALT-II), on the TEXTOR tokamak at Juelich, FRG is anticipated for early 1986. This paper discusses the preliminary mechanical design and materials considerations undertaken as part of the feasibility study phase for ALT-II. Since the actively cooled limiter blade is the component in direct contact with the plasma edge, and thus subject to the severe plasma environment, most preliminary design efforts have concentrated on analysis of the blade. The screening process which led to the recommended preliminary design consisting of a dispersion strenghthened copper or OFHC copper cover plate over an austenitic stainless steel base plate is discussed. A 1 to 3 mm thick low atomic number coating consisting of a graded plasma-sprayed Silicon Carbide-Aluminium composite is recommended subject to further experiment and evaluation. Thermal-hydraulic and stress analyses of the limiter blade are also discussed. (orig.)

Study of tokamaks carbon deposits after heat treatment

International Nuclear Information System (INIS)

Richou, M.; Martin, C.; Roubin, P.; Delhaes, P.; Couzi, M.; Brosset, C.; Pegourie, B.

2006-01-01

One of the most important problem of tokamak is the interaction plasma-wall. The wall component is the graphite. Meanwhile it is submitted to erosion phenomena, deposition and co-deposition with the hydrogen. This carbon deposits have been studied and show an oval shape. In order to obtain more information on the structure and the growth of these deposits, the authors heated them till 2500 C. Raman spectroscopy, transmission microscopy, magnetic and density measurements have been realized and compared for two types of samples: from Tore Supra and from Textor. (A.L.B.)

Modular pump limiter systems for large tokamaks

International Nuclear Information System (INIS)

Uckan, T.; Klepper, C.C.; Mioduszewski, P.K.; McGrath, R.T.

1987-09-01

Long-pulse (>10-s) operation of large tokamaks with high-power (>10-MW) heating and extensive external fueling will require correspondingly efficient particle exhaust for density control. A pump limiter can provide the needed exhaust capability by removing a small percentage of the particles, which would otherwise be recycled. Single pump limiter modules have been operated successfully on ISX-B, PDX, TEXTOR, and PLT. An axisymmetric pump limiter is now being installed and will be studied in TEXTOR. A third type of pump limiter is a system that consists of several modules and exhibits performance different from that of a single module. To take advantage of the flexibility of a modular pump limiter system in a high-power, long-pulse device, the power load must be distributed among a number of modules. Because each added module changes the performance of all the others, a set of design criteria must be defined for the overall limiter system. The design parameters for the modules are then determined from the system requirements for particle and power removal. Design criteria and parameters are presented, and the impact on module design of the state of the art in engineering technology is discussed. The relationship between modules are considered from the standpoint of flux coverage and shadowing effects. The results are applied to the Tore Supra tokamak. A preliminary conceptual design for the Tore Supra pump limiter system is discussed, and the design parameters of the limiter modules are presented. 21 refs., 12 figs

Measurement of Dα sources for particle confinement time determination in TEXTOR

International Nuclear Information System (INIS)

Gray, D.S.; Boedo, J.A.; Conn, R.W.; Finken, K.H.; Mank, G.; Pospieszczyk, A.; Samm, U.

1993-01-01

An important quantity in the study of tokamak discharges is the global particle confinement time, defined for each ionic species i by the equation below, where N i is the total population of the species in the plasma and S i is the source rate (ionization rate) of the species: τ pi N i /(S i - dN i /dt). Of particular significance is the confinement time of the main plasma component, deuterium; here, in most cases of interest, the time derivative is negligible and the confinement time is given by N/S. The deuterium content N can be estimated from the electron content, measured by interferometry, if Z eff is known. A common method of estimating the fueling rate S is to measure the emission of D α light from recycling neutrals in the plasma boundary, since collisional-radiative modeling has shown that, for plasma conditions typical in the tokamak edge, the rate of ionization of D atoms and the rate of emission of D α photons are related by a factor that varies only weakly with electron density and temperature. This paper describes the use of a CCD video camera at TEXTOR for the purpose of spatially resolving the D α light in order to measure more accurately the total emission so that τ p can be determined reliably. (author) 5 refs., 5 figs

Charge exchange spectroscopy as a fast ion diagnostic on TEXTOR

International Nuclear Information System (INIS)

Delabie, E.; Jaspers, R. J. E.; Hellermann, M. G. von; Nielsen, S. K.; Marchuk, O.

2008-01-01

An upgraded charge exchange spectroscopy diagnostic has been taken into operation at the TEXTOR tokamak. The angles of the viewing lines with the toroidal magnetic field are close to the pitch angles at birth of fast ions injected by one of the neutral beam injectors. Using another neutral beam for active spectroscopy, injected counter the direction in which fast ions injected by the first beam are circulating, we can simultaneously measure a fast ion tail on the blue wing of the D α spectrum while the beam emission spectrum is Doppler shifted to the red wing. An analysis combining the two parts of the spectrum offers possibilities to improve the accuracy of the absolute (fast) ion density profiles. Fast beam modulation or passive viewing lines cannot be used for background subtraction on this diagnostic setup and therefore the background has to be modeled and fitted to the data together with a spectral model for the slowing down feature. The analysis of the fast ion D α spectrum obtained with the new diagnostic is discussed.

Transport through dissipative trapped electron mode and toroidal ion temperature gradient mode in TEXTOR

International Nuclear Information System (INIS)

Rogister, A.; Hasselberg, G.; Waelbroeck, F.; Weiland, J.

1987-12-01

A self-consistent transport code is used to evaluate how plasma confinement in tokamaks is influenced by the microturbulent fields which are excited by the dissipative trapped electron (DTE) instability. As shown previously, the saturation theory on which the code is based has been developed from first principles. The toroidal coupling resulting from the ion magnetic drifts is neglected; arguments are presented to justify this approximation. The numerical results reproduce well the neo-Alcator scaling law observed experimentally - e.g. in TEXTOR - in non detached ohmic discharges, the confinement degradation which results when auxiliary heating is applied, as well as a large number of other experimental observations. We also assess the possible impact of the toroidal ion temperature gradient mode on energy confinement by estimating the ion thermal flux with the help of the mixing length approximation. (orig./GG)

Detritiation of tiles from tokamaks by laser cleaning

International Nuclear Information System (INIS)

Coad, J. Paul; Widdowson, Anna; Farcage, Daniel; Semerok, Alexander; Thro, P.-Y.; Likonen, Jari; Renvall, Tommi

2007-01-01

Laser ablation has been used to clean surfaces or to decontaminate hot cells by removing paint, and has been tested on deposited carbon layers from the TEXTOR tokamak. This paper reports on successful trials in the Beryllium Handling Facility of a pulsed laser cleaning system to remove H-isotope containing carbon deposits on tiles from the JET tokamak. The laser beam is rastered over the surface of the tiles to remove the deposit. Two types of JET carbon-fibre composite (CFC) tiles were treated. The first was covered with carbon-based deposits up to 300 μm thick with high H-isotope content, the other was covered with a mixed Be/C film ∼ 50 microns thick. One scan of the laser was sufficient to completely change the appearance and expose the fibre planes. From cross-sectional micrographs, it was found that overall three scans provided the most effective settings for complete film removal. An area 250 cm 2 of the second tile was cleaned in 20 minutes, clearly demonstrating the efficiency of laser cleaning for the removal of tokamak deposits such as likely to occur in ITER. (authors)

Manufacturing and testing of actively cooled test limiters for TEXTOR made of the brazed joint SEPCARB-N11/TZM

International Nuclear Information System (INIS)

Hohenauer, W.; Bolt, H.; Koppitz, T.; Linke, J.; Lison, R.; You, J.H.; Nickel, H.

1998-01-01

To investigate the erosion and redepositon phenomena of fusion-related materials under stationary conditions, actively cooled test limiters were developed for TEXTOR (Tokamak Experiment for Technology Orientated Research). They allow experiments under stationary conditions within the plasma pulse length of 10 s. Heat loads of typically 10 MW m<-2 are removed by pressurised water: volume flow is 10 m 3 h -1, pressure 15 bar and the minimum coefficient of heat transfer is about 75000 W m-2 K. Prototype limiters were built as brazed composites of a C/C material (SEPCARB-N11) and a TZM substrate. The samples were successfully tested in screening tests in the ion beam facility MARION (Material Research Ion Beam Test Facility) with hydrogen beams. Maximum heat loads of up to 22 MW m<-2 were applied without any failure of the cooling system. Steady state of the surface temperature was measured within 5 s. An advanced brazing technique enabled the joining of hemispherically shaped C/C shells to a TZM heat sink without failure. An optimised test limiter was tested in TEXTOR. Analytical and numerical models describing the effects of the heat load distribution, spatial temperatures and stresses were experimentally verified. (orig.)

Runaway snakes in TEXTOR-94

NARCIS (Netherlands)

Entrop, I.; R. Jaspers,; Cardozo, N. J. L.; Finken, K.H.

1999-01-01

Observations of a runaway beam confined in an island-like structure, a so-called runaway snake, are reported. The observations are made in TEXTOR-94 by measurement of synchrotron radiation emitted by these runaways. A full poloidal View allows for the study of the synchrotron pattern of the snake to

Calculation of poloidal rotation in the edge plasma of limiter tokamaks

International Nuclear Information System (INIS)

Gerhauser, H.; Claassen, H.A.

1987-05-01

The existing 2-d two-fluid code for computing the plasma profiles in the scrape-off layer of limiter tokamaks has been further developed to include the effect of poloidal rotation in the basic equations. This rotation is produced by radial electric fields which arise in the limiter shadow due to radial gradients in the Langmuir sheath potential in front of the limiter. As a consequence slight deviations from ambipolar motion must occur. A strong increase of rotation near the separatrix is connected with an electric current circuit closed via the limiter edge. The 2-d profiles of all relevant quantities are calculated and discussed for TEXTOR-typical parameters including also the effect of limiter recycled neutrals. The results agree well with the known experimental evidence on poloidal rotation and should be transferable to all limiter tokamaks. (orig.)

Overview of magnetic structure induced by the TEXTOR-DED and the related transport

International Nuclear Information System (INIS)

Abdullaev, S.S.; Finken, K.H.; Kobayashi, M.; Reiser, D.; Reiter, D.; Jakubowski, M.W.; Runov, A.M.

2003-01-01

The Dynamic Ergodic Divertor (DED), a new concept of the ergodic divertor, is presently installed for the TEXTOR tokamak. Beside the conventional ergodic divertor operation the DED also permits the operation with a rotating magnetic field which allows, in particular, to broaden the heat deposition pattern on the divertor plates. Since its first proposal of the DED in 1996 the structure of magnetic field, especially, the onset of ergodic zone of field lines and related transport in the DED operation has been extensively studied using different theoretical and numerical methods. New methods to study the magnetic field, in particular, the field line mapping have been developed. The presentation gives the overview of the studies on the structure of magnetic field in the DED, the formation of the ergodic and laminar zones of field lines at the plasma edge. It also includes studies on the modelling efforts of the transport of heat and particles in the ergodic and laminar zones. (author)

Elastic-plastic cyclic deformation of the TEXTOR 94 modified liner under conditions of heating and plasma disruption

International Nuclear Information System (INIS)

Bohn, F.H.; Czymek, G.; Giesen, B.; Bondarchuk, E.; Doinikov, N.; Kozhukhovskaja, N.; Panin, A.

2001-01-01

The present liner of the TEXTOR 94 tokamak installed inside the vacuum vessel represents the thin toroidal shell that is rested on the vessel inner surface. In order to integrate the dynamic ergodic divertor into the tokamak the liner design has been drastically changed. The 120 deg. sector of the liner shell facing the ergodic coils system is removed and some additional holes in the liner are provisioned. This demands a new liner supporting system allowing for the liner thermal expansion and taking the electromagnetic load occurring in the liner during plasma disruption. The cyclic elasto-plastic deformations of the liner caused by the electromagnetic forces and temperature rise have been studied. It is shown that the local plastic deformations occur in the liner elements after the first heating and electromagnetic loading. The most thermal stresses take place in the reinforcing structures around the holes because of the thermal expansion difference of the inconel shell and the steel reinforcements. These stresses are coupled with the bending stress due to the electromagnetic loading. Subsequent repetitive loading does not lead to any significant increment of the plastic deformation. After the materials' hardening the structure cyclically works mostly in the elastic range

BRIEF COMMUNICATION: Fast-ion redistribution due to sawtooth crash in the TEXTOR tokamak measured by collective Thomson scattering

Science.gov (United States)

Nielsen, S. K.; Bindslev, H.; Salewski, M.; Bürger, A.; Delabie, E.; Furtula, V.; Kantor, M.; Korsholm, S. B.; Leipold, F.; Meo, F.; Michelsen, P. K.; Moseev, D.; Oosterbeek, J. W.; Stejner, M.; Westerhof, E.; Woskov, P.; TEXTOR Team

2010-09-01

Here we present collective Thomson scattering measurements of 1D fast-ion velocity distribution functions in neutral beam heated TEXTOR plasmas with sawtooth oscillations. Up to 50% of the fast ions in the centre are redistributed as a consequence of a sawtooth crash. We resolve various directions to the magnetic field. The fast-ion distribution is found to be anisotropic as expected. For a resolved angle of 39° to the magnetic field we find a drop in the fast-ion distribution of 20-40%. For a resolved angle of 83° to the magnetic field the drop is no larger than 20%.

In situ deuterium inventory measurements of a-C:D layers on tungsten in TEXTOR by laser induced ablation spectroscopy

International Nuclear Information System (INIS)

Gierse, N; Brezinsek, S; Coenen, J W; Huber, A; Laengner, M; Möller, S; Nonhoff, M; Philipps, V; Pospieszczyk, A; Schweer, B; Sergienko, G; Xiao, Q; Zlobinski, M; Samm, U; Giesen, T F

2014-01-01

Laser induced ablation spectroscopy (LIAS) is a diagnostic to provide temporally and spatially resolved in situ measurements of tritium retention and material migration in order to characterize the status of the first wall in future fusion devices. In LIAS, a ns-laser pulse ablates the first nanometres of the first wall plasma-facing components into the plasma edge. The resulting line radiation by plasma excitation is observed by spectroscopy. In the case of the full ionizing plasma and with knowledge of appropriate photon efficiencies for the corresponding line emission the amount of ablated material can be measured in situ. We present the photon efficiency for the deuterium Balmer α-line resulting from ablation in TEXTOR by performing LIAS on amorphous hydrocarbon (a-C:D) layers deposited on tungsten substrate of thicknesses between 0.1 and 1.1 μm. An experimental inverse photon efficiency of [(D/(XB))] D α (EXP) a-C:D→ LIAS D =75.9±23.4 was determined. This value is a factor 5 larger than predicted values from the ADAS database for atomic injection of deuterium under TEXTOR plasma edge conditions and about twice as high, assuming normal wall recycling and release of molecular deuterium and break-up of D 2 via the molecular ion which is usually observed at the high temperature tokamak edge (T e  > 30 eV). (paper)

Plasma rotation effect on interaction of low frequency fields with plasmas at the rational surfaces in tokamaks

International Nuclear Information System (INIS)

Rondan, E.R.; Elfimov, A.G.; Galvao, R.M.O.; Pires, C.J.A.

2006-01-01

The effect of plasma rotation on low frequency (LF) field penetration, absorption and ponderomotive forces in TEXTOR and in Tokamak Chauffage Alfven Bresilien (TCABR) is investigated in the frequency band of 1-10 kHz. The LF fields are driven by the dynamic ergodic divertor in TEXTOR and the ergodic magnetic limiter in TCABR. Alfven wave mode conversion is responsible for the LF field absorption at the rational magnetic surface where q = -M/N is the integer. Analytical and numerical calculations show the maxima of the LF field absorption at the local Alfven wave resonance ω - k · U = k parallel c A , where ω and k are the frequency and the wave vector, respectively, and c A is the Alfven velocity at the rational magnetic surface q = 2, 3 in TEXTOR and TCABR. The rotation velocity U along the magnetic surfaces, taken into account in the dielectric tensor, can strongly modify the LF field and dissipated power profiles. The absorption in the local AW resonances begins to be non-symmetric in relation to the resonance surface. Calculations show that coil impedance has a maximum related to excitation of some stable (possibly Suydam) modes for waves travelling in the direction of plasma rotation

Electrical conductivity in tokamaks and extended neoclassical theory

International Nuclear Information System (INIS)

Segre, S.E.; Zanza, V.

1992-01-01

The electrical conductivity measurements reported from various tokamaks (D-III, PLT, TEXT, ASDEX, JT-60, TEXTOR, JET, TFTR) and compared with the usual neoclassical theory are here also compared with the extended neoclassical theory where the electron-electron collision rate is anomalous while the electron-ion collision rate remains Coulombian. It is found that, out of the 14 experiments considered, three are consistent with both the neoclassical and the extended neoclassical theories, four are consistent only with the extended neoclassical theory, and four are consistent with the neoclassical theory and also, within the experimental errors, not inconsistent with the extended neoclassical theory; the remaining three experiments appear to be incompatible with both theories. It is concluded that the extended neoclassical theory is in better agreement with conductivity experiments than the conventional neoclassical theory and, indeed, the extended theory is a serious candidate for explaining tokamak behaviour, since it accommodates naturally an anomalous electron thermal transport, which the conventional neoclassical theory is unable to do. (author). 31 refs, 1 fig

Status of electron temperature and density measurement with beam emission spectroscopy on thermal helium at TEXTOR

International Nuclear Information System (INIS)

Schmitz, O; Schweer, B; Pospieszczyk, A; Lehnen, M; Samm, U; Unterberg, B; Beigman, I L; Vainshtein, L A; Kantor, M; Xu, Y; Krychowiak, M

2008-01-01

Beam emission spectroscopy on thermal helium is used at the TEXTOR tokamak as a reliable method to obtain radial profiles of electron temperature T e (r, t) and electron density n e (r, t). In this paper the experimental realization of this method at TEXTOR and the status of the atomic physics employed as well as the major factors for the measurement's accuracy are evaluated. On the experimental side, the hardware specifications are described and the impact of the beam atoms on the local plasma parameters is shown to be negligible. On the modeling side the collisional-radiative model (CRM) applied to infer n e and T e from the measured He line intensities is evaluated. The role of proton and deuteron collisions and of charge exchange processes is studied with a new CRM and the impact of these so far neglected processes appears to be of minor importance. Direct comparison to Thomson scattering and fast triple probe data showed that for high densities n e > 3.5 x 10 19 m -3 the T e values deduced with the established CRM are too low. However, the new atomic data set implemented in the new CRM leads in general to higher T e values. This allows us to specify the range of reliable application of BES on thermal helium to a range of 2.0 x 10 18 e 19 m -3 and 10 eV e < 250 eV which can be extended by routine application of the new CRM.

Simultaneous Microwave Imaging System for Density and Temperature Fluctuation Measurements on TEXTOR

International Nuclear Information System (INIS)

Park, H.; Mazzucato, E.; Munsat, T.; Domier, C.W.; Johnson, M.; Luhmann, N.C. Jr.; Wang, J.; Xia, Z.; Classen, I.G.J.; Donne, A.J.H.; Pol, M.J. van de

2004-01-01

Diagnostic systems for fluctuation measurements in plasmas have, of necessity, evolved from simple 1-D systems to multi-dimensional systems due to the complexity of the MHD and turbulence physics of plasmas illustrated by advanced numerical simulations. Using the recent significant advancements in millimeter wave imaging technology, Microwave Imaging Reflectometry (MIR) and Electron Cyclotron Emission Imaging (ECEI), simultaneously measuring density and temperature fluctuations, are developed for TEXTOR. The MIR system was installed on TEXTOR and the first experiment was performed in September, 2003. Subsequent MIR campaigns have yielded poloidally resolved spectra and assessments of poloidal velocity. The new 2-D ECE Imaging system (with a total of 128 channels), installed on TEXTOR in December, 2003, successfully captured a true 2-D images of Te fluctuations of m=1 oscillation (''sawteeth'') near the q ∼ 1 surface for the first time

The Bragg-polarimeter at TEXTOR 94

International Nuclear Information System (INIS)

Herzog, O.; Weinheimer, J.; Rosmej, F.B.; Kunze, H.-J.; Bertschinger, G.; Bitter, M.; Urnov, A.

1998-01-01

In TEXTOR 94 line polarisation was measured which shows that during current steps in ohmic discharges non-thermal electrons exist. The density and the energy have to be calculated by theoretical models. A systematical analysis especially for discharges with other heating scenarios like neutral beam heating, ion cyclotron heating or combinations of both is very important and will be done. (author)

ERO-TEXTOR. 3D-Monte Carlo code for local impurity-modeling in the scrape-off-layer of TEXTOR. Version 2.0

International Nuclear Information System (INIS)

Koegler, U.; Winter, J.

1997-03-01

The ERO-TEXTOR code is described in detail. The code solves the kinetic equations of impurities in the scrape-off layer of a tokamak plasma in the vicinity of material surfaces like limiters or divertors. A relaxation time ansatz in the traced impurity limit is chosen, taking the gyro-motion of the particles into account. Since the background plasma is slightly non-maxwellian at the plasma edge higher order corrections (thermal forces) to the relaxation time ansatz are also considered. Background plasma parameters are calculated from a simple plasma model, i.e. the one dimensional continuity and momentum equations are used to derive the local electron density, the local flow velocity and the pre-sheath and sheath electric fields. Since these calculations are not done in a selfconsistent way, the measured values of electron density and temperature are used as basic input to derive the dependency of these quantities. The regarded magnetic topology is still straight and uniform. Also detailed account is given to the plasma surface interaction and the erosion/deposition processes. A linear differential equation model for multi species impact on a material surface has been developed and is used in a discrete time step approximation. External databases include the ionization rates for atomic species, molecular processes of methane and silane molecules and the sputtering and reflection yields, which are taken from binary collision calculation codes (e.g. TRIM) or from semi-empirical fits (e.g. the Bodhansky and Yamamura fits). (orig.)

Plasma edge physics in the TEXTOR tokamak with poloidal and toroidal limiters

International Nuclear Information System (INIS)

Samm, U.; Bogen, P.; Hartwig, H.; Hintz, E.; Hoethker, K.; Lie, Y.T.; Pospieszczyk, A.; Rusbueldt, D.; Schweer, B.; Yu, Y.J.

1989-01-01

Investigations of the plasma edge in TEXTOR are presented on the one hand by comparing results obtained with poloidal and toroidal limiters and on the other hand by discussing general problems of plasma edge physics which are independent of the limiter configuration. The characteristic properties of plasma flow to the different limiters are analyzed and show e.g. that the fraction of total ion flow to the limiter is much larger in the case of a toroidal limiter (80%). Density and heat flux profiles are presented which demonstrate that for both types of limiters a significant steepening of the scrape-off layer (SOL) occurs close to the limiter, leading to a small heat load e-folding length of 5-8 mm. The velocity distribution of recycled neutral hydrogen at a main limiter has been determined from the Doppler broadening of the H α line. The data clearly show that a large fraction of particles (30-50%) is reflected at the limiter surface having energies of about the sheath potential. Significant isotopic effects (H/D) concerning the plasma edge properties and the plasma core are presented and their relation to enhanced particle and energy transport in hydrogen compared to deuterium is discussed. A decrease of the cross field diffusion coefficient with increasing density can be deduced from density profile measurements in the SOL and a comparison with density fluctuations is given. The role of oxygen for impurity release is demonstrated. A new type of wall conditioning - boronization - is described, with two major improvements for quasi stationary conditions: reduction of oxygen and better density control. Best results with ICRH have been obtained under these conditions. (orig.)

First results from the dynamic ergodic divertor at TEXTOR

International Nuclear Information System (INIS)

Lehnen, M.; Abdullaev, S.S.; Biel, W.; Brezinsek, S.; Finken, K.H.; Harting, D.; Hellermann, M. von; Jakubowski, M.; Jaspers, R.; Kobayashi, M.; Koslowski, H.R.; Kraemer-Flecken, A.; Matsunaga, G.; Pospieszczyk, A.; Reiter, D.; Van Rompuy, T.; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Wolf, R.; Zimmermann, O.

2005-01-01

Experimental results from the dynamic ergodic divertor (DED) at TEXTOR are given, describing the complex structure of the edge plasma and the properties of the divertor as well as its influence on the plasma rotation

Disruption generated runaway electrons in TEXTOR and ITER

NARCIS (Netherlands)

R. Jaspers,; Cardozo, N. J. L.; Schüller, F. C.; Finken, K.H.; Grewe, T.; Mank, G.

1996-01-01

Runaway generation during a major disruption has been observed in TEXTOR. Measurements of the synchrotron radiation yielded number, energy and pitch angle of the runaways. A simple model, which assumes that the runaways take over the current density in the centre of the discharge, successfully

Experimental Investigation of Runaway Electron Generation in Textor

NARCIS (Netherlands)

R. Jaspers,; Finken, K.H.; Mank, G.; Hoenen, F.; Boedo, J. A.; Cardozo, N. J. L.; Schüller, F. C.

1993-01-01

An experimental study of the generation of runaway electrons in TEXTOR has been performed. From the infrared synchrotron radiation emitted by relativistic electrons, the number of runaway electrons can be obtained as a function of time. In low density discharges (n(e)BAR < 1 X 10(19) m-3)

Dust Studies in DIII-D and TEXTOR

International Nuclear Information System (INIS)

Rudakov, D.L.; Litnovsky, A.; West, W.P.; Yu, J.H.; Boedo, J.A.; Bray, B.D.; Brezinsek, S.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Hollmann, E.M.; Huber, A.; Hyatt, A.W.; Krasheninnikov, S.I.; Lasnier, C.J.; Moyer, R.A.; Pigarov, A.Y.; Philipps, V.; Pospieszczyk, A.; Smirnov, R.D.; Sharpe, J.P.; Solomon, W.M.; Watkins, J.G.; Wong, C.C.

2009-01-01

Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Direct heating of the dust particles by the neutral beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust. Large flakes or debris falling into the plasma may result in a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micron-size dust in plasma discharges. In DIII-D, a sample holder filled with ∼30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure (∼0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. At the given configuration of the launch, the dust did not penetrate the core plasma and only moderately perturbed the edge plasma, as evidenced by an increase of the edge carbon content.

Deposition and erosion in local shadow regions of TEXTOR-94

International Nuclear Information System (INIS)

Wienhold, P.; Mayer, M.; Kirschner, A.; Rubel, M.; Hildebrandt, D.; Schneider, W.

2001-01-01

Carbon erosion and deposition were investigated on the surface of a flat target covered with an a-C:H film and exposed for 197 s in the SOL of TEXTOR-94. The target was declined by 20 with respect to the toroidal direction and partly protected by an aluminum (3 mm) plate which created an 8 mm wide local shadow. Thickness changes were measured by colorimetry after each plasma discharge. Carbon is eroded from surface areas near the plasma edge (LCFS +1 cm) and transported into the local shadow regions. Accumulation rates up to ∼7 nm/s were found. The erosion in the local shadow regions (about -0.1 nm/s) is due to charge exchange neutrals. The observations are confirmed by ion beam analyses and by preliminary calculations with the B2-EIRENE and ERO-TEXTOR code. (orig.)

Theoretical and experimental studies of runaway electrons in the TEXTOR tokamak

International Nuclear Information System (INIS)

Abdullaev, S.S.; Finken, K.H.; Wongrach, K.; Willi, O.

2016-01-01

Theoretical and experimental studies of runaway electrons in tokamaks and their mitigations, particularly the recent studies performed by a group of the Heinrich-Heine University Duesseldorf in collaboration with the Institute of Energy and Climate Research of the Research Centre (Forschungszentrum) of Juelich are reviewed. The main topics focus on (i) runaway generation mechanisms, (ii) runaway orbits in equilibrium plasma, (iii) transport in stochastic magnetic fields, (iv) diagnostics and investigations of transport of runaway electron and their losses in low density discharges (v) runaway electrons during plasma disruptions, and (vi) runaway mitigation methods. The development of runaway diagnostics enables the measurement of runaway electrons in both the centre and edge of the plasma. The diagnostics provide an absolute runaway energy resolved measurement, the radial decay length of runaway electrons and, the structure and dynamics of runaway electron beams. The new mechanism of runaway electron formation during plasma disruptions is discussed.

Theoretical and experimental studies of runaway electrons in the TEXTOR tokamak

Energy Technology Data Exchange (ETDEWEB)

Abdullaev, S.S.; Finken, K.H.; Wongrach, K.; Willi, O.

2016-07-01

Theoretical and experimental studies of runaway electrons in tokamaks and their mitigations, particularly the recent studies performed by a group of the Heinrich-Heine University Duesseldorf in collaboration with the Institute of Energy and Climate Research of the Research Centre (Forschungszentrum) of Juelich are reviewed. The main topics focus on (i) runaway generation mechanisms, (ii) runaway orbits in equilibrium plasma, (iii) transport in stochastic magnetic fields, (iv) diagnostics and investigations of transport of runaway electron and their losses in low density discharges (v) runaway electrons during plasma disruptions, and (vi) runaway mitigation methods. The development of runaway diagnostics enables the measurement of runaway electrons in both the centre and edge of the plasma. The diagnostics provide an absolute runaway energy resolved measurement, the radial decay length of runaway electrons and, the structure and dynamics of runaway electron beams. The new mechanism of runaway electron formation during plasma disruptions is discussed.

Sawtooth stability in neutral beam heated plasmas in TEXTOR

NARCIS (Netherlands)

Chapman, I.T.; Pinches, S. D.; Koslowski, H. R.; Liang, Y.; Kramer-Flecken, A.; De Bock, M.

2008-01-01

The experimental sawtooth behaviour in neutral beam injection (NBI) heated plasmas in TEXTOR is described. It is found that the sawtooth period is minimized with a low NBI power oriented in the same direction as the plasma current. As the beam power is increased in the opposite direction to the

Dust Studies in DIII-D and TEXTOR

International Nuclear Information System (INIS)

Rudakov, D.; Litnovsky, A.; West, W.; Yu, J.; Boedo, J.; Bray, B.; Brezinsek, S.; Brooks, N.; Fenstermacher, M.; Groth, M.; Hollmann, E.; Huber, A.; Hyatt, A.; Krasheninnikov, S.; Lasnier, C.; Moyer, R.; Pigarov, A.; Philipps, V.; Pospieszezyk, A.; Smirnov, R.; Sharpe, J.; Solomon, W.; Watkins, J.; Wong, C.

2008-01-01

Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Energetic plasma disruptions produce significant amounts of dust. However, dust production by disruptions alone is insufficient to account for the estimated in-vessel dust inventory in DIII-D. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by injecting micron-size dust in plasma discharges. In DIII-D, a sample holder filled with ∼30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure (∼0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. Individual dust particles are observed moving at velocities of 10-100 m/s, predominantly in the toroidal direction, consistent with the drag force from the deuteron flow and in agreement with modeling by the 3D DustT code. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. Dust is launched either in the beginning of a discharge or at the initiation of NBI, preferentially in a direction perpendicular to the toroidal magnetic field. At the given configuration of the launch, the dust did not penetrate

Pellet fuelling into radiative improved confinement discharges in TEXTOR-94

NARCIS (Netherlands)

Hobirk, J.; Messiaen, A. M.; Finken, K.H.; Ongena, J.; Brix, M.; R. Jaspers,; Koslowski, H. R.; Kramer-Flecken, A.; Mank, G.; Rapp, J.; Telesca, G.; Unterberg, B.

2000-01-01

Normally pellet injection in strongly heated discharges leads at most to a relatively short improvement of the energy and particle confinement times. In contrast to this finding, the radiative improved (RI) mode plasma of TEXTOR-94 is a very well suited target for pellet injection: the interaction

Carbon deposition at the bottom of gaps in TEXTOR experiments

International Nuclear Information System (INIS)

Matveev, D.; Kirschner, A.; Esser, H.G.; Freisinger, M.; Kreter, A.; Van Hoey, O.; Borodin, D.; Litnovsky, A.; Wienhold, P.; Coenen, J.W.; Stoschus, H.; Philipps, V.; Brezinsek, S.; Van Oost, G.

2013-01-01

Results of a new dedicated experiment addressing the problem of impurity deposition at the bottom in gaps are presented along with modelling. A test limiter with an isolated gap was exposed to the scrape-off layer plasma in TEXTOR. The exposure was accompanied by injection of 13 C-marked methane in the vicinity of the gap. Deposition at the bottom of the gap was monitored in situ with Quartz-Microbalance diagnostics. The 13 C deposition efficiency of about 2.6 × 10 −5 was measured. Post mortem analysis of resulting deposited layers performed with SIMS and EPMA techniques yields about a factor 2 smaller value corresponding to approximately 10% contribution of the gap bottom to the total 13 C deposition in the gap. This measured contribution is effectively much smaller than observed earlier in TEXTOR, taking the difference in geometry into account, and is in reasonable agreement with modelling performed with ERO and 3D-GAPS codes

Study on sawtooth and transport in part of Japan-TEXTOR collaboration 1995

International Nuclear Information System (INIS)

Itoh, K.

1996-02-01

A collaboration programme 'physics of sawtooth and transport' has been performed in the frame work of the Japan-TEXTOR collaboration. The summary of the workshops and collaborations in 1995 is reported. (author)

Observation of self-organized criticality (SOC) behavior during edge biasing experiment on TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Xu, Y.H.; Jachmich, S.; Weynants, R.R. [Ecole Royale Militaire/Koninklijke Militaire School, Laboratory for Plasma Physics, Euratom-Belgian State Association, Brussels, Belgium, Partner in the Trilateral Euregio Cluster (Belgium)

2004-07-01

The self-organized criticality (SOC) behavior of the edge plasma transport has been investigated using the fluctuation data measured in the plasma edge and the scrape-off layer of TEXTOR tokamak before and during the edge electrode biasing experiments. In the 'non-shear' discharge phase before biasing, both the potential and density fluctuations clearly exhibit some of the characteristics associated with SOC: (1) existence of f{sup -1} power-law dependence in the frequency spectrum, (2) slowly decaying long tails in the autocorrelation function, (3) values of Hurst parameters larger than 0.5 at all the detected radial locations, (4) non-Gaussian probability density function of fluctuations and (5) radial propagation of avalanche-like events in the edge plasma area. During the biasing phase, with the generation of an edge radial electric field E{sub r} and hence a sheared E{sub r} x B flow, the local turbulence is found to be well de-correlated by the E{sub r} x B velocity shear, consistent with theoretical predictions. Nevertheless, it is concomitantly found that the Hurst parameters are substantially enhanced in the negative flow shear region and in the scrape-off layer as well, which is contrary to theoretical expectation. Implication of these observations to our understanding of plasma transport mechanisms is discussed. (authors)

Prototype tests on the ion source power supplies of the TEXTOR NI-system

International Nuclear Information System (INIS)

Goll, O.; Braunsberger, U.; Schwarz, U.

1987-01-01

The PINI ion source for the TEXTOR neutral injector is fed by a new modular transistorized power supply. All modules are located in a high voltage cage on 55 kV dc against ground. The normal operation of the injectors includes frequent grid breakdowns causing transient high voltage stresses on the ion source power supplies. These stresses must not disturb the safe operation of the power supplies. The paper describes the set up for extensive testing of a supply prototype module under the expected operating conditions. The main features of this test program are reviewed and the measures taken for a safe operation are discussed. As a result of the investigations, recommendations for the installation of the power supplies at the TEXTOR NI system are given

High Confinement and High Density with Stationary Plasma Energy and Strong Edge Radiation Cooling in Textor-94

Science.gov (United States)

Messiaen, A. M.

1996-11-01

A new discharge regime has been observed on the pumped limiter tokamak TEXTOR-94 in the presence of strong radiation cooling and for different scenarii of additional hearing. The radiated power fraction (up to 90%) is feedback controlled by the amount of Ne seeded in the edge. This regime meets many of the necessary conditions for a future fusion reactor. Energy confinement increases with increasing densities (reminiscent of the Z-mode obtained at ISX-B) and as good as ELM-free H-mode confinement (enhancement factor verus ITERH93-P up to 1.2) is obtained at high densities (up to 1.2 times the Greenwald limit) with peaked density profiles showing a peaking factor of about 2 and central density values around 10^14cm-3. In experiments where the energy content of the discharges is kept constant with an energy feedback loop acting on the amount of ICRH power, stable and stationary discharges are obtained for intervals of more than 5s, i.e. 100 times the energy confinement time or about equal to the skin resistive time, even with the cylindrical q_α as low as 2.8 β-values up to the β-limits of TEXTOR-94 are achieved (i.e. β n ≈ 2 of and β p ≈ 1.5) and the figure of merit for ignition margin f_Hqa in these discharges can be as high as 0.7. No detrimental effects of the seeded impurity on the reactivity of the plasma are observed. He removal in these discharges has also been investigated. [1] Laboratoire de Physique des Plasmas-Laboratorium voor Plasmafysica, Association "EURATOM-Belgian State", Ecole Royale Militaire-Koninklijke Militaire School, Brussels, Belgium [2] Institut für Plasmaphysik, Forschungszentrum Jülich, GmbH, Association "EURATOM-KFA", Jülich, Germany [3] Fusion Energy Research Program, Mechanical Engineering Division, University of California at San Diego, La Jolla, USA [4] FOM Institüt voor Plasmafysica Rijnhuizen, Associatie "FOM-EURATOM", Nieuwegein, The Netherlands [*] Researcher at NFSR, Belgium itemize

Plasma performance, boundary studies and first experiments with ICRH in TEXTOR

International Nuclear Information System (INIS)

Waidmann, G.; Bay, H.L.; Bertschinger, G.

1985-01-01

The TEXTOR plasma serves as a test bed for plasma/wall interaction studies and ICRH experiments. Reproducible and long-lasting discharges with soft termination were generated in the internal disruptive mode. The operational regime for Ohmic heating is shown in a 1/q versus n-barsub(e)R/Bsub(T) diagram. A comparison of electrical conductivity derived from current density measurements with calculated values favours neoclassical theory. A pump limiter installed on TEXTOR demonstrated a particle removal rate of 6x10 20 particles per second out of the boundary layer. It could decrease the central electron density by 50%. The pump limiter was used to control fuelling and recycling characteristics of stable discharges. First experiments with additional ICRH showed a strong influence on the plasma boundary and scrape-off layer. The interaction of the radiofrequency with the boundary layer at present limits the power input to the plasma. Plasma boundary parameters have been measured by optical methods combined with neutral particle beams. (author)

Multi-pulse 20 kHz TV Thomson scattering with high spatial resolution on TEXTOR-94

International Nuclear Information System (INIS)

Meiden, H.J.V.D.; Barth, C.J.; Oyevaar, T.

2001-01-01

This article describes the first high repetition rate TVTS system in the world. It will be implemented on TEXTOR-94, with the aim to study the dynamic behaviour of meso scale plasma phenomena, like MHD modes, filaments, transport barriers and edge phenomena. To reach this, a 20 kHz intracavity laser system is combined with an ultra fast CCD camera. During one discharge of TEXTOR-94 three bursts of 40 pulses can be extracted from the laser system with a time separation of 0.5 s between the bursts. This new equipment will be implemented on the beam line and spectrometer of the present double pulse TVTS system of TEXTOR-94. The new TVTS system will be capable of producing three times 40 electron temperature- and density profiles along a laser chord of 900 mm with a spatial resolution of 7.5 mm for the full plasma diameter and 2 mm for the edge region, respectively. An observational error of 6% on T e and 3% on n e is expected for n e = 3.5x10 19 m -3 , using a laser pulse energy of typical 16 J. (author)

Three-dimensional plasma transport in open chaotic magnetic fields. A computational assessment for tokamak edge layers

International Nuclear Information System (INIS)

Frerichs, Heinke Gerd

2010-04-01

The development of nuclear fusion as an alternative energy source requires the research on magnetically confined, high temperature plasmas. In particular, the quantification of plasma flows in the domain near exposed material surfaces of the plasma container by computer simulations is of key importance, both for guiding interpretation of present fusion experiments and for aiding the ongoing design activities for large future devices such as ITER, W7-X or the DEMO reactor. There is a large number of computational issues related to the physics of hot, fully ionized and magnetized plasmas near surfaces of the vacuum chamber. This thesis is dedicated to one particular such challenge, namely the numerical quantification of self-consistent kinetic neutral gas and plasma fluid flows in very complex 3D (partially chaotic) magnetic fields, in the absence of any common symmetries for plasma and neutral gas dynamics. Such magnetic field configurations are e.g. generated by externally applied magnetic perturbations at the plasma edge, and are of great interest for the control of particle and energy exhausts. In the present thesis the 3D edge plasma and neutral particle transport code EMC3-EIRENE is applied to two distinct configurations of open chaotic magnetic system: at the TEXTOR and DIII-D tokamaks. Improvements of the edge transport model and extensions of the transport code are presented, which have allowed such simulations for the first time for 3D scenarios at DIII-D with ITER similar plasmas. A strong 3D effect of the chaotic magnetic field on the DIII-D edge plasma is found and analyzed in detail. It is found that a pronounced striation pattern of target particle and heat fluxes at DIII-D can only be obtained up to a certain upper limiting level of anomalous cross-field transport. Hence, in comparison to experimental data, these findings allow to narrow down the range of this model parameter. One particular interest at TEXTOR is the achievement of a regime with

Light impurity production in tokamaks

International Nuclear Information System (INIS)

Philipps, V.; Vietzke, E.; Erdweg, M.

1989-01-01

A review is given of the different erosion processes of carbon materials with special emphasis on conditions relevant to plasma surface interactions. New results on the chemical erosion and radiation enhanced sublimation of boron-carbon layers are presented. The chemical hydrocarbon formation produced by the interaction of the TEXTOR scrape-off plasma with a carbon target has been investigated up to temperatures of 1500K using a Sniffer probe. The chemical interaction of the plasma with the carbon walls in TEXTOR is also analysed by measuring the hydrocarbon and CO and CO 2 partial pressures built up on the surrounding walls during the discharges. The recycling of oxygen impurities in an all carbon surrounding occurs predominantly in the form of CO and Co 2 molecules and the analysis of both neutral pressures during the discharges has been used as an additional diagnostic for the oxygen impurity situation in TEXTOR. These data are discussed in view of spectroscopic measurements on the influx of carbon and oxygen atoms from the walls and impurity line radiation. CD-band spectroscopy in addition is employed to identify the hydrocarbon chemical carbon erosion. Our present understanding of the oxygen impurity recycling and the oxygen sources are described. Particle induced release of CO molecules from the entire first wall is believed to be the dominant influx process of oxygen in the SOL of plasmas with carbon facing materials. The influence of coating the TEXTOR first wall with a boron-carbon film (B/C ≅1) on the light impurity behaviour is shown. (author)

Transport in the tokamak plasma edge

International Nuclear Information System (INIS)

Vold, E.L.

1989-01-01

Experimental observations characterize the edge plasma or boundary layer in magnetically confined plasmas as a region of great complexity. Evidence suggests the edge physics plays a key role in plasma confinement although the mechanism remains unresolved. This study focuses on issues in two areas: observed poloidal asymmetries in the Scrape Off Layer (SOL) edge plasma and the physical nature of the plasma-neutral recycling. A computational model solves the coupled two dimensional partial differential equations governing the plasma fluid density, parallel and radial velocities, electron and ion temperatures and neutral density under assumptions of toroidal symmetry, ambipolarity, anomalous diffusive radial flux, and neutral-ion thermal equilibrium. Drift flow and plasma potential are calculated as dependent quantities. Computational results are compared to experimental data for the CCT and TEXTOR:ALT-II tokamak limiter cases. Comparisons show drift flux is a major component of the poloidal flow in the SOL along the tangency/separatrix. Plasma-neutral recycling is characterized in several tokamak divertors, including the C-MOD device using magnetic flux surface coordinates. Recycling is characterized by time constant, τ rc , on the order of tens of milliseconds. Heat flux transients from the core into the edge on shorter time scales significantly increase the plasma temperatures at the target and may increase sputtering. Recycling conditions in divertors vary considerably depending on recycled flux to the core. The high density, low temperature solution requires that the neutral mean free path be small compared to the divertor target to x-point distance. The simulations and analysis support H-mode confinement and transition models based on the recycling divertor solution bifurcation

Oxygen collection in the limiter shadow of TEXTOR depending on wall conditioning with boron

International Nuclear Information System (INIS)

Wienhold, P.; Seggern, J. v.; Kuenzli, H.

1991-01-01

One of the major consequences of the boronization of TEXTOR compared to the carbonized machine was the further and remaining decrease of the oxygen contamination of the plasma. This has lowered also the carbon chemical sputtering by a factor of two in spite of higher radiative power loads to the graphite limiters and made auxiliary heating up to 6 MW possible. The fact, that oxygen did not reoccur as it happened during operation with carbonized walls caused the suggestion of gettering by the formation of a stable bond to the boron. Therefore, a period (May/June 89) where different conditioning treatments with boron were applied to TEXTOR gave ideal circumstances for collection experiments in the SOL and the subsequent analysis of the deposits aiming at the understanding of this hypothesis. (author) 10 refs., 2 figs

Application of tungsten for plasma limiters in TEXTOR

International Nuclear Information System (INIS)

Tanabe, T.; Wada, M.; Ohgo, T.; Philipps, V.; Rubel, M.; Huber, A.; Seggern, J. von; Ohya, K.; Pospieszczyk, A.; Schweer, B.

2000-01-01

Three different types of W limiters were exposed in the TEXTOR plasma and the response of the plasma and materials performance of the limiters were investigated. - A W bulk limiter operated with preheating above 800 K withstood a plasma heat load of about ∼20 MW/m 2 for a few seconds with some slight surface melting during the highest heat load shot. However, it was severely damaged when operated at around 500 K. - A C/W twin test limiter, half made of bulk W and the other half of graphite (EK-98) gave very useful information on how low- and high-Z materials behave under conditions of simultaneous utilization as PFM such as cross-contamination and the influence of a large mass difference on hydrogen reflection and deposition. - Two sets of main poloidal W limiters made of vacuum vapor sprayed (VPS)-W deposited on graphite (IG-430U) with a Re interlayer could absorb about 60% of the total convection heat and the ohmic plasma with a density as high as 5 x 10 13 cm -3 was sustained. Most of the VPS-W coated limiters tolerated a heat load of ∼20 MW/m 2 . This series of W limiters experiments in TEXTOR has shown that W is applicable as a PFM, if its central accumulation is avoided by NBI and/or ICRH heating. Nevertheless, some concerns still remain, including difficulty of plasma start-up, W behavior in higher temperature plasmas, and materials' selection

Observation of magnetic field perturbations during sawtooth activity in tokamak plasmas

International Nuclear Information System (INIS)

Soltwisch, H.; Koslowski, H.R.

1997-01-01

Sawtooth activity is a prominent example of a global plasma instability which is observed in virtually all tokamak devices. Despite numerous experimental and theoretical investigations, the phenomenon is still barely understood. As far as experimental effort is concerned, much attention has been paid to soft X-ray emission from the plasma and to its analysis in terms of two-dimensional contour plots, because it is thought to reflect the shape and temporal behaviour of magnetic flux surfaces during a sawtooth cycle. Recently, more direct methods of detecting sawtooth-related changes in the magnetic field structure have become available and have added new facets to the general picture. In this picture, some observations made on the Juelich tokamak TEXTOR by means of a Faraday rotation diagnostic technique will be reported. First, in correlation with the sawtooth collapse a localized periodic perturbation of the magnetic field with principal mode numbers m = 1 and n = 0 has been detected which, in the presence of an m = n = 1 island, may give rise to magnetic field line stochastization and thereby contribute significantly to a rapid expulsion of electronic energy from the plasma core region. Second, the so-called precursor oscillations prior to a sawtooth crash have been investigated and estimates have been obtained for the growth rate and width of a magnetic island forming immediately before the collapse. (Author)

Upgrading a TEXTOR Data acquisition system for remote participation using Java and Corba

NARCIS (Netherlands)

Korten, M.; Becks, B.; Blom, H.; Busch, P.; Kemmerling, G.; Kooijman, W.; Krom, J. G.; de Laat, C. T. A. M.; Lourens, W.; van der Meer, E.; Niderost, B.; Oomens, A. A. M.; Wijnoltz, F.; Samm, U.

2000-01-01

The partners in the Trilateral Euregio Cluster (TEC) are implementing and developing Remote Participation technologies that are expected to support a joint research programme on the experimental facility TEXTOR-94. A common TEC architecture for our heterogeneous data acquisition and storage systems

Stochastization of Magnetic Field Surfaces in Tokamaks by an Inner Coil

International Nuclear Information System (INIS)

Chavez-Alarcon, Esteban; Herrera-Velazquez, J. Julio E.; Braun-Gitler, Eliezer

2006-01-01

A 3-D code has been developed in order to simulate the magnetic field lines in circular cross-section tokamaks. The toroidal magnetic field can be obtained from the individual fields of circular coils arranged around the torus, or alternatively, as a ripple-less field. The poloidal field is provided by a given toroidal current density profile. Proposing initial conditions for a magnetic filed line, it is integrated along the toroidal angle coordinate, and Poincare maps can be obtained at any desired cross section plane. Following this procedure, the code allows the mapping of magnetic field surfaces for the axisymmetric case. For this work, the density current profile is chosen to be bell-shaped, so that realistic safety factor profiles can be obtained. This code is used in order to study the braking up of external surfaces when the symmetry is broken by an inner coil with tilted circular loops, with the purpose of modelling the behaviour of ergodic divertors, such as those devised for TEXTOR

Upgrading the power supplies of TEXTOR for three Tesla operation

International Nuclear Information System (INIS)

Giesen, B.; Veiders, E.; Petree, F.; Fink, R.; Wagnitz, R.

1986-01-01

The toroidal magnetic system of TEXTOR can tolerate a magnetic field load of up to 2.6 Tesla routinely at full plasma current, and of up to 3 Tesla under certain boundary conditions and for a restricted number of discharges. The original power supply which can generate a toroidal magnetic field of 2 Tesla has been upgraded to operate at a field strength of 3 Tesla, by adding a new, controlled rectifier, with its own independent control, connected in parallel with the first. Studies were undertaken to determine its behaviour where control is lost, such as when a circuit breaker trips or in ''freewheel'' operation. This paper analyzes this asymmetrical arrangement and discusses the danger of damaging the smaller unit by commutating a large current into it. Moreover, in order to improve the availability of TEXTOR, the new controlled rectifier is redundant to two other units that control the vertical field and the ohmic heating coil currents. For this purpose the two bridges of this 12-pulse system are to be changed from a parallel to a series connexion, the free-wheeling diodes are disconnected and redeployed to block the large voltage pulses that are induced at plasma initiation, and 3-phase ''freewheeling'' thyristors are added that serve to reduce reactive power consumption

Impurity sources in TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Pospieszczyk, A; Bay, H L; Bogen, P; Hartwig, H; Hintz, E; Koenen, L; Ross, G G; Rusbueldt, D; Samm, U; Schweer, B

1987-02-01

The deuterium, oxygen and carbon fluxes from the main limiter and the deuterium fluxes from the wall are measured in TEXTOR for an 'all carbon' surrounding as a function of central density n/sub e/, of applied ICRH-power and of different wall conditions (carbonization). For this purpose, emission spectroscopy both with filter systems and spectrometers has been used. It is found that a major release mechanism for light impurities is via the formation of molecules. Oxygen seems to enter the discharge from the liner via O-D containing molecules, whereas the limiter acts as the main carbon source by the release of hydro-carbons as indicated by the observed CD-band spectra. Both oxygen and carbon fluxes are reduced by about a factor of two after a fresh carbonization. Above a certain critical density the plasma detaches from the limiter and forms a stable discharge with a radiation cooled boundary layer and with a major fraction of particles now reaching the wall instead of the limiter. The critical density rises with decreasing impurity fluxes or with increasing heating powers.

Impurity sources in TEXTOR

International Nuclear Information System (INIS)

Pospieszczyk, A.; Bay, H.L.; Bogen, P.; Hartwig, H.; Hintz, E.; Koenen, L.; Ross, G.G.; Rusbueldt, D.; Samm, U.; Schweer, B.

1987-01-01

The deuterium, oxygen and carbon fluxes from the main limiter and the deuterium fluxes from the wall are measured in TEXTOR for an 'all carbon' surrounding as a function of central density n e , of applied ICRH-power and of different wall conditions (carbonization). For this purpose, emission spectroscopy both with filter systems and spectrometers has been used. It is found that a major release mechanism for light impurities is via the formation of molecules. Oxygen seems to enter the discharge from the liner via O-D containing molecules, whereas the limiter acts as the main carbon source by the release of hydro-carbons as indicated by the observed CD-band spectra. Both oxygen and carbon fluxes are reduced by about a factor of two after a fresh carbonization. Above a certain critical density the plasma detaches from the limiter and forms a stable discharge with a radiation cooled boundary layer and with a major fraction of particles now reaching the wall instead of the limiter. The critical density rises with decreasing impurity fluxes or with increasing heating powers. (orig.)

Non-resonant magnetic braking on JET and TEXTOR

DEFF Research Database (Denmark)

Sun, Y.; Liang, Y.; Shaing, K.C.

2012-01-01

The non-resonant magnetic braking effect induced by a non-axisymmetric magnetic perturbation is investigated on JET and TEXTOR. The collisionality dependence of the torque induced by the n = 1, where n is the toroidal mode number, magnetic perturbation generated by the error field correction coils...... in the 1/ν regime. The strongest NTV torque on JET is also located near the plasma core. The magnitude of the NTV torque strongly depends on the plasma response, which is also discussed in this paper. There is no obvious braking effect with n = 2 magnetic perturbation generated by the dynamic ergodic...

ICRF/edge physics research on TEXTOR

International Nuclear Information System (INIS)

Oost, G. van; Nieuwenhove, R. van; Koch, R.; Messiaen, A.M.; Vandenplas, P.E.; Weynants, R.R.; Dippel, K.H.; Finken, K.H.; Lie, Y.T.; Pospieszczyk, A.; Samm, U.; Schweer, B.; Conn, R.W.; Corbett, W.J.; Goebel, D.M.; Moyer, R.A.; California Univ., Los Angeles

1990-01-01

Extensive investigations of ICRF-induced effects on the edge plasma and on plasma-wall interaction were conducted on TEXTOR under different wall- and limiter as well as plasma- and heating conditions. Several strong effects of ICRF on the edge parameters were observed on TEXTOR, such as density rise, instantaneous electron heating, modification of SOL profiles, influx of ligth and/or heavy impurities, increased heat flux to the limiters, and production of energetic ions in the SOL. The fast response time of some of the changes and the observation of a maximum in the SOL profile of electron temperature, heat flux and metal sputtering clearly demonstrated that RF power is directly absorbed in the SOL. Estimates of this power amount to several percent of the total RF power launched into the plasma. Plasma-wall interaction during ICRF was substantially reduced by an appropriate choice of the wall conditioning procedures (wall carbonization with liner at 400degC or, above all, boronization). As a result record low values of the radiated power fraction were achieved during ICRF and long pulse, high power, low impurity operation was possible. Further improvement was obtained by ICRF antenna phasing. When ICRF power is coupled to the plasma, several effects on the core and edge plasma influence the operation of the toroidal pump limiter ALT-II. Experimental and theoretical studies were performed to elucidate the mechanisms responsible for the ICRF-induced effects, including the propagation of plasma waves in the edge plasma and nonlinear phenomena such as parametric decay, important changes in the DC current between the antenna structure and the liner due to the sheath effect at the antennas, and the generation of waves at harmonics of the RF generator frequency. Radial profiles of the DC radial and poloidal electric fields as well as a localized RF electric field structure were measured in the SOL using a fast scanning probe. (orig.)

Runaway snakes in TEXTOR-94

International Nuclear Information System (INIS)

Entrop, I.; Jaspers, R.; Lopes Cardozo, N.J.; Finken, K.H.

1999-01-01

Observations of a runaway beam confined in an island-like structure, a so-called runaway snake, are reported. The observations are made in TEXTOR-94 by measurement of synchrotron radiation emitted by these runaways. A full poloidal view allows for the study of the synchrotron pattern of the snake to estimate runaway energy, pitch angle and the radius, shift and safety factor of the drift surface q D at which the runaway beam has developed. The runaway snake parameters are investigated under different current and magnetic field strength conditions. Examples are found of a runaway snake at the q D =1 and the q D =2 drift surface. The radial diffusion coefficient of runaways inside a snake is D r approx. 0.01m 2 s -1 . The rapid runaway losses in regions of (macroscopic) magnetic perturbations outside a snake and the good confinement inside an island assumed to consist of perfect nested surfaces are consistent with magnetic turbulence as the main cause for runaway transport. (author)

Neutral iron densities in front of a reference limiter in TEXTOR

International Nuclear Information System (INIS)

Schweer, B.; Bay, H.L.

1983-09-01

Preliminary measurements of the time and space resolved densities of neutral iron in front of a reference limiter in Textor have been performed using laser induced fluorescence. The limiter was made of stainless steel (SS 316) and formed as a half sphere, 10 cm in diameter. Neutral iron densities up to 5 x 10 9 atoms/cm 3 were found. The detection limit was below 10 7 atoms/cm 3 . (orig.)

Link between self-consistent pressure profiles and electron internal transport barriers in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Razumova, K A [Nuclear Fusion Institute, RRC ' Kurchatov Institute' , 123182 Moscow (Russian Federation); Andreev, V F [Nuclear Fusion Institute, RRC ' Kurchatov Institute' , 123182 Moscow (Russian Federation); Donne, A J H [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Hogeweij, G M D [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Lysenko, S E [Nuclear Fusion Institute, RRC ' Kurchatov Institute' , 123182 Moscow (Russian Federation); Shelukhin, D A [Nuclear Fusion Institute, RRC ' Kurchatov Institute' , 123182 Moscow (Russian Federation); Spakman, G W [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Vershkov, V A [Nuclear Fusion Institute, RRC ' Kurchatov Institute' , 123182 Moscow (Russian Federation); Zhuravlev, V A [Nuclear Fusion Institute, RRC ' Kurchatov Institute' , 123182 Moscow (Russian Federation)

2006-09-15

Tokamak plasmas have a tendency to self-organization: the plasma pressure profiles obtained in different operational regimes and even in various tokamaks may be represented by a single typical curve, called the self-consistent pressure profile. About a decade ago local zones with enhanced confinement were discovered in tokamak plasmas. These zones are referred to as internal transport barriers (ITBs) and they can act on the electron and/or ion fluid. Here the pressure gradients can largely exceed the gradients dictated by profile consistency. So the existence of ITBs seems to be in contradiction with the self-consistent pressure profiles (this is also often referred to as profile resilience or profile stiffness). In this paper we will discuss the interplay between profile consistency and ITBs. A summary of the cumulative information obtained from T-10, RTP and TEXTOR is given, and a coherent explanation of the main features of the observed phenomena is suggested. Both phenomena, the self-consistent profile and ITB, are connected with the density of rational magnetic surfaces, where the turbulent cells are situated. The distance between these cells determines the level of their interaction, and therefore the level of the turbulent transport. This process regulates the plasma pressure profile. If the distance is wide, the turbulent flux may be diminished and the ITB may be formed. In regions with rarefied surfaces the steeper pressure gradients are possible without instantaneously inducing pressure driven instabilities, which force the profiles back to their self-consistent shapes. Also it can be expected that the ITB region is wider for lower dq/d{rho} (more rarefied surfaces)

Investigation of eddy currents in the components of the dynamic ergodic divertor of TEXTOR using analytical and numerical approaches

International Nuclear Information System (INIS)

Giesen, B.; Neubauer, O.; Bondarchuk, E.; Doinikov, N.; Kitaev, B.; Obidenko, T.; Panin, A.

2003-01-01

Analytical and numerical approaches for the calculation of eddy currents in mechanical structures of the TEXTOR tokamak in view of operating the dynamic ergodic divertor (DED) coil system fed with the alternating current up to 15 kA at frequencies up to 10 kHz are described. The design of the in-vessel components located close to the DED coils requires detailed investigation of eddy current effects to avoid unacceptable heating and forces. Different approaches depending on skin-layer depths compared with the body dimensions are analyzed. The applied algorithms are based on analytical and simplified numerical methods. Precision and application range of these algorithms have been checked by a numerical code. The simplified technique is rather effective for first step engineering estimation and gives a good understanding for the problem. In a certain parameter range, it results in even precise values and can be used for design optimization of the structures without huge efforts in numerical modeling. After modification of the component's shape prototypes have been manufactured and successfully tested in a full-scale model under the real DED field. The design recommendations resulting from the eddy current studies contributed significantly to the optimized lay out of the DED in-vessel components

Runaway electrons in disruptions and perturbed magnetic topologies of tokamak plasmas

International Nuclear Information System (INIS)

Forster, Michael

2012-01-01

Nuclear fusion represents a valuable perspective for a safe and reliable energy supply from the middle of the 21st century on. Currently, the tokamak is the most advanced principle of confining a man-made fusion plasma. The operation of future, reactor sized tokamaks like ITER faces a crucial difficulty in the generation of runaway electrons. The runaway of electrons is a free fall acceleration into the relativistic regime which is known in various kinds of plasmas including astrophysical ones, thunderbolts and fusion plasmas. The tokamak disruption instability can include the conversion of a substantial part of the plasma current into a runaway electron current. When the high energetic runaways are lost, they can strike the plasma facing components at localised spots. Due to their high energies up to a few tens of MeV, the runaways carry the potential to reduce the lifetimes of wall components and even to destroy sensitive, i.e. actively cooled parts. The research for effective ways to suppress the generation of runaway electrons is hampered by the lack of a complete understanding of the physics of the runaways in disruptions. As it is practically impossible to use standard electron detectors in the challenging environment of a tokamak, the experimental knowledge about runaways is limited and it relies on rather indirect techniques of measurement. The main diagnostics used for this PhD work are three reciprocating probes which measure the runaway electrons directly at the plasma edge of the tokamak TEXTOR. A calorimetric probe and a material probe which exploits the signature that a runaway beam impact leaves in the probe were developed in the course of the PhD work. Novel observations of the burst-like runaway electron losses in tokamak disruptions are reported. The runaway bursts are temporally resolved and first-time measurements of the corresponding runaway energy spectra are presented. A characteristic shape and typical burst to burst variations of the

Collective Thomson scattering measurements with high frequency resolution at TEXTOR

DEFF Research Database (Denmark)

Stejner Pedersen, Morten; Nielsen, Stefan Kragh; Korsholm, Søren Bang

2010-01-01

We discuss the development and first results of a receiver system for the collective Thomson scattering (CTS) diagnostic at TEXTOR with frequency resolution in the megahertz range or better. The improved frequency resolution expands the diagnostic range and utility of CTS measurements in general ...... and is a prerequisite for measurements of ion Bernstein wave signatures in CTS spectra. The first results from the new acquisition system are shown to be consistent with theory and with simultaneous measurements by the standard receiver system. © 2010 EURATOM...

Characterization of redeposited carbon layers on TEXTOR limiter by Laser Raman spectroscopy

International Nuclear Information System (INIS)

Egashira, K.; Tanabe, T.; Yoshida, M.; Nakazato, H.; Philipps, V.; Brezinsek, S.; Kreter, A.

2011-01-01

Highlights: ► Laser Raman technique has applied to analyze the deposited carbon layers on TEXTOR test limiters of C and W. ► The carbon deposited layers showed the Raman spectra composed of G-peak and D-peak. ► For W limiter, hydrogen concentrations in the deposited carbon layers and their thicknesses correlated to the two peaks. ► The Laser Raman spectroscopy is a promising tool for in situ analysis of carbon redeposit layers on plasma facing W materials. - Abstract: Laser Raman spectroscopy is quite sensitive to detect the changes of graphite structure. In this study, the Laser Raman technique was applied to analyze the deposited carbon layers on TEXTOR test limiters of carbon (C) and tungsten (W) produced by intentional carbon deposition experiments by methane gas puffing. The carbon deposited layers showed the Raman spectra composed of two broad peaks, G-peak and D-peak, centered at around 1580 and 1355 cm −1 respectively. For W limiter, the G-peak position and the integrated intensity of the two peaks well correlate to hydrogen concentrations in the deposited carbon layers and their thicknesses, respectively. Hence Laser Raman spectroscopy is a promising tool for the in situ analysis of carbon redeposit layers on plasma facing W materials and probably on Be materials.

Influence of nonlinear effects on the neutral gas transport in tokamaks

International Nuclear Information System (INIS)

Behringer, T.

1992-06-01

The linear Monte Carlo computer code EIRENE for calculation of free molecular flow of neutral gases through a background plasma has been extended to the non-linear transition flow regime (Knudsen number 0.1-10). Motivation arose from higher gas densities in the range of 10 13 -10 15 cm -3 appearing in the srape-off layer and in parts of the vacuum system of advanced tokamak experiments. To treat the problem, the Direct Monte Carlo Simulation Method after Bird, a kinetic approach, was chosen, since the conditions for application of continuum theory are not met. First results with the extended code were obtained in calculating the conductance of plasma-free short cylindrical ducts and elbows. A steady increase in conductance with decreasing Knudsen number was found, which is in good agreement with experimental data. Further calculations for transition flows through fixed background plasmas were made. In these, solutions obtained were represented as differences from solutions obtained by linear calculations. Simulation of a 1-D plasma slab configuration (related to the gaseous divertor concept) revealed markedly varying neutral gas profiles due to neutral-neutral collisions. In addition, in these runs neutral-neutral inelastic collision processes turned out to be negligible. Finally, neutral gas behaviour at higher densities in pump limiter geometries was studied, related to experiments on the tokamak TEXTOR. An increase in conductance in the direction to the pumps of up to 25% relative to linear results was found. Recently obtained experimental data on the impact of non-linear neutral effects upon conductance could be confirmed. (orig.) [de

Modification of the collective Thomson scattering radiometer in the search for parametric decay on TEXTOR

DEFF Research Database (Denmark)

Nielsen, Stefan Kragh; Salewski, Mirko; Bongers, W.

2012-01-01

Strong scattering of high-power millimeter waves at 140 GHz has been shown to take place in heating and current-drive experiments at TEXTOR when a tearing mode is present in the plasma. The scattering signal is at present supposed to be generated by the parametric decay instability. Here we descr...

Real-time digital control of plasma position and shape on the TEXTOR tokamak

International Nuclear Information System (INIS)

Mitri, Mikhael

2009-01-01

Beside the objective of contributing to the controlled thermonuclear fusion research and ultimately the development of a fusion based power plant, the main objectives of the thesis are a substantial improvement of plasma vertical position control and plasma shape control as well as a better understanding of formerly unexplained effects, e.g. disturbance fields. As for the vertical position control, a deep analysis has to be undertaken to identify the problem sources. Accurate control of the plasma position is very difficult to achieve. This is mainly due to the complexity of the tokamak and the difficulty in measuring or modelling all relevant discharge variables. Any models would be highly nonlinear and time varying. Thus, for simulation and controller design, a simplified, but nevertheless accurate model has to be developed, based on physics and measured data of the process. Furthermore, the quality of the measured position has to be improved by using new inductive sensors, integrators, and hardware. The integration drift problem has to be analysed and resolved by developing a drift-free integration method. Concerning the shape control, a better understanding of the relation between the stray fields and the iron core saturation is required. Furthermore, the influence on the plasma elongation has to be determined. Upon this, a shape compensation algorithm has to be developed accordingly. The accuracy of the shape control has to be better than 1%. (orig.)

Modelling of local carbon deposition on rough test limiter exposed to the edge plasma of TEXTOR

International Nuclear Information System (INIS)

Dai Shuyu; Sun Jizhong; Wang Dezhen; Kirschner, A.; Matveev, D.; Borodin, D.; Bjoerkas, C.

2013-01-01

A Monte-Carlo code called SURO has been developed to study the influence of surface roughness on the impurity deposition characteristic in fusion experiments. SURO uses the test particle approach to describe the impact of background plasma and the deposition of impurity particles on a sinusoidal surface. The local impact angle and dynamic change of surface roughness as well as surface concentrations of different species due to erosion and deposition are taken into account. Coupled with 3D Monte-Carlo code ERO, SURO was used to study the impact of surface roughness on 13 C deposition in 13 CH 4 injection experiments in TEXTOR. The simulations showed that the amount of net deposited 13 C species increases with surface roughness. Parameter studies with varying 12 C and 13 C fluxes were performed to gain insight into impurity deposition characteristic on the rough surface. Calculations of the exposure time needed for surface smoothing for TEXTOR and ITER were also carried out for different scenarios. (author)

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-03-01

This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

Exposure of metal mirrors in the scrape-off layer of TEXTOR

International Nuclear Information System (INIS)

Wienhold, P.; Litnovsky, A.; Philipps, V.; Schweer, B.; Sergienko, G.; Oelhafen, P.; Ley, M.; De Temmerman, G.; Schneider, W.; Hildebrandt, D.; Laux, M.; Rubel, M.; Emmoth, B.

2005-01-01

Large molybdenum mirrors have been exposed in the SOL of TEXTOR in order to simulate conditions relevant for ITER optical components. Distortions of the reflectivity - increase as well as decrease - are found in the erosion and deposition dominated areas, respectively. The changes are most pronounced in the near UV and level off in the IR and can partly be attributed to observed surface changes. A novel periscope system was installed and mirrors exposed in a pilot experiment to simulate the transmission of light to distant sensors in ITER

Beam-guiding system for Rutherford-scattering diagnostic at TEXTOR

International Nuclear Information System (INIS)

Cosler, A; Bertschinger, G.; Kemmereit, E.; Ven, H.W. van der; Barbian, E.P.; Blokland, A.A.E. van

1988-01-01

A beam-guiding system for a neutral beam probe diagnostic has been developed for implementation at TEXTOR. Energetic helium atoms scattered on the plasma ions provide information about the local ion temperature. Time resolution is attained by sampling scattered particles measured individually by a time-of-flight analyser. The mechanical supports have been designed for lateral and angular movement of the beam-guiding system to be used for radial scanning of the torus and for optimization of the scattering angle. The parameters of the probing beam itself can be controlled jby a small beam profile diagnsotic. Provisions are made to observe separately the radial or axial component of the ion velocity distribution. (author). 10 refs.; 7 figs

The arc power supply for the TEXTOR neutral injectors

International Nuclear Information System (INIS)

Schwarz, U.; Pfister, U.; Goll, O.; Wurslin, R.; Scherer, J.; Haubmann, S.

1986-01-01

The 24 single arcs in the plasma source of the TEXTOR neutral injector are supplied with an overall current of 1800 A at an arc voltage of 150 V DC. The current is switched on and off in less than 1 msec. The paper presents a new modular solution for such a power supply. Each arc is powered by a separately switched mode supply module. One single module consists of a diode rectifier bridge with a filter, a fast semiconductor switch, an inductance in series for stabilizing the current and a free-wheeling path. The layout of this power supply system is described in detail based on test results. Design features and technical data are given

Overview of material migration and mixing, fuel retention and cleaning of ITER-like castellated structures in TEXTOR

International Nuclear Information System (INIS)

Litnovsky, A.; Philipps, V.; Wienhold, P.; Kreter, A.; Kirschner, A.; Matveev, D.; Brezinsek, S.; Sergienko, G.; Pospieszczyk, A.; Schweer, B.; Schulz, C.; Schmitz, O.; Coenen, J.W.; Samm, U.; Krieger, K.; Hirai, T.; Emmoth, B.; Rubel, M.; Bazylev, B.; Breuer, U.

2011-01-01

Plasma-facing components (PFCs) in ITER will be castellated by splitting them into small-size blocks to maintain the thermo-mechanical stability. However, there are concerns in particular on retention of codeposited radioactive fuel in the gaps. An R and D program is underway in TEXTOR addressing this acute issue of castellation. Material migration and fuel inventory are investigated using long- and short-term discharge-resolved experiments with castellated structures in TEXTOR. Significant impurity transport to the gaps was detected and results were in part quantitatively reproduced with 3D-GAPS code. Deposits containing up to 70 at.% of tungsten on the gap areas closest to the plasma were detected in recent experiments. Deposition in the gaps accompanied by metal mixing demand for development of effective cleaning techniques. In experiments with ITER-like castellation, the gaps were cleaned from carbonaceous deposits using oxygen plasmas at 350 deg. C. This contribution contains an overview of experimental and modeling results along with recommendations for PFCs in ITER.

Characterization of a source of carbon particles produced by laser ablation and used for the calibration of erosion measurement made by spectroscopy in a tokamak

International Nuclear Information System (INIS)

Naiim Habib, M.

2011-12-01

In a tokamak, plasma-wall interactions lead to the erosion of plasma facing components, which can be detrimental to plasma operation and to the safety of the tokamak. In order to fulfill the safety requirements imposed to the ITER project, it is necessary to monitor the amount of eroded material. Optical emission spectroscopy in the visible range is traditionally used to measure particle fluxes from the wall to the plasma. These measurements are done thanks to a collisional radiative model based on atomic physics data. However, these data don't take into account the observation geometry of the spectroscopic diagnostic, and suffer from relatively large uncertainties. Furthermore, transport, deposition and re-erosion phenomena, as well as the evolution of the transmission or the reflection of optical components can lead to an incorrect estimation of the amount of effectively eroded material. An in situ calibration technique, which consists in injecting by laser a known carbon particle source in the line of sight of the spectroscopic diagnostic during plasma operation, is proposed. The experimental study of laser ablation of carbon allowed to determine the optimal conditions for the constitution of this source, and to characterise the ablated species. These experiments are completed by a modelling of the emission spectrum of the laser induced plasma, in order to obtain information on its ionisation degree. Finally, results of the first validation experiments realised in the German TEXTOR tokamak are presented and discussed. (author)

Overview of TEXTOR results

International Nuclear Information System (INIS)

Schueller, F.C.; Barth, C.J.; Abdullaev, S.S.

2003-01-01

From March 2001 to November 2002 TEXTOR undergoes a rebuild to install the Dynamic Ergodic Divertor (DED). Details of the DED will be described. During the experimental campaigns preceding the DED-shutdown the position of ITBs in L-mode-, shallow-shear- and RI-mode-plasmas were highlighted by focussed ECRH. The RI-mode can keep H-mode confinement up to N GW = 1.4 provided that the neutral gas influx is kept as low as possible. The absence of power degradation in case of central ECH may be ascribed to an ITB at q=1. Impurity studies with repeated Argon puffs indicated anomalous diffusion coefficients. The impurity transport changes if the concentrations increase from 'trace' to 'bulk' levels. Observations on T e - and n e -fluctuations due to broadband turbulence in the presence of large m=2 islands show an enhanced fluctuation spectrum around the X-point compared to island fluctuations with improved confinement inside islands. PWI-studies concentrated on erosion/deposition of carbon, fuel recycling and retention and atomistic processes in the vicinity of limiters. The toroidal limiter is the dominant source of carbon, which is only partly re-deposited in the vicinity of the erosion. The majority is deposited at the neutraliser plates in the toroidal pump-limiters and other obstacles in the SOL. The C-layers directly facing the plasma impact have a low H-retention ( -2 H/C) due to thermal release. (author)

Tokamak

International Nuclear Information System (INIS)

Wesson, John.

1996-01-01

This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

Measurements of ion temperature and plasma hydrogenic composition by collective Thomson scattering in neutral beam heated discharges at TEXTOR

DEFF Research Database (Denmark)

Stejner Pedersen, Morten; Salewski, Mirko; Korsholm, Søren Bang

2013-01-01

A method is developed to perform plasma composition and ion temperature measurements across the plasma minor radius in TEXTOR based on ion cyclotron structures in collective Thomson scattering spectra. By gradually moving the scattering volume, we obtain measurements across the outer midplane of ...

Chemical impurity production under boronized wall conditions in TEXTOR

International Nuclear Information System (INIS)

Philipps, V.; Vietzke, E.; Erdweg, M.

1992-01-01

The TEXTOR SNIFFER probe has been used to analyse the chemical impurity production under various plasma and boronized wall conditions. Methane formation has been observed to 0.6-1 x 10 -2 CH 4 /H at room temperature, increasing slightly with increasing density in the SOL. The hydrocarbon formation yields increase from R.T. to the maximum at about 500 o C by a factor of 1.5-2.5. Increasing the impact energy by biasing the graphite plate leads to a decrease of the hydrocarbon yield at room temperature but to an increase at 500 o C. Chemical CO formation due interaction of oxygen impurities with the graphite reaches ratios between 0.5 and 3 x 10 -2 CO/H,D increasing with increasing distance to the limiter edge. (author) 10 refs., 6 figs

Properties of the TEXTOR boundary layer

International Nuclear Information System (INIS)

Bogen, P.; Hartwig, H.; Hintz, E.; Hoethker, K.; Lie, Y.T.; Pospieszczyk, A.; Samm, U.

1984-01-01

First measurements on the TEXTOR boundary layer are reported. The hydrogen recycling in front of the four limiter segments has been studied by means of a CCD-camera, which proved to be a good instrument to center the discharge for symmetric plasma-limiter contact. The composition of the neutral fluxes from the limiter have been measured: oxygen fluxes are about a factor of ten higher than the metal fluxes; within the error limits the composition does not change with varying limiter radius. Electron densities in the scrape-off layer away from the limiter have been determined by injecting an Li-atom beam from a thermal source and by observing its emission as a function of radius. Similar measurements have been made in front of the limiter with sputtered Cr and O atoms. Both methods gave for the magnetic surface of the limiter radius nsub(e) approx.= 1 x 10 12 /cm 23 . Infrared observations of a test limiter with a CCD-camera and a PbSe-detector have been performed to record the thermal loads. About 10% of the input power flows to the limiter. (orig.)

In-situ observation of the chemical erosion of graphite in the scrape-off-layer of TEXTOR

International Nuclear Information System (INIS)

Philipps, V.; Vietzke, E.; Erdweg, M.

1989-01-01

A sniffer probe system has been used to investigate the chemical erosion during interaction of the TEXTOR scrape-off plasma with a pyrolytic graphite plate at temperatures up to 1400 0 C. Floating potential conditions as well as 200 V bias has been applied at plasma ion fluxes of about 10 18 ions/cm 2 sec. Methane formation was found to be 8x10 -3 CH 4 /H and 1.5x10 -2 CD 4 /D + for room temperature graphite and floating potential increasing by a factor of two at temperature around 500 0 C. Biasing the graphite decreases the methane yield at room temperature and increase it in the maximum temperature range. CO formation due to chemical interaction of oxygen ions with the graphite reaches ratios between 3 and 6x10 -2 CO/D(H) near the limiter edge under normal TEXTOR scrape-off conditions and exceeds the chemical hydro-(deutero-)carbon formation significantly. The results are discussed in view of the present status of hydro-(deutero-)carbon formation on graphite and carbon impurity observations made in fusion experiments. (orig.)

ICRF Mode Conversion Current Drive for Plasma Stability Control in Tokamaks

International Nuclear Information System (INIS)

Grekov, D.; Kock, R.; Lyssoivan, A.; Noterdaeme, J. M.; Ongena, J.

2007-01-01

There is a substantial incentive for the International Thermonuclear Experimental Reactor (ITER) to operate at the highest attainable beta (plasma pressure normalized to magnetic pressure), a point emphasized by requirements of attractive economics in a reactor. Recent experiments aiming at stationary high beta discharges in tokamak plasmas have shown that maximum achievable beta value is often limited by the onset of instabilities at rational magnetic surfaces (neoclassical tearing modes). So, methods of effective control of these instabilities have to be developed. One possible way for neoclassical tearing modes control is an external current drive in the island to locally replace the missing bootstrap current and thus to suppress the instability. Also, a significant control of the sawtooth behaviour was demonstrated when the magnetic shear was modified by driven current at the magnetic surface where safety factor equals to 1. In the ion cyclotron range of frequencies (ICRF), the mode conversion regime can be used to drive the local external current near the position of the fast-to-slow wave conversion layer, thus providing an efficient means of plasma stability control. The slow wave energy is effectively absorbed in the vicinity of mode conversion layer by electrons with such parallel to confining magnetic field velocities that the Landau resonance condition is satisfied. For parameters of present day tokamaks and for ITER parameters the slow wave phase velocity is rather low, so the large ratio of momentum to energy content would yield high current drive efficiency. In order to achieve noticeable current drive effect, it is necessary to create asymmetry in the ICRF power absorption between top and bottom parts of the plasma minor cross-section. Such asymmetric electron heating may be realized using: - shifted from the torus midplane ICRF antenna in TEXTOR tokamak; - plasma displacement in vertical direction that is feasible in ASDEX-Upgrade; - the

Upgrading a TEXTOR Data Acquisition system for remote participation using Java and Corba

International Nuclear Information System (INIS)

Korten, M.; Becks, B.; Blom, H.; Busch, P.; Kemmerling, G.; Kooijman, W.; Krom, J.G.; Laat, C.T.A.M. de; Lourens, W.; Meer, E. van der; Nideroest, B.; Oomens, A.A.M.; Wijnoltz, F.; Samm, U.

2000-01-01

The partners in the Trilateral Euregio Cluster (TEC) are implementing and developing Remote Participation technologies that are expected to support a joint research programme on the experimental facility TEXTOR-94. A common TEC architecture for our heterogeneous data acquisition and storage systems is seen to be one of the major issues. As a consequence, legacy systems will be affected and have to be upgraded for optimised wide area network communication, platform independent data access and display. The object oriented redesign of the system to be described follows theses guidelines. The architecture of the system under development uses Java as programming environment and CORBA as Client/Server communication standard. It is described in this paper, how an operational Data Acquisition CAMAC subsystem of TEXTOR-94 based on OpenVMS and Decnet communications could be redesigned into an open, object oriented architecture in a platform independent way. A suitable Web Browser is required on the client side without further installation of application software to run the server. CORBA static method invocations are used for the communication between the client and server. At the server side, there is only Java code on top of the existing commercial OpenVMS CAMAC device driver. A modular object oriented software design permitted to eliminate dependencies of the generic module levels from the underlying bus systems. Porting of the Java code to other platforms like Windows NT and Linux has proven to be successful

Spectroscopic determination of inverse photon efficiencies of W atoms in the scrape-off layer of TEXTOR

Science.gov (United States)

Brezinsek, S.; Laengner, M.; Coenen, J. W.; O'Mullane, M. G.; Pospieszczyk, A.; Sergienko, G.; Samm, U.

2017-12-01

Optical emission spectroscopy can be applied to determine in situ tungsten particle fluxes from erosion processes at plasma-facing materials. Inverse photon efficiencies convert photon fluxes of WI and WII line transitions into W and {{{W}}}+ particle fluxes, respectively, dependening on the local plasma conditions. Experiments in TEXTOR were carried out to determine effective conversion factors for different WI and WII transitions with the aid of WF6 injection into deuterium scrape-off layer plasmas in the electron temperature T e range between {T}{e}=20 {eV} and {T}{e}=82 {eV}. The inverse photon efficiencies or so-called effective \\tfrac{S}{{XB}}-values have been determined for WI lines at λ =400.9 {nm}, 429.5 nm, 488.7 nm, 498.3 nm, and 522.5 nm as well as for WII at λ =434.6 {nm} and compared with theoretical calculations from the ADAS data base. Moreover, a multi-machine scaling for the \\tfrac{S}{{XB}}-value in the range of T e between 2...100 {eV} has been determined for the most prominent WI line at λ =400.9 {nm} to \\tfrac{S}{{XB}}({T}{e})=53.63-56.07× {e}(0.045× {T{e}[{eV}])} considering experimental data from TEXTOR, ASDEX Upgrade, PSI and PISCES. Comparison with ADAS calculations for the same transition reveal a good qualitative agreement with the dependence on T e , but an underestimation of ADAS calculations of less than 25% over the full covered range of experimentally accessible T e in the multi-machine scaling. A good agreement within the experimental uncertainties is found between TEXTOR and ADAS \\tfrac{S}{{XB}}-values for WI at λ =429.5 {nm} and λ =488.7 {nm} whereas an underestimation of up to a factor two of ADAS values for WI at λ =522.5 {nm} and λ =498.3 {nm} was measured. Potentially, reasons for the discrepancy are an overestimation of applied ionisation rate coefficients in ADAS for neutral W and a stronger electron dependence n e for these transitions.

Experimental investigations of castellated monoblock structures in TEXTOR

International Nuclear Information System (INIS)

Litnovsky, A.; Philipps, V.; Wienhold, P.; Sergienko, G.; Emmoth, B.; Rubel, M.; Breuer, U.; Wessel, E.

2005-01-01

To insure the thermo-mechanical durability of ITER it is planned to manufacture the castellated armour of the divertor i.e. to split the armour into cells [W. Daener et al., Fusion Eng. Des. 61 and 62 (2002) 61]. This will cause an increase of the surface area and may lead to carbon deposition and tritium accumulation in the gaps in between cells. To investigate the processes of deposition and fuel accumulation in gaps, a castellated test-limiter was exposed to the SOL plasma of TEXTOR. The geometry of castellation used was the same as proposed for the vertical divertor target in ITER [W. Daener et al., Fusion Eng. Des. 61 and 62 (2002) 61]. After exposure the limiter was investigated with various surface diagnostic techniques. Deposited layers containing carbon, hydrogen, deuterium and boron were found both on top plasma-facing surfaces and in the gaps. The amount of deuterium in the gaps was at least 30% of that found on the top surfaces

Experimental investigations of castellated monoblock structures in TEXTOR

Science.gov (United States)

Litnovsky, A.; Philipps, V.; Wienhold, P.; Sergienko, G.; Emmoth, B.; Rubel, M.; Breuer, U.; Wessel, E.

2005-03-01

To insure the thermo-mechanical durability of ITER it is planned to manufacture the castellated armour of the divertor i.e. to split the armour into cells [W. Daener et al., Fusion Eng. Des. 61&62 (2002) 61]. This will cause an increase of the surface area and may lead to carbon deposition and tritium accumulation in the gaps in between cells. To investigate the processes of deposition and fuel accumulation in gaps, a castellated test-limiter was exposed to the SOL plasma of TEXTOR. The geometry of castellation used was the same as proposed for the vertical divertor target in ITER [W. Daener et al., Fusion Eng. Des. 61&62 (2002) 61]. After exposure the limiter was investigated with various surface diagnostic techniques. Deposited layers containing carbon, hydrogen, deuterium and boron were found both on top plasma-facing surfaces and in the gaps. The amount of deuterium in the gaps was at least 30% of that found on the top surfaces.

Control of runaway electron secondary generation by changing Z(eff)

NARCIS (Netherlands)

Pankratov, I. M.; R. Jaspers,; Finken, K.H.; Entrop, I.; Mank, G.

1998-01-01

The effect of Z(eff) on the runaway generation process by close collisions has been studied experimentally in the TEXTOR-94 tokamak in ohmic low density discharges. It is shown that the effective avalanching time increases with increasing Z(eff). This opens the possibility of controlling the runaway

Feedback control of tearing modes through ECRH with launcher mirror steering and power modulation using a line-of-sight ECE diagnostic

NARCIS (Netherlands)

Hennen, B.A.; Westerhof, E.; Nuij, P.W.J.M.; Ayten, B.; Baar, de M.R.; Bongers, W.A.; Bürger, A.; Lazzari, De D.; Oosterbeek, J.W.; Thoen, D.J.; Steinbuch, M.

2010-01-01

A demonstration of real-time feedback control for autonomous tracking and stabilization of m/n = 2/1 tearing modes in a tokamak using Electron Cyclotron Resonance Heating and Current Drive (ECRH/ECCD) is reported. The prototype system on TEXTOR combines in the same sight-line an Electron Cyclotron

Kalman filters for real-time magnetic island phase tracking

NARCIS (Netherlands)

Borgers, D. P.; Lauret, M.; M.R. de Baar,

2013-01-01

For control of neoclassical tearing modes (NTMs) and the resulting rotating magnetic islands in tokamak plasmas, the frequency and phase of the magnetic islands need to be accurately tracked in real-time. In previous experiments on TEXTOR, this was achieved using a phase-locked loop (PLL). For ASDEX

Continuous tokamaks

International Nuclear Information System (INIS)

Peng, Y.K.M.

1978-04-01

A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

Impact of screening of resonant magnetic perturbations in three dimensional edge plasma transport simulations for DIII-D

Czech Academy of Sciences Publication Activity Database

Frerichs, H.; Reiter, D.; Schmitz, O.; Cahyna, Pavel; Evans, T.; Feng, Y.; Nardon, E.

2012-01-01

Roč. 19, č. 5 (2012), 052507-052507 ISSN 1070-664X R&D Projects: GA ČR GAP205/11/2341 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * TEXTOR * divertors * plasma boundary layers * plasma density * plasma magnetohydrodynamics * plasma simulation * plasma temperature * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.376, year: 2012 http://pop.aip.org/resource/1/phpaen/v19/i5/p052507_s1

Relevance, Realization and stability of a cold layer at the plasma edge for fusion reactors

International Nuclear Information System (INIS)

1990-09-01

The workshop was dedicated to the realization and stability of a cold layer at the plasma edge for fusion reactors. The subjects of the communications presented were: impurity transport, and control, plasma boundary layers, power balance, radiation control and modifications, limiter discharges, tokamak density limit, Asdex divertor discharges, thermal stability of a radiating diverted plasma, plasma stability, auxiliary heating in Textor, detached plasma in Tore Supra, poloidal divertor tokamak, radiation cooling, neutral-particle transport, plasma scrape-off layer, edge turbulence

Kalman filters for real-time magnetic island phase tracking

International Nuclear Information System (INIS)

Borgers, D.P.; Lauret, M.; Baar, M.R. de

2013-01-01

Highlights: • We propose two Kalman filters for tracking of NTMs on ASDEX Upgrade. • The Kalman filters can track NTMs in a much larger frequency range than PLLs. • The filters are tested on synthetic and experimental data from TEXTOR and TCV. • We conclude that the unscented Kalman filter can be useful for NTM control. -- Abstract: For control of neoclassical tearing modes (NTMs) and the resulting rotating magnetic islands in tokamak plasmas, the frequency and phase of the magnetic islands need to be accurately tracked in real-time. In previous experiments on TEXTOR, this was achieved using a phase-locked loop (PLL). For ASDEX Upgrade however, the desired frequency range in which the islands are to be tracked (100 Hz–10 kHz) is much larger than is possible with a PLL. In this contribution, an extended Kalman filter (EKF) and an unscented Kalman filter (UKF) are proposed for real-time frequency, phase and amplitude tracking of sinusoidal signals, based on noisy measurements. Compared to PLLs, the EKF and UKF are able to track sinusoidal signals in a much larger frequency range. The filters are applied on synthetic data and on experimental data from the TEXTOR and TCV tokamaks, from which we conclude that the UKF can be useful for real-time control of magnetic islands on ASDEX Upgrade

Kalman filters for real-time magnetic island phase tracking

Energy Technology Data Exchange (ETDEWEB)

Borgers, D.P. [Hybrid and Networked Systems, Department of Mechanical Engineering – Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Lauret, M., E-mail: M.Lauret@tue.nl [FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, Nieuwegein (Netherlands); Control Systems Technology, Department of Mechanical Engineering – Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Baar, M.R. de [FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, Nieuwegein (Netherlands); Control Systems Technology, Department of Mechanical Engineering – Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands)

2013-11-15

Highlights: • We propose two Kalman filters for tracking of NTMs on ASDEX Upgrade. • The Kalman filters can track NTMs in a much larger frequency range than PLLs. • The filters are tested on synthetic and experimental data from TEXTOR and TCV. • We conclude that the unscented Kalman filter can be useful for NTM control. -- Abstract: For control of neoclassical tearing modes (NTMs) and the resulting rotating magnetic islands in tokamak plasmas, the frequency and phase of the magnetic islands need to be accurately tracked in real-time. In previous experiments on TEXTOR, this was achieved using a phase-locked loop (PLL). For ASDEX Upgrade however, the desired frequency range in which the islands are to be tracked (100 Hz–10 kHz) is much larger than is possible with a PLL. In this contribution, an extended Kalman filter (EKF) and an unscented Kalman filter (UKF) are proposed for real-time frequency, phase and amplitude tracking of sinusoidal signals, based on noisy measurements. Compared to PLLs, the EKF and UKF are able to track sinusoidal signals in a much larger frequency range. The filters are applied on synthetic data and on experimental data from the TEXTOR and TCV tokamaks, from which we conclude that the UKF can be useful for real-time control of magnetic islands on ASDEX Upgrade.

Effects of an RF limiter on TEXTOR's edge plasmas

International Nuclear Information System (INIS)

Boedo, J.A.; Sakawa, Y.; Gray, D.S.; Mank, G.; Noda, N.

1997-01-01

Studies directed towards the reduction of particle and heat fluxes to plasma facing components by the application of ponderomotive forces generated by radio frequency (RF) are being conducted in TEXTOR. A modified poloidal limiter is used as an antenna with up to 3 kW of RF power; the data obtained show that the plasma is repelled by the RF ponderomotive potential. The density is reduced by a factor of 2-4 and the radial decay length is substantially altered. The density near the limiter decays exponentially with RF power. The electron temperature profile changes, with the decay length becoming longer (almost flat) during the RF. The temperature in the scrape off layer (SOL) increases and its increase is roughly proportional to the RF power until it saturates, suggesting that the heating efficiency drops with power, and that improved performance is to be expected at higher powers. (orig.)

Tokamak Systems Code

International Nuclear Information System (INIS)

Reid, R.L.; Barrett, R.J.; Brown, T.G.

1985-03-01

The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

Hydrogen recycling and transport in the helical divertor of TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Clever, Meike

2010-07-01

The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm{+-}0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

Hydrogen recycling and transport in the helical divertor of TEXTOR

International Nuclear Information System (INIS)

Clever, Meike

2010-01-01

The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm±0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

Tokamak experiments

International Nuclear Information System (INIS)

Robinson, D.C.

1987-01-01

With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

PPPL tokamak program

International Nuclear Information System (INIS)

Furth, H.P.

1984-10-01

The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

Tokamak engineering mechanics

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Weiyue; Du, Shijun

2014-01-01

Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

Tokamak concept innovations

International Nuclear Information System (INIS)

1986-04-01

This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

Suppression of Runaway Electrons by Resonant Magnetic Perturbations in TEXTOR Disruptions

International Nuclear Information System (INIS)

Lehnen, M.; Bozhenkov, S. A.; Abdullaev, S. S.; TEXTOR Team,; Jakubowski, M. W.

2008-01-01

The generation of runaway electrons in the international fusion experiment ITER disruptions can lead to severe damage at plasma facing components. Massive gas injection might inhibit the generation process, but the amount of gas needed can affect, e.g., vacuum systems. Alternatively, magnetic perturbations can suppress runaway generation by increasing the loss rate. In TEXTOR disruptions runaway losses were enhanced by the application of resonant magnetic perturbations with toroidal mode number n=1 and n=2. The disruptions are initiated by fast injection of about 3x10 21 argon atoms, which leads to a reliable generation of runaway electrons. At sufficiently high perturbation levels a reduction of the runaway current, a shortening of the current plateau, and the suppression of high energetic runaways are observed. These findings indicate the suppression of the runaway avalanche during disruptions

Joint research using small tokamaks

International Nuclear Information System (INIS)

Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

2005-01-01

Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

Joint research using small tokamaks

International Nuclear Information System (INIS)

Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

2005-01-01

Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

Tokamaks. 2. ed.

International Nuclear Information System (INIS)

Wesson, John; Campbell, D.J.; Connor, J.W.

1997-01-01

It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

Tokamak devices: towards controlled fusion

International Nuclear Information System (INIS)

Trocheris, M.

1975-01-01

The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

Experimental study of the plasma structure and characterization of the transport behaviour in the laminar zone of a stochastized plasma edge; Experimentelle Untersuchung der Plasmastruktur und Charakterisierung des Transportverhaltens in der laminaren Zone einer stochastisierten Plasmarandschicht

Energy Technology Data Exchange (ETDEWEB)

Schmitz, O.

2006-07-15

For a detailed study of the plasma structure and the transport characteristics of a stochastized plasma edge at the tokamak TEXTOR the dynamic ergodic divertor (DED) was constructed, by which differently shaped external disturbing fields are statically and dynamically generated. Aim of this thgesis is to study experimentally the radial and poloidal structure of the plasma edge stochastized by the DED disturbing field and to analyze its transport characteristics. For this spatially highly resolved radial profiles of the electron density and temperature were measured by means of radiation-emission spectroscopy on thermal helium at the high- and low-field side of TEXTOR. These experimental results yield a good stating base for the validation and further development of three-dimensional transport codes.

Status of the tokamak program

Science.gov (United States)

Sheffield, J.

1981-08-01

For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

Task III: UCSD/DIII-D/Textor FY-97-98 Accomplishments

International Nuclear Information System (INIS)

Boedo, J.A.

2000-01-01

OAK (B204) Task III: UCSD/DIII-D/Textor FY-97-98 Accomplishments. A comprehensive report on the physics of pump limiters and particularly, the characterization of ALT-II, was published in Nuclear Fusion, bringing the project to a closure. The performance of the toroidal pump limiter was characterized under full auxiliary heating of 7 MW of NBI and ICRH and full pumping, as stated in the project milestones. Relevant highlights are: (1) Pumping with ALT-II allows for density control. (2) The achieved exhaust efficiency is 4% during NBI operation and near 2% during OH or ICRH operation. (3) We have shown that an exhaust efficiency of 2% is sufficient to satisfy the ash removal requirements of fusion reactors. (4) The plasma particle efflux and the pumped flux both increase with density and heating power. (5) The particle confinement time is less than the energy confinement time by a factor of 4. In summary, pumped belt limiters could provide the density control and ash exhaust requirements of fusion reactors

Present status of Tokamak research

International Nuclear Information System (INIS)

Basu, Jayanta

1991-01-01

The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

Tokamak engineering mechanics

CERN Document Server

Song, Yuntao; Du, Shijun

2013-01-01

Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

The tokamak as a neutron source

International Nuclear Information System (INIS)

Hendel, H.W.; Jassby, D.L.

1989-11-01

This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

Thermal load distribution on the ALT-II limiter of TEXTOR-94 during RI mode operation and during disruptions

International Nuclear Information System (INIS)

Finken, K.H.; Denner, T.; Mank, G.

2000-01-01

Thermographic measurements using an IR scanner have been performed at the pump limiter ALT-II of TEXTOR-94 during RI mode discharges and during disruptions. The measurements on the RI mode discharges were done to complete the TEXTOR database which had shown a structured decay pattern of the deposited power. It was found that the underlying radial heat flux can be described by two exponential decay functions. This structure, which generates an unexpected heat component close to the tangent line, has been observed in all discharge conditions including the RI mode. During disruptions, the heat is released in short pulses with a typical duration of 0.01-0.1 ms. The radial decay length of these pulses has a similar shape to the heat flux during normal discharges: it consists again of a strong component close to the tangent line with a radial decay length of 2-5 mm and probably one with a decay length of the order of 1 cm. The heat is released at the time when the edge electron temperature of the plasma drops, when intense hydrogen and carbon fluxes occur near the walls, and when electrical currents in the limiter blades are excited. In a tentative interpretation, the temporal and spatial structure of the heat pulse is attributed to the presence and growth of a laminar zone at the plasma edge, which is connected with the ergodization of the plasma edge during a disruption. (author)

First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

Science.gov (United States)

Bin, W.; Bruschi, A.; D'Arcangelo, O.; Castaldo, C.; De Angeli, M.; Figini, L.; Galperti, C.; Garavaglia, S.; Granucci, G.; Grosso, G.; Korsholm, S. B.; Lontano, M.; Mellera, V.; Minelli, D.; Moro, A.; Nardone, A.; Nielsen, S. K.; Rasmussen, J.; Simonetto, A.; Stejner, M.; Tartari, U.

2015-10-01

Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under certain circumstances, the use of the ECRH in fusion devices. An accurate characterization of the conditions for the occurrence of this phenomenon and of its consequences is thus of primary importance. Exploiting the front-steering configuration available with the real-time launcher, the implementation of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system during the very first operations in 2014. The present work has been carried out under an EUROfusion Enabling Research project. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

Status of tokamak research

International Nuclear Information System (INIS)

Rawls, J.M.

1979-10-01

An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

Tokamak ARC damage

International Nuclear Information System (INIS)

Murray, J.G.; Gorker, G.E.

1985-01-01

Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

Tokamak ARC damage

Energy Technology Data Exchange (ETDEWEB)

Murray, J.G.; Gorker, G.E.

1985-01-01

Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

International Nuclear Information System (INIS)

Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka

2014-03-01

The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

Tokamak confinement scaling laws

International Nuclear Information System (INIS)

Connor, J.

1998-01-01

The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

Radiative edge layers in limiter tokamaks

International Nuclear Information System (INIS)

Monier-Garbet, P.

1997-01-01

The characteristics of the highly radiative edge layers produced in the limiter configuration and with an open ergodic divertor are reviewed, with emphasis on the results obtained in TEXTOR and Tore Supra. In these two experiments an impurity injection technique is used to obtain highly radiating homogeneous peripheral layers. This requires that the peripheral radiation capability be maximized, while at the same time avoiding plasma core contamination; it is also necessary to insure the stability of the radiating layer. These physics issues, governing the success of the highly radiative edge scenario, are discussed. (orig.)

Large Aspect Ratio Tokamak Study

International Nuclear Information System (INIS)

Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

1980-06-01

The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

Installation, testing and first results of TEXTOR's new ICRH system

International Nuclear Information System (INIS)

Durodie, F.; Delvigne, T.; Descamps, P.; Koch, R.; Ongena, J.; Vandenplas, P.E.; Van Nieuwenhove, R.; Van Oost, G.; Weynants, R.R.; Shen, X.M.; Ecole Royale Militaire, Brussels; Messiaen, A.M.; Ecole Royale Militaire, Brussels; Huetteman, P.; Kohlhaas, W.; Stickelman, C.; Cosler, A.

1989-01-01

The new ICRH system for TEXTOR, presented at the previous SOFT conference, has been tested and installed during spring and summer of 1987. Pulses of up to 2.8 MW have been achieved representing a power density at the antenna of about 3.1 MW/m 2 and over 90% of the installed RF power. Taking into account the already achieved volttages in the system one could extrapolate that a power density of 10 MW/m 2 with an transmission efficiency well over 90% would be technically feasible. First results, such as the interesting property that, in contrast with other experiments, the two antennae in each pair operate with zero and with π phase difference with nearly the same coupling efficiency, are discussed. The testing and conditioning procedures are described. RF-leak problems encounterd at the behinning of the experimental phase are discussed. Antenna and transmission line diagnostics as well as related tuning procedures are also described. During the installation of the neutral beam injectors, from beginning of April to about end of August 1988, several modifications to the whole of the ICRH system are being implemented. (author). 5 refs.; 6 figs.; 1 tab

Fusion potential for spherical and compact tokamaks

International Nuclear Information System (INIS)

Sandzelius, Mikael

2003-02-01

The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

Fusion potential for spherical and compact tokamaks

Energy Technology Data Exchange (ETDEWEB)

Sandzelius, Mikael

2003-02-01

The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

Microwave Tokamak Experiment

International Nuclear Information System (INIS)

Anon.

1988-01-01

The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

Advanced Tokamak Stability Theory

Science.gov (United States)

Zheng, Linjin

2015-03-01

The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

2009-01-01

Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

Tokamak control simulator

International Nuclear Information System (INIS)

Edelbaum, T.N.; Serben, S.; Var, R.E.

1976-01-01

A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks

Survey of Tokamak experiments

International Nuclear Information System (INIS)

Bickerton, R.J.

1977-01-01

The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

Tokamak building-design considerations for a large tokamak device

International Nuclear Information System (INIS)

Barrett, R.J.; Thomson, S.L.

1981-01-01

Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

International Nuclear Information System (INIS)

Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P

2004-01-01

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values

Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

Energy Technology Data Exchange (ETDEWEB)

Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

2004-09-01

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

Advanced commercial tokamak study

International Nuclear Information System (INIS)

Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

1985-12-01

Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

D-D tokamak reactor assessment

International Nuclear Information System (INIS)

Baxter, D.C.; Dabiri, A.E.

1983-01-01

A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

Investigation of advanced materials for fusion alpha particle diagnostics

Energy Technology Data Exchange (ETDEWEB)

Bonheure, G., E-mail: g.bonheure@fz-juelich.de [Laboratory for Plasma Physics, Association “Euratom-Belgian State”, Royal Military Academy, Avenue de la Renaissance, 30 Kunstherlevinglaan, B-1000 Brussels (Belgium); Van Wassenhove, G. [Laboratory for Plasma Physics, Association “Euratom-Belgian State”, Royal Military Academy, Avenue de la Renaissance, 30 Kunstherlevinglaan, B-1000 Brussels (Belgium); Hult, M.; González de Orduña, R. [Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Strivay, D. [Centre Européen d’Archéométrie, Institut de Physique Nucléaire, Atomique et de Spectroscopie, Université de Liège (Belgium); Vermaercke, P. [SCK-CEN, Boeretang, B-2400 Mol (Belgium); Delvigne, T. [DSI SPRL, 3 rue Mont d’Orcq, Froyennes B-7503 (Belgium); Chene, G.; Delhalle, R. [Centre Européen d’Archéométrie, Institut de Physique Nucléaire, Atomique et de Spectroscopie, Université de Liège (Belgium); Huber, A.; Schweer, B.; Esser, G.; Biel, W.; Neubauer, O. [Forschungszentrum Jülich GmbH, Institut für Plasmaphysik, EURATOM-Assoziation, Trilateral Euregio Cluster, D-52425 Jülich (Germany)

2013-10-15

Highlights: ► We examine the feasibility of alpha particle measurements in ITER. ► We test advanced material detectors borrowed from the GERDA neutrino experiment. ► We compare experimental results on TEXTOR tokamak with our detector response model. ► We investigate the detector response in ITER full power D–T plasmas. ► Advanced materials show good signal to noise ratio and alpha particle selectivity. -- Abstract: Fusion alpha particle diagnostics for ITER remain a challenging task. Standard escaping alpha particle detectors in present tokamaks are not applicable to ITER and techniques suitable for fusion reactor conditions need further research and development [1,2]. The activation technique is widely used for the characterization of high fluence rates inside neutron reactors. Tokamak applications of the neutron activation technique are already well developed [3] whereas measuring escaping ions using this technique is a novel fusion plasma diagnostic development. Despite low alpha particle fluence levels in present tokamaks, promising results using activation technique combined with ultra-low level gamma-ray spectrometry [4] were achieved before in JET [5,6]. In this research work, we use new advanced detector materials. The material properties beneficial for alpha induced activation are (i) moderate neutron cross-sections (ii) ultra-high purity which reduces neutron-induced background activation and (iii) isotopic tailoring which increases the activation yield of the measured activation product. Two samples were obtained from GERDA[7], an experiment aimed at measuring the neutrinoless double beta decay in {sup 76}Ge. These samples, made of highly pure (9 N) germanium highly enriched to 87% in isotope Ge-76, were irradiated in real D–D fusion plasma conditions inside the TEXTOR tokamak. Comparison of the calculated and the experimentally measured activity shows good agreement. Compared to previously investigated high temperature ceramic material [8

First fusion proton measurements in TEXTOR plasmas using

Czech Academy of Sciences Publication Activity Database

Bonheure, G.; Mlynář, Jan; Van Wassenhove, G.; Hult, M.; González de Orduña, R.; Lutter, G.; Vermaercke, P.; Huber, A.; Schweer, B.; Esser, G.; Biel, W.

2012-01-01

Roč. 83, č. 10 (2012), 10D318 ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * fusion * activation * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://rsi.aip.org/resource/1/rsinak/v83/i10/p10D318_s1

Summary discussion: An integrated advanced tokamak reactor

International Nuclear Information System (INIS)

Sauthoff, N.R.

1994-01-01

The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

Material mixing on W/C twin limiter in TEXTOR-94

International Nuclear Information System (INIS)

Tanabe, T.; Ohgo, T.; Wada, M.; Rubel, M.; Philipps, V.; Seggern, J. von; Ohya, K.; Huber, A.; Pospieszczyk, A.; Schweer, B.

2000-01-01

In order to investigate the effect of mutual contamination between tungsten (W) and carbon (C) and its influence on the plasma, a W-C twin test limiter, half made of W and the other half of C, was inserted into the edge plasma of TEXTOR-94 under ohmic and NBI heating conditions. The contamination process was observed by spectroscopy, and the intensity distribution of WI showed migration of W onto the C side by the successive cycles of sputtering and prompt redeposition. On the other hand, the deposition of C on the W surface was not obvious. Most of the hydrogen (deuterium) on the limiter was found to be retained in the deposited layers and that in the deposited C layer much higher than that in the deposited W layer. This indicates that tritium retention is smaller in metallic deposits above 500 K. The AES analysis conducted after the exposure of the test limiter showed that W deposited on C reacted with the substrate to form carbides at higher temperatures. The thickness of carbide layer, and/or the content of W in C were influenced by the temperature and flux distributions, and no carbide layer was formed at the limiter edge where the temperature was relatively low

Progress of the ECH·ECCD experiments. Research progress of the ECH·ECCD experiments in tokamaks and spherical tokamaks

International Nuclear Information System (INIS)

Isayama, Akihiko; Tanaka, Hitoshi

2009-01-01

Recent progress in the ECH·ECCD study in tokamak and spherical tokamak devices is described. As for the tokamak study, results on the control of neoclassical tearing modes and sawtooth oscillations, the current profile, the internal transport barrier, the plasma start-up and the discharge cleaning are given. As for the spherical tokamak study, the plasma start-up by ECH·ECCD and the electron-Bernstein-wave heating and the current drive are described. (T.I.)

Preliminary Design of Alborz Tokamak

Science.gov (United States)

Mardani, M.; Amrollahi, R.; Saramad, S.

2012-04-01

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

Moving Divertor Plates in a Tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Zhang, H.

2009-01-01

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

Moving Divertor Plates in a Tokamak

Energy Technology Data Exchange (ETDEWEB)

S.J. Zweben, H. Zhang

2009-02-12

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

Paleoclassical transport explains electron transport barriers in RTP and TEXTOR

NARCIS (Netherlands)

Hogeweij, G. M. D.; Callen, J.D.

2008-01-01

The recently developed paleoclassical transport model sets the minimum level of electron thermal transport in a tokamak. This transport level has proven to be in good agreement with experimental observations in many cases when fluctuation-induced anomalous transport is small, i.e. in (near-) ohmic

The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1989-10-01

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

Summary report on tokamak confinement experiments

International Nuclear Information System (INIS)

1982-03-01

There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

Research using small tokamaks

International Nuclear Information System (INIS)

1991-01-01

The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

Accelerator technology in tokamaks

International Nuclear Information System (INIS)

Kustom, R.L.

1977-01-01

This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1993-04-01

This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

START: the creation of a spherical tokamak

International Nuclear Information System (INIS)

Sykes, Alan

1992-01-01

The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)

Tokamaks (Second Edition)

Energy Technology Data Exchange (ETDEWEB)

Stott, Peter [JET, UK (United Kingdom)

1998-10-01

The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

Tokamaks (Second Edition)

International Nuclear Information System (INIS)

Stott, Peter

1998-01-01

The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

Modelling dust transport in tokamaks

International Nuclear Information System (INIS)

Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

2008-01-01

The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

Compact tokamak reactors

International Nuclear Information System (INIS)

Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

1997-01-01

The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

Magnetic confinement experiment -- 1: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1994-01-01

This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

Microwave Tokamak Experiment: Overview and status

International Nuclear Information System (INIS)

1990-05-01

The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-01-01

This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

Magnetic ''islandography'' in tokamaks

International Nuclear Information System (INIS)

Callen, J.D.; Waddell, B.V.; Hicks, H.R.

1978-09-01

Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

The dynamic ergodic divertor in TEXTOR-A novel tool for studying magnetic perturbation field effects

International Nuclear Information System (INIS)

Neubauer, O.; Czymek, G.; Finken, K.H.; Giesen, B.; Huettemann, P.W.; Lambertz, H.T.; Schruff, J.

2005-01-01

Recently TEXTOR has been upgraded by the installation of the dynamic ergodic divertor (DED). The purpose of the DED is to influence transport parameters in plasma edge and core and to study the resulting effects on heat exhaust, edge cooling, impurity screening, plasma confinement and stability. Alternatively, the DED creates static or rotating multipolar helical magnetic perturbation fields of different mode patterns. A set of 16 helical coils has been installed on the inboard high-field side of the vacuum vessel. Rotating fields of up to 10 kHz can be generated. A novel coil design has been developed which fulfills the various mechanical, electrical, high frequency, thermal, and vacuum requirements. In addition to the various technical aspects of the DED design, implementation and commissioning, highlights of recent experiments will be presented. In particular the impact of the perturbation field on MHD stability and plasma rotation will be addressed

Plasma influence on throat conductance of the TEXTOR pump limiter ALT-I

International Nuclear Information System (INIS)

Hardtke, A.; Finken, K.H.; Reiter, D.; Dippel, K.H.; Goebel, D.M.; McGrath, R.T.; Sagara, A.

1989-01-01

On the TEXTOR pump limiter ALT-I conductance measurements for the backstreaming of gas from the pump limiter vessel through the pump limiter entrance have been performed. In these experiments neutral gas has been injected into the pump limiter plenum during a short pulse. The influence of the instreaming plasma results in a reduction of the conductance of the outstreaming gas. For helium the conductance is reduced to about 40% of the molecular conductance when a plasma flux of 0.8 A/cm 2 (T e =T i =11 eV) is streaming into the pump limiter throat. The reduction of the conductance for backstreaming hydrogen and deuterium under the same plasma conditions is smaller; about 70% of the molecular conductance is obtained. This reduction can be explained by an increased recycling of ions which have been produced in the throat back to the neutralizer plate. The experimental results can be reproduced by Monte Carlo neutral transport code calculations if the recycling coefficient is about 0.85 for hydrogen and deuterium and about 0.95 for helium ions. Processes causing these high recycling coefficients are discussed and their influence is estimated. (orig.)

A real time 155 GHz millimeter wave interferometer module for electron density measurement in large plasma devices

International Nuclear Information System (INIS)

Huettemann, P.W.; Waidmann, G.

1982-09-01

A homodyne, real time 155 GHz interferometer channel is described which is one module of a multichannel system for use on TEXTOR tokamak. A standing sine wave is generated in a phase bridge by transmitting a frequency modulated millimeter wave down two unequal interferometer branches. The presence of plasma produces a phase slip of the sine wave with respect to a reference signal. The phase shift is linear proportional to plasma density for expected TEXTOR plasmas. Long plasma paths give multiradian phase shifts which are recorded by a digital fringe counting system. The accuracy of phase measurement is ΔPHI = 2π/16. Phase changes of 7π/8 are accepted per modulation period. The microwave in the measurement branch of the interferometer is transmitted using a quasioptical technique. Components and technical details are described. The interferometer was tested in a simulation set-up and in two different plasma experiments. Experimental results are presented. (orig.)

Current drive by spheromak injection into a tokamak

International Nuclear Information System (INIS)

Brown, M.R.; Bellan, P.M.

1990-01-01

The authors report the first observation of current drive by spheromak injection into a tokamak due to the process of helicity injection. Current drive is observed in Caltech's ENCORE tokamak (30% increase, ΔI > 1 kA) only when both the tokamak and injected spheromak have the same sign of helicity (where helicity is defined as positive if current flows parallel to magnetic field lines and negative if anti-parallel). The initial increase (decrease) in current is accompanied by a sharp decrease (increase) in loop voltage and the increase in tokamak helicity is consistent with the helicity content of the injected spheromak. In addition, the injection of the spheromak raises the tokamak central density by a factor of six. The introduction of cold spheromak plasma causes sudden cooling of the tokamak discharge from 12 eV to 4 eV which results in a gradual decline in tokamak plasma current by a factor of three. In a second experiment, the authors inject spheromaks into the magnetized toroidal vacuum vessel (with no tokamak plasma). An m = 1 magnetic structure forms in the vessel after the spheromak undergoes a double tilt; once in the cylindrical entrance between gun and tokamak, then again in the tokamak vessel. A horizontal shift of the spheromak equilibrium is observed in the direction opposite that of the static toroidal field. In the absence of net toroidal flux, the structure develops a helical pitch as predicted by theory. Experiments with a number of refractory metal coatings have shown that tungsten and chrome coatings provide some improvement in spheromak parameters. They have also designed and will soon construct a larger, higher current spheromak gun with a new accelerator section for injection experiments on the Phaedrus-T tokamak

A need for non-tokamak approaches to magnetic fusion energy

International Nuclear Information System (INIS)

Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

1992-01-01

Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

1985-01-01

The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

1986-01-01

Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

The role of the spherical tokamak in clarifying tokamak physics

International Nuclear Information System (INIS)

Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.

1999-01-01

The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)

The effect of resonant magnetic perturbations on the impurity transport in TEXTOR-DED plasmas

International Nuclear Information System (INIS)

Greiche, Albert Josef

2009-01-01

Thermonuclear fusion provides a new mechanism for the generation of electrical power which has the perspective to serve humanity for several millions of years. One possibility to implement fusion on earth is to magnetically confine hot deuterium tritium plasmas in so called tokamaks. The fusion reactions take place in the hot plasma core. Each of the fusion reactions between deuterium and tritium yields 17.6 MeV which can be used in the process of generating electrical power. Impurities contaminate the plasma which then is cooled down and diluted. This leads to a reduction of the fusion reactions and in consequence the energy yield. The transport behaviour of the impurities in the plasma is not fully understood up to now. Nevertheless, experiments have shown that the application of resonant magnetic perturbations (RMP) can control the impurity content in the plasma. The dynamic ergodic divertor (DED) on the tokamak Textor is able to induce static and dynamic RMPs. During the application of RMPs transient impurity transport experiments with argon have been performed and the time evolution of the impurity concentrations have been monitored. The line emission intensity of the impurities in the plasma is measured in the vacuum ultraviolet (VUV) and in the soft X-ray (SXR) with the absolutely calibrated VUV spectrometer Hexos and SXR PIN diodes, respectively. The analysis of the transient impurity transport experiments is performed with the help of the transport code Strahl. The impurity flows in Strahl are described by a combination of a diffusive and a convective flow. In the computing process the code solves the coupled set of continuity equations of each of the ionization stages of an impurity. With this method the time evolution of the impurity ion densities and the line emission intensities of the ionization stages can be computed. The adaption to the experimental measurements is performed with the help of the diffusion coefficient and the drift velocity which

Advanced tokamak burning plasma experiment

International Nuclear Information System (INIS)

Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.

2001-01-01

A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

Computational studies of tokamak plasmas

International Nuclear Information System (INIS)

Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

1981-02-01

Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

Combined confinement system applied to tokamaks

International Nuclear Information System (INIS)

Ohkawa, Tihiro

1986-01-01

From particle orbit point of view, a tokamak is a combined confinement configuration where a closed toroidal volume is surrounded by an open confinement system like a magnetic mirror. By eliminating a cold halo plasma, the energy loss from the plasma becomes convective. The H-mode in diverted tokamaks is an example. Because of the favorable scaling of the energy confinement time with temperature, the performance of the tokamak may be significantly improved by taking advantage of this effect. (author)

Research into controlled fusion in tokamaks

International Nuclear Information System (INIS)

Zacek, F.

1992-01-01

During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

On the merits of heating and current drive for tearing mode stabilization

International Nuclear Information System (INIS)

De Lazzari, D.; Westerhof, E.

2009-01-01

Neoclassical tearing modes (NTMs) are magnetohydrodynamic modes that can limit the performance of high β discharges in a tokamak, leading eventually to a plasma disruption. A NTM is sustained by the perturbation of the 'bootstrap' current, which is a consequence of the pressure flattening across a magnetic island. Control and suppression of this mode can be achieved by means of electron cyclotron waves (ECWs) which allow the deposition of highly localized power at the island location. The ECW power replenishes the missing bootstrap current by generating a current perturbation either inductively, through a temperature perturbation (electron cyclotron resonance heating), or non-inductively by direct current drive (electron cyclotron current drive). Although both methods have been applied successfully to experiments showing a predominance of ECRH for medium-sized limiter tokamaks (TEXTOR, T-10) and of ECCD for mid-to-large-sized divertor tokamaks (AUG, DIII-D, JT-60), conditions determining their relative importance are still unclear. We address this problem with a numerical study focused on the contributions of heating and current drive to the temporal evolution of NTMs as described by the modified Rutherford equation. For the effects of both heating as well as current drive, simple analytical expressions have been found in terms of an efficiency fore-factor times a 'geometrical' term depending on the power deposition width w dep , location and modulation. When the magnetic island width w equals the width of the deposition profile, w ∼ w dep , both geometric terms are practically identical. Whereas for current drive the geometric term approaches a constant for small island widths and is inversely proportional to (w/w dep ) 2 for large island widths, the heating term approaches a constant for large island widths and is proportional to (w/w dep ) for small island widths. For medium-sized tokamaks (TEXTOR, AUG) the heating and current drive efficiencies are of the

The density limit in Tokamaks

International Nuclear Information System (INIS)

Alladio, F.

1985-01-01

A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs

Development of Operation Scenario for Spherical Tokamak at SNU

International Nuclear Information System (INIS)

Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.

2009-01-01

Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed

Economic considerations of commercial tokamak options

International Nuclear Information System (INIS)

Dabiri, A.E.

1986-05-01

Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m 2 , which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m 2 , will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e)

Large aspect ratio tokamak study

International Nuclear Information System (INIS)

Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Sardella, C.; Wiseman, G.W.

1979-01-01

The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured

Plasma boundary phenomena in tokamaks

International Nuclear Information System (INIS)

Stangeby, P.C.

1989-06-01

The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (n e and T e ) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

Nuclear fusion research at Tokamak Energy Ltd

International Nuclear Information System (INIS)

Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

2017-01-01

Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

Tokamak simulation code manual

International Nuclear Information System (INIS)

Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

1995-01-01

The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

Mercier criterion for high-β tokamaks

International Nuclear Information System (INIS)

Galvao, R.M.O.

1984-01-01

An expression, for the application of the Mercier criterion to numerical studies of diffuse high-β tokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the high-β tokamak approximation is derived. (L.C.) [pt

Tokamak Physics Experiment (TPX) design

International Nuclear Information System (INIS)

Schmidt, J.A.

1995-01-01

TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

Steady State Advanced Tokamak (SSAT): The mission and the machine

International Nuclear Information System (INIS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

Summary of the 1982 small tokamak users meeting

International Nuclear Information System (INIS)

Sprott, J.C.

1982-11-01

On November 1, 1982, the sixth in a series of approximately annual meetings of the users of small tokamaks was held in conjunction with the APS Division of Plasma Physics meeting at New Orleans. The meeting lasted three hours, with 34 people attending. The interest was on strengthening the ties between the small tokamaks and the large tokamaks. Accordingly, the latest meeting was dedicated to this theme, and in contrast to previous meetings, a few representatives from the large tokamaks were invited to attend and make presentations. Summaries of the various talks are included

Magnetic confinement by Tokamak: physical aspects

International Nuclear Information System (INIS)

Tachon, J.

1980-01-01

After describing the Tokamak configuration concept, the author provides an analysis of the principal physical aspects of this type of installation and concludes by estimating that the Tokamak concept is a 'plausible candidate' as a means of producing controlled thermonuclear fusion [fr

Tokamak research in the Soviet Union

International Nuclear Information System (INIS)

Strelkov, V.S.

1981-01-01

Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

Neutral beam in ALVAND IIC tokamak

International Nuclear Information System (INIS)

Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

1992-01-01

Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

Diagnostics for the Rijnhuizen Tokamak Project

NARCIS (Netherlands)

Donne, A. J. H.

1994-01-01

The research program of the Rijnhuizen Tokamak Project is concentrated on the study of plasma transport processes. The RTP tokamak is therefore equipped with an extensive set of multichannel diagnostics, including a 19-channel FIR interferometer, a 20-channel heterodyne ECE system, an 80-channel

Plasma position control in TCABR Tokamak

International Nuclear Information System (INIS)

Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

1998-01-01

The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

Global gyrokinetic simulation of tokamak transport

International Nuclear Information System (INIS)

Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T.

1998-10-01

A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or η i (η i ≡ ∂ ell nT i /∂ ell n n i ) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling

Fast IR diodes thermometer for tokamak

International Nuclear Information System (INIS)

Chen Xiangbo

2001-01-01

A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices

Prospects for Tokamak Fusion Reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.

1995-01-01

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

Presheath profiles in simulated tokamak edge plasmas

International Nuclear Information System (INIS)

LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

1988-04-01

The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

Robust Sliding Mode Control for Tokamaks

Directory of Open Access Journals (Sweden)

I. Garrido

2012-01-01

Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

Three novel tokamak plasma regimes in TFTR

International Nuclear Information System (INIS)

Furth, H.P.

1985-10-01

Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

Internal disruption in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1990-01-01

A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described

Internal disruptions in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1990-01-01

Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described

ICRF antenna and feedthrough development at ORNL

International Nuclear Information System (INIS)

Baity, F.W.; Bryan, W.E.; Hoffman, D.J.; Owens, T.L.; Rettig, C.L.; Schechter, D.E.

1985-01-01

The rf technology program at Oak Ridge National Laboratory is highlighted. Simply stated, the objective of the program is to develop the technology for ion cyclotron range of frequencies heating of the fusion machines leading up to a reactor. Results from an investigation of the importance of current strap shaping in compact antenna design are presented. Designs of the Doublet III-D and Advanced Toroidal Facility compact loop launchers are described, as are the vacuum feedthroughs for the West German tokamak TEXTOR and the Tandem Mirror Experiment Upgrade (TMX-U)

Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak; Projeto e analise dos circuitos de producao de campo magnetico toroidal e de formacao do plasma do Tokamak ETE (Experimento Tokamak Esferico)

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Luis Filipe F.P.W.; Bosco, Edson del

1994-12-31

This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs.

Estimation of Zeff in Novillo Tokamak

International Nuclear Information System (INIS)

Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.

1996-01-01

We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)

Time resolved observation of the erosion of boron containing protective coatings on wall elements of TEXTOR-94 by means of colorimetry

International Nuclear Information System (INIS)

Wienhold, P.; Esser, H.G.; Winter, J.

1997-01-01

The paper describes the investigation of the progressive erosion of an a-B:D coated test piece during 22 pulses in the SOL of TEXTOR-94. Time resolved observations by colorimetry reveal that the erosion proceeds in steps: during an intermediate phase the rates do not exceed ∼-1.5 nm/s. Thereafter they jump to about -6 nm/s. This is due to carbon incorporation and triggered when the concentration approaches ∼40%. The changing composition may influence the ratio of the BII/CII emission near the surface. The process ends with a carbon rich layer on the remnants of the boron film. Combination of different investigations (AES, NRA, EPMA) results in a preliminary model description. (orig.)

Study of density fluctuations during MHD activity, soft landing discharges and major disruptions in TEXTOR using CO2 laser collective scattering

International Nuclear Information System (INIS)

Boileau, A.; Van Andel, H.W.H.; Hellermann, M. von; Rogister, A.

1987-01-01

A modulation of microturbulence is observed in TEXTOR during low mode number MHD activity using CO 2 laser collective scattering. This is accomplished by a strong enhancement of density fluctuations near ka s approx. = 3 at the end of soft landing discharges and a displacement of the frequency spectrum towards lower frequencies. The increase is most significant for rapid rampdown of the plasma current accompanied by strong MHD activity but also occurs when the latter is not detected. The evolution of microturbulence is also studied during major plasma disruptions. It was found that disruptions without MHD precursor oscillations are characterized by a rapid increase in the density fluctuations starting approx. 100 ms before plasma disruption. (author)

Numerical Tokamak Project code comparison

International Nuclear Information System (INIS)

Waltz, R.E.; Cohen, B.I.; Beer, M.A.

1994-01-01

The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

Confinement and diffusion in tokamaks

International Nuclear Information System (INIS)

McWilliams, R.

1988-01-01

The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

Numerical studies of edge localized instabilities in tokamaks

International Nuclear Information System (INIS)

Wilson, H.R.; Snyder, P.B.; Huysmans, G.T.A.; Miller, R.L.

2002-01-01

A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code

An overview on plasma disruption mitigation and avoidance in tokamak

International Nuclear Information System (INIS)

He Kaihui; Pan Chuanhong; Feng Kaiming

2002-01-01

Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

Tokamak COMPASS

Czech Academy of Sciences Publication Activity Database

Řípa, Milan; Křenek, Petr

2011-01-01

Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

An enhanced tokamak startup model

Science.gov (United States)

Goswami, Rajiv; Artaud, Jean-François

2017-01-01

The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

Spherical tokamak development in Brazil

International Nuclear Information System (INIS)

Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group

2003-01-01

This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Enhancement of confinement in tokamaks

International Nuclear Information System (INIS)

Furth, H.P.

1986-01-01

The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

Spherical tokamak development in Brazil

Energy Technology Data Exchange (ETDEWEB)

Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

2003-07-01

The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Spherical tokamak development in Brazil

Energy Technology Data Exchange (ETDEWEB)

Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

2003-12-01

This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Spherical tokamak development in Brazil

International Nuclear Information System (INIS)

Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes

2003-01-01

The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Bibliography of fusion product physics in tokamaks

International Nuclear Information System (INIS)

Hively, L.M.; Sigmar, D.J.

1989-09-01

Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

International Nuclear Information System (INIS)

Sheffield, J.

1985-01-01

During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma

MAST: a Mega Amp Spherical Tokamak

International Nuclear Information System (INIS)

Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.

1995-01-01

The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)

The ARIES-I tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

Economic evaluation of tokamak power plants

International Nuclear Information System (INIS)

Reid, R.L.; Steiner, D.

1977-01-01

This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

International Nuclear Information System (INIS)

Azizov, E A

2012-01-01

The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined. (conferences and symposia)

Overview of Tokamak Results

International Nuclear Information System (INIS)

Unterberg, Bernhard; Samm, Ulrich

2004-01-01

An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas

Plea for stellarator funding raps tokamaks

International Nuclear Information System (INIS)

Blake, M.

1992-01-01

The funding crunch in magnetic confinement fusion development has moved the editor of a largely technical publication to speak out on a policy issue. James A. Rome, who edits Stellarator News from the Fusion Energy Division at Oak Ridge National Laboratory, wrote an editorial that appeared on the front page of the May 1992 issue. It was titled open-quotes The US Stellarator Program: A Time for Renewal,close quotes and while it focused chiefly on that subject (and lamented the lack of funding for the operation of the existing ATF stellarator at Oak Ridge), it also cited some of the problems inherent in the mainline MCF approach--the tokamak--and stated that if the money can be found for further tokamak design upgrades, it should also be found for stellarators. Rome wrote, open-quotes There is growing recognition in the US, and elsewhere, that the conventional tokamak does not extrapolate to a commercially competitive energy source except with very high field coils ( 1000 MWe).close quotes He pointed up open-quotes the difficulty of simultaneously satisfying conflicting tokamak requirements for efficient current drive, high bootstrap-current fraction, complete avoidance of disruptions, adequate beta limits, and edge-plasma properties compatible with improved (H-mode) confinement and acceptable erosion of divertor plates.close quotes He then called for support for the stellarator as open-quotes the only concept that has performance comparable to that achieved in tokamaks without the plasma-current-related limitations listed above.close quotes

Electron thermal transport in tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Konings, J A

1994-11-30

The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

1997-12-31

This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

1998-12-31

This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

Analysis of tokamak plasma confinement modes using the fast

Indian Academy of Sciences (India)

The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and ...

Empirical scaling for present Ohmically heated tokamaks

International Nuclear Information System (INIS)

Daughney, C.

1975-01-01

Experimental results from the Adiabatic Toroidal Compressor (ATC) tokamak are used to obtain empirical scaling laws for the average electron temperature and electron energy confinement time as functions of the average electron density, the effective ion charge, and the plasma current. These scaling laws are extended to include dependence upon minor and major plasma radius and toroidal field strength through a comparison of the various tokamaks described in the literature. Electron thermal conductivity is the dominant loss process for the ATC tokamak. The parametric dependences of the observed electron thermal conductivity are not explained by present theoretical considerations. The electron temperature obtained with Ohmic heating is shown to be a function of current density - which will not be increased in the next generation of large tokamaks. However, the temperature dependence of the electron energy confinement time suggests that significant improvement in confinement time will be obtained with supplementary electron heating. (author)

Isotope exchange experiments on TEXTOR and TORE SUPRA using Ion Cyclotron Wall Conditioning and Glow Discharge Conditioning

International Nuclear Information System (INIS)

Wauters, T.; Douai, D.; Lyssoivan, A.; Philipps, V.; Bremond, S.; Freisinger, M.; Kreter, A.; Lombard, G.; Marchuk, O.; Mollard, P.; Paul, M.K.; Pegourie, B.; Reimer, H.; Sergienko, G.; Tsitrone, E.; Vervier, M.; Van Wassenhove, G.; Wuenderlich, D.; Van Schoor, M.; Van Oost, G.

2011-01-01

This contribution reports on isotope exchange studies with both Ion Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) in TEXTOR and TORE SUPRA. The discharges have been carried out in H 2 , D 2 (ICWC and GDC) and He/H 2 mixtures (ICWC). The higher reionization probability in ICWC compared to GDC, following from the 3 to 4 orders of magnitude higher electron density, leads to a lower pumping efficiency of wall desorbed species. GDC has in this analysis (5-10) times higher removal rates of wall desorbed species than ICWC, although the wall release rate is 10 times higher in ICWC. Also the measured high retention during ICWC can be understood as an effect of the high reionization probability. The use of short RF pulses (∼1 s) followed by a larger pumping time significantly improves the ratio of implanted over recovered particles, without severely lowering the total amount of removed particles.

Simulation of a major tokamak disruption

International Nuclear Information System (INIS)

White, R.B.; Monticello, D.A.; Rosenbluth, M.N.

1977-08-01

It is known that the internal tokamak disruption leads to a current profile which is flattened inside the surface where the safety factor equals unity. It is shown that such a profile can lead to m = 2 magnetic islands which grow to fill a substantial part of the tokamak cross section in a time consistent with the observations of the major disruption

Edge plasma diagnostics on Tore Supra tokamak

International Nuclear Information System (INIS)

Fujita, Junji

1991-01-01

From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)

Integral torque balance in tokamaks

International Nuclear Information System (INIS)

Pustovitov, V.D.

2011-01-01

The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.

Tokamak and RFP ignition requirements

International Nuclear Information System (INIS)

Werley, K.A.

1991-01-01

A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

Physics design of an ultra-long pulsed tokamak reactor

International Nuclear Information System (INIS)

Ogawa, Y.; Inoue, N.; Wang, J.; Yamamoto, T.; Okano, K.

1993-01-01

A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

System assessment of helical reactors in comparison with tokamaks

International Nuclear Information System (INIS)

Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

2002-10-01

A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)

Magnetic field structure of experimental high beta tokamak equilibria

International Nuclear Information System (INIS)

Deniz, A.V.

1986-01-01

The magnetic field structure of several low and high β tokamaks in the Columbia High Beta Tokamak (HBT) was determined by high-impedance internal magnetic probes. From the measurement of the magnetic field, the poloidal flux, toroidal flux, toroidal current, and safety factor are calculated. In addition, the plasma position and cross-sectional shape are determined. The extent of the perturbation of the plasma by the probe was investigated and was found to be acceptably small. The tokamaks have major radii of approx.0.24 m, minor radii of approx.0.05 m, toroidal plasma current densities of approx.10 6 A/m 2 , and line-integrated electron densities of approx.10 20 m -2 . The major difference between the low and high β tokamaks is that the high β tokamak was observed to have an outward shift in major radius of both the magnetic center and peak of the toroidal current density. The magnetic center moves inward in major radius after 20 to 30 μsec, presumably because the plasma maintains major radial equilibrium as its pressure decreases from radiation due to impurity atoms. Both the equilibrium and the production of these tokamaks from a toroidal field stabilized z-pinch are modeled computationally. One tokamak evolves from a state with low β features, through a possibly unstable state, to a state with high β features

Reduction of the turbulent blob transport in the scrape-off layer by a resonant magnetic perturbation in TEXTOR

International Nuclear Information System (INIS)

Xu, Y.; Weynants, R.R.; Van Schoor, M.; Vergote, M.; Jachmich, S.; Jakubowski, M.W.; Mitri, M.; Schmitz, O.; Unterberg, B.; Reiser, D.; Finken, K.H.; Lehnen, M.; Beyer, P.

2009-01-01

During the static 6/2 Dynamic Ergodic Divertor experiments in TEXTOR, a significant influence of the edge resonant magnetic perturbation (RMP) on the turbulent blob transport in the scrape-off layer (SOL) has been observed. In ohmic discharges without the RMP, the blobs extend 4-5 cm deep into the SOL with a radially outward moving speed of about 1 km s -1 and hence constitute a strong outflow of mass. With the application of the RMP, the blob amplitudes and their radially moving velocity are both reduced, resulting in a significant reduction of the blob transport in the SOL. The reduction effect of the RMP on blobs is found to be robust to changes in the operational regime and to phasing variations of the RMP as well. The blob dynamics appears to be consistent with the paradigm of the radial motions of the blob structures driven by the interchange instability.

INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

International Nuclear Information System (INIS)

HUMPHREYS, D.A.; FERRON, J.R.; JOHNSON, R.D; LEUER, J.A.; PENAFLOR, B.G.; WALKER, M.L.; WELANDER, A.S.; KHAYRUTDINOV, R.R; DOKOUKA, V.; EDGELL, D.H.; FRANSSON, C.M.

2004-03-01

OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

Full power in the European tokamak

International Nuclear Information System (INIS)

Lallia, P.P.; Hugon, M.

1987-01-01

A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years

ARIES tokamak reactor study

International Nuclear Information System (INIS)

Steiner, D.; Embrechts, M.

1990-07-01

This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

Runaway electron generation as possible trigger for enhancement of magnetohydrodynamic plasma activity and fast changes in runaway beam behavior

International Nuclear Information System (INIS)

Pankratov, I. M.; Zhou, R. J.; Hu, L. Q.

2015-01-01

Peculiar phenomena were observed during experiments with runaway electrons: rapid changes in the synchrotron spot and its intensity that coincided with stepwise increases in the electron cyclotron emission (ECE) signal (cyclotron radiation of suprathermal electrons). These phenomena were initially observed in TEXTOR (Tokamak Experiment for Technology Oriented Research), where these events only occurred in the current decay phase or in discharges with thin stable runaway beams at a q = 1 drift surface. These rapid changes in the synchrotron spot were interpreted by the TEXTOR team as a fast pitch angle scattering event. Recently, similar rapid changes in the synchrotron spot and its intensity that coincided with stepwise increases in the non-thermal ECE signal were observed in the EAST (Experimental Advanced Superconducting Tokamak) runaway discharge. Runaway electrons were located around the q = 2 rational magnetic surface (ring-like runaway electron beam). During the EAST runaway discharge, stepwise ECE signal increases coincided with enhanced magnetohydrodynamic (MHD) activity. This behavior was peculiar to this shot. In this paper, we show that these non-thermal ECE step-like jumps were related to the abrupt growth of suprathermal electrons induced by bursting electric fields at reconnection events during this MHD plasma activity. Enhancement of the secondary runaway electron generation also occurred simultaneously. Local changes in the current-density gradient appeared because of local enhancement of the runaway electron generation process. These current-density gradient changes are considered to be a possible trigger for enhancement of the MHD plasma activity and the rapid changes in runaway beam behavior

Runaway electron generation as possible trigger for enhancement of magnetohydrodynamic plasma activity and fast changes in runaway beam behavior

Energy Technology Data Exchange (ETDEWEB)

Pankratov, I. M., E-mail: pankratov@kipt.kharkov.ua, E-mail: rjzhou@ipp.ac.cn [Institute of Plasma Physics, NSC Kharkov Institute of Physics and Technology, Academicheskaya Str. 1, 61108 Kharkov (Ukraine); Zhou, R. J., E-mail: pankratov@kipt.kharkov.ua, E-mail: rjzhou@ipp.ac.cn; Hu, L. Q. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2015-07-15

Peculiar phenomena were observed during experiments with runaway electrons: rapid changes in the synchrotron spot and its intensity that coincided with stepwise increases in the electron cyclotron emission (ECE) signal (cyclotron radiation of suprathermal electrons). These phenomena were initially observed in TEXTOR (Tokamak Experiment for Technology Oriented Research), where these events only occurred in the current decay phase or in discharges with thin stable runaway beams at a q = 1 drift surface. These rapid changes in the synchrotron spot were interpreted by the TEXTOR team as a fast pitch angle scattering event. Recently, similar rapid changes in the synchrotron spot and its intensity that coincided with stepwise increases in the non-thermal ECE signal were observed in the EAST (Experimental Advanced Superconducting Tokamak) runaway discharge. Runaway electrons were located around the q = 2 rational magnetic surface (ring-like runaway electron beam). During the EAST runaway discharge, stepwise ECE signal increases coincided with enhanced magnetohydrodynamic (MHD) activity. This behavior was peculiar to this shot. In this paper, we show that these non-thermal ECE step-like jumps were related to the abrupt growth of suprathermal electrons induced by bursting electric fields at reconnection events during this MHD plasma activity. Enhancement of the secondary runaway electron generation also occurred simultaneously. Local changes in the current-density gradient appeared because of local enhancement of the runaway electron generation process. These current-density gradient changes are considered to be a possible trigger for enhancement of the MHD plasma activity and the rapid changes in runaway beam behavior.

Runaway electron generation as possible trigger for enhancement of magnetohydrodynamic plasma activity and fast changes in runaway beam behavior

Science.gov (United States)

Pankratov, I. M.; Zhou, R. J.; Hu, L. Q.

2015-07-01

Peculiar phenomena were observed during experiments with runaway electrons: rapid changes in the synchrotron spot and its intensity that coincided with stepwise increases in the electron cyclotron emission (ECE) signal (cyclotron radiation of suprathermal electrons). These phenomena were initially observed in TEXTOR (Tokamak Experiment for Technology Oriented Research), where these events only occurred in the current decay phase or in discharges with thin stable runaway beams at a q = 1 drift surface. These rapid changes in the synchrotron spot were interpreted by the TEXTOR team as a fast pitch angle scattering event. Recently, similar rapid changes in the synchrotron spot and its intensity that coincided with stepwise increases in the non-thermal ECE signal were observed in the EAST (Experimental Advanced Superconducting Tokamak) runaway discharge. Runaway electrons were located around the q = 2 rational magnetic surface (ring-like runaway electron beam). During the EAST runaway discharge, stepwise ECE signal increases coincided with enhanced magnetohydrodynamic (MHD) activity. This behavior was peculiar to this shot. In this paper, we show that these non-thermal ECE step-like jumps were related to the abrupt growth of suprathermal electrons induced by bursting electric fields at reconnection events during this MHD plasma activity. Enhancement of the secondary runaway electron generation also occurred simultaneously. Local changes in the current-density gradient appeared because of local enhancement of the runaway electron generation process. These current-density gradient changes are considered to be a possible trigger for enhancement of the MHD plasma activity and the rapid changes in runaway beam behavior.

Improvement of tokamak performance by injection of electrons

International Nuclear Information System (INIS)

Ono, Masayuki.

1992-12-01

Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas

Neutronic analysis of fusion tokamak devices by PHITS

International Nuclear Information System (INIS)

Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

2011-01-01

A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

Multi-mode remote participation on the GOLEM tokamak

International Nuclear Information System (INIS)

Svoboda, V.; Huang, B.; Mlynar, J.; Pokol, G.I.; Stoeckel, J.; Vondrasek, G.

2011-01-01

The GOLEM tokamak (formerly CASTOR) at Czech Technical University is demonstrated as an educational tokamak device for domestic and foreign students. Remote participation of several foreign universities (in Hungary, Belgium, Poland and Costa Rica) has been successfully performed. A unique feature of the GOLEM device is functionality which enables complete remote participation and control, solely through Internet access. Basic remote control is possible either in online mode via WWW/SSH interface or offline mode using batch processing code. Discharge parameters are set in each case to configure the tokamak for a plasma discharge. Using the X11 protocol it is possible to control in an advanced mode many technological aspects of the tokamak operation, including: i) vacuum pump initialization, ii) chamber baking, iii) charging of power supplies, iv) plasma discharge scenario, v) data acquisition system.

Tokamak Physics Experiment (TPX) power supply design and development

International Nuclear Information System (INIS)

Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.

1995-01-01

The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes

High performance operational limits of tokamak and helical systems

International Nuclear Information System (INIS)

Yamazaki, Kozo; Kikuchi, Mitsuru

2003-01-01

The plasma operational boundaries of tokamak and helical systems are surveyed and compared with each other. Global confinement scaling laws are similar and gyro-Bohm like, however, local transport process is different due to sawtooth oscillations in tokamaks and ripple transport loss in helical systems. As for stability limits, achievable tokamak beta is explained by ideal or resistive MHD theories. On the other hand, beta values obtained so far in helical system are beyond ideal Mercier mode limits. Density limits in tokamak are often related to the coupling between radiation collapse and disruptive MHD instabilities, but the slow radiation collapse is dominant in the helical system. The pulse length of both tokamak and helical systems is on the order of hours in small machines, and the longer-pulsed good-confinement plasma operations compatible with radiative divertors are anticipated in both systems in the future. (author)

Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak

International Nuclear Information System (INIS)

Barbosa, Luis Filipe F.P.W.; Bosco, Edson del.

1994-01-01

This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs

LHCD experiments on tokamak CASTOR

International Nuclear Information System (INIS)

Zacek, F.; Badalec, J.; Jakubka, J.; Kryska, L.; Preinhaelter, J.; Stoeckel, J.; Valovic, M.; Nanobashvili, S.; Weixelbaum, L.; Wenzel, U.; Spineanu, F.; Vlad, M.

1990-10-01

A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

Measurement of the hydrogen recombination coefficient in the TEXT tokamak as a function of outgassing and power radiated during tokamak discharges

International Nuclear Information System (INIS)

Langley, R.A.; Rowan, W.L.; Bravenec, R.V.; Nelin, K.

1986-10-01

The global recombination rate coefficient k/sub r/ for hydrogen has been measured in the TEXT tokamak vacuum vessel for various surface conditions. An attempt was made to correlate the measured values of k/sub r/ with residual gas analyzer (RGA) data taken before each measurement of k/sub r/ and with the power radiated during tokamak discharges produced after each measurement of k/sub r/. The results show that k/sub r/ increases during a series of tokamak discharges, k/sub r/ is relatively insensitive to power radiated during tokamak discharges, and k/sub r/ increases with the RGA measurements of mass 28 and 40 but not with those of mass 18. In addition, it was found that the mass 18 (H 2 O) signal decreases as glow discharge experiments with hydrogen were performed

Behavior of oxygen impurities in tokamak. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

El-Sharif, R N; Beket, A H [Plasma and Nuclear Fusion Department, Nuclear Research Center, Atomic Energy Aurhority, Cairo (Egypt)

1996-03-01

Impurity transport in tokamak plasma is a subject of great importance in present day tokamak experiments. The transport of oxygen as an impurity element in small tokamak was studied theoretically. The viscosity coefficient of oxygen has been calculated in different approximation 13 and 21 moment approximation, taking into consideration {chi}>>1,{chi}{omega}{sub c} {tau}. It was found that in 21 moment approximation additional terms added to the perturbation from equilibrium leads to increase in viscosity coefficients than in 13 moments approximation. 9 figs.

Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

Indian Academy of Sciences (India)

The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...

Comparative studies of stellarator and tokamak transport

Energy Technology Data Exchange (ETDEWEB)

Stroth, U; Burhenn, R; Geiger, J; Giannone, L.; Hartfuss, H J; Kuehner, G; Ledl, L; Simmet, E E; Walter, H [Max-Planck-Inst. fuer Plasmaphysik, IPP-Euratom Association, Garching (Germany); ECRH Team; W7-AS Team

1997-09-01

Transport properties in the W7-AS stellarator and in tokamaks are compared. The parameter dependences and the absolute values of the energy confinement time are similar. Indications are found that the density dependence, which is usually observed in stellarator confinement, can vanish above a critical density. The density dependence in stellarators seems to be similar to that in the linear ohmic confinement regime, which, in small tokamaks, extends to high density values, too. Because of the similarity in the gross confinement properties, transport in stellarators and tokamaks should not be dominated by the parameters which are very different in the two concepts, i.e. magnetic shear, major rational values of the rotational transform and plasma current. A difference in confinement is that there exists evidence for pinches in the particle and, possibly, energy transport channels in tokamaks whereas in stellarators no pinches have been observed, so far. In order to study the effect of plasma current and toroidal electric fields, stellarator discharges were carried out with an increasing amount of plasma current. From these experiments, no clear evidence of a connection of pinches with these parameters is found. The transient response in W7-AS plasmas can be described in terms of a non-local model. As in tokamaks, also cold pulse experiments in W7-AS indicate the importance of non-local transport. (author). 8 refs, 5 figs.

Spherical tokamak without external toroidal fields

International Nuclear Information System (INIS)

Kaw, P.K.; Avinash, K.; Srinivasan, R.

2001-01-01

A spherical tokamak design without external toroidal field coils is proposed. The tokamak is surrounded by a spheromak shell carrying requisite force free currents to produce the toroidal field in the core. Such equilibria are constructed and it is indicated that these equilibria are likely to have robust ideal and resistive stability. The advantage of this scheme in terms of a reduced ohmic dissipation is pointed out. (author)

Tokamak residual zonal flow level in near-separatrix region

International Nuclear Information System (INIS)

Bing-Ren, Shi

2010-01-01

Residual zonal flow level is calculated for tokamak plasmas in the near-separatrix region of a diverted tokamak. A recently developed method is used to construct an analytic divertor tokamak configuration. It is shown that the residual zonal flow level becomes smaller but still keeps finite near the separatrix because the neoclassical polarisation mostly due to the trapped particles goes larger in this region. (fluids, plasmas and electric discharges)

Effects of boronisation on the plasma parameters in TCA

International Nuclear Information System (INIS)

Dudok de Wit, Th.; Duval, B.P.; Hollenstein, Ch.; Joye, B.

1990-01-01

Wall conditioning and deposition of low Z materials on the first wall and limiters play an important role in plasma impurity control. Carbon film deposition (carbonisation) is already used on many Tokamaks. As proposed by Veprek, a film containing boron and carbon would be more resistant to chemical erosion and could also getter the oxygen. This procedure (boronisation) has been tried on Textor, Asdex and recently on TCA. The TCA vacuum vessel, the 8 rf antenna groups and 4 antenna screens are stainless steel and there are 4 carbon limiters placed in one poloidal plane. (author) 6 refs., 3 figs

The scientific program of the Tokamak de Varennes

International Nuclear Information System (INIS)

Daughney, C.C.

1989-01-01

The Tokamak de Varennes (TdeV) is the principal research tool of the Centre canadien de fusion magnetique (CCFM). This article places the Tokamak de Varennes within the framework of the Canadian National Fusion Program (NFP) and describes the scientific program of the TdeV as it was presented at the April 1989 meeting of the CCFM Advisory Committee. The CCFM scientific plant contains three main elements: tokamak development, research on transport and equilibrium in plasmas, and research on the plasma-wall problem. Phase I of the experimental program, commissioning the tokamak and the diagnostic systems, has been completed. Phase II of the experimental program will begin in December 1989 with the plasma boundary defined by a magnetic divertor and the power supplies and vacuum system capable of creating a sequence of one-second plasma pulses. (3 figs., 3 refs.) (L.L.)

Second regime tokamak operation at large aspect ratio

International Nuclear Information System (INIS)

Navratil, G.A.

1989-01-01

This paper reviews the need for high beta in economic tokamak reactors and summarizes recent results on the scaling of the second regime beta limit for high-n ballooning modes using optimized pressure profiles as well as results on low-n mode stability at the first regime beta limit from the Columbia HBT tokamak. While several experiments have studied ballooning limits using high εβ p plasmas, the most important question for the use of the second stability regime for tokamak reactor improvement is how to achieve these high values of εβ p while at the same time increasing the value of beta to several times the Troyon beta limit. An approach to the study of this key question on beta limits using modest sized, large aspect ratio tokamaks is described. (author). 28 refs, 7 figs, 1 tab

Hard X-ray studies on the Castor tokamak

International Nuclear Information System (INIS)

Mlynar, J.

1990-04-01

The electron runaway processes in tokamaks are discussed with regard to hard X radiation measurements. The origin and confinement of runaway electrons, their bremsstrahlung spectra and the influence of lower hybrid current drive on the distribution of high-energy electrons are analyzed for the case of the Castor tokamak. The hard X-ray spectrometer designed for the Castor tokamak is also described and preliminary qualitative results of hard X-ray measurements are presented. The first series of integral measurements made it possible to map the azimuthal dependence of the hard X radiation

Real-time horizontal position control for Aditya-upgrade tokamak

International Nuclear Information System (INIS)

Kumar, Rohit; Ghosh, Joydeep; Tanna, Rakesh L.

2015-01-01

Position of plasma column is required to be controlled in real time for improved operation of any tokamak. A PID based system for real-time horizontal plasma position control has been designed for Aditya Upgrade tokamak. Modelling of transfer functions of actuators, plasma and diagnostic system are carried out for ADITYA-U tokamak. The PID controller is optimized using MATLAB-SIMULINK for horizontal position control. Further feed-forward loop is implemented where disturbance due to density variation is suppressed, which results in improved performance as compared to conventional PID operation. In this paper the detailed design of the whole system for real time control of plasma horizontal position in Aditya Upgrade tokamak is presented. (author)

Varennes Tokamak

International Nuclear Information System (INIS)

Cumyn, P.B.

A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

Tokamak WEST připraven ke startu!

Czech Academy of Sciences Publication Activity Database

Řípa, Milan

Květen (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * WEST * Tora Supra * divertor Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/2014-tokamak-west-pripraven-ke- start u

Nuclear micro-beam analysis of deuterium distribution in carbon fibre composites for controlled fusion devices

International Nuclear Information System (INIS)

Petersson, P.; Kreter, A.; Possnert, G.; Rubel, M.

2010-01-01

Probes made of carbon fibre composite NB41 were exposed to deuterium plasmas in the TEXTOR tokamak and in a simulator of plasma-wall interactions, PISCES. The aim was to assess the deuterium retention and its lateral and depth distribution. The analysis was performed by means of D( 3 He, p) 4 He and 12 C( 3 He, p) 14 N nuclear reactions analysis using a standard (1 mm spot) and micro-beam (20 μm resolution). The measurements have revealed non uniform distribution of deuterium atoms in micro-regions: differences by a factor of 3 between the maximum and minimum deuterium concentrations. The differences were associated with the orientation and type of fibres for samples exposed in PICSES. For surface structure in the erosion zone of samples exposed to a tokamak plasma the micro-regions were more complex. Depth profiling has indicated migration of fuel into the bulk of materials.

Self-organized Te Redistribution during Driven Reconnection Processes in High Temperature Plasmas

International Nuclear Information System (INIS)

Park, H.K.; Mazzucato, E.; Luhmann, N.C. Jr.; Domier, C.W.; Xia, Z.; Munsat, T.; Donne, A.J.H.; Classen, I.G.J.; van de Pol, M.J.

2005-01-01

Two-dimensional (2-D) images of electron temperature fluctuations with a high temporal and spatial resolution were employed to study the sawtooth oscillation in TEXTOR tokamak plasmas. The new findings are: (1) 2-D images revealed that the reconnection is localized and permitted the determination of the physical dimensions of the reconnection zone in the poloidal and toroidal planes. (2) The combination of a pressure driven mode and a kink instability leads to an 'X-point' reconnection process. (3) Reconnection can take place anywhere along the q∼1 rational magnetic surface (both high and low field sides). (4) Heat flow from the core to the outside of the inversion radius during the reconnection time is highly asymmetric and the behavior is collective. These new findings are compared with the characteristics of various theoretical models and experimental results for the study of the sawtooth oscillation in tokamak plasmas

Tokamak reactor startup power

International Nuclear Information System (INIS)

Weldon, D.M.; Murray, J.G.

1983-01-01

Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

Conceptual design of a commercial tokamak reactor using resistive magnets

International Nuclear Information System (INIS)

LeClaire, R.J. Jr.

1988-01-01

The future of the tokamak approach to controlled thermonuclear fusion depends in part on its potential as a commercial electricity-producing device. This potential is continually being evaluated in the fusion community using parametric, system, and conceptual studies of various approaches to improving tokamak reactor design. The potential of tokamaks using resistive magnets as commercial electricity-producing reactors is explored. Parametric studies have been performed to examine the major trade-offs of the system and to identify the most promising configurations for a tokamak using resistive magnets. In addition, a number of engineering issues have been examined including magnet design, blanket/first-wall design, and maintenance. The study indicates that attractive design space does exist and presents a conceptual design for the Resistive Magnet Commercial Tokamak Reactor (RCTR). No issue has been identified, including recirculating power, that would make the overall cost of electricity of RCTR significantly different from that of a comparably sized superconducting tokamak. However, RCTR may have reliability and maintenance advantages over commercial superconducting magnet devices

Time - resolved thermography at Tokamak T-10

International Nuclear Information System (INIS)

Grunow, C.; Guenther, K.; Lingertat, J.; Chicherov, V.M.; Evstigneev, S.A.; Zvonkov, S.N.

1987-01-01

Thermographic experiments were performed at T-10 tokamak to investigate the thermic coupling of plasma and the limiter. The limiter is an internal equipment of the vacuum vessel of tokamak-type fusion devices and the interaction of plasma with limiter results a high thermal load of limiter for short time. In according to improve the limiter design the temperature distribution on the limiter surface was measured by a time-resolved thermographic method. Typical isotherms and temperature increment curves are presented. This measurement can be used as a systematic plasma diagnostic method because the limiter is installed in the tokamak whereas special additional probes often disturb the plasma discharge. (D.Gy.) 3 refs.; 7 figs

Study on assembly techniques and procedures for ITER tokamak device

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi; Ue, Koichi; Shimizu, Katsusuke; Onozuka, Masanori

2006-06-01

The International Thermonuclear Experimental Reactor (ITER) tokamak is mainly composed of a doughnut-shaped vacuum vessel (VV), four types of superconducting coils such as toroidal field coils (TF coils) arranged around the VV, and in-vessel components, such as blanket and divertor. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of the VV and the TF coil are required to be a high accuracy of ±3 mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements as well as the configuration of the tokamak with large size and heavy weight. Based on the above backgrounds, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The tokamak assembly operations are categorized into six work break down structures (WBS), i.e., (1) preparation for assembly operations, (2) sub-assembly of the 40deg sector composed of 40deg VV sector, two TF coils and thermal shield between VV and TF coil at the assembly hall, (3) completion of the doughnut-shaped tokamak assembly composed of nine 40deg sectors in the cryostat at the tokamak pit, (4) measurement of positioning and accuracy after the completion of the tokamak assembly, (5) installation of the ex-vessel components, and (6) installation of in-vessel components. In the present report, two assembly operations of (2) and (3) in the above six WBS, which are the most critical in the tokamak assembly, are mainly described. The report describes the following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology

Trajectory planning of tokamak flexible in-vessel inspection robot

Energy Technology Data Exchange (ETDEWEB)

Wang, Hesheng [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Lai, Yinping; He, Tao [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China)

2015-10-15

Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

Trajectory planning of tokamak flexible in-vessel inspection robot

International Nuclear Information System (INIS)

Wang, Hesheng; Chen, Weidong; Lai, Yinping; He, Tao

2015-01-01

Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

Magnetic confinement experiment. I: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1995-08-01

Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

International Nuclear Information System (INIS)

Schultz, J.H.; Montgomery, D.B.

1983-04-01

HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

A generic access to shot-based data for European Tokamaks

Energy Technology Data Exchange (ETDEWEB)

Signoret, J.; Imbeaux, F. [Association EURATOM-CEA, CEA / DSM / Institut de Recherche sur la Fusion par confinement Magnetique, CEA-Cadarache, 13 - ST-Paul-Lez-Durance (France)

2009-07-01

The EFDA Integrated Tokamak Modeling Task Force has defined a data structure offering a generic representation of the properties of physics problems and tokamak subsystem characteristics. It gathers the hardware description, modeling results and data measured during experiments, structured in terms of Consistent Physical Objects (CPOs). A generic tool has been developed to retrieve shot-based data from the various European tokamak databases: Exp2ITM. A tokamak specific XML 'mapping file' is used to map the local data formats to the ITM (Integrated Tokamak Modeling) data format. Exp2ITM is then dynamically generated from the ITM data structure and uses generic procedures to import the shot-based data. Successful tests show we have managed to import into the ITM DB experimental data from Jet and Tore-Supra. This document is a poster. (authors)

On the HL-1M tokamak plasma confinement time

International Nuclear Information System (INIS)

Qin Yunwen

2001-01-01

Emphasizing that the tokamak plasma confinement time is the plasma particle or thermal energy loss characteristic time, the relevant physical concept and HL-1M tokamak experimental data analyses are reviewed

Digital control of plasma position in Damavand tokamak

Energy Technology Data Exchange (ETDEWEB)

Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.

2002-03-01

Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)

A computer code for Tokamak reactor concepts evaluation

International Nuclear Information System (INIS)

Rosatelli, F.; Raia, G.

1985-01-01

A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

3He functions in tokamak-pumped laser systems

International Nuclear Information System (INIS)

Jassby, D.L.

1986-10-01

3 He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the 3 He(n,p)T reaction, and thereby excite gaseous lasants mixed with the 3 He while simultaneously breeding tritium. The total 3 He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak

Analysis of EAST tokamak cryostat anti-seismic performance

International Nuclear Information System (INIS)

Chen Wei; Kong Xiaoling; Liu Sumei; Ni Xiaojun; Wang Zhongwei

2014-01-01

A 3-D finite element model for EAST tokamak cryostat is established by using ANSYS. On the basis of the modal analysis, the seismic response of the EAST tokamak cryostat structure is calculated according to an input of the design seismic response spectrum referring to code for seismic design of nuclear power plants. Calculation results show that EAST cryostat displacement and stress response is small under the action of earthquake. According to the standards, EAST tokamak cryostat structure under the action of design seismic can meet the requirements of anti-seismic design intensity, and ensure the anti-seismic safety of equipment. (authors)

Characterization of the Tokamak Novillo in cleaning regime

International Nuclear Information System (INIS)

Lopez C, R.; Melendez L, L.; Valencia A, R.; Chavez A, E.; Colunga S, S.; Gaytan G, E.

1992-02-01

In this work the obtained results of the investigation about the experimental characterization of those low energy pulsed discharges of the Tokamak Novillo are reported. With this it is possible to fix the one operation point but appropriate of the Tokamak to condition the chamber in the smallest possible time for the cleaning discharges regime before beginning the main discharge. The characterization of the cleaning discharges in those Tokamaks is an unique process and characteristic of each device, since the good points of operation are consequence of those particularities of the design of the machine. In the case of the Tokamak Novillo, besides characterizing it a contribution is made to the cleaning discharges regime which consists on the one product of the current peak to peak of plasma by the duration of the discharge Ip t like reference parameter for the optimization of the operation of the device in the cleaning discharge regime. The maximum value of the parameter I (p) t, under different work conditions, allowed to find the good operation point to condition the discharges chamber of the Tokamak Novillo in short time and to arrive to a regime in which is not necessary the preionization for the obtaining of the cleaning discharges. (Author)

Advanced tokamak physics in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)

2000-12-01

Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)

Technology and physics in the Tokamak Program: The need for an integrated, steady-state RandD tokamak experiment

International Nuclear Information System (INIS)

1988-05-01

The Steady-state Tokamak (STE) Experiment is a proposed superconducting-coil, hydrogen-plasma tokamak device intended to address the integrated non-nuclear issues of steady state, high-power tokamak physics and technology. Such a facility has been called for in the US program plan for the mid 1990's, and will play a unique role in the world-wide fusion effort. Information from STE on steady-state current drive, plasma control, and high power technology will contribute significantly to the operating capabilities of future steady-state devices. This paper reviews preliminary designs and expected technological contributions to the US and world fusion reactor research from each of the above mentioned reactor systems. This document is intended as a proposal and feasibility discussion and does not include exhaustive technical reviews. 12 figs., 3 tabs

The ARIES tokamak fusion reactor study

International Nuclear Information System (INIS)

Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

1989-01-01

The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

STARFIRE: a commercial tokamak reactor

International Nuclear Information System (INIS)

1979-12-01

The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

Operating tokamaks with steady-state toroidal current

International Nuclear Information System (INIS)

Fisch, N.J.

1981-04-01

Continuous operation of a tokamak requires, among other things, a means of continuously providing the toroidal current. Various methods have been proposed to provide this current including methods which utilize radio-frequency waves in any of several frequency regimes. Here we elaborate on the prospects of incorporating these current-drive techniques in tokamak reactors, concentrating on the theoretical minimization of the power requirements

Effect of impurity radiation on tokamak equilibrium

International Nuclear Information System (INIS)

Rebut, P.H.; Green, B.J.

1977-01-01

The energy loss from a tokamak plasma due to the radiation from impurities is of great importance in the overall energy balance. Taking the temperature dependence of this loss for two impurities characteristic of those present in existing tokamak plasmas, the condition for radial power balance is derived. For the impurities considered (oxygen and iron) it is found that the radiation losses are concentrated in a thin outer layer of the plasma and the equilibrium condition places an upper limit on the plasma paraticle number density in this region. This limiting density scales with mean current density in the same manner as is experimentally observed for the peak number density of tokamak plasmas. The stability of such equilibria is also discussed. (author)

Activation analysis of the compact ignition tokamak

International Nuclear Information System (INIS)

Selcow, E.C.

1986-01-01

The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak

Ripple induced trapped particle loss in tokamaks

International Nuclear Information System (INIS)

White, R.B.

1996-05-01

The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetric orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor and International Thermonuclear Experimental Reactor equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks

Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

NARCIS (Netherlands)

J.W. Haverkort (Willem)

2013-01-01

htmlabstractOne of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma

Theory of tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.

1989-01-01

The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.

Axisymmetric tokamak scapeoff transport

International Nuclear Information System (INIS)

Singer, C.E.; Langer, W.D.

1982-08-01

We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed

Realizing steady-state tokamak operation for fusion energy

International Nuclear Information System (INIS)

Luce, T. C.

2011-01-01

Continuous operation of a tokamak for fusion energy has clear engineering advantages but requires conditions beyond those sufficient for a burning plasma. The fusion reactions and external sources must support both the pressure and the current equilibrium without inductive current drive, leading to demands on stability, confinement, current drive, and plasma-wall interactions that exceed those for pulsed tokamaks. These conditions have been met individually, and significant progress has been made in the past decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are operated routinely without disruptions near pressure limits, as needed for steady-state operation. Fully noninductive sustainment with more than half of the current from intrinsic currents has been obtained for a resistive time with normalized pressure and confinement approaching those needed for steady-state conditions. One remaining challenge is handling the heat and particle fluxes expected in a steady-state tokamak without compromising the core plasma performance.

Gamma ray imager on the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.; Van Zeeland, M. A.; Watkins, M. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Cooper, C. M. [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California-San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Riso, V. [State University of New York-Buffalo, 12 Capen Hall, Buffalo, New York 14260-1660 (United States)

2016-04-15

A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1–60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.

Investigation of E x B transport with a multi-electrode probe in the plasma boundary of TEXTOR

International Nuclear Information System (INIS)

Ivanov, R.S.; Moyer, R.A.; Nieuwenhove, R. van; Oost, G. van; Fuchs, G.; Hoethker, K.; Samm, U.

1991-01-01

A movable multi-element Langmuir probe was implemented in TEXTOR in order to study properties of the edge and scrape-off plasma. The probe has five graphite electrode pins allowing the simultaneous measurement of main parameters such as plasma densities, electron temperatures, floating potentials, poloidal and radial electric fields. Both time-averaged and fluctuating quantities have been considered in order to evaluate the DC and turbulence-driven cross-field particle fluxes. The spectral analysis of the fluctuating floating potentials at spatially separated probe pins allows to determine the velocity associated with the rotations of the boundary plasma. The investigations have been focused on the variations of plasma boundary properties in plasmas with pure ohmic heating as well as auxiliary heating (ICRH). Special attention has been paid to the change of transport properties with the transition to a detached plasma. In particular, a significant reduction of the poloidal phase velocity at the limited edge has been observed for detached plasmas. Preliminary data on physical effects near the plasma boundary, which occur when the toroidal belt limiter (ALT-II) is biased, are reported. (orig.)

Control strategy for plasma equilibrium in a tokamak

International Nuclear Information System (INIS)

Miskell, R.V.

1975-08-01

Dynamic control of the plasma position within the torus of a TOKAMAK fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. (auth)

Alcator C-Mod Tokamak

Data.gov (United States)

Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...

Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1

International Nuclear Information System (INIS)

Saxena, Y.C.

2000-01-01

SST-1 is a steady state superconducting tokamak for studying the physics of the plasma processes in tokamak under steady state conditions and to learn technologies related to the steady state operation of the tokamak. SST-1 will have superconducting magnets made from NbTi based conductors operating at 4.5 K temperature. The design of the superconducting magnets and the cryogenic system of SST-1 tokamak are described. (author)

Diagnostics systems for the TBR-E tokamak

International Nuclear Information System (INIS)

Ueda, M.; Ferreira, J.L.; Aso, Y.; Ferreira, J.G.

1992-08-01

A general view of the several diagnostics systems proposed for the TBR-E tokamak is given. This project is a joint undertaking of INPE, USP and UNICAMP plasma laboratories. The requirements for the measurements of the plasma produced parameters are described. Special attention is given for diagnostics used to investigate new physical issues on a low aspect ratio tokamak such as TBR-E. (author)

Engineering Design of KSTAR tokamak main structure

International Nuclear Information System (INIS)

Im, K.H.; Cho, S.; Her, N.I.

2001-01-01

The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

System studies for quasi-steady-state advanced physics tokamak

International Nuclear Information System (INIS)

Reid, R.L.; Peng, Y.K.M.

1983-11-01

Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1990-04-01

This paper discusses the following work on the text tokamak: data systems; particle confinement; impurity transport; plasma rotation; runaway electrons; electron cyclotron heating; FIR system; transient transport; internal turbulence; edge turbulence; ion temperature; EML experiments; impurity pellet experiments; MHD experiments and analysis; TEXT Upgrade; and Upgrade diagnostics

Nonneutralized charge effects on tokamak edge magnetohydrodynamic stability

International Nuclear Information System (INIS)

Zheng, Linjin; Horton, W.; Miura, H.; Shi, T.H.; Wang, H.Q.

2016-01-01

Owing to the large ion orbits, excessive electrons can accumulate at tokamak edge. We find that the nonneutralized electrons at tokamak edge can contribute an electric compressive stress in the direction parallel to magnetic field by their mutual repulsive force. By extending the Chew–Goldburger–Low theory (Chew et al., 1956 [13]), it is shown that this newly recognized compressive stress can significantly change the plasma average magnetic well, so that a stabilization of magnetohydrodynamic modes in the pedestal can result. This linear stability regime helps to explain why in certain parameter regimes the tokamak high confinement can be rather quiet as observed experimentally.

The ICRH tokamak fusion test reactor

International Nuclear Information System (INIS)

Perkins, F.W.

1976-01-01

A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

Spherical tokamak power plant design issues

International Nuclear Information System (INIS)

Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.

2000-01-01

The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed

Issues for the electric utilities posed by DT tokamak fusion powerplants

International Nuclear Information System (INIS)

Roth, J.R.

1990-01-01

The DT tokamak is the mainline approach to magnetic fusion energy in all industrialized countries with a major commitment to fusion research. It achieved this status largely through historical accident and not as the result of considered choice among alternatives. After twenty-five years of intensive tokamak research, it is appropriate to ask whether the path down which the tokamak concept is leading the fusion community is the way to an acceptable powerplant for the electric utilities, or an aberration which should be replaced with an approach more promising in the long term. Issues surrounding the DT tokamak can be grouped in three broad areas: physics; safety/environmental; and engineering/economic. In addition to these problems, detailed engineering design studies of DT tokamak fusion powerplants over a twenty year period have revealed a number of additional problems. Most of thee are related to the presence of tritium and energetic neutron fluxes, which tend to make the cost of electricity of DT tokamaks higher than that of fossil or fission powerplants. These safety and economic issues of the DT tokamak powerplant also appear to be intractable, and have not been made to go away by twenty years of progressively more detailed and extensive engineering design studies

Magnet design considerations for Tokamak fusion reactors

International Nuclear Information System (INIS)

Purcell, J.R.; Chen, W.; Thomas, R.

1976-01-01

Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)

A programmatic framework for the Tokamak Physics Experiment (TPX)

International Nuclear Information System (INIS)

Thomassen, K.I.; Goldston, R.J.; Neilson, G.H.

1993-01-01

Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the open-quotes power breakevenclose quotes level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program

Joint Czechoslovak-Soviet workshop on current drive in tokamaks

International Nuclear Information System (INIS)

1985-10-01

At the Joint Czechoslovak-Soviet Workshop on Current Drive in Tokamaks, five papers dealing with issues of general interest were presented. In a theoretical paper by Klima and Pavlo a one-dimensional model of the lower-hybrid current drive is described and the results of its analysis are used in a numerical simulation using T-7 tokamak parameters. In the second theoretical paper by Vojtsekhovich, Parail and Pereverzev the influence of the LH wave spectrum on the efficiency of the current drive is studied. Two papers deal with a new microwave system designed for experiments on LHCD in the T-7 tokamak. In particular, the power spectra of new four-waveguide grills are computed. In the last paper the non-inductive start-up of the discharge in the T-7 tokamak by means of electron cyclotron waves is investigated. (J.U.)

Energy storage for tokamak reactor cycles

International Nuclear Information System (INIS)

Buchanan, C.H.

1979-01-01

The inherent characteristic of a tokamak reactor requiring periodic plasma quench and reignition introduces the problem of energy storage to permit continuous electrical output to the power grid. The cycle under consideration in this paper is a 1000 second burn followed by a 100 second reignition phase. The physical size of a typical toroidal plasma reaction chamber for a tokamak reactor has been described earlier. The thermal energy storage requirements described in this reference will serve as a basis for much of the ensuing discussion

Technology and plasma-materials interaction processes of tokamak disruptions

International Nuclear Information System (INIS)

McGrath, R.T.; Kellman, A.G.

1992-01-01

A workshop on the technology and plasma-materials interaction processes of tokamak disruptions was held April 3, 1992 in Monterey, California, as a satellite meeting of the 10th International Conference on Plasma-Surface Interactions. The objective was to bring together researchers working on disruption measurements in operating tokamaks, those performing disruption simulation experiments using pulsed plasma gun, electron beam and laser systems, and computational physicists attempting to model the evolution and plasma-materials interaction processes of tokamak disruptions. This is a brief report on the workshop. 4 refs

Particle injection into the Castor tokamak by electric arcs

International Nuclear Information System (INIS)

Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.

1989-01-01

The influence of arcing on the tokamak discharge was investigated in the Castor tokamak. A special calibrated gun which emitted tantalum by artificially ignited electric arcs, was used to study the transport of the injected tantalum ions, neutrals and droplets. The injection of tantalum led to an increase in electron density and to a change of plasma position only if the transported charge was higher than 0.01 C. As the naturally occurring arcs are well below this limit, the arcing in tokamaks is rather the consequence than the reason of instabilities. (J.U.)

The residual zonal dynamics in a toroidally rotating tokamak

International Nuclear Information System (INIS)

Zhou Deng

2015-01-01

Zonal flows, initially driven by ion-temperature-gradient turbulence, may evolve due to the neoclassic polarization in a collisionless tokamak plasma. In this presentation, the form of the residual zonal flow is presented for tokamak plasmas rotating toroidally at arbitrary velocity. The gyro-kinetic equation is analytically solved to give the expression of residual zonal flows with arbitrary rotating velocity. The zonal flow level decreases as the rotating velocity increases. The numerical evaluation is in good agreement with the previous simulation result for high aspect ratio tokamaks. (author)

Optimization design for SST-1 Tokamak insulators

International Nuclear Information System (INIS)

Zhang Yuanbin; Pan Wanjiang

2012-01-01

With the help of ANSYS FEA technique, high voltage and cryogenic proper- ties of the SST-1 Tokamak insulators were obtained, and the structure of the insulators was designed and modified by taking into account the simulation results. The simulation results indicate that the optimization structure has better high voltage insulating property and cryogenic mechanics property, and also can fulfill the qualification criteria of the SST-1 Tokamak insulators. (authors)

Equilibrium reconstruction in the TCA/Br tokamak; Reconstrucao do equilibrio no tokamak TCA/BR

Energy Technology Data Exchange (ETDEWEB)

Sa, Wanderley Pires de

1996-12-31

The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author) 68 refs., 31 figs., 16 tabs.

Configuration studies for a small-aspect-ratio tokamak stellarator hybrid

International Nuclear Information System (INIS)

Carreras, B.A.; Lynch, V.E.; Ware, A.

1996-08-01

The use of modulated toroidal coils offers a new path to the tokamak-stellarator hybrids. Low-aspect-ratio configurations can be found with robust vacuum flux surfaces and rotational transform close to the transform of a reverse-shear tokamak. These configurations have clear advantages in minimizing disruptions and their effect and in reducing tokamak current drive needs. They also allow the study of low-aspect-ratio effects on stellarator confinement in small devices

Compact fusion energy based on the spherical tokamak

Science.gov (United States)

Sykes, A.; Costley, A. E.; Windsor, C. G.; Asunta, O.; Brittles, G.; Buxton, P.; Chuyanov, V.; Connor, J. W.; Gryaznevich, M. P.; Huang, B.; Hugill, J.; Kukushkin, A.; Kingham, D.; Langtry, A. V.; McNamara, S.; Morgan, J. G.; Noonan, P.; Ross, J. S. H.; Shevchenko, V.; Slade, R.; Smith, G.

2018-01-01

Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor magnets as a possible route to fusion power using relatively small devices. We present an overview of the development programme including details of the enabling technologies, the key modelling methods and results, and the remaining challenges on the path to compact fusion.

Conceptual Design of Alborz Tokamak Poloidal Coils System

Science.gov (United States)

Mardani, M.; Amrollahi, R.

2013-04-01

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.

Compact tokamak reactors. Part 1 (analytic results)

International Nuclear Information System (INIS)

Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

1996-01-01

We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

Tokamak power systems studies, FY 1985

International Nuclear Information System (INIS)

Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

1985-12-01

The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs

Tokamak Engineering Technology Facility scoping study

Energy Technology Data Exchange (ETDEWEB)

Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

1976-03-01

A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.

Ballooning stable high beta tokamak equilibria

International Nuclear Information System (INIS)

Tuda, Takashi; Azumi, Masafumi; Kurita, Gen-ichi; Takizuka, Tomonori; Takeda, Tatsuoki

1981-04-01

The second stable regime of ballooning modes is numerically studied by using the two-dimensional tokamak transport code with the ballooning stability code. Using the simple FCT heating scheme, we find that the plasma can locally enter this second stable regime. And we obtained equilibria with fairly high beta (β -- 23%) stable against ballooning modes in a whole plasma region, by taking into account of finite thermal diffusion due to unstable ballooning modes. These results show that a tokamak fusion reactor can operate in a high beta state, which is economically favourable. (author)

Electronic system of TBR tokamak device

International Nuclear Information System (INIS)

Silva, R.P. da.

1980-01-01

The electronics developed as a part of the TBR project, which involves the construction of a small tokamak at the Physics Institute of the University of Sao Paulo, is described. On the basis of tokamak parameter values, the electronics for the toroidal field, ohmic/heating and vertical field systems is presented, including capacitors bank, switches, triggering circuits and power supplies. A controlled power oscilator used in discharge cleaning and pre-ionization is also described. The performance of the system as a function of the desired plasma parameters is discussed. (Author) [pt

Tokamak power systems studies, FY 1985

Energy Technology Data Exchange (ETDEWEB)

Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

1985-12-01

The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

Tokamak Engineering Technology Facility scoping study

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

1976-03-01

A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR

Can better modelling improve tokamak control?

International Nuclear Information System (INIS)

Lister, J.B.; Vyas, P.; Ward, D.J.; Albanese, R.; Ambrosino, G.; Ariola, M.; Villone, F.; Coutlis, A.; Limebeer, D.J.N.; Wainwright, J.P.

1997-01-01

The control of present day tokamaks usually relies upon primitive modelling and TCV is used to illustrate this. A counter example is provided by the successful implementation of high order SISO controllers on COMPASS-D. Suitable models of tokamaks are required to exploit the potential of modern control techniques. A physics based MIMO model of TCV is presented and validated with experimental closed loop responses. A system identified open loop model is also presented. An enhanced controller based on these models is designed and the performance improvements discussed. (author) 5 figs., 9 refs

Helicity content and tokamak applications of helicity

International Nuclear Information System (INIS)

Boozer, A.H.

1986-05-01

Magnetic helicity is approximately conserved by the turbulence associated with resistive instabilities of plasmas. To generalize the application of the concept of helicity, the helicity content of an arbitrary bounded region of space will be defined. The definition has the virtues that both the helicity content and its time derivative have simple expressions in terms of the poloidal and toroidal magnetic fluxes, the average toroidal loop voltage and the electric potential on the bounding surface, and the volume integral of E-B. The application of the helicity concept to tokamak plasmas is illustrated by a discussion of so-called MHD current drive, an example of a stable tokamak q profile with q less than one in the center, and a discussion of the possibility of a natural steady-state tokamak due to the bootstrap current coupling to tearing instabilities

Study and optimization of magnetized ICRF discharges for tokamak wall conditioning and assessment of the applicability to ITER

International Nuclear Information System (INIS)

Wauters, T.

2011-11-01

This work is devoted to the study and optimization of the Ion Cyclotron Wall Conditioning (ICWC) technique. ICWC, operated in presence of the toroidal magnetic field, makes use of four main tokamak systems: the ICRF antennas to initiate and sustain the conditioning discharge, the gas injection valves to provide the discharge gas, the machine pumps to remove the wall desorbed particles, and the poloidal magnetic field system to optimize the discharge homogeneity. Additionally neutral gas and plasma diagnostics are required to monitor the discharge and the conditioning efficiency. In chapter 2 a general overview on ICWC is given. Chapter 3 treats the ICRF discharge homogeneity and the confinement properties of the employed magnetic field. In the first part we will discuss experimental facts on plasma homogeneity, and how experimental optimization led to its improvement. In the second part of the chapter the confinement properties of a partially ionized plasma in a toroidal magnetic field configuration with additional small vertical component are discussed. Chapter 4 gives an overview of experimental results on the efficiency of ICWC, obtained on TORE SUPRA, TEXTOR, JET and ASDEX Upgrade. In chapter 5 a 0D kinetic description of hydrogen-helium RF plasmas is outlined. The model, describing the evolution of ICRF plasmas from discharge initiation to the (quasi) steady state plasma stage, is developed to obtain insight on ICRF plasma parameters, particle fluxes to the walls and the main collisional processes. Chapter 6 presents a minimum structure for a 0D reservoir model of the wall to investigate in deeper detail the ICWC plasma wall interaction during isotopic exchange experiments. The hypothesis used to build up the wall model is that the same model structure should be able to describe the wall behavior during normal plasmas and conditioning procedures. Chapter 7 extrapolates the results to the envisaged application of ICWC on ITER

Numerical simulation of edge plasma in tokamak

International Nuclear Information System (INIS)

Chen Yiping; Qiu Lijian

1996-02-01

The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)

The ETE spherical Tokamak project. IAEA report

Energy Technology Data Exchange (ETDEWEB)

Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br

2002-07-01

This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Total magnetic reconnection during a tokamak major disruption

International Nuclear Information System (INIS)

Goetz, J.A.

1990-09-01

Magnetic reconnection has long been considered to be the cause of sawtooth oscillations and major disruptions in tokamak experiments. Experimental confirmation of reconnection models has been hampered by the difficulty of direct measurement of reconnection, which would involve tracing field lines for many transits around the tokamak. Perhaps the most stringent test of reconnection in a tokamak involves measurement of the safety factor q. Reconnection arising from a single helical disturbance with mode numbers m and n should raise q to m/n everywhere inside of the original resonant surface. Total reconnection should also flatten the temperature and current density profiles inside of this surface. Disruptive instabilities have been studied in the Tokapole 2, a poloidal divertor tokamak. When Tokapole 2 is operated in the material limiter configuration, a major disruption results in current termination as in most tokamaks. However, when operated in the magnetic limiter configuration current termination is suppressed and major disruptions appear as giant sawtooth oscillations. The objective of this thesis is to determine if total reconnection is occurring during major disruptions. To accomplish this goal, the poloidal magnetic field has been directly measured in Tokapole 2 with internal magnetic coils. A full two-dimensional measurement over the central current channel has been done. From these measurements, the poloidal magnetic flux function is obtained and the magnetic surfaces are plotted. The flux-surface-averaged safety factor is obtained by integrating the local magnetic field line pitch over the experimentally obtained magnetic surface

Transport and turbulence in a magnetized plasma (application to tokamak plasmas); Transport et turbulence dans un plasma magnetise (application aux plasmas de tokamaks)

Energy Technology Data Exchange (ETDEWEB)

Sarazin, Y

2004-03-01

This document gathers the lectures made in the framework of a Ph.D level physics class dedicated to plasma physics. This course is made up of 3 parts : 1) collisions and transport, 2) transport and turbulence, and 3) study of a few exchange instabilities. More precisely the first part deals with the following issues: thermonuclear fusion, Coulomb collisions, particles trajectories in a tokamak, neo-classical transport in tokamaks, the bootstrap current, and ware pinch. The second part involves: particle transport in tokamaks, quasi-linear transport, resonance islands, resonance in tokamaks, from quasi to non-linear transport, and non-linear saturation of turbulence. The third part deals with: shift velocities in fluid theory, a model for inter-change instabilities, Rayleigh-Benard instability, Hasegawa-Wakatani model, and Hasegawa-Mima model. This document ends with a series of appendices dealing with: particle-wave interaction, determination of the curvature parameter G, Rossby waves.

TPX tokamak construction management

International Nuclear Information System (INIS)

Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

1995-01-01

A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly

High beta tokamaks

International Nuclear Information System (INIS)

Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.

1978-01-01

MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby

/sup 3/He functions in tokamak-pumped laser systems

Energy Technology Data Exchange (ETDEWEB)

Jassby, D.L.

1986-10-01

/sup 3/He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the /sup 3/He(n,p)T reaction, and thereby excite gaseous lasants mixed with the /sup 3/He while simultaneously breeding tritium. The total /sup 3/He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak.

Options for an ignited tokamak

International Nuclear Information System (INIS)

Sheffield, J.

1984-02-01

It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon β/sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed

Starfire: a commercial tokamak reactor

International Nuclear Information System (INIS)

Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

1979-01-01

The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

Tokamak fusion reactor exhaust

International Nuclear Information System (INIS)

Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

1981-08-01

This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

The spheric tokamak programme at Culham

International Nuclear Information System (INIS)

Sykes, A.

1999-01-01

The Spherical Tokamak (ST) is the low aspect ratio limit of the conventional tokamak, and appears to offer attractive physics properties in a simpler device. The START (Small Tight Aspect Ratio Tokamak) experiment provided the world's first demonstration of the properties of hot plasmas in an ST configuration, and was operational at Culham from January 1991 to March 1998, obtaining plasma current of up to 300 kA and pulse durations of ∼ 50 ms. Its successor, MAST is scheduled to obtain first plasma in Autumn 1998 and is a purpose built, high vacuum machine designed to have a tenfold increase in plasma volume with plasma currents up to 2 MA. Current drive and heating will be by a combination of induction-compression as on START, a high-performance central solenoid, 1.5 MW ECRH and 5 MW of Neutral Beam Injection. The promising results from START are reviewed, and the many challenges posed for the next generation of purpose-built STs (such as MAST) are described. (author)

Dust limit management strategy in tokamaks

Science.gov (United States)

Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Delaporte, P.; Douai, D.; Garnier, D.; Gauthier, E.; Gulden, W.; Hong, S. H.; Pitcher, S.; Rodriguez, L.; Taylor, N.; Tesini, A.; Vartanian, S.; Vatry, A.; Wykes, M.

2009-06-01

Dust is produced in tokamaks by the interaction between the plasma and the plasma facing components. Dust has not yet been of a major concern in existing tokamaks mainly because the quantity is small and these devices are not nuclear facilities. However, in ITER and in future reactors, it will represent operational and potential safety issues. From a safety point of view, in order to control the potential dust hazard, the current ITER strategy is based on a defense in depth approach designed to provide reliable confinement systems, to avoid failures, and to measure and minimise the dust inventory. In addition, R&D is put in place for optimisation of the proposed methods, such as improvement of measurement, dust cleaning and the reduction of dust production. The aim of this paper is to present the approach for the control of the dust inventory, relying on the monitoring of envelope values and the development of removal techniques already developed in the existing tokamaks or plasma dedicated devices or which will need further research and development in order to be integrated in ITER.

Accessibility of high β tokamak states

International Nuclear Information System (INIS)

Hogan, J.T.

1978-05-01

Encouraging results with neutral beam heating and adiabatic compression of tokamak plasmas have prompted new experiments which will study the approach to high β states. As projected tokamak β values become nonnegligible (average β of 4% is the goal), the models previously used for transport calculations will become inadequate. These models will be required to account for the evolution of the magnetic geometry, along with the change in plasma parameters. We present an axisymmetric transport model which should be useful for studying the approach to higher β values in tokamak experiments. Results from transport calculations with this model allow us to draw a parallel between observed behavior in seemingly unrelated experiments: electron heating by neutral injection in the ORMAK device and adiabatic compression in the ATC experiment. Finally, we find that the nature of cross-field transport may be expected to change as significant β values are reached. Enhanced transport from ballooning instabilities is likely to play a role as important as that now played by sawtooth (m = 1) and saturated (m = 2) instabilities. New techniques for describing this transport are required

Dust limit management strategy in tokamaks

International Nuclear Information System (INIS)

Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Delaporte, P.; Douai, D.; Garnier, D.; Gauthier, E.; Gulden, W.; Hong, S.H.; Pitcher, S.; Rodriguez, L.; Taylor, N.; Tesini, A.; Vartanian, S.; Vatry, A.; Wykes, M.

2009-01-01

Dust is produced in tokamaks by the interaction between the plasma and the plasma facing components. Dust has not yet been of a major concern in existing tokamaks mainly because the quantity is small and these devices are not nuclear facilities. However, in ITER and in future reactors, it will represent operational and potential safety issues. From a safety point of view, in order to control the potential dust hazard, the current ITER strategy is based on a defense in depth approach designed to provide reliable confinement systems, to avoid failures, and to measure and minimise the dust inventory. In addition, R and D is put in place for optimisation of the proposed methods, such as improvement of measurement, dust cleaning and the reduction of dust production. The aim of this paper is to present the approach for the control of the dust inventory, relying on the monitoring of envelope values and the development of removal techniques already developed in the existing tokamaks or plasma dedicated devices or which will need further research and development in order to be integrated in ITER.

The design of the KSTAR tokamak

International Nuclear Information System (INIS)

Lee, G.S.; Kim, J.; Hwang, S.M.

1999-01-01

The Korea superconducting tokamak advanced research (KSTAR) project is the major effort of the Korean national fusion program (KNFP) to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. Major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 Tesla, and plasma current 2 MA with a strongly shaped plasma cross-section and double-null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beam, ion cyclotron waves, lower hybrid waves, and electron-cyclotron waves for flexible profile control. A comprehensive set of diagnostics is planned for plasma control and performance evaluation and physics understanding. The project has completed its conceptual design phase and moved to the engineering design phase. The target date of the first plasma is set for year 2002. (orig.)

The Tokamak IST-TOK

International Nuclear Information System (INIS)

Varandas, C.A.F.; Cabral, J.A.C.; Manso, M.E.

1991-01-01

A small tokamak is under construction at the Portuguese Technical Superior Institute. The main objective is to create a home based laboratory in which an independent scientific program might be developed. (L.C.J.A.). 14 refs, 6 figs

Recent progress on the Compact Ignition Tokamak (CIT)

International Nuclear Information System (INIS)

Ignat, D.W.

1987-01-01

This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule

Lower hybrid heating experiments in tokamaks: an overview

International Nuclear Information System (INIS)

Porkolab, M.

1985-10-01

Lower hybrid wave propagation theory relevant to heating fusion grade plasmas (tokamaks) is reviewed. A brief discussion of accessibility, absorption, and toroidal ray propagation is given. The main part of the paper reviews recent results in heating experiments on tokamaks. Both electron and ion heating regimes will be discussed. The prospects of heating to high temperatures in reactor grade plasmas will be evaluated

A flexible software architecture for tokamak discharge control systems

International Nuclear Information System (INIS)

Ferron, J.R.; Penaflor, B.; Walker, M.L.; Moller, J.; Butner, D.

1995-01-01

The software structure of the plasma control system in use on the DIII-D tokamak experiment is described. This system implements control functions through software executing in real time on one or more digital computers. The software is organized into a hierarchy that allows new control functions needed to support the DIII-D experimental program to be added easily without affecting previously implemented functions. This also allows the software to be portable in order to create control systems for other applications. The tokamak operator uses an X-windows based interface to specify the time evolution of a tokamak discharge. The interface provides a high level view for the operator that reduces the need for detailed knowledge of the control system operation. There is provision for an asynchronous change to an alternate discharge time evolution in response to an event that is detected in real time. Quality control is enhanced through off-line testing that can make use of software-based tokamak simulators

Development in Diagnostics Application to Control Advanced Tokamak Plasma

International Nuclear Information System (INIS)

Koide, Y.

2008-01-01

For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)

Electric conductivity and bootstrap current in tokamak

International Nuclear Information System (INIS)

Mao Jianshan; Wang Maoquan

1996-12-01

A modified Ohm's law for the electric conductivity calculation is presented, where the modified ohmic current can be compensated by the bootstrap current. A comparison of TEXT tokamak experiment with the theories shows that the modified Ohm's law is a more close approximation to the tokamak experiments than the classical and neoclassical theories and can not lead to the absurd result of Z eff <1, and the extended neoclassical theory would be not necessary. (3 figs.)

Preliminary results of the TBR small tokamak

International Nuclear Information System (INIS)

Nascimento, I.C.; Fagundes, A.N.; Da Silva, R.P.; Galvao, R.M.O.; Del Bosco, E.; Vuolo, J.H.; Sanada, E.K.; Dellaqua, R.

1982-01-01

The paper gives a short description of the TBR - small Brazilian tokamak and the first results obtained for plasma formation and equilibrium. Measured breakdown curves for hydrogen are shown to be confined within analytically calculated limits and to depend strongly on stray vertical magnetic fields. Time profiles of plasma current in equilibrium are shown and compared with the predictions of a simple analytical model for tokamak discharges. Reasonable agreement is obtained taking Zsub(eff) as a free parameter. (author)

Plasma diagnostics using synchrotron radiation in tokamaks

International Nuclear Information System (INIS)

Fidone, I.; Giruzzi, G.; Granata, G.

1995-09-01

This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs

Heat load material studies: Simulated tokamak disruptions

International Nuclear Information System (INIS)

Gahl, J.M.; McDonald, J.M.; Zakharov, A.; Tserevitinov, S.; Barabash, V.; Guseva, M.

1991-01-01

It is clear that an improved understanding of the effects of tokamak disruptions on plasma facing component materials is needed for the ITER program. very large energy fluxes are predicted to be deposited in ITER and could be very damaging to the machine. During 1991, Sandia National Laboratories and the University of New Mexico conducted cooperative tokamak disruption simulation experiments at several Soviet facilities. These facilities were located at the Efremov Institute in Leningrad, the Kurchatov Atomic Energy Institute (Troisk and Moscow) and the Institute for Physical Chemistry of the Soviet Adademy of Sciences in Moscow. Erosion of graphite from plasma stream impact is seen to be much less than that observed with laser or electron beams with similar energy fluxes. This, along with other data obtained, seem to suggest that the ''vapor shielding'' effect is a very important phenomenon in the study of graphite erosion during tokamak disruption

Sawtooth phenomena in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1989-01-01

A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

Tokamak impurity-control techniques

International Nuclear Information System (INIS)

Schmidt, J.A.

1980-01-01

A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)

Plasma features and alpha particle transport in low-aspect ratio tokamak reactor

International Nuclear Information System (INIS)

Xu Qiang; Wang Shaojie

1997-06-01

The results of the experiment and theory from low-aspect ratio tokamak devices have proved that the MHD stability will be improved. Based on present plasma physics and extrapolation to reduced aspect ratio, the feature of physics of low-aspect ratio tokamak reactor is discussed primarily. Alpha particle confinement and loss in the self-justified low-aspect ratio tokamak reactor parameters and the effect of alpha particle confinement and loss for different aspect ratio are calculated. The results provide a reference for the feasible research of compact tokamak reactor. (9 refs., 2 figs., 3 tabs.)

Economic trends of tokamak power plants independent of physics scaling models

International Nuclear Information System (INIS)

Reid, R.L.; Steiner, D.

1978-01-01

This study examines the effects of plasma radius, field on axis, plasma impurity level, and aspect ratio on power level and unit capital cost, $/kW/sub e/, of tokamak power plants sized independent of plasma physics scaling models. It is noted that tokamaks sized in this manner are thermally unstable based on trapped particle scaling relationships. It is observed that there is an economic advantage for larger power level tokamaks achieved by physics independent sizing; however, the incentive for increased power levels is less than that for fission reactors. It is further observed that the economic advantage of these larger power level tokamaks is decreased when plasma thermal stability measures are incorporated, such as by increasing the plasma impurity concentration. This trend of economy with size obtained by physics independent sizing is opposite to that observed when the tokamak designs are constrained to obey the trapped particle and empirical scaling relationships

Two-ion ICRF heating in Tokamaks

International Nuclear Information System (INIS)

Tennfors, E.

1985-03-01

The practical consequences for tokamak plasma heating in the ion cyclotron frequency regime of the two-dimensional treatment of the two-ion mode conversion layer are analyzed. The problem of evaluation of the condition for fast wave resonance is analyzed, as well as the limitations imposed by warm plasma effects. Simple ways to find the mode conversion surfaces when they exist are presented. Also for large tokamaks, it is possible to obtain mode conversion conditions for realistic antenna spectra provided species concentration and frequency are chosen such that the surface Epsilon = 0 intersects the plasma midplane just outside of the magnetic axis. (Author)

Reconnection in tokamaks

International Nuclear Information System (INIS)

Pare, V.K.

1983-01-01

Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

Effects of isotropic alpha populations on tokamak ballooning stability

International Nuclear Information System (INIS)

Spong, D.A.; Sigmar, D.J.; Tsang, K.T.; Ramos, J.J.; Hastings, D.E.; Cooper, W.A.

1986-12-01

Fusion product alpha populations can significantly influence tokamak stability due to coupling between the trapped alpha precessional drift and the kinetic ballooning mode frequency. Careful, quantitative evaluations of these effects are necessary in burning plasma devices such as the Tokamak Fusion Test Reactor and the Joint European Torus, and we have continued systematic development of such a kinetic stability model. In this model we have considered a range of different forms for the alpha distribution function and the tokamak equilibrium. Both Maxwellian and slowing-down models have been used for the alpha energy dependence while deeply trapped and, more recently, isotropic pitch angle dependences have been examined

Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

International Nuclear Information System (INIS)

Abiteboul, J.

2012-10-01

The goal of magnetic confinement devices such as tokamaks is to produce energy from nuclear fusion reactions in plasmas at low densities and high temperatures. Experimentally, toroidal flows have been found to significantly improve the energy confinement, and therefore the performance of the machine. As extrinsic momentum sources will be limited in future fusion devices such as ITER, an understanding of the physics of toroidal momentum transport and the generation of intrinsic toroidal rotation in tokamaks would be an important step in order to predict the rotation profile in experiments. Among the mechanisms expected to contribute to the generation of toroidal rotation is the transport of momentum by electrostatic turbulence, which governs heat transport in tokamaks. Due to the low collisionality of the plasma, kinetic modeling is mandatory for the study of tokamak turbulence. In principle, this implies the modeling of a six-dimensional distribution function representing the density of particles in position and velocity phase-space, which can be reduced to five dimensions when considering only frequencies below the particle cyclotron frequency. This approximation, relevant for the study of turbulence in tokamaks, leads to the so-called gyrokinetic model and brings the computational cost of the model within the presently available numerical resources. In this work, we study the transport of toroidal momentum in tokamaks in the framework of the gyrokinetic model. First, we show that this reduced model is indeed capable of accurately modeling momentum transport by deriving a local conservation equation of toroidal momentum, and verifying it numerically with the gyrokinetic code GYSELA. Secondly, we show how electrostatic turbulence can break the axisymmetry and generate toroidal rotation, while a strong link between turbulent heat and momentum transport is identified, as both exhibit the same large-scale avalanche-like events. The dynamics of turbulent transport are

Automated Fault Detection for DIII-D Tokamak Experiments

International Nuclear Information System (INIS)

Walker, M.L.; Scoville, J.T.; Johnson, R.D.; Hyatt, A.W.; Lee, J.

1999-01-01

An automated fault detection software system has been developed and was used during 1999 DIII-D plasma operations. The Fault Identification and Communication System (FICS) executes automatically after every plasma discharge to check dozens of subsystems for proper operation and communicates the test results to the tokamak operator. This system is now used routinely during DIII-D operations and has led to an increase in tokamak productivity

Erosion of the first wall of Tokamaks

International Nuclear Information System (INIS)

Guseva, M.I.; Ionova, E.S.; Martynenko, Yu.V.

1980-01-01

An estimate of the rate of erosion of the wall due to sputtering and blistering requires knowledge of the fluxes and energies of the particles which go from the plasma to the wall, of the sputtering coefficients S, and of the erosion coefficients S* for blistering. The overall erosion coefficient is equal to the sum of the sputtering coefficient and the erosion coefficient for blistering. Here the T-20 Tokamak is examined as an example of a large-scale Tokamak. 18 refs

Thermonuclear ignition in the next generation tokamaks

International Nuclear Information System (INIS)

Johner, J.

1989-04-01

The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aB t x of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition

New directions in tokamak reactors

International Nuclear Information System (INIS)

Baker, C.C.

1985-01-01

New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs

High-β steady-state advanced tokamak regimes for ITER and FIRE

International Nuclear Information System (INIS)

Meade, D.M.; Sauthoff, N.R.; Kessel, C.E.; Budny, R.V.; Gorelenkov, N.; Jardin, S.C.; Schmidt, J.A.; Navratil, G.A.; Bialek, J.; Ulrickson, M.A.; Rognlein, T.; Mandrekas, J.

2005-01-01

An attractive tokamak-based fusion power plant will require the development of high-β steady-state advanced tokamak regimes to produce a high-gain burning plasma with a large fraction of self-driven current and high fusion-power density. Both ITER and FIRE are being designed with the objective to address these issues by exploring and understanding burning plasma physics both in the conventional H-mode regime, and in advanced tokamak regimes with β N ∼ 3 - 4, and f bs ∼50-80%. ITER has employed conservative scenarios, as appropriate for its nuclear technology mission, while FIRE has employed more aggressive assumptions aimed at exploring the scenarios envisioned in the ARIES power-plant studies. The main characteristics of the advanced scenarios presently under study for ITER and FIRE are compared with advanced tokamak regimes envisioned for the European Power Plant Conceptual Study (PPCS-C), the US ARIES-RS Power Plant Study and the Japanese Advanced Steady-State Tokamak Reactor (ASSTR). The goal of the present work is to develop advanced tokamak scenarios that would fully exploit the capability of ITER and FIRE. This paper will summarize the status of the work and indicate critical areas where further R and D is needed. (author)

Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

International Nuclear Information System (INIS)

Giunta, A.S.; Henderson, S.; O'Mullane, M.; Summers, H.P.; Harrison, J.; Doyle, J.G.

2016-01-01

This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.

Fast ion dynamics in ASDEX upgrade and TEXTOR measured by collective Thomson scattering

International Nuclear Information System (INIS)

Moseev, D.

2011-11-01

Fast ions are an essential ingredient in burning nuclear fusion plasmas: they are responsible for heating the bulk plasma, carry a significant amount of plasma current and moreover interact with various magnetohydrodynamic (MHD) instabilities. The collective Thomson scattering (CTS) diagnostic is sensitive to the projection of fast ion velocity distribution function. This thesis is mainly devoted to investigations of fast ion physics in tokamak plasmas by means of CTS. (Author)

Fast ion dynamics in ASDEX upgrade and TEXTOR measured by collective Thomson scattering

Energy Technology Data Exchange (ETDEWEB)

Moseev, D.

2011-11-15

Fast ions are an essential ingredient in burning nuclear fusion plasmas: they are responsible for heating the bulk plasma, carry a significant amount of plasma current and moreover interact with various magnetohydrodynamic (MHD) instabilities. The collective Thomson scattering (CTS) diagnostic is sensitive to the projection of fast ion velocity distribution function. This thesis is mainly devoted to investigations of fast ion physics in tokamak plasmas by means of CTS. (Author)

Improvement of the tokamak concept

Energy Technology Data Exchange (ETDEWEB)

Laurent, L

1994-12-31

Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.

Analysis of toroidal rotation data for the DIII-D tokamak

International Nuclear Information System (INIS)

John, H.St.; Burrell, K.H.; Groebner, R.; DeBoo, J.; Gohil, P.

1989-01-01

Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner et al. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Similar studies have been previously reported for Doublet III, ASDEX, TFTR, JET and other tokamaks. (author) 13 refs., 4 figs

Engineering design of TFTR and it's impact on future tokamaks

International Nuclear Information System (INIS)

Sabado, M.M.

1981-01-01

TFTR is a second generation tokamak whose key objective is scientific break-even. TFTR is expected to be the first machine to demonstrate proper combination of plasma confinement time, density, and temperature to obtain this objective. A summary of major TFTR design parameters, including TFM, is presented, and their potential impact on future tokamaks discussed. Details of the updated engineering design and analysis of components are described. Status of major hardware fabrication, assembly installation and test are reviewed. TFTR features, technology, predicted performance and their potential implication for future tokamaks are summarized

Effects of alpha populations on tokamak ballooning stability

International Nuclear Information System (INIS)

Spong, D.A.; Sigmar, D.J.; Tsang, K.T.; Ramos, J.J.; Hastings, D.E.; Cooper, W.A.

1986-01-01

Fusion product alpha populations can significantly influence tokamak stability due to coupling between the trapped alpha precessional drift and the kinetic ballooning mode frequency. This effect is of particular importance in parameter regimes where the alpha pressure gradient begins to constitute a sizable fraction of the thermal plasma pressure gradient. Careful, quantitative evaluations of these effects are necessary in burning plasma devices such as the Tokamak Fusion Test Reactor and the Joint European Torus, and we have continued systematic development of such a kinetic stability model. In this model we have considered a range of different forms for the alpha distribution function and the tokamak equilibrium. Both Maxwellian and slowing-down models have been used for the alpha energy dependence while deeply trapped and, more recently, isotropic pitch angle dependence have been examined

Compact toroid fueling of the TdeV tokamak

International Nuclear Information System (INIS)

Martin, F.; Raman, R.; Xiao, C.; Thomas, J.

1993-01-01

Compact toroids have been proposed as a means of centrally fueling tokamak reactors because of the high velocity to which they can be accelerated. These are cold (T e ∼ 10 eV), high density (n e > 10 20 m -3 ) spheromak plasmoids that are accelerated in a magnetized Marshall gun. As a proof of principle experiment, a compact toroid fueler (CTF) has been developed for injection into the TdeV tokamak. The engineering goals of the experiment are to measure and minimize the impurity content of the CT plasma and the neutral gas remaining after CT formation. Also of importance is the effect of CT central fueling on the tokamak density profile and bootstrap current, and the relaxation rate of the density profile providing information on the confinement time of the CT fuel

Super high field ohmically heated tokamak operation

International Nuclear Information System (INIS)

Cohn, D.R.; Bromberg, L.; Leclaire, R.J.; Potok, R.E.; Jassby, D.L.

1986-01-01

The authors discuss a super high field mode of tokamak operation that uses ohmic heating or near ohmic heating to ignition. The super high field mode of operation uses very high values of Β/sup 2/α, where Β is the magnetic field and a is the minor radius (Β/sup 2/α > 100 T/sup 2/m). We analyze copper magnet devices with major radii from 1.7 to 3.0 meters. Minimizing or eliminating the need for auxiliary heating has the potential advantages of reducing uncertainty in extrapolating the energy confinement time of current tokamak devices, and reducing engineering problems associated with large auxiliary heating requirements. It may be possible to heat relatively short pulse, inertially cooled tokamaks to ignition with ohmic power alone. However, there may be advantages in using a very small amount of auxiliary power (less than the ohmic heating power) to boost the ohmic heating and provide a faster start-up, expecially in relatively compact devices

Analytic description of tokamak equilibrium sustained by high fraction bootstrap current

International Nuclear Information System (INIS)

Shi Bingren

2002-01-01

Recently, to save the current drive power and to obtain more favorable confinement merit for tokamak reactor, large faction bootstrap current sustained equilibrium has attracted great interests both theoretically and experimentally. An powerful expanding technique and the tokamak ordering are used to expand the Grad-Shafranov equation to obtain a series of ordinary differential equations which allow for different sets of input parameters. The fully bootstrap current sustained tokamak equilibria are then solved analytically

JUST: Joint Upgraded Spherical Tokamak

International Nuclear Information System (INIS)

Azizov, E.A.; Dvorkin, N.Ya.; Filatov, O.G.

1997-01-01

The main goals, ideas and the programme of JUST, spherical tokamak (ST) for the plasma burn investigation, are presented. The place and prospects of JUST in thermonuclear investigations are discussed. (author)

Design of Tokamak plasma with high Tc superconducting coils

International Nuclear Information System (INIS)

Uchimoto, T.; Miya, K.; Yoshida, Y.; Yamada, T.

1999-01-01

This paper presents a design of tokamak plasma in light of how the small ignited tokamak is possible with use of the HTSC coils as plasma stabilizer. The same data base and formulas as ITER are here used and any innovative technology other than the HTSC stabilizing coils is not assumed. (author)

Natural current profiles in tokamaks

International Nuclear Information System (INIS)

Biskamp, D.

1986-01-01

It is proposed that a certain class of equilibrium, which follow from an elementary variational principle, are the natural current profiles in tokamaks, to which actual discharge profiles tend to relax. (orig.)

Radial electric fields for improved tokamak performance

International Nuclear Information System (INIS)

Downum, W.B.

1981-01-01

The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

Impurity control in near-term tokamak reactors

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Smith, D.L.; Brooks, J.N.

1976-10-01

Several methods for reducing impurity contamination in near-term tokamak reactors by modifying the first-wall surface with a low-Z or low-sputter material are examined. A review of the sputtering data and an assessment of the technological feasibility of various wall modification schemes are presented. The power performance of a near-term tokamak reactor is simulated for various first-wall surface materials, with and without a divertor, in order to evaluate the likely effect of plasma contamination associated with these surface materials

Diffusion in a tokamak with helical magnetic cells

International Nuclear Information System (INIS)

Wakatani, Masahiro

1975-05-01

In a tokamak with helical magnetic cells produced by a resonant helical magnetic field, diffusion in the collisional regime is studied. The diffusion coefficient is greatly enhanced near the resonant surface even for a weak helical magnetic field. A theoretical model for disruptive instabilities based on the enhanced transport due to helical magnetic cells is discussed. This may explain experiments of the tokamak with resonant helical fields qualitatively. (author)

HT-7U superconducting tokamak: Physics design, engineering progress and schedule

International Nuclear Information System (INIS)

Wan Yuanxi

2002-01-01

The superconducting tokamak research program begun in China in ASIPP since 1994. The program is included in existent superconducting tokamak HT-7 and the next new superconducting tokamak HT-7U which is one of national key research projects in China. With the elongation cross-section, divertor and higher plasma parameter the main objectives of HT-7U are widely investigation both of the physics and technology for steady state advanced tokamak as well as the investigation of power and particle handle under steady-state operation condition. The physics and engineering design have been completed and significant progresses on R and D and fabrication have been achieved. HT-7U will begin assembly at 2003 and possible to get first plasma around 2004. (author)

Dynamics and feedback control of plasma equilibrium position in a tokamak

International Nuclear Information System (INIS)

Burenko, O.

1983-01-01

A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems

Time-dependent analysis of the resistivity of post-disruption tokamak plasmas

International Nuclear Information System (INIS)

Bakhtiari, M.; Whyte, D. G.

2006-01-01

The effect of neutrals on plasma resistivity due to electron-neutral collisions is studied with respect to its effect on tokamak disruptions. The resistivity of the tokamak plasma after the thermal quench is critical in determining the current quench rate, the plasma temperature, and runaway electron generation in tokamaks through the electric field, all features which are important for mitigating the damaging effect of disruptions. It is shown that the plasma resistivity during tokamak disruptions is a time-dependent parameter which may vary with disruption time scales due to the increasing fraction of neutrals. However the effect of neutrals on resistivity is found to be small for the expected neutral fraction, mostly due to power balance considerations between radiation and Ohmic heating in the plasma

A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

Directory of Open Access Journals (Sweden)

Mitsuru Kikuchi

2010-11-01

Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

Discharge cleaning for a tokamak

International Nuclear Information System (INIS)

Ishii, Shigeyuki

1983-01-01

Various methods of discharge cleaning for tokamaks are described. The material of the first walls of tokamaks is usually stainless steel, inconel, titanium and so on. Hydrogen is exclusively used as the discharge gas. Glow discharge cleaning (GDC), Taylor discharge cleaning (TDC), and electron cyclotron resonance discharge cleaning (ECR-DC) are discussed in this paper. The cleaning by GDC is made by moving a movable anode to the center of a tokamak vassel. Taylor found the good cleaning effect of induced discharge by high pressure and low power discharge. This is called TDC. When the frequency of high frequency discharge in a magnetic field is equal to that of the electron cyclotron resonance, the break down potential is lowered if the pressure is sufficiently low. The ECR-CD is made by using this effect. In TDC and ECR-DC, the electron temperature, which has a close relation to the production rate of H 0 , can be controlled by the pressure. In GDC, the operating pressure was improved by the radio frequency glow (RG) method. However, there is still the danger of arcing. In case of GDC and ECR-DC, the position of plasma can be controlled, but not in case of TDC. The TDC is accepted by most of takamak devices in the world. (Kato, T.)

Methods for the design and optimization of shaped tokamaks

International Nuclear Information System (INIS)

Haney, S.W.

1988-05-01

Two major questions associated with the design and optimization of shaped tokamaks are considered. How do physics and engineering constraints affect the design of shaped tokamaks? How can the process of designing shaped tokamaks be improved? The first question is addressed with the aid of a completely analytical procedure for optimizing the design of a resistive-magnet tokamak reactor. It is shown that physics constraints---particularly the MHD beta limits and the Murakami density limit---have an enormous, and sometimes, unexpected effect on the final design. The second question is addressed through the development of a series of computer models for calculating plasma equilibria, estimating poloidal field coil currents, and analyzing axisymmetric MHD stability in the presence of resistive conductors and feedback. The models offer potential advantages over conventional methods since they are characterized by extremely fast computer execution times, simplicity, and robustness. Furthermore, evidence is presented that suggests that very little loss of accuracy is required to achieve these desirable features. 94 refs., 66 figs., 14 tabs

Laser detritiation and co-deposited layer characterisation for future ITER Installation

International Nuclear Information System (INIS)

Semerok, Alexandre; Brygo, Francois; Fomichev, Sergey V.; Champonnois, Francois; Weulersse, Jean-Marc; Thro, Pierre-Yves; Fichet, Pascal; Grisolia, Christian

2006-01-01

The experimental equipment in combination with pulsed Nd-YAG lasers was developed and applied to investigate co-deposited layer characterisation and ablation. Heating and ablation regimes were distinguished by ablation threshold fluence that was determined experimentally for graphite samples from TexTor (Germany) and TORE SUPRA (France) tokamaks. With 100 ns pulses, the ablation threshold for graphite substrate (2.5±0.5 J/cm 2 ) was much higher than the one for co-deposited layer (0.4±0.1 J cm -2 ). These threshold features are very promising to ensure self-controlled laser cleaning without substrate surface damage. The obtained optimal conditions (laser fluence F=1-2 J/cm 2 , 10-20 kHz repetition rate) were applied for co-deposited layer cleaning. The TexTor 50 μm thickness layer was almost completely removed after a single scanning without any damage of the graphite substrate. Cleaning rate of 0.2 m 2 /hour was demonstrated experimentally for 20 W mean laser power. A theoretical model of a complex surface heating (graphite or metal with a co-deposited layer) was developed to explain the experimental results and to obtain laser cleaning optimisation. A good agreement of the theoretical data with the experimental results was obtained. The studies on LIBS method for co-deposited layer characterisation have determined the analytical spectral lines for hydrogen, carbon, and other impurities (B, Fe, Si, and Cu) in TexTor graphite tile. The obtained results should be regarded optimistic for co-deposited layers characterisation by LIBS method. The development of certain laser methods and their application for in-situ detritiation and co-deposited layer characterisation are presented and discussed. (authors)

ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

International Nuclear Information System (INIS)

HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

2002-01-01

A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

Review of tokamak power reactor and blanket designs in the United States

International Nuclear Information System (INIS)

Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

1986-01-01

The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

Software development of the KSTAR Tokamak Monitoring System

International Nuclear Information System (INIS)

Kim, K.H.; Lee, T.G.; Baek, S.; Lee, S.I.; Chu, Y.; Kim, Y.O.; Kim, J.S.; Park, M.K.; Oh, Y.K.

2008-01-01

The Korea Superconducting Tokamak Advanced Research (KSTAR) project, which is constructing a superconducting Tokamak, was launched in 1996. Much progress in instrumentation and control has been made since then and the construction phase will be finished in August 2007. The Tokamak Monitoring System (TMS) measures the temperatures of the superconducting magnets, bus-lines, and structures and hence monitors the superconducting conditions during the operation of the KSTAR Tokamak. The TMS also measures the strains and displacements on the structures in order to monitor the mechanical safety. There are around 400 temperature sensors, more than 240 strain gauges, 10 displacement gauges and 10 Hall sensors. The TMS utilizes Cernox sensors for low temperature measurement and each sensor has its own characteristic curve. In addition, the TMS needs to perform complex arithmetic operations to convert the measurements into temperatures for each Cernox sensor for this large number of monitoring channels. A special software development effort was required to reduce the temperature conversion time and multi-threading to achieve the higher performance needed to handle the large number of channels. We have developed the TMS with PXI hardware and with EPICS software. We will describe the details of the implementations in this paper

Study of a compact reversed shear Tokamak reactor

International Nuclear Information System (INIS)

Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.

1998-01-01

A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)

The physics of magnetic confinement configurations : Tokamak theory and experiment

International Nuclear Information System (INIS)

Robinson, D.C.

1982-01-01

Several aspects, both theoretical and experimental, in plasma physics are discussed. The problem of magnetic confinement in Tokamak devices is treated. A discussion on the history of the development and on the future problems to be solved in Tokamaks is made. (L.C.) [pt

Control and Data Acquisition for the Spherical Tokamak MEDUSA-CR

Science.gov (United States)

Soto, Christian; Gonzalez, Jeferson; Carvajal, Johan; Ribeiro, Celso

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R loan to our laboratory via NI-Costa Rica. The interface with the energy, gas fueling, and security systems are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.; Miller, J.R.

1988-01-01

The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

Flux surface shaping effects on tokamak edge turbulence and flows

Energy Technology Data Exchange (ETDEWEB)

Kendl, A. [Innsbruck Univ., Institut fuer Theoretische Physik, Association EURATOM (Austria); Scott, B.D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching bei Muenchen (Germany)

2004-07-01

The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 {<=} {kappa} {>=} 2 and triangularity 0 {<=} {delta} {<=} 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)

Magnetic analysis of tokamak plasma with approximate MHD equilibrium solution

International Nuclear Information System (INIS)

Moriyama, Shin-ichi; Hiraki, Naoji

1993-01-01

A magnetic analysis method for determining equilibrium configuration parameters (plasma shape, poloidal beta and internal inductance) on a non-circular tokamak is described. The feature is to utilize an approximate MHD equilibrium solution which explicitly relates the configuration parameters with the magnetic fields picked up by magnetic sensors. So this method is suitable for the real-time analysis performed during a tokamak discharge. A least-squares fitting procedure is added to the analytical algorithm in order to reduce the errors in the magnetic analysis. The validity is investigated through the numerical calculation for a tokamak equilibrium model. (author)

Flux surface shaping effects on tokamak edge turbulence and flows

International Nuclear Information System (INIS)

Kendl, A.; Scott, B.D.

2004-01-01

The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 ≤ κ ≥ 2 and triangularity 0 ≤ δ ≤ 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)

Comprehensive numerical modelling of tokamaks

International Nuclear Information System (INIS)

Cohen, R.H.; Cohen, B.I.; Dubois, P.F.

1991-01-01

We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell

Plasma diagnostics on large tokamaks

International Nuclear Information System (INIS)

Orlinskij, D.V.; Magyar, G.

1988-01-01

The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs

Advanced commercial Tokamak optimization studies

International Nuclear Information System (INIS)

Whitley, R.H.; Berwald, D.H.; Gordon, J.D.

1985-01-01

Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units

Tokamak start-up with electron-cyclotron heating

International Nuclear Information System (INIS)

Holly, D.J.; Prager, S.C.; Shepard, D.A.; Sprott, J.C.

1981-01-01

Experiments are described in which the start-up voltage in a tokamak is reduced by about a factor of two by the use of a modest amount of electron cyclotron resonance heating power for pre-ionization. The solution of the zero-dimensional start-up equations indicates that the effect is due to the high initial density which increases the rate at which the conductivity increases in the neutral-dominated initial plasma. The effect extrapolates favourably to larger tokamaks. A 50% reduction in the start-up volt-second requirement and impurity reflux is also observed. (author)

Tokamak start-up with electron-cyclotron heating

Energy Technology Data Exchange (ETDEWEB)

Holly, D J; Prager, S C; Shepard, D A; Sprott, J C [Wisconsin Univ., Madison (USA)

1981-11-01

Experiments are described in which the start-up voltage in a tokamak is reduced by about a factor of two by the use of a modest amount of electron cyclotron resonance heating power for pre-ionization. The solution of the zero-dimensional start-up equations indicates that the effect is due to the high initial density which increases the rate at which the conductivity increases in the neutral-dominated initial plasma. The effect extrapolates favourably to larger tokamaks. A 50% reduction in the start-up volt-second requirement and impurity reflux is also observed.

Proposed tokamak poloidal field system development program

Energy Technology Data Exchange (ETDEWEB)

Rogers, J.D.; Vogel, H.F.; Warren, R.W.; Weldon, D.M.

1977-05-01

A program is proposed to develop poloidal field components for TNS and EPR size tokamak devices and to test these components in realistic circuits. Emphasis is placed upon the development of the most difficult component, the superconducting ohmic-heating coil. Switches must also be developed for testing the coils, and this switching technology is to be extended to meet the requirements for the large scale tokamaks. Test facilities are discussed; power supplies, including a homopolar to drive the coils, are considered; and poloidal field systems studies are proposed.

Gas blanket fueling of a tokamak reactor

International Nuclear Information System (INIS)

Gralnick, S.L.

1978-01-01

The purpose of this paper is a speculative investigation of the potential of fueling a Tokamak by introducing a sufficiently large quantity of gaseous deuterium and tritium at the vacuum wall boundary. It is motivated by two factors: current generation tokamaks are, in a manner of speaking, fueled from the edge quite successfully as is evidenced by pulse lengths that are long compared to particle recycling times, and by rapid plasma density increase produced by gas puffing, alternative, deep penetration fueling techniques that have been proposed possess severe technological problems and large costs

Increase in beta limit in tokamak plasmas

International Nuclear Information System (INIS)

Kamada, Yutaka

2003-01-01

This paper reviews recent studies of tokamak MHD stability towards the achievement of a high beta steady-state, where the profile control of current, pressure, and rotation, and the optimization of the plasma shape play fundamental roles. The key instabilities include the neoclassical tearing mode, the resistive wall mode, the edge localized mode, etc. In order to demonstrate an economically attractive tokamak reactor, it is necessary to increase the beta value simultaneously with a sufficiently high integrated plasma performance. Towards this goal, studies of stability control in self-regulating plasma systems are essential. (author)

Simulation study on dynamics of runaways in tokamaks

International Nuclear Information System (INIS)

Liu Jian; Qin Hong; Fisch, Nathaniel J.

2014-01-01

Electrons with high velocities can be accelerated to very high energies by a strong electric field to form runaway electrons. In tokamak, runaway electrons are produced in many different processes, including the acceleration from the high-energy tail of thermal distribution, through the runaway avalanche, during the rf wave heating and other non-Ohmic current drive, and even in the magnetic reconnection. This proceeding focus on different dynamical problems of runaway electrons in tokamaks. (author)

Design studies of Tokamak power reactor in JAERI

International Nuclear Information System (INIS)

Tone, T.; Nishikawa, M.; Tanaka, Y.

1985-01-01

Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance

First Results from Tests of High Temperature Superconductor Magnets on Tokamak

Energy Technology Data Exchange (ETDEWEB)

Gryaznevich, M.; Todd, T.T., E-mail: mikhail.gryaznevich@ccfe.ac.uk [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Svoboda, V.; Markovic, T.; Ondrej, G. [Czech Technical University, Prague (Czech Republic); Stockel, J.; Duran, I.; Kovarik, K. [IPP Prague, Czech Technical University, Prague (Czech Republic); Sykes, A.; Kingham, D. [Tokamak Solutions, Culham Science Centre, Abingdon (United Kingdom); Melhem, Z.; Ball, S.; Chappell, S. [Oxford Instruments, Abingdon (United Kingdom); Lilley, M. K.; De Grouchy, P.; Kim, H. -T. [Imperial College, London (United Kingdom)

2012-09-15

Full text: It has long been known that high temperature superconductors (HTS) could have an important role to play in the future of tokamak fusion research. Here we report on first results of the use of HTS in a tokamak magnet and on the progress in design and construction of the first fully-HTS tokamak. In the experiment, the two copper vertical field coils of the small tokamak GOLEM were replaced by two coils each with 6 turns of HTS (Re)BCO tape. Liquid nitrogen was used to cool the coils to below the critical temperature at which HTS becomes superconducting. Little effect on the HTS critical current has been observed for perpendicular field up to 0.5 T and superconductivity has been achieved at {approx} 90.5K during bench tests. There had been concerns that the plasma pulses and pulsed magnetic fields might cause a 'quench' in the HTS, i.e., a sudden and potentially damaging transition from superconductor to normal conductor. However, many plasma pulses were fired without any quenches even when disruptions occurred with corresponding induced electrical fields. In addition, experiments without plasma have been performed to study properties of the HTS in a tokamak environment, i.e., critical current and its dependence on magnetic and electrical fields generated in a tokamak both in DC and pulsed operations, maximum current ramp-up speed, performance of the HTS tape after number of artificially induced quenches etc. No quench has been observed at DC currents up to 200 A (1.2 kA-turns through the coil). In short pulses, current up to 1 kA through the tape (6 kA-turns) has been achieved with no subsequent degradation of the HTS performance with a current ramp rate up to 0.6 MA/s. In future experiments, increases in both the plasma current and pulse duration are planned. Considerable experience has been gained during design and fabrication of the cryostat, coils, isolation and insulation, feeds and cryosystems, and GOLEM is now routinely operated with HTS coils. The

The basics of spherical tokamaks and progress in European research

International Nuclear Information System (INIS)

Gusev, V K; Alladio, F; Morris, A W

2003-01-01

When the aspect ratio of a tokamak (A = R/a) decreases significantly, there is a transformation of the well studied tokamak toroidal magnetic configuration into the spherical tokamak (ST) configuration. This configuration has high natural plasma elongation and triangularity and other unique equilibrium and stability properties of ST configuration, which are discussed in this paper. European research into ST physics is well advanced in spite of the young age of this branch of fusion science. An overview of selected experimental and theoretical results obtained at Ioffe, Culham and Frascati is given with the emphasis on their complementarity and links to the main stream of tokamak research, such as ITER. An outline of the basic ST advantages and the potential of ST research for new insights into magnetic confinement is also given. More detailed descriptions of recent advances in ST theory and experiment may be found in the invited papers by Akers and Ono in the proceedings of this conference

Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo

Energy Technology Data Exchange (ETDEWEB)

Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

1998-07-01

Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

Current drive by Alfven waves in elongated cross section tokamak

International Nuclear Information System (INIS)

Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A.; Assis, A.S. de

1997-01-01

Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves 9GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory

Current drive by Alfven waves in elongated cross section tokamak

Energy Technology Data Exchange (ETDEWEB)

Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Inst. de Fisica; Assis, A.S. de [Universidade Federal Fluminense, Niteroi, RJ (Brazil). Inst. de Fisica

1997-12-31

Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves (GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory

Deposit of thin films for Tokamaks conditioning; Deposito de peliculas delgadas para acondicionar Tokamaks

Energy Technology Data Exchange (ETDEWEB)

Valencia A, R

2006-07-01

As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (<10 eV), both in noble and reactive gases, as well as the conditioning by thin film deposits of hydrogen rich amorphous carbon (carbonization) leading to a reduction in the plasma resistivity from 8.99 x 10{sup -6} to 4.5 x 10{sup -6} {omega}-m, thus taking the Z{sub ef} value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission

Control-oriented Automatic System for Transport Analysis (ASTRA)-Matlab integration for Tokamaks

International Nuclear Information System (INIS)

Sevillano, M.G.; Garrido, I.; Garrido, A.J.

2011-01-01

The exponential growth in energy consumption has led to a renewed interest in the development of alternatives to fossil fuels. Between the unconventional resources that may help to meet this energy demand, nuclear fusion has arisen as a promising source, which has given way to an unprecedented interest in solving the different control problems existing in nuclear fusion reactors such as Tokamaks. The aim of this manuscript is to show how one of the most popular codes used to simulate the performance of Tokamaks, the Automatic System For Transport Analysis (ASTRA) code, can be integrated into the Matlab-Simulink tool in order to make easier and more comfortable the development of suitable controllers for Tokamaks. As a demonstrative case study to show the feasibility and the goodness of the proposed ASTRA-Matlab integration, a modified anti-windup Proportional Integral Derivative (PID)-based controller for the loop voltage of a Tokamak has been implemented. The integration achieved represents an original and innovative work in the Tokamak control area and it provides new possibilities for the development and application of advanced control schemes to the standardized and widely extended ASTRA transport code for Tokamaks. -- Highlights: → The paper presents a useful tool for rapid prototyping of different solutions to deal with the control problems arising in Tokamaks. → The proposed tool embeds the standardized Automatic System For Transport Analysis (ASTRA) code for Tokamaks within the well-known Matlab-Simulink software. → This allows testing and combining diverse control schemes in a unified way considering the ASTRA as the plant of the system. → A demonstrative Proportional Integral Derivative (PID)-based case study is provided to show the feasibility and capabilities of the proposed integration.

Stability of tokamak magnetic configuration with a poloidal divertor

International Nuclear Information System (INIS)

Bazaeva, A.V.; Bykov, V.E.; Georgievskii, A.V.; Kaminskii, A.O.; Peletminskaya, V.G.; Pyatov, V.H.

1979-02-01

This paper investigates instabilities in the preseparatrix region of a tokamak magnetic configuration with a poloidal divertor with respect to perturbations produced by various irregularities in the manufacturing of tokamak magnetic systems. A computer solution, a system of differential equations describing the behavior of a force line, showed that small perturbation amplitudes may be the cause of the stochastic instability of force lines in the preseparatrix region. This instability is responsible for a number of demands on the accuracy in the manufacturing of tokamak magnetic systems. In particular, the misalignment in the divertor ring must not be larger than 0.5 0 , its displacement must be less than Δ/R = 10 -2 (Δ/R -2 ). This study can be used in the design of large thermonuclear installations

Electron cyclotron current drive efficiency in an axisymmetric tokamak

Energy Technology Data Exchange (ETDEWEB)

Gutierrez-Tapia, C.; Beltran-Plata, M. [Instituto Nacional de Investigaciones Nucleares, Dept. de Fisica, Mexico D.F. (Mexico)

2004-07-01

The neoclassical transport theory is applied to calculate electron cyclotron current drive (ECCD) efficiency in an axisymmetric tokamak in the low-collisionality regime. The tokamak ordering is used to obtain a system of equations that describe the dynamics of the plasma where the nonlinear ponderomotive (PM) force due to high-power radio-frequency (RF) waves is included. The PM force is produced around an electron cyclotron resonant surface at a specific poloidal location. The ECCD efficiency is analyzed in the cases of first and second harmonics (for different impinging angles of the RF waves) and it is validated using experimental parameter values from TCV and T-10 tokamaks. The results are in agreement with those obtained by means of Green's function techniques. (authors)

First experiments with SST-1 tokamak

International Nuclear Information System (INIS)

Saxena, Y.C.

2005-01-01

SST-1, a steady state superconducting tokamak, is undergoing commissioning tests at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in a tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. The auxiliary current drive is based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. Detailed commissioning tests on the cryogenic system and experiments on the hydraulic characters and cool down features of single TF coils have been completed prior to the cool down of the entire superconducting system. Results of the single TF magnet cool down, and testing of the magnet system are presented. First experiments related to the breakdown and the current ramp up will subsequently be carried out. (author)

Experimental and theoretical basis for advanced tokamaks

International Nuclear Information System (INIS)

Chan, V.S.

1994-09-01

In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

Behaviour of metallic droplets in a tokamak plasma

International Nuclear Information System (INIS)

Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.

1989-01-01

Micrometre sized tantalum droplets were injected into a tokamak plasma by a controllable arcing gun located behind the wall. The trajectories of the ablating particles were photographed by a high speed camera. Various possible mechanisms which may explain the observed curvature of the particle paths are discussed. The migration of the ablated material in the tokamak was studied by post-mortem analysis of collector probes and limiters. (author). Letter-to-the-editor. 12 refs, 9 figs

Models for impurity effects in tokamaks

International Nuclear Information System (INIS)

Hogan, J.T.

1980-03-01

Models for impurity effects in tokamaks are described with an emphasis on the relationship between attainment of high β and impurity problems. We briefly describe the status of attempts to employ neutral beam heating to achieve high β in tokamaks and propose a qualitative model for the mechanism by which heavy metal impurities may be produced in the startup phase of the discharge. We then describe paradoxes in impurity diffusion theory and discuss possible resolutions in terms of the effects of large-scale islands and sawtooth oscillations. Finally, we examine the prospects for the Zakharov-Shafranov catastrophe (long time scale disintegration of FCT equilibria) in the context of present and near-term experimental capability

LIDAR Thomson scattering for advanced tokamaks. Final report

International Nuclear Information System (INIS)

Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

1996-01-01

The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured

Conceptual design of Remote Control System for EAST tokamak

International Nuclear Information System (INIS)

Sun, X.Y.; Wang, F.; Wang, Y.; Li, S.

2014-01-01

Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication

Conceptual design of Remote Control System for EAST tokamak

Energy Technology Data Exchange (ETDEWEB)

Sun, X.Y., E-mail: xysun@ipp.ac.cn; Wang, F.; Wang, Y.; Li, S.

2014-05-15

Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.

Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

International Nuclear Information System (INIS)

Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

2016-01-01

Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

Energy Technology Data Exchange (ETDEWEB)

Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

2016-11-01

Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

Initial DEMO tokamak design configuration studies

Energy Technology Data Exchange (ETDEWEB)

Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

2015-10-15

Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

Liquid tin limiter for FTU tokamak

Energy Technology Data Exchange (ETDEWEB)

Vertkov, A., E-mail: avertkov@yandex.ru [JSC “Red Star”, Moscow (Russian Federation); Lyublinski, I. [JSC “Red Star”, Moscow (Russian Federation); NRNU MEPhI, Moscow (Russian Federation); Zharkov, M. [JSC “Red Star”, Moscow (Russian Federation); Mazzitelli, G.; Apicella, M.L.; Iafrati, M. [Associazione EURATOM-ENEA sulla Fusione, C. R. Frascati, Frascati, Rome, Italy, (Italy)

2017-04-15

Highlights: • First steady state operating liquid tin limiter TLL is under study on FTU tokamak. • The cooling system with water spray coolant for TLL has been developed and tested. • High corrosion resistance of W and Mo in molten Sn confirmed up to 1000 °C. • Wetting process with Sn has been developed for Mo and W. - Abstract: The liquid Sn in a matrix of Capillary Porous System (CPS) has a high potential as plasma facing material in steady state operating fusion reactor owing to its physicochemical properties. However, up to now it has no experimental confirmation in tokamak conditions. First steady state operating limiter based on the CPS with liquid Sn installed on FTU tokamak and its experimental study is in progress. Several aspects of the design, structural materials and operation parameters of limiter based on tungsten CPS with liquid Sn are considered. Results of investigation of corrosion resistance of Mo and W in Sn and their wetting process are presented. The heat removal for limiter steady state operation is provided by evaporation of flowing gaswater spray. The effectiveness of such heat removal system is confirmed in modelling tests with power flux up to 5 MW/m2.