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Sample records for sunist spherical tokamak

  1. Design and calibration of high-frequency magnetic probes for the SUNIST spherical tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Yangqing; Tan, Yi; Pan, Ou; Ke, Rui; Wang, Wenhao; Gao, Zhe

    2014-11-01

    A new high-frequency magnetic diagnostic system is designed, installed, and calibrated in the Sino-United Spherical Tokamak (SUNIST) to investigate Alfvén waves (AWs). The system consists of a fixed toroidal array and a movable radial array of high-frequency magnetic probes (HFMPs) with 21 and 60 probes, respectively. Based on the method of vacuum enameled wire wound on ceramic bobbins, the fixed toroidal array is located as near as possible to the plasma and carefully shielded to reduce the attenuation of high-frequency magnetic field. Meanwhile, by using the technology of commercial chip inductors mounted on printed circuit boards, the movable radial array is inserted into a thin quartz tube that allows positioning along radial direction. A Helmholtz coil is utilized to calibrate the effective areas as well as the frequency response of each HFMP. The calibration results are consistent with the calculated results of an equivalent probe-and-cable circuit model. High-frequency magnetic signals related to AW are detected with these HFMPs. These HFMPs are expected to play a key role in analyzing Alfvén eigenmodes excited by AW antenna in the SUNIST.

  2. Design and calibration of high-frequency magnetic probes for the SUNIST spherical tokamak.

    PubMed

    Liu, Yangqing; Tan, Yi; Pan, Ou; Ke, Rui; Wang, Wenhao; Gao, Zhe

    2014-11-01

    A new high-frequency magnetic diagnostic system is designed, installed, and calibrated in the Sino-United Spherical Tokamak (SUNIST) to investigate Alfvén waves (AWs). The system consists of a fixed toroidal array and a movable radial array of high-frequency magnetic probes (HFMPs) with 21 and 60 probes, respectively. Based on the method of vacuum enameled wire wound on ceramic bobbins, the fixed toroidal array is located as near as possible to the plasma and carefully shielded to reduce the attenuation of high-frequency magnetic field. Meanwhile, by using the technology of commercial chip inductors mounted on printed circuit boards, the movable radial array is inserted into a thin quartz tube that allows positioning along radial direction. A Helmholtz coil is utilized to calibrate the effective areas as well as the frequency response of each HFMP. The calibration results are consistent with the calculated results of an equivalent probe-and-cable circuit model. High-frequency magnetic signals related to AW are detected with these HFMPs. These HFMPs are expected to play a key role in analyzing Alfvén eigenmodes excited by AW antenna in the SUNIST. PMID:25430367

  3. Movable multi-probes for plasma boundary measurement in sino-united spherical tokamak

    SciTech Connect

    Chai, Song Wang, Wenhao; Tan, Yi; Gao, Zhe

    2014-11-15

    A novel movable multi-probes is developed to get local magnetic and electrostatic profiles on Sino-UNIted Spherical Tokamak (SUNIST). This multi-probes combines a four-tips Langmuir probe, a magnetic coil, and a retarding field energy analyzer (RFEA). It can be used to simultaneously measure the poloidal magnetic field B{sub p}, electric field E{sub r}, electron temperature T{sub e}, electron density n{sub e}, and ion temperature T{sub i}. Its small overall size (20 × 20 × 38 mm{sup 3}) enables the movable multi-probes to measure the magnetic and electrostatic profiles in high spatial resolution, with negligible impact to plasma in SUNIST. This paper presents the design of the movable multi-probes, in particular, details of RFEA for reliable ion energy measurements. Preliminary experimental results of the movable multi-probes are given as well.

  4. Assessment of the beam path deflection for a vertically installed microwave interferometer in SUNIST.

    PubMed

    Zhong, H; Ling, B L; Tan, Y; Gao, Z

    2016-08-01

    Microwave interferometry has been widely employed to provide reliable line averaged electron density measurement on plasma devices. For a vertically installed interferometer on a tokamak, the refraction problem, which distorts the beam path and aggravates power loss at the receiving antenna, may become significant if taking the cross section shape into account. Increasing the frequency of the probing microwave can alleviate the distortion, but at the expense of losing the density resolution. To seek for an optimized frequency, previous calculations are mainly based on the cylindrical column geometry which grossly underestimates the deflection of the beam path induced by the plasma shape, and empirical suggestions indicating ne0/nc = 1/2 ∼ 1/3 may not always be the appropriate option. Here a single ray tracing method is applied to estimate the final horizontal deviation at the receiving antenna, which is supposed to represent the level of power loss. The calculation is carried out under the real tokamak geometry in Sino-UNIted Spherical Tokamak (SUNIST) with the cross section parameters obtained from the equilibrium reconstruction, and the result indicates that for a target density of 1.2 × 10(19) m(-3), a frequency of at least 100 GHz is desirable to reduce the power loss to an acceptable level. This would be helpful for the design of a vertically installed interferometer on SUNIST. PMID:27587117

  5. Assessment of the beam path deflection for a vertically installed microwave interferometer in SUNIST

    NASA Astrophysics Data System (ADS)

    Zhong, H.; Ling, B. L.; Tan, Y.; Gao, Z.

    2016-08-01

    Microwave interferometry has been widely employed to provide reliable line averaged electron density measurement on plasma devices. For a vertically installed interferometer on a tokamak, the refraction problem, which distorts the beam path and aggravates power loss at the receiving antenna, may become significant if taking the cross section shape into account. Increasing the frequency of the probing microwave can alleviate the distortion, but at the expense of losing the density resolution. To seek for an optimized frequency, previous calculations are mainly based on the cylindrical column geometry which grossly underestimates the deflection of the beam path induced by the plasma shape, and empirical suggestions indicating ne0/nc = 1/2 ˜ 1/3 may not always be the appropriate option. Here a single ray tracing method is applied to estimate the final horizontal deviation at the receiving antenna, which is supposed to represent the level of power loss. The calculation is carried out under the real tokamak geometry in Sino-UNIted Spherical Tokamak (SUNIST) with the cross section parameters obtained from the equilibrium reconstruction, and the result indicates that for a target density of 1.2 × 1019 m-3, a frequency of at least 100 GHz is desirable to reduce the power loss to an acceptable level. This would be helpful for the design of a vertically installed interferometer on SUNIST.

  6. Microtearing modes in spherical and conventional tokamaks

    NASA Astrophysics Data System (ADS)

    Moradi, S.; Pusztai, I.; Guttenfelder, W.; Fülöp, T.; Mollén, A.

    2013-06-01

    The onset and characteristics of microtearing modes (MTM) in the core of spherical (NSTX) and conventional tokamaks (ASDEX Upgrade and JET) are studied through local linear gyrokinetic simulations with GYRO (Candy and Belli 2011 General Atomics Report GA-A26818). For experimentally relevant core plasma parameters in the NSTX and ASDEX Upgrade tokamaks, in agreement with previous works, we find MTMs as the dominant linear instability. Also, for JET-like core parameters considered in our study an MTM is found as the most unstable mode. In all of these plasmas, finite collisionality is needed for MTMs to become unstable and the electron temperature gradient is found to be the fundamental drive. However, a significant difference is observed in the dependence of the linear growth rate of MTMs on electron temperature gradient. While it varies weakly and non-monotonically in JET and ASDEX Upgrade plasmas, in NSTX it increases with the electron temperature gradient.

  7. Spherical tokamaks with plasma centre-post

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso

    2013-10-01

    The metal centre-post (MCP) in tokamaks is a structure which carries the total toroidal field current and also houses the Ohmic heating solenoid in conventional or low aspect ratio (Spherical)(ST) tokamaks. The MCP and solenoid are critical components for producing the toroidal field and for the limited Ohmic flux in STs. Constraints for a ST reactor related to these limitations lead to a minimum plasma aspect ratio of 1.4 which reduces the benefit of operation at higher betas in a more compact ST reactor. Replacing the MCP is of great interest for reactor-based ST studies since the device is simplified, compactness increased, and maintenance reduced. An experiment to show the feasibility of using a plasma centre-post (PCP) is being currently under construction and involves a high level of complexity. A preliminary study of a very simple PCP, which is ECR(Electron Cyclotron Resonance)-assisted and which includes an innovative fuelling system based on pellet injection, has recently been reported. This is highly suitable for an ultra-low aspect ratio tokamak (ULART) device. Advances on this PCP ECR-assisted concept within a ULART and the associated fuelling system are presented here, and will include the field topology for the PCP ECR-assisted scheme, pellet ablation modeling, and a possible global equilibrium simulation. VIE-ITCR, IAEA-CRP contr.17592, National Instruments-Costa Rica.

  8. The Spherical Tokamak MEDUSA for Mexico

    NASA Astrophysics Data System (ADS)

    Ribeiro, C.; Salvador, M.; Gonzalez, J.; Munoz, O.; Tapia, A.; Arredondo, V.; Chavez, R.; Nieto, A.; Gonzalez, J.; Garza, A.; Estrada, I.; Jasso, E.; Acosta, C.; Briones, C.; Cavazos, G.; Martinez, J.; Morones, J.; Almaguer, J.; Fonck, R.

    2011-10-01

    The former spherical tokamak MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14m, a < 0.10m, BT < 0.5T, Ip < 40kA, 3ms pulse) is currently being recomissioned at the Universidad Autónoma de Nuevo León, Mexico, as part of an agreement between the Faculties of Mech.-Elect. Eng. and Phy. Sci.-Maths. The main objective for having MEDUSA is to train students in plasma physics & technical related issues, aiming a full design of a medium size device (e.g. Tokamak-T). Details of technical modifications and a preliminary scientific programme will be presented. MEDUSA-MX will also benefit any developments in the existing Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

  9. First Results of the ETE Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    Ludwig, G. O.; del Bosco, E.; Ferreira, J. G.; Barroso, J. J.; Berni, L. A.; Oliveira, R. M.

    2001-10-01

    First tokamak plasma discharges were obtained in the ETE Spherical Tokamak Experiment (Experimento Tokamak Esferico) by November 2000. ETE is a medium size machine (major radius R=0.30m) with a compact design and good access for diagnostics. During the first phase of operation a plasma current of 200kA (t=15ms) will be produced in a 1.5 aspect ratio configuration with a toroidal magnetic field up to 0.4T. The ultimate values are limited by mechanical stresses in the joints of the toroidal field coil (B<0.8T), and by stresses and heating in the solenoid (0.24Wb, 30kA, 0.2s) for a maximum plasma current of about 400kA. Presently the experiments are focused on plasma formation, vacuum conditioning and diagnostics implementation. The machine was constructed in accordance with stringent design specifications. The assembly has an overall precision better than 2mm. Vacuum conditioning is being improved with baking, glow discharge cleaning and usual tokamak operation. Breakdown is easily obtained even without the pre-ionization provided by a hot filament electron emitter and by an ultraviolet lamp. Preliminary measurements of stray magnetic fields were carried out and eddy current effects are being modeled. Energy of the capacitor banks is being continuously increased to achieve the design parameters. A fast neutral lithium beam probe for plasma edge studies and a 6.7GHz high-power monotron for pre-ionization and electron cyclotron resonance plasma heating experiments are under development.

  10. The Spherical Tokamak MEDUSA for Costa Rica

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso; Vargas, Ivan; Guadamuz, Saul; Mora, Jaime; Ansejo, Jose; Zamora, Esteban; Herrera, Julio; Chaves, Esteban; Romero, Carlos

    2012-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R<0.14m, a<0.10m, BT<0.5T, Ip<40kA, 3ms pulse)[1] is in a process of donation to Costa Rica Institute of Technology. The main objective of MEDUSA is to train students in plasma physics /technical related issues which will help all tasks of the very low aspect ratio stellarator SCR-1(A≡R/>=3.6, under design[2]) and also the ongoing activities in low temperature plasmas. Courses in plasma physics at undergraduate and post-graduate joint programme levels are regularly conducted. The scientific programme is intend to clarify several issues in relevant physics for conventional and mainly STs, including transport, heating and current drive via Alfv'en wave, and natural divertor STs with ergodic magnetic limiter[3,4]. [1] G.D.Garstka, PhD thesis, University of Wisconsin at Madison, 1997 [2] L.Barillas et al., Proc. 19^th Int. Conf. Nucl. Eng., Japan, 2011 [3] C.Ribeiro et al., IEEJ Trans. Electrical and Electronic Eng., 2012(accepted) [4] C.Ribeiro et al., Proc. 39^th EPS Conf. Contr. Fusion and Plasma Phys., Sweden, 2012

  11. Probing spherical tokamak plasmas using charged fusion products

    NASA Astrophysics Data System (ADS)

    Boeglin, Werner U.; Perez, Ramona V.; Darrow, Douglass S.; Cecconello, Marco; Klimek, Iwona; Allan, Scott Y.; Akers, Rob J.; Jones, Owen M.; Keeling, David L.; McClements, Ken G.; Scannell, Rory

    2015-11-01

    The detection of charged fusion products, such as protons and tritons resulting from D(d,p)t reactions, can be used to determine the fusion reaction rate profile in large spherical tokamak plasmas with neutral beam heating. The time resolution of a diagnostic of this type makes it possible to study the slowly-varying beam density profile, as well as rapid changes resulting from MHD instabilities. A 4-channel prototype proton detector (PD) was installed and operated on the MAST spherical tokamak in August/September 2013, and a new 6-channel system for the NSTX-U spherical tokamak is under construction. PD and neutron camera measurements obtained on MAST will be compared with TRANSP calculations, and the design of the new NSTX-U system will be presented, together with the first results from this diagnostic, if available. Supported in part by DOE DE-SC0001157.

  12. Thermal Fluid Multiphysics Optimization of Spherical Tokamak

    SciTech Connect

    Lumsdaine, Arnold; Tipton, Joseph B; Peng, Yueng Kay Martin

    2012-01-01

    An experimental Fusion Nuclear Science Facility (FNSF) is required that will create the environment that simultaneously achieves high energy neutrons and high ion fluence necessary in order to bridge the gaps from ITER to the realization of a fusion nuclear power plant. One concept for achieving this is a high duty cycle spherical torus. This study will focus on thermal modeling of the spherical torus centerpost using computational fluid dynamics to effectively model the thermal transfer of the cooling fluid to the centerpost. The design of the fluid channels is optimized in order to minimize the temperature in the centerpost. Results indicate the feasibility of water cooling for a long-pulse spherical torus FNSF.

  13. The Science of Spherical Tokamak Plasmas: Progress and Promise

    NASA Astrophysics Data System (ADS)

    Sykes, Alan

    2008-11-01

    The talk will summarize the development of the low aspect ratio `Spherical' Tokamak (ST) from early linear magnetic confinement devices, through toroidal pinches, to the emergence of the tokamak in the 1960's. Theoretical predictions given by Peng and Strickler of the exciting physics of extreme low aspect ratio tokamaks (supported by early experiments involving centre rods inserted into existing Rotamaks, Spheromaks and other small-scale experiments), led to the pioneering START experiment at Culham which convincingly demonstrated the potential of the ST concept. There are now many STs world-wide. The largest among these are MA-scale devices NSTX and MAST with plasmas of cross-section comparable to DIII-D and Asdex-Upgrade. The major results include development of start-up methods; the refinement of scaling laws; improved understanding of general tokamak phenomena such as Edge Localised Modes and development of heating and current drive schemes. ST research on over 20 devices has extended the tokamak plasma regime in many ways, notably a factor 4 increase in stable toroidal average beta, and large increases in the Alfven Mach number and ExB flow shear. By exploiting such features, joint experiments with tokamaks of conventional aspect ratio are resolving several key degeneracies of interest to ITER, DEMO and larger future ST devices. Present STs have low toroidal fields sufficient for most physics studies, but for high fusion yield or energy production higher fields are required; importantly, studies on both NSTX and MAST indicate a stronger than expected improvement of performance with toroidal field. Both devices are planning exciting upgrades which feature a considerable increase of toroidal field. Recent designs for a D-T Component Test Facility based on the Spherical Tokamak show the promise of low Tritium consumption and minimum build cost. Such a facility would provide valuable R&D on the scientific and technical issues of fusion power.

  14. The role of spherical torus in clarifying tokamak physics

    SciTech Connect

    Morris, A. W.; Peng, Yueng Kay Martin

    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.

  15. Estimation of Electron Temperature on Glass Spherical Tokamak (GLAST)

    NASA Astrophysics Data System (ADS)

    Hussain, S.; Sadiq, M.; Shah, S. I. W.; GLAST Team

    2015-03-01

    Glass Spherical Tokamak (GLAST) is a small spherical tokamak indigenously developed in Pakistan with an insulating vacuum vessel. A commercially available 2.45 GHz magnetron is used as pre-ionization source for plasma current startup. Different diagnostic systems like Rogowski coils, magnetic probes, flux loops, Langmuir probe, fast imaging and emission spectroscopy are installed on the device. The plasma temperature inside of GLAST, at the time of maxima of plasma current, is estimated by taking into account the Spitzer resistivity calculations with some experimentally determined plasma parameters. The plasma resistance is calculated by using Ohm's law with plasma current and loop voltage as experimentally determined inputs. The plasma resistivity is then determined by using length and area of the plasma column. Finally, the average plasma electron temperature is predicted to be 12.65eV for taking neon (Ne) as a working gas.

  16. Solenoid-free plasma start-up in spherical tokamaks

    NASA Astrophysics Data System (ADS)

    Raman, R.; Shevchenko, V. F.

    2014-10-01

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

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

    NASA Astrophysics Data System (ADS)

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

    2009-05-01

    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 CO2 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. CO2 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.

  18. A charged fusion product diagnostic for a spherical tokamak

    NASA Astrophysics Data System (ADS)

    Perez, Ramona Leticia Valenzuela

    Designs for future nuclear fusion power reactors rely on the ability to create a stable plasma (hot ionized gas of hydrogen isotopes) as a medium with which to sustain nuclear fusion reactions. My dissertation work involves designing, constructing, testing, installing, operating, and validating a new diagnostic for spherical tokamaks, a type of reactor test facility. Through detecting charged particles emitted from the plasma, this instrument can be used to study fusion reaction rates within the plasma and how they are affected by plasma perturbations. Quantitatively assessing nuclear fusion reaction rates at specific locations inside the plasma and as a function of time can provide valuable data that can be used to evaluate theory-based simulations related to energy transport and plasma stability. The Proton Detector (PD), installed in the Mega Amp Spherical Tokamak (MAST) at the Culham Centre for Fusion Energy (CCFE) in Abingdon, England, was the first instrument to experimentally detect 3 MeV Protons and 1 MeV Tritons created from deuterium- deuterium (hydrogen isotopes) nuclear fusion reactions inside a spherical tokamak's plasma. The PD consists of an array of particle detectors with a protective housing and the necessary signal conditioning electronics and readout. After several years of designing (which included simulations for detector orientations), fabricating, and testing the PD, it was installed in MAST and data were collected over a period of two months in the summer of 2013. Proton and triton rates as high as 200 kHz were measured and an initial radial profile of these fusion reaction rates inside the plasma was extracted. These results will be compared to a complementary instrument at MAST as well as theory-based simulations and form the knowledge basis for developing a larger future instrument. The design and performance of all instrument components (electrical, computational, mechanical), and subsequent data analysis methods and results are

  19. Centre-solenoid-free merging start-up of spherical tokamak plasmas in UTST

    NASA Astrophysics Data System (ADS)

    Inomoto, M.; Watanabe, T. G.; Gi, K.; Yamasaki, K.; Kamio, S.; Imazawa, R.; Yamada, T.; Guo, X.; Ushiki, T.; Ishikawa, H.; Nakamata, H.; Kawakami, N.; Sugawara, T.; Matsuyama, K.; Noma, K.; Kuwahata, A.; Tanabe, H.

    2015-03-01

    A centre-solenoid-free merging start-up scheme for spherical tokamak plasmas was developed in a University of Tokyo spherical tokamak (UTST) experiment by using outer poloidal field coils. Torus breakdown was initiated at null points and two spherical tokamak plasmas with a total current up to 80 kA were generated inductively. Their merging process provided substantial ion and electron heating by magnetic reconnection. The obtained dependence of heating on plasma current suggests that high-temperature and high-current plasma suitable for neutral beam injection is attainable under the realistic conditions in the merging start-up method.

  20. Review of Globus-M spherical tokamak results

    NASA Astrophysics Data System (ADS)

    Gusev, V. K.; Bakharev, N. N.; Belyakov, V. A.; Ber, B. Ya.; Bondarchuk, E. N.; Bulanin, V. V.; Bykov, A. S.; Chernyshev, F. V.; Demina, E. V.; Dyachenko, V. V.; Goncharov, P. R.; Gorodetsky, A. E.; Gusakov, E. Z.; Iblyaminova, A. D.; Ivanov, A. A.; Irzak, M. A.; Kaveeva, E. G.; Khitrov, S. A.; Khokhlov, M. V.; Khromov, N. A.; Kolmogorov, V. V.; Kornev, V. A.; Krasnov, S. V.; Kurskiev, G. S.; Labusov, A. N.; Lepikhov, S. A.; Litunovsky, N. V.; Mazul, I. V.; Melnik, A. D.; Mikov, V. V.; Minaev, V. B.; Mineev, A. B.; Mironov, M. I.; Miroshnikov, I. V.; Mukhin, E. E.; Novokhatsky, A. N.; Ovsyannikov, A. D.; Patrov, M. I.; Petrov, A. V.; Petrov, Yu. V.; Rozhansky, V. A.; Sakharov, N. V.; Saveliev, A. N.; Senichenkov, I. Yu.; Sergeev, V. Yu.; Shchegolev, P. B.; Shcherbinin, O. N.; Shikhovtsev, I. V.; Tanaev, V. S.; Tanchuk, V. N.; Tolstyakov, S. Yu.; Varfolomeev, V. I.; Vekshina, E. O.; Voronin, A. V.; Voskoboinikov, S. P.; Wagner, F.; Yashin, A. Yu.; Zadvitskiy, G. V.; Zakharov, A. P.; Zalavutdinov, R. Kh.; Zhilin, E. G.

    2015-10-01

    The first experiments on noninductive current drive (CD) using lower hybrid waves in a spherical tokamak are described. Waves at 2.45 GHz were launched by a 10 waveguide grill with 120° phase shift between neighbouring waveguides. The experimental results for a novel poloidal slowing-down scheme are described. The CD efficiency is found to be somewhat larger than that predicted theoretically whilst at the same time being somewhat less than that for standard tokamak lower hybrid CD. Geodesic acoustic modes (GAM) have been discovered in Globus-M. GAMs are localized 2-3 cm inside the separatrix. The GAM frequency agrees with theory. The mode structures of plasma density and magnetic field oscillation at the GAM frequency have been studied. Fast particle confinement during neutral beam injection has been investigated and numerically simulated. Alfvén instabilities excited by fast particles were detected by a toroidal Mirnov probe array. Their excitation conditions are discussed and the dynamics of fast ion losses induced by Alfvén eigenmodes is presented. Preliminary experiments on the isotopic effect influence on global confinement in the ohmic heating (OH) regime are described. Scrape-off layer (SOL) parameters were measured and compared with results from self-consistent integrated transport modelling. Results showed that SOL width scales inversely proportional to plasma current. The behaviour of an a priori damaged tungsten divertor plate mock-up exposed to plasma flows was investigated. Preliminary conclusions are that the initial damage gives rise to a loose layer formation with low thermal conductivity right beneath the surface. Finally, engineering design issues of the next step—Globus-M2 (1 T, 500 kA) and the status of component manufacture are described.

  1. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator.

    SciTech Connect

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, Thomas W; Dix, D.; El-GuebalyUniv. Wisco, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D. G.; Zarnstorff, M. C.

    2011-01-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  2. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    NASA Astrophysics Data System (ADS)

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  3. Resolving electron scale turbulence in spherical tokamaks with flow shear

    SciTech Connect

    Guttenfelder, W.; Candy, J.

    2011-02-15

    This paper presents nonlinear gyrokinetic simulations of electron temperature gradient (ETG) turbulence based on spherical tokamak (ST) parameters. Most significantly the simulations include the strong toroidal flow and flow shear present in STs that suppress ion-scale turbulence while using kinetic ions at full mass ratio (m{sub i}/m{sub e}=3600). The flow shear provides a physical long-wavelength cutoff mechanism that aids saturation of the simulations, which has previously been demonstrated to be problematic depending on magnetic shear. As magnetic shear varies widely in STs we systematically demonstrate saturation and convergence of the ETG simulations with respect to grid resolution, physical domain size, and boundary conditions. While using reduced ion mass or adiabatic ions can lessen computational expense they do not always provide reliable results. The resulting spectra from converged simulations are anisotropic everywhere in contrast to previous ETG simulations without flow shear. These results have implications for interpreting turbulence measurements, and represent an important step in determining when and where ETG turbulence is expected to be relevant in ST plasmas. They are also important in the context of validating simulations with both experimental transport analysis and turbulence measurements.

  4. EBW H&CD Potential for Spherical Tokamaks

    NASA Astrophysics Data System (ADS)

    Urban, J.; Decker, J.; Peysson, Y.; Preinhaelter, J.; Shevchenko, V.; Taylor, G.; Vahala, L.; Vahala, G.

    2011-12-01

    Spherical tokamaks (STs), which feature relatively high neutron flux and good economy, operate generally in high-ß regimes, in which the usual EC O- and X- modes are cut-off. In this case, electron Bernstein waves (EBWs) seem to be the only option that can provide features similar to the EC waves—controllable localized heating and current drive (H&) that can be utilized for core plasma heating as well as for accurate plasma stabilization. We first derive an analytical expression for Gaussian beam OXB conversion efficiency. Then, an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX) is performed. Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

  5. CXRS measurements of ion temperature in NBI discharges on Globus-M spherical tokamak

    NASA Astrophysics Data System (ADS)

    Avdeeva, G. F.; Miroshnikov, I. V.; Bakharev, N. N.; Kurskiev, G. S.; Patrov, M. I.; Sergeev, V. Yu; Schegolev, P. B.

    2016-01-01

    The thesis describes the Charge-eXchange Recombination Spectroscopy (CXRS) diagnostics setup and its application to the ion temperature determination in the NBI-heated discharges in the Globus-M spherical tokamak. Measurements of the spectral line shape of the radiation emitted by the hydrogen-like carbon ion CVI (5290.5 Å) are presented. Comparison of CXRS and Neutral Particle Analyser (NPA) data demonstrates good agreement.

  6. Bounce Precession Fishbones in the National Spherical Tokamak Experiment

    SciTech Connect

    Eric Fredrickson; Liu Chen; Roscoe White Eric Fredrickson; Roscoe White

    2003-06-27

    Bursting modes are observed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40 (2000) 557], which are identified as bounce-precession-frequency fishbone modes. They are predicted to be important in high-current, low-shear discharges with a significant population of trapped particles with a large mean-bounce angle, such as produced by near-tangential beam injection into a large aspect-ratio device. Such a distribution is often stable to the usual precession-resonance fishbone mode. These modes could be important in ignited plasmas, driven by the trapped-alpha-particle population.

  7. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    SciTech Connect

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  8. Fast Ion Non-adiabaticity in Spherical Tokamaks

    SciTech Connect

    V.A. Yavorskij; D. Darrow; V.Ya. Goloborod'ko; S.N. Reznik; U. Holzmueller-Steinacker; N. Gorelenkov; K. Schoepf

    2002-08-01

    Transport processes of fast ions in axisymmetric low-aspect-ratio spherical torus (ST) plasmas are investigated, which are induced by the non-conservation of the magnetic moment {mu}. The reason for non-conservation of {mu} of fast ions in ST's is the relatively large adiabaticity parameter epsilon typically exceeding the value 0.1 (epsilon = ratio of ion gyroradius to the gradient scale length of the magnetic field). Both analytical and numerical evaluations of the magnitude of nonadiabatic variations of {mu} are performed. Nonadiabaticity effects are shown to be most significant for fast ions for which the bounce oscillations are in resonance with the gyromotion, i.e., for ions with omega(subscript)B - lomega(subscript)b = 0, where omega(subscript)B and omega(subscript)b represent the bounce-averaged gyrofrequency and the bounce frequency, respectively, and l is an integer. The critical threshold of the adiabaticity parameter, epsilon(subscript)cr, to be exceeded for the transition to stochastic behavior of fast ions in axisymmetric ST's is inspected. Nonadiabatic variations of {mu} are shown to lead to collisionless transformation of trapped orbits into circulating ones and vice versa. For the case of strong nonadiabaticity, epsilon > epsilon(subscript)cr, we assess the transport coefficients describing intense collisionless pitch-angle diffusion, whereas, in the case of weak nonadiabaticity, epsilon > epsilon(subscript)cr, the more substantial coefficients of enhanced collisional radial diffusion and convection of fast ions gyrating resonantly with the bounce oscillations are estimated.

  9. Fast scanning probe for the NSTX spherical tokamak

    NASA Astrophysics Data System (ADS)

    Boedo, J. A.; Crocker, N.; Chousal, L.; Hernandez, R.; Chalfant, J.; Kugel, H.; Roney, P.; Wertenbaker, J.; NSTX Team

    2009-12-01

    We describe a fast reciprocating Langmuir probe and drive system, which has four main new features: (1) use of high-temperature, vacuum, circuit boards instead of cables to reduce weight and increase to 21 the number of possible connections, (2) rotatable and removable shaft, (3) 10 tip construction with designed hardware bandwidth up to 10 MHz, and (4) a detachable and modular tip assembly for easy maintenance. The probe is mounted in a fast pneumatic drive capable of speeds ˜7 m/s and ˜20g's acceleration in order to reach the scrape-off layer (SOL) and pedestal regions and remain inserted long enough to obtain good statistics while minimizing the heat deposition to the tips and head in a power density environment of 1-10 MW/m2. The National Spherical Torus Experiment SOL features electron temperature, Te˜10-30 eV, and electron density, ne˜0.1-5×1012 cm-3 while the pedestal features ne˜0.5-1.5×1013 cm-3 and Te˜30-150 eV. The probe described here has ten tips which obtain a wide spectrum of plasma parameters: electron temperature profile Te(r), electron density profile ne(r) and Mach number profile M(r ), floating potential Vf(r), poloidal and radial electric field profiles Eθ(r) and Eρ(r), saturation current profile Isat(r), and their fluctuations up to 3 MHz. We describe the probe and show representative radial profiles of various parameters.

  10. Fast scanning probe for the NSTX spherical tokamak.

    PubMed

    Boedo, J A; Crocker, N; Chousal, L; Hernandez, R; Chalfant, J; Kugel, H; Roney, P; Wertenbaker, J

    2009-12-01

    We describe a fast reciprocating Langmuir probe and drive system, which has four main new features: (1) use of high-temperature, vacuum, circuit boards instead of cables to reduce weight and increase to 21 the number of possible connections, (2) rotatable and removable shaft, (3) 10 tip construction with designed hardware bandwidth up to 10 MHz, and (4) a detachable and modular tip assembly for easy maintenance. The probe is mounted in a fast pneumatic drive capable of speeds approximately 7 m/s and approximately 20g's acceleration in order to reach the scrape-off layer (SOL) and pedestal regions and remain inserted long enough to obtain good statistics while minimizing the heat deposition to the tips and head in a power density environment of 1-10 MW/m2. The National Spherical Torus Experiment SOL features electron temperature, T(e) approximately 10-30 eV, and electron density, n(e) approximately 0.1-5x10(12) cm(-3) while the pedestal features n(e) approximately 0.5-1.5x10(13) cm(-3) and T(e) approximately 30-150 eV. The probe described here has ten tips which obtain a wide spectrum of plasma parameters: electron temperature profile T(e)(r), electron density profile n(e)(r) and Mach number profile M(r), floating potential V(f)(r), poloidal and radial electric field profiles E(theta)(r) and E(rho)(r), saturation current profile I(sat)(r), and their fluctuations up to 3 MHz. We describe the probe and show representative radial profiles of various parameters. PMID:20073119

  11. Nonlinear MHD simulation of DC helicity injection in the Pegasus spherical tokamak

    NASA Astrophysics Data System (ADS)

    Bayliss, Adam; Sovinec, Carl

    2006-10-01

    DC helicity injection has been successfully employed in spherical tokamaks (ST's) to produce a tokamak-like plasma with either a poloidal-gap voltage known as coaxial helicity injection [HIT-II, NSTX] or a biased cathode gun configuration [CDX, PEGASUS]. In PEGASUS, the tokamak-like plasma which is subsequently ohmically driven is the product of a reversal of vacuum poloidal flux and a merger of gun-injected current filaments. A 3D nonlinear MHD computation using the NIMROD code [Sovinec et al. JCP 195, 355 (2004)] simulates the formation, merger, and relaxation of the gun-injected current filaments to the tokamak-like plasma. The reversal of poloidal flux due to the field induced by the helicity drive is reproduced and the MHD processes leading to the merger and relaxation of the current filaments are described. Over the lifetime of a helically-driven experimental shot (approximately 10ms), the extent to which the merged plasma exhibits amplication of poloidal flux and the injected current in the relaxed state, reported in PEGASUS, is explored. The results are compared with simulations of current drive in NSTX via coaxial helicity injection which exhibit an n=1 open field-line kink [Tang and Boozer, Phys. Plasmas 11, 2679 (2004)].

  12. Theory Issues for Induced Plasma Convection Experiments in the Divertor of the MAST Spherical Tokamak

    SciTech Connect

    Cohen, R H; Fielding, S; Helander, P; Ryutov, D D

    2001-09-05

    This paper surveys theory issues associated with inducing convective cells through divertor tile biasing in a tokamak to broaden the scrape-off layer (SOL). The theory is applied to the Mega-Ampere Spherical Tokamak (MAST), where such experiments are planned in the near future. Criteria are presented for achieving strong broadening and for exciting shear-flow turbulence in the SOL; these criteria are shown to be attainable in practice. It is also shown that the magnetic shear present in the vicinity of the X-point is likely to confine the potential perturbations to the divertor region below the X-point, leaving the part of the SOL that is in direct contact with the core plasma intact. The current created by the biasing and the associated heating power are found to be modest.

  13. Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR

    NASA Astrophysics Data System (ADS)

    Gonzalez, Jeferson; Soto, Christian; Carvajal, Johan; Ribeiro, Celso

    2013-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5 T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the energy, vacuum, gas fueling, and security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

  14. Control and Data Acquisition for the Spherical Tokamak MEDUSA-CR

    NASA Astrophysics Data System (ADS)

    Soto, Christian; Gonzalez, Jeferson; Carvajal, Johan; Ribeiro, Celso

    2013-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5 T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the control and data acquisition systems for MEDUSA-CR device which are based on National Instruments (NI) software (LabView) and hardware on 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.

  15. Theoretical interpretation of frequency sweeping observations in the Mega-Amp Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    Vann, R. G. L.; Dendy, R. O.; Gryaznevich, M. P.

    2005-03-01

    Frequency sweeping (chirping) of high frequency magnetohydrodynamic modes is widely observed in tokamak plasmas. In this paper observations of chirping in neutral-beam-heated plasmas in the Mega-Amp Spherical Tokamak (MAST) [A. Sykes, R. J. Akers, L. C. Appel et al., Nucl. Fusion 41, 1423 (2001)] are considered, and it is shown that these may be interpreted using the Berk-Breizman augmentation of the Vlasov-Maxwell equations. This model includes an energetic particle source: it leads not only to a single chirp but also to a series of bursting events. This repetitious behavior is characteristic of the chirping seen in experiments such as MAST. The similarity between features in velocity space and features in frequency space reinforces the theory that hole-clump pair formation is responsible for the observed frequency sweeping.

  16. Natural Divertor Spherical Tokamak Plasmas with bean shape and ergodic limiter

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso; Herrera, Julio; Chavez, Esteban; Tritz, Kevin

    2013-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We report here improvements in the self-consistency of these equilibrium comparisons and a preliminary study of their MHD stability beta limits. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

  17. Two-fluid simulations of driven reconnection in the mega-ampere spherical tokamak

    SciTech Connect

    Stanier, A.; Browning, P.; Gordovskyy, M.; McClements, K. G.; Gryaznevich, M. P.

    2013-12-15

    In the merging-compression method of plasma start-up, two flux-ropes with parallel toroidal current are formed around in-vessel poloidal field coils, before merging to form a spherical tokamak plasma. This start-up method, used in the Mega-Ampere Spherical Tokamak (MAST), is studied as a high Lundquist number and low plasma-beta magnetic reconnection experiment. In this paper, 2D fluid simulations are presented of this merging process in order to understand the underlying physics, and better interpret the experimental data. These simulations examine the individual and combined effects of tight-aspect ratio geometry and two-fluid physics on the merging. The ideal self-driven flux-rope dynamics are coupled to the diffusion layer physics, resulting in a large range of phenomena. For resistive MHD simulations, the flux-ropes enter the sloshing regime for normalised resistivity η≲10{sup −5}. In Hall-MHD, three regimes are found for the qualitative behaviour of the current sheet, depending on the ratio of the current sheet width to the ion-sound radius. These are a stable collisional regime, an open X-point regime, and an intermediate regime that is highly unstable to tearing-type instabilities. In toroidal axisymmetric geometry, the final state after merging is a MAST-like spherical tokamak with nested flux-surfaces. It is also shown that the evolution of simulated 1D radial density profiles closely resembles the Thomson scattering electron density measurements in MAST. An intuitive explanation for the origin of the measured density structures is proposed, based upon the results of the toroidal Hall-MHD simulations.

  18. Scale-up of spherical tokamak solenoid-free startup by coaxial helicity injection

    SciTech Connect

    Tang, X. Z.; Boozer, A. H.

    2007-10-15

    Current multiplication and flux amplification are two critical measures in assessing the usefulness of magnetic helicity injection for forming the plasma confining magnetic field in laboratory spheromak and spherical tokamak (ST) experiments. While the two concepts are closely related for spheromaks, they are independent for a Taylor-relaxed ST plasma, and negatively correlated for a more realistic, partially relaxed ST plasma. An important application of this understanding leads to the so-called relaxed transient coaxial helicity injection scheme for solenoid-free ST startup, which can deliver reactor-grade high current multiplication and flux amplification.

  19. Coupled Ray-tracing and Fokker-Planck EBW Modeling for Spherical Tokamaks

    NASA Astrophysics Data System (ADS)

    Urban, J.; Decker, J.; Peysson, Y.; Preinhaelter, J.; Taylor, G.; Vahala, L.; Vahala, G.

    2009-11-01

    The AMR (Antenna—Mode-conversion—Ray-tracing) code [1, 2] has been recently coupled with the LUKE [3] Fokker-Planck code. This modeling suite is capable of complex simulations of electron Bernstein wave (EBW) emission, heating and current drive. We employ these codes to study EBW heating and current drive performance under spherical tokamak (ST) configurations—typical NSTX discharges are employed. EBW parameters, such as frequency, antenna position and direction, are varied and optimized for particular configurations and objectives. In this way, we show the versatility of EBWs.

  20. Distinct turbulence sources and confinement features in the spherical tokamak plasma regime

    SciTech Connect

    Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.

    2015-10-30

    New turbulence contributions to plasma transport and confinement in the spherical tokamak (ST) regime are identified through nonlinear gyrokinetic simulations. The drift wave Kelvin-Helmholtz (KH) mode characterized by intrinsic mode asymmetry is shown to drive significant ion thermal transport in strongly rotating national spherical torus experiment (NSTX) L-modes. The long wavelength, quasi-coherent dissipative trapped electron mode (TEM) is destabilized in NSTX H-modes despite the presence of strong ExB shear, providing a robust turbulence source dominant over collisionless TEM. Dissipative trapped electron mode (DTEM)-driven transport in the NSTX parametric regime is shown to increase with electron collision frequency, offering one possible source for the confinement scaling observed in experiments. There exists a turbulence-free regime in the collision-induced collisionless trapped electron mode to DTEM transition for ST plasmas. This predicts a natural access to a minimum transport state in the low collisionality regime that future advanced STs may cover.

  1. Modeling divertor concepts for spherical tokamaks NSTX-U and ST-FNSF

    NASA Astrophysics Data System (ADS)

    Meier, E. T.; Gerhardt, S.; Menard, J. E.; Rognlien, T. D.; Soukhanovskii, V. A.

    2015-08-01

    The compact nature of the spherical tokamak (ST) presents an economically attractive path to fusion commercialization, but concentrates power exhaust, threatening the integrity of plasma-facing components. To address this challenge, experimentally constrained divertor modeling in the National Spherical Torus Experiment (NSTX) is extrapolated to investigate divertor concepts for future ST devices. Analysis of NSTX Upgrade with UEDGE shows that the secondary snowflake X-point position can be adjusted for favorable neutral transport, enabling stable partial detachment at reduced core densities. For a notional ST-based Fusion Nuclear Science Facility, divertor concepts are identified that provide heat flux mitigation (<10 MW m-2) and low temperatures (<10 eV) compatible with high-Z targets. This research provides guidance for upcoming experiments and a basis for continued development of predictive capability for divertor performance in STs.

  2. Optimization of a mega-ampere spherical tokamak for beta-limit and confinement studies

    SciTech Connect

    McCool, S.C.; Wootton, A.J.; Bravenec, R.V.

    1994-10-01

    Recent favorable results on the START experiment have caused renewed interest in the potential of low aspect ratio tokamaks. To aid in designing a next-step spherical tokamak to study confinement scaling, high beta, and high normalized beta plasmas for minimal cost, the authors have developed a transport scaling and device optimization code. This code STOP, benchmarked against START, includes 10 empirical confinement scaling laws and essential tokamak physics such as stability limits. Parameters are optimized separately for each scaling law and physical goal. Using STOP the authors find for R/a = 1.2 to 2.0 one can achieve {beta}{sub N} = 5, ({beta}) = 31--44%, and easily study predicted confinement degradation with auxiliary heating with just two neutral beams (P{sub NB} < 3.5 MW) for I{sub p} {ge} 0.75 MA, and R{sub 0} {ge} 0.6 m. In contrast, if one insists on using the nominal device parameters, i.e. I{sub p} = 1 MA and R{sub 0} = 0.8 m, with each scaling law, achieving {beta}{sub N} = 5 requires typically P{sub NB} {approx} 7.5 MW. They also conclude that while running D{sup 0} {yields} D{sup +} increases {tau}{sub E} {approx} 25%, with {tau}{sub E} already uncertain by 2--3{times}, one incurs restricted machine access and compromised physics operation.

  3. Conceptual design study of the moderate size superconducting spherical tokamak power plant

    NASA Astrophysics Data System (ADS)

    Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

    2015-06-01

    A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

  4. The National Spherical Tokamak Experiment at the Princeton Plasma Physics Laboratory

    SciTech Connect

    1995-12-01

    The Department of Energy (DOE) has prepared an Environmental Assessment (EA), DOE/EA-1108, evaluating the environmental effects of the proposed construction and operation of the National Spherical Tokamak Experiment (NSTX) within the existing C-Stellarator (CS) Building at the Princeton Plasma Physics Laboratory, Princeton, New Jersey. The purpose of the NSTX is to investigate the physics of spherically shaped plasmas as an alternative path to conventional tokamaks for development of fusion energy. Fusion energy has the potential to help compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Construction of the NSTX in the CS Building would require the dismantling and removal of the existing unused Princeton Large Torus (PLT) device, part of which would be reused to construct the NSTX. Based on the analyses in the EA, the DOE has determined that the proposed action does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, 42 U.S.C. 4,321 et seq. The preparation of an Environmental Impact Statement is not required. Thus, the DOE is issuing a FONSI pursuant to the Council on Environmental Quality regulations implementing NEPA (40 CFR Parts 1500--1508) and the DOE NEPA implementing regulations (10 CFR Part 1021).

  5. Ideal MHD stability of a spherical tokamak power plant and a component test facility.

    SciTech Connect

    Hole, M. J.; Wilson, H. R.; Abeysuriya, R.; Larson, J. W.

    2010-12-01

    We have investigated ideal MHD stability of two advanced spherical tokamak confinement concepts: the spherical tokamak power plant (STPP), a 3 GW concept fusion power plasma producing 1 GW of electric power, and the component test facility (CTF), a concept designed for in situ materials testing for ITER and beyond. Detailed stability studies for toroidal mode number n = 1, 2, 3 displacements are presented as a function of conformal wall radius R{sub w} and on-axis safety factor q{sub 0}. For the STPP marginal stability scans held the current profile fixed, but varied the total plasma current. For the CTF we have extended and parallelized earlier marginal stability scans to scan over both the plasma beta and q{sub 0} by varying the current profile to preserve the total plasma current. These confirm that both concepts are stable provided that the wall is sufficiently close and q{sub 0} sufficiently large (q{sub 0} > 2.8 for the power plant and q{sub 0} > 2.1 for the CTF). Both power plant and CTF configurations are found to be ballooning stable.

  6. A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

    NASA Astrophysics Data System (ADS)

    Urban, Jakub; Decker, Joan; Peysson, Yves; Preinhaelter, Josef; Shevchenko, Vladimir; Taylor, Gary; Vahala, Linda; Vahala, George

    2011-08-01

    The electron Bernstein wave (EBW) is typically the only wave in the electron cyclotron (EC) range that can be applied in spherical tokamaks for heating and current drive (H&CD). Spherical tokamaks (STs) operate generally in high-β regimes, in which the usual EC O- and X-modes are cut off. In this case, EBWs seem to be the only option that can provide features similar to the EC waves—controllable localized H&CD that can be used for core plasma heating as well as for accurate plasma stabilization. The EBW is a quasi-electrostatic wave that can be excited by mode conversion from a suitably launched O- or X-mode; its propagation further inside the plasma is strongly influenced by the plasma parameters. These rather awkward properties make its application somewhat more difficult. In this paper we perform an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX). Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions, which are the fundamental EBW parameters that can be chosen and controlled. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

  7. Identification of new turbulence contributions to plasma transport and confinement in spherical tokamak regime

    SciTech Connect

    Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.

    2015-10-01

    Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E x B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transport that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around k(theta)rho(s) similar to 0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E x B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E x B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Moreover, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport

  8. Identification of new turbulence contributions to plasma transport and confinement in spherical tokamak regime

    NASA Astrophysics Data System (ADS)

    Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.

    2015-10-01

    Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E ×B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transport that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around kθρs˜0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E ×B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E ×B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Moreover, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport in advanced ST

  9. Identification of new turbulence contributions to plasma transport and confinement in spherical tokamak regime

    DOE PAGESBeta

    Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.

    2015-10-15

    Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E x B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transportmore » that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around kθρs ~ 0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E x B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E x B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Furthermore, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport in

  10. Identification of new turbulence contributions to plasma transport and confinement in spherical tokamak regime

    SciTech Connect

    Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.

    2015-10-15

    Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E x B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transport that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around kθρs ~ 0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E x B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E x B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Furthermore, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma

  11. High-beta spherical tokamak startup in TS-4 merging experiment by use of toroidal field ramp-up

    NASA Astrophysics Data System (ADS)

    Kaminou, Yasuhiro; , Toru, II; Kato, Joji; Inomoto, Michiaki; Ono, Yasushi; TS Group Team; National InstituteFusion Science Collaboration

    2014-10-01

    We demonstrated the formation method of an ultrahigh-beta spherical tokamak by use of a field-reversed configuration and a spheromak in TS-4 device (R ~ 0.5 m, A ~ 1.5, Ip ~ 30-100 kA, B ~ 100 mT). This method is composed of the following steps: 1. Two spheromaks are merged together and a high-beta spheromak or FRC is formed by reconnection heating. 2. External toroidal magnetic field is added (current rising time ~50 μs), and spherical tokamak-like configuration is formed. In this way, the ultrahigh-beta ST is formed. The ultrahigh-beta ST formed by FRC has a diamagnetic toroidal field, and it presumed to be in a second-stable state for ballooning stability, and the one formed by spheromak has a weak paramagnetic toroidal magnetic field, while a spheormak has a strong paramagnetic toroidal magnetic field. This diamagnetic current derives from inductive electric field by ramping up the external toroidal magnetic field, and the diamagnetic current sustains high thermal pressure of the ultrahigh-beta spherical tokamak. And the beta of the ultrahigh-beta ST formed by FRC reaches about 50%. To sustain the high-beta state, 0.6 MW neutral beam injection and center solenoid coils are installed to the TS-4 device. In the poster, we report the experimental results of ultrahigh-beta spherical tokamak startup and sustainment by NBI and CS current driving experiment.

  12. Reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak

    SciTech Connect

    Sakharov, N. V. Voronin, A. V.; Gusev, V. K.; Kavin, A. A.; Kamenshchikov, S. N.; Lobanov, K. M.; Minaev, V. B.; Novokhatsky, A. N.; Patrov, M. I. Petrov, Yu. V.; Shchegolev, P. B.

    2015-12-15

    The results of reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak by means of the EFIT code and by the method of movable filaments with the use of the data from magnetic measurements are compared. The EFIT code allows one to completely reconstruct the magnetic configuration by solving the Grad−Shafranov equation. In the method of movable filaments, the distribution of the toroidal current flowing through the plasma is described by a set of infinitely thin current-carrying rings. In this method, the last closed magnetic surface (LCMS) and the open surfaces lying beyond the LCMS are calculated. Using both methods, the coordinates of the regions where the separatrix strikes the divertor plates were determined. The results obtained agree well with the distributions of the temperature over the tungsten divertor tiles measured using an IR camera.

  13. Reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak

    NASA Astrophysics Data System (ADS)

    Sakharov, N. V.; Voronin, A. V.; Gusev, V. K.; Kavin, A. A.; Kamenshchikov, S. N.; Lobanov, K. M.; Minaev, V. B.; Novokhatsky, A. N.; Patrov, M. I.; Petrov, Yu. V.; Shchegolev, P. B.

    2015-12-01

    The results of reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak by means of the EFIT code and by the method of movable filaments with the use of the data from magnetic measurements are compared. The EFIT code allows one to completely reconstruct the magnetic configuration by solving the Grad-Shafranov equation. In the method of movable filaments, the distribution of the toroidal current flowing through the plasma is described by a set of infinitely thin current-carrying rings. In this method, the last closed magnetic surface (LCMS) and the open surfaces lying beyond the LCMS are calculated. Using both methods, the coordinates of the regions where the separatrix strikes the divertor plates were determined. The results obtained agree well with the distributions of the temperature over the tungsten divertor tiles measured using an IR camera.

  14. Noninductive plasma generation and current drive in the Globus-M spherical tokamak

    SciTech Connect

    D'yachenko, V. V.; Gusev, V. K.; Larionov, M. M.; Mel'nik, A. D.; Novokhatskii, A. N.; Petrov, Yu. V.; Rozhdestvenskii, V. V.; Sakharov, N. V.; Stepanov, A. Yu.; Khitrov, S. A.; Khromov, N. A.; Chernyshev, F. V.; Shevelev, A. E.; Shcherbinin, O. N.; Bender, S. E.; Kavin, A. A.; Lobanov, K. M.

    2013-03-15

    Experimental results on the generation and maintenance of the toroidal current in the Globus-M spherical tokamak by using waves in the lower hybrid frequency range without applying an inductive vortex electric field are presented. For this purpose, the original ridge guide antennas forming a field distribution similar to that produced by multiwaveguide grills were used. The high-frequency field (900 MHz) was used for both plasma generation and current drive. The magnitude of the generated current reached 21 kA, and its direction depended on the direction of the vertical magnetic field. Analysis of the experimental results indicates that the major fraction of the current is carried by the suprathermal electron beam.

  15. Simulation of current-filament dynamics and relaxation in the Pegasus Spherical Tokamak

    SciTech Connect

    O'Bryan, J. B.; Sovinec, C. R.; Bird, T. M.

    2012-08-15

    Nonlinear numerical computation is used to investigate the relaxation of non-axisymmetric current-channels from washer-gun plasma sources into 'tokamak-like' plasmas in the Pegasus toroidal experiment [Eidietis et al. J. Fusion Energy 26, 43 (2007)]. Resistive MHD simulations with the NIMROD code [Sovinec et al. Phys. Plasmas 10(5), 1727-1732 (2003)] utilize ohmic heating, temperature-dependent resistivity, and anisotropic, temperature-dependent thermal conduction corrected for regions of low magnetization to reproduce critical transport effects. Adjacent passes of the simulated current-channel attract and generate strong reversed current sheets that suggest magnetic reconnection. With sufficient injected current, adjacent passes merge periodically, releasing axisymmetric current rings from the driven channel. The current rings have not been previously observed in helicity injection for spherical tokamaks, and as such, provide a new phenomenological understanding for filament relaxation in Pegasus. After large-scale poloidal-field reversal, a hollow current profile and significant poloidal flux amplification accumulate over many reconnection cycles.

  16. Diamagnetic Fishbone Mode Associated with Circulating Fast Ions in Spherical Tokamaks

    SciTech Connect

    Ya.I. Kolesnichenko; V.S. Marchenko; R.B. White

    2001-06-19

    Recently it was shown theoretically that high beta (beta is the ratio of the plasma pressure to the magnetic field pressure) inherent to plasmas of Spherical Tokamaks (ST) stabilizes the fishbone mode associated with the trapped particles. This prediction agrees with the experimental observations of the fishbone behavior on the Small Tight Aspect Ratio Tokamak (START). However, in the mentioned experiments the circulating particles rather than the trapped ones were dominant in the energetic ion population. Therefore, the theory of Kolesnichenko, et al. in Phys. Rev. Lett. 82 (1999) 3260 and Nuclear Fusion 40 (2000) 1731 is not sufficient to explain the START experiment and predict the behavior of the circulating-particle-induced fishbone mode in future experiments on STs. Thus, a new theory is required, which stimulated the fulfillment of this present work. There are two fishbone branches: the high-frequency (precession) branch and the low-frequency (diamagnetic) one. In this work, we restrict ourselves with the study of the low-frequency branch. The stability of this branch associated with the circulating particles in a low-beta plasma was studied by Betti, et al. in Phys. Rev. Lett. 70 (1993) 3428; no attempts to consider high beta plasmas were done yet.

  17. Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe.

    PubMed

    Walkden, N R; Adamek, J; Allan, S; Dudson, B D; Elmore, S; Fishpool, G; Harrison, J; Kirk, A; Komm, M

    2015-02-01

    The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ∼1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the ER measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak. PMID:25725845

  18. Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe

    SciTech Connect

    Walkden, N. R.; Adamek, J.; Komm, M.; Allan, S.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Dudson, B. D.

    2015-02-15

    The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ∼1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the E{sub R} measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak.

  19. Fast neutral lithium beam probing of the edge region of the spherical tokamak ETE

    SciTech Connect

    Oliveira, R.M.; Ueda, M.; Berni, L.A.; Iguchi, H.

    2004-10-01

    The close relationship between the behavior of the edge plasma characteristics and the global confinement in discharges in tokamaks has been extensively explored in recent years. A 15 keV lithium beam with 80% neutralization efficiency, using a high ion current density emitter and a modified Pierce gun design for extraction of ions, was developed for edge studies on the spherical tokamak ETE--(major radius R=0.30 m, aspect ration A=1.5, designed plasma current of 200 kA with a torodial magnetic field up to 0.4 T). Already in the calibration phase this diagnostics confirmed to be an efficient experimental tool, allowing accurate measurements of the fast varying pressure of the discharges in ETE. The measurement of the temporal evolution of the density of the plasma edge with fast neutral lithium beam is in accordance with data from Langmuir probe for the region near the limiter and the one made by the Thomson scattering for the inner region.

  20. Parametric studies of next-step spherical tokamaks using high-temperature superconductors

    NASA Astrophysics Data System (ADS)

    Menard, Jonathan; Brown, Tom; El-Guebaly, Laila; Zhai, Yuhu

    2015-11-01

    High-temperature superconducting (HTS) magnets are potentially attractive for compact spherical tokamak (ST) applications due to higher operating temperature which could reduce thermal shielding requirements and reduce device size relative to configurations that utilize low-temperature superconductors (LTS). HTS conductors can also operate with very high current densities and high magnetic fields. Recent studies have shown that for elongation and kink stability dependencies on aspect ratio consistent with NSTX data, accounting for the engineering limits of HTS magnets, and having only a modest central solenoid, the optimal aspect ratio for an HTS tokamak pilot plant is between A = 1.7 and 2.3 depending on inboard shielding thickness. These results point to the interesting finding that the optimal aspect ratio for a compact HTS pilot plant may be near A = 2 which is an unexplored configuration in the present fusion program. The potential implications for required energy confinement, plasma formation and sustainment, power exhaust, and other performance parameters will be discussed. Work Supported by U.S.D.O.E. Contract No. DE-AC02-09CH11466.

  1. High kinetic energy plasma jet generation and its injection into the Globus-M spherical tokamak

    NASA Astrophysics Data System (ADS)

    Voronin, A. V.; Gusev, V. K.; Petrov, Yu. V.; Sakharov, N. V.; Abramova, K. B.; Sklyarova, E. M.; Tolstyakov, S. Yu.

    2005-09-01

    Progress in the theoretical and experimental development of the plasma jet source and injection of hydrogen plasma and neutral gas jets into the Globus-M spherical tokamak is discussed. An experimental test bed is described for investigation of intense plasma jets that are generated by a double-stage plasma gun consisting of an intense source for neutral gas production and a conventional pulsed coaxial accelerator. A procedure for optimizing the accelerator parameters so as to achieve the maximum possible flow velocity with a limited discharge current and a reasonable length of the coaxial electrodes is presented. The calculations are compared with experiment. Plasma jet parameters, among them pressure distribution across the jet, flow velocity, plasma density, etc, were measured. Plasma jets with densities of up to 1022 m-3, total numbers of accelerated particles (1-5) × 1019, and flow velocities of 50-100 km s-1 were successfully injected into the plasma column of the Globus-M tokamak. Interferometric and Thomson scattering measurements confirmed deep jet penetration and a fast density rise (<0.5 ms) at all spatial points up to a radius rap 0.3a. The plasma particle inventory increase by ~50% (from 0.65 × 1019 to 1 × 1019) did not result in plasma degradation.

  2. Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe

    NASA Astrophysics Data System (ADS)

    Walkden, N. R.; Adamek, J.; Allan, S.; Dudson, B. D.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Komm, M.

    2015-02-01

    The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ˜1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the ER measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak.

  3. Simulation of current-filament dynamics and relaxation in the Pegasus Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    O'Bryan, J. B.; Sovinec, C. R.; Bird, T. M.

    2012-08-01

    Nonlinear numerical computation is used to investigate the relaxation of non-axisymmetric current-channels from washer-gun plasma sources into "tokamak-like" plasmas in the Pegasus toroidal experiment [Eidietis et al. J. Fusion Energy 26, 43 (2007)]. Resistive MHD simulations with the NIMROD code [Sovinec et al. Phys. Plasmas 10(5), 1727-1732 (2003)] utilize ohmic heating, temperature-dependent resistivity, and anisotropic, temperature-dependent thermal conduction corrected for regions of low magnetization to reproduce critical transport effects. Adjacent passes of the simulated current-channel attract and generate strong reversed current sheets that suggest magnetic reconnection. With sufficient injected current, adjacent passes merge periodically, releasing axisymmetric current rings from the driven channel. The current rings have not been previously observed in helicity injection for spherical tokamaks, and as such, provide a new phenomenological understanding for filament relaxation in Pegasus. After large-scale poloidal-field reversal, a hollow current profile and significant poloidal flux amplification accumulate over many reconnection cycles.

  4. Distinct turbulence sources and confinement features in the spherical tokamak plasma regime

    DOE PAGESBeta

    Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.

    2015-10-30

    New turbulence contributions to plasma transport and confinement in the spherical tokamak (ST) regime are identified through nonlinear gyrokinetic simulations. The drift wave Kelvin-Helmholtz (KH) mode characterized by intrinsic mode asymmetry is shown to drive significant ion thermal transport in strongly rotating national spherical torus experiment (NSTX) L-modes. The long wavelength, quasi-coherent dissipative trapped electron mode (TEM) is destabilized in NSTX H-modes despite the presence of strong E x B shear, providing a robust turbulence source dominant over collisionless TEM. Dissipative trapped electron mode (DTEM)-driven transport in the NSTX parametric regime is shown to increase with electron collision frequency, offeringmore » one possible source for the confinement scaling observed in experiments. There exists a turbulence-free regime in the collision-induced collisionless trapped electron mode to DTEM transition for ST plasmas. In conclusion, this predicts a natural access to a minimum transport state in the low collisionality regime that future advanced STs may cover.« less

  5. Distinct turbulence sources and confinement features in the spherical tokamak plasma regime

    SciTech Connect

    Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.

    2015-10-30

    New turbulence contributions to plasma transport and confinement in the spherical tokamak (ST) regime are identified through nonlinear gyrokinetic simulations. The drift wave Kelvin-Helmholtz (KH) mode characterized by intrinsic mode asymmetry is shown to drive significant ion thermal transport in strongly rotating national spherical torus experiment (NSTX) L-modes. The long wavelength, quasi-coherent dissipative trapped electron mode (TEM) is destabilized in NSTX H-modes despite the presence of strong E x B shear, providing a robust turbulence source dominant over collisionless TEM. Dissipative trapped electron mode (DTEM)-driven transport in the NSTX parametric regime is shown to increase with electron collision frequency, offering one possible source for the confinement scaling observed in experiments. There exists a turbulence-free regime in the collision-induced collisionless trapped electron mode to DTEM transition for ST plasmas. In conclusion, this predicts a natural access to a minimum transport state in the low collisionality regime that future advanced STs may cover.

  6. Measurement of eddy-current distribution in the vacuum vessel of the Sino-UNIted Spherical Tokamak.

    PubMed

    Li, G; Tan, Y; Liu, Y Q

    2015-08-01

    Eddy currents have an important effect on tokamak plasma equilibrium and control of magneto hydrodynamic activity. The vacuum vessel of the Sino-UNIted Spherical Tokamak is separated into two hemispherical sections by a toroidal insulating barrier. Consequently, the characteristics of eddy currents are more complex than those found in a standard tokamak. Thus, it is necessary to measure and analyze the eddy-current distribution. In this study, we propose an experimental method for measuring the eddy-current distribution in a vacuum vessel. By placing a flexible printed circuit board with magnetic probes onto the external surface of the vacuum vessel to measure the magnetic field parallel to the surface and then subtracting the magnetic field generated by the vertical-field coils, the magnetic field due to the eddy current can be obtained, and its distribution can be determined. We successfully applied this method to the Sino-UNIted Spherical Tokamak, and thus, we obtained the eddy-current distribution despite the presence of the magnetic field generated by the external coils. PMID:26329187

  7. Measurement of eddy-current distribution in the vacuum vessel of the Sino-UNIted Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    Li, G.; Tan, Y.; Liu, Y. Q.

    2015-08-01

    Eddy currents have an important effect on tokamak plasma equilibrium and control of magneto hydrodynamic activity. The vacuum vessel of the Sino-UNIted Spherical Tokamak is separated into two hemispherical sections by a toroidal insulating barrier. Consequently, the characteristics of eddy currents are more complex than those found in a standard tokamak. Thus, it is necessary to measure and analyze the eddy-current distribution. In this study, we propose an experimental method for measuring the eddy-current distribution in a vacuum vessel. By placing a flexible printed circuit board with magnetic probes onto the external surface of the vacuum vessel to measure the magnetic field parallel to the surface and then subtracting the magnetic field generated by the vertical-field coils, the magnetic field due to the eddy current can be obtained, and its distribution can be determined. We successfully applied this method to the Sino-UNIted Spherical Tokamak, and thus, we obtained the eddy-current distribution despite the presence of the magnetic field generated by the external coils.

  8. Collisional Damping of Electron Bernstein Waves and its Mitigation by Evaporated Lithium Conditioning in Spherical-Tokamak Plasmas

    SciTech Connect

    Diem, S. J.; Caughman, J. B.; Taylor, G.; Efthimion, P. C.; Kugel, H.; LeBlanc, B. P.; Phillips, C. K.; Preinhaelter, J.; Urban, J.

    2009-07-03

    The first experimental verification of electron Bernstein wave (EBW) collisional damping, and its mitigation by evaporated Li conditioning, in an overdense spherical-tokamak plasma has been observed in the National Spherical Torus Experiment (NSTX). Initial measurements of EBW emission, coupled from NSTX plasmas via double-mode conversion to O-mode waves, exhibited <10% transmission efficiencies. Simulations show 80% of the EBW energy is dissipated by collisions in the edge plasma. Li conditioning reduced the edge collision frequency by a factor of 3 and increased the fundamental EBW transmission to 60%.

  9. Dynamics of the ballooning mode and the relation to edge-localized modes in a spherical tokamak

    SciTech Connect

    Khan, R.; Mizuguchi, N.; Nakajima, N.; Hayashi, T.

    2007-06-15

    Nonlinear simulations based on the magnetohydrodynamic model have been executed to reveal the dynamics of the ballooning mode in the spherical tokamak plasma. The simulation results have reproduced the characteristic features of the edge-localized mode crash phase, where the filamentary structures are formed along the magnetic field in the edge region, and separated from the core plasma. Moreover, the finite Larmor radius effect is addressed.

  10. Review of ICR Heating on the Spherical Tokamak Globus-M

    NASA Astrophysics Data System (ADS)

    Dyachenko, V. V.; Ayushin, B. B.; Gusev, V. K.; Khitrov, S. A.; Chernyshev, F. V.; Mironov, M. I.; Petrov, Yu. V.; Sakharov, N. V.; Shcherbinin, O. N.

    2009-11-01

    The results of experiments on the RF heating of hydrogen-deuterium plasma at ion cyclotron frequencies in the Globus-M spherical tokamak are presented. RF power up to 200 kW at frequencies about 7.5-9 MHz fed via a single-loop antenna led to an increase in the ion temperature from 180 eV up to 400 eV in the best case. High heating efficiency is maintained at high H-concentration (up to 70%). Characteristic times of the ion temperature buildup and decay corresponded to the ion energy confinement time. If the second hydrogen harmonic is arranged in front of antenna, on-axis ion heating becomes worse. It is shown, that in Globus conditions graphite shielding of an inner surface of a vacuum chamber decreases heating due to surface absorption of RF power. Simulation of processes of RF power absorption is performed also. For modeling 1-D wave code was used including all possible mechanisms of RF power absorption.

  11. Electron and Ion Heating Characteristics during Magnetic Reconnection in the MAST Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    Tanabe, H.; Yamada, T.; Watanabe, T.; Gi, K.; Kadowaki, K.; Inomoto, M.; Imazawa, R.; Gryaznevich, M.; Michael, C.; Crowley, B.; Conway, N. J.; Scannell, R.; Harrison, J.; Fitzgerald, I.; Meakins, A.; Hawkes, N.; McClements, K. G.; O'Gorman, T.; Cheng, C. Z.; Ono, Y.

    2015-11-01

    Electron and ion heating characteristics during merging reconnection start-up on the MAST spherical tokamak have been revealed in detail using a 130 channel yttrium aluminum garnet (YAG) and a 300 channel Ruby-Thomson scattering system and a new 32 chord ion Doppler tomography diagnostic. Detailed 2D profile measurements of electron and ion temperature together with electron density have been achieved for the first time and it is found that electron temperature forms a highly localized hot spot at the X point and ion temperature globally increases downstream. For the push merging experiment when the guide field is more than 3 times the reconnecting field, a thick layer of a closed flux surface form by the reconnected field sustains the temperature profile for longer than the electron and ion energy relaxation time ˜4 - 10 ms , both characteristic profiles finally forming a triple peak structure at the X point and downstream. An increase in the toroidal guide field results in a more peaked electron temperature profile at the X point, and also produces higher ion temperatures at this point, but the ion temperature profile in the downstream region is unaffected.

  12. Electron and Ion Heating Characteristics during Magnetic Reconnection in the MAST Spherical Tokamak.

    PubMed

    Tanabe, H; Yamada, T; Watanabe, T; Gi, K; Kadowaki, K; Inomoto, M; Imazawa, R; Gryaznevich, M; Michael, C; Crowley, B; Conway, N J; Scannell, R; Harrison, J; Fitzgerald, I; Meakins, A; Hawkes, N; McClements, K G; O'Gorman, T; Cheng, C Z; Ono, Y

    2015-11-20

    Electron and ion heating characteristics during merging reconnection start-up on the MAST spherical tokamak have been revealed in detail using a 130 channel yttrium aluminum garnet (YAG) and a 300 channel Ruby-Thomson scattering system and a new 32 chord ion Doppler tomography diagnostic. Detailed 2D profile measurements of electron and ion temperature together with electron density have been achieved for the first time and it is found that electron temperature forms a highly localized hot spot at the X point and ion temperature globally increases downstream. For the push merging experiment when the guide field is more than 3 times the reconnecting field, a thick layer of a closed flux surface form by the reconnected field sustains the temperature profile for longer than the electron and ion energy relaxation time ~4-10 ms, both characteristic profiles finally forming a triple peak structure at the X point and downstream. An increase in the toroidal guide field results in a more peaked electron temperature profile at the X point, and also produces higher ion temperatures at this point, but the ion temperature profile in the downstream region is unaffected. PMID:26636857

  13. Toroidal ripple transport of beam ions in the mega-ampere spherical tokamak

    SciTech Connect

    McClements, K. G.

    2012-07-15

    The transport of injected beam ions due to toroidal magnetic field ripple in the mega-ampere spherical tokamak (MAST) is quantified using a full orbit particle tracking code, with collisional slowing-down and pitch-angle scattering by electrons and bulk ions taken into account. It is shown that the level of ripple losses is generally rather low, although it depends sensitively on the major radius of the outer midplane plasma edge; for typical values of this parameter in MAST plasmas, the reduction in beam heating power due specifically to ripple transport is less than 1%, and the ripple contribution to beam ion diffusivity is of the order of 0.1 m{sup 2} s{sup -1} or less. It is concluded that ripple effects make only a small contribution to anomalous transport rates that have been invoked to account for measured neutron rates and plasma stored energies in some MAST discharges. Delayed (non-prompt) losses are shown to occur close to the outer midplane, suggesting that banana-drift diffusion is the most likely cause of the ripple-induced losses.

  14. A high resolution Mirnov array for the Mega Ampere Spherical Tokamak

    SciTech Connect

    Hole, M. J.; Appel, L. C.; Martin, R.

    2009-12-15

    Over the past two decades, the increase in neutral-beam heating and {alpha} particle production in magnetically confined fusion plasmas has led to an increase in energetic particle driven mode activity, much of which has an electromagnetic signature which can be detected by the use of external Mirnov coils. Typically, the frequency and spatial wave number band of such oscillations increase with increasing injection energy, offering new challenges for diagnostic design. In particular, as the frequency approaches the megahertz range, care must be taken to model the stray capacitance of the coil, which limits the resonant frequency of the probe; model transmission line effects in the system, which if unchecked can produce system resonances; and minimize coil conductive shielding, so as to minimize skin currents which limit the frequency response of the coil. As well as optimizing the frequency response, the coils should also be positioned to confidently identify oscillations over a wide wave number band. This work, which draws on new techniques in stray capacitance modeling and coil positioning, is a case study of the outboard Mirnov array for high-frequency acquisition in the Mega Ampere Spherical Tokamak, and is intended as a roadmap for the design of high frequency, weak field strength magnetic diagnostics.

  15. Characteristics of a novel lower hybrid wave antenna for the TST-2 spherical tokamak

    SciTech Connect

    Takase, Y.; Shinya, T.; Wakatsuki, T.; Ejiri, A.; Furui, H.; Hiratsuka, J.; Imamura, K.; Inada, T.; Kakuda, H.; Nakamura, K.; Nakanishi, A.; Oosako, T.; Sonehara, M.; Togashi, H.; Tsuda, S.; Tsujii, N.; Yamaguchi, T.; Moeller, C. P.

    2014-02-12

    A new type of traveling wave antenna which excites the lower hybrid wave directly was developed. This antenna is similar to the inductively-coupled combline antenna in that only the first element of the antenna array is excited externally, and subsequent elements are excited passively by mutual coupling between adjacent elements. The main difference is that whereas the inductively-coupled combline antenna makes use of mutual inductance, the presently proposed antenna makes use of mutual capacitance. The radiating elements are located at the voltage maximum, and the electric field induced in the plasma is in the toroidal direction rather than the poloidal direction, matching the polarization of the lower hybrid wave. Optimization studies were carried out to obtain a band-pass characteristic centered around 200 MHz, and a unidirectional wavenumber spectrum with the parallel index of refraction corresponding to approximately 5. Plasma current ramp-up to 2 kA has been achieved on the TST-2 spherical tokamak with 12 kW of RF power at 200 MHz during the initial experimental period using this antenna. Further optimization studies are being performed.

  16. Linear stability and nonlinear dynamics of the fishbone mode in spherical tokamaks

    SciTech Connect

    Wang, Feng; Liu, J. Y.; Fu, G. Y.; Breslau, J. A.

    2013-10-15

    Extensive linear and nonlinear simulations have been carried out to investigate the energetic particle-driven fishbone instability in spherical tokamak plasmas with weakly reversed q profile and the q{sub min} slightly above unity. The global kinetic-MHD hybrid code M3D-K is used. Numerical results show that a fishbone instability is excited by energetic beam ions preferentially at higher q{sub min} values, consistent with the observed appearance of the fishbone before the “long-lived mode” in MAST and NSTX experiments. In contrast, at lower q{sub min} values, the fishbone tends to be stable. In this case, the beam ion effects are strongly stabilizing for the non-resonant kink mode. Nonlinear simulations show that the fishbone saturates with strong downward frequency chirping as well as radial flattening of the beam ion distribution. An (m, n) = (2, 1) magnetic island is found to be driven nonlinearly by the fishbone instability, which could provide a trigger for the (2, 1) neoclassical tearing mode sometimes observed after the fishbone instability in NSTX.

  17. Plasma current start-up using the lower hybrid wave on the TST-2 spherical tokamak

    NASA Astrophysics Data System (ADS)

    Takase, Y.; Ejiri, A.; Inada, T.; Moeller, C. P.; Shinya, T.; Tsujii, N.; Yajima, S.; Furui, H.; Homma, H.; Imamura, K.; Nakamura, K.; Nakamura, K.; Sonehara, M.; Takeuchi, T.; Togashi, H.; Tsuda, S.; Yoshida, Y.

    2015-12-01

    Non-inductive plasma current start-up, ramp-up and sustainment by waves in the lower hybrid wave (LHW) frequency range at 200 MHz were investigated on the TST-2 spherical tokamak (R0 ≤ 0.38 m, a ≤ 0.25 m, Bt0 ≤ 0.3T, Ip ≤ 0.14 MA). Experimental results obtained using three types of antenna were compared. Both the highest plasma current (Ip = 18 kA) and the highest current drive figure of merit ηCD≡n¯eIpR0/PRF=1.4 ×1017 A/W/m2 were achieved using the capacitively-coupled combline (CCC) antenna, designed to excite the LHW with a sharp and highly directional wavenumber spectrum. For Ip greater than about 5 kA, high energy electrons accelerated by the LHW become the dominant carrier of plasma current. The low value of ηCD observed so far are believed to be caused by a rapid loss of energetic electrons and parasitic losses of the LHW energy in the plasma periphery. ηCD is expected to improve by an order of magnitude by increasing the plasma current to improve energetic electron confinement. In addition, edge power losses are expected to be reduced by increasing the toroidal magnetic field to improve wave accessibility to the plasma core, and by launching the LHW from the inboard upper region of the torus to achieve better single-pass absorption.

  18. Experimental studies of neoclassical tearing modes on the MAST spherical tokamak

    NASA Astrophysics Data System (ADS)

    Snape, Jack

    Neoclassical tearing modes (NTMs) are plasma instabilities that can limit the performance of tokamaks and cause a termination of the plasma if allowed to grow. Systems to mitigate NTMs exist but have significant power requirements, which motivates further study of the mechanisms that lead to their growth in order to assist in the development of NTM avoidance strategies. NTMs typically require a seed magnetic island, above some threshold width, before they become unstable. The best available description of this threshold is the modified Rutherford equation (MRE) for NTM evolution; a combination of different models, which includes the effect of transport on NTM stability. Finite transport across magnetic field lines means that magnetic islands smaller than a critical width, w_c, do not completely flatten the pressure profiles and have a reduced bootstrap current perturbation, which leads to a threshold width, w_th. This thesis describes novel measurements of NTMs with mode structure m/n=2/1 on the MAST spherical tokamak (ST), which have allowed a direct evaluation of the effect of transport on island behaviour for the first time on an ST. Temperature profiles obtained with the upgraded Thomson scattering system on MAST have been used to constrain the solutions of a heat transport equation for a magnetic island, allowing the experimental determination of w_c, an important parameter in the MRE. The measured value of w_c = 0.7pm 0.2cm obtained for an ensemble of MAST discharges is used in an analysis of the MRE for 2/1 NTM onset and saturation on MAST. By using a probabilistic method for parameter and error estimation, which takes account of the experimental uncertainty on measured equilibrium parameters, it is found that the temporal evolution of the island size is well described by marginally, classically unstable NTMs (that is, Delta'>0) with strongly destabilising bootstrap current and stabilising curvature terms. Finally, an analysis of two beta ramp

  19. Compact, battery powered, wireless digitizers for in situ data acquisitions in the sino-united spherical tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Yangqing; Tan, Yi; Ke, Rui; Yang, Hao; Wang, Wenhao; Gao, Zhe

    2015-07-01

    Potential isolation and long cable drive are very important in acquiring certain signals from tokamak diagnostics. Compact, battery powered, wireless digitizers for in situ data acquisition have been developed and routinely used in the sino-united spherical tokamak to solve the problems of isolation and long cables. The wireless digitizers utilize the integrated analog to digital converters and the static random access memory of microcontrollers but transfer data wirelessly. They consist of simple and concise circuits but have considerable performances of 12-16 bit in resolution and 500-1000 kS/s in sample rate. Wireless triggering and energy saving are two major challenges of the wireless digitizers. Wireless transceivers in the data link layer are used as trigger and can reduce the trigger jitters to be smaller than 1 μs. In order to reduce the energy consumption, the wireless digitizers are waken only when the tokamak is about to discharge. After discharges, they turn to a periodic checking mode with current consumption smaller than 200 μA. Because of low duty cycle, the wireless digitizers have a battery life of up to four weeks. In general, the wireless digitizers have better performance than normal isolation amplifiers and can greatly simplify the cable connections. They are very suitable for the data acquisition of dangerous and/or susceptible analog signals in tokamaks.

  20. Compact, battery powered, wireless digitizers for in situ data acquisitions in the sino-united spherical tokamak.

    PubMed

    Liu, Yangqing; Tan, Yi; Ke, Rui; Yang, Hao; Wang, Wenhao; Gao, Zhe

    2015-07-01

    Potential isolation and long cable drive are very important in acquiring certain signals from tokamak diagnostics. Compact, battery powered, wireless digitizers for in situ data acquisition have been developed and routinely used in the sino-united spherical tokamak to solve the problems of isolation and long cables. The wireless digitizers utilize the integrated analog to digital converters and the static random access memory of microcontrollers but transfer data wirelessly. They consist of simple and concise circuits but have considerable performances of 12-16 bit in resolution and 500-1000 kS/s in sample rate. Wireless triggering and energy saving are two major challenges of the wireless digitizers. Wireless transceivers in the data link layer are used as trigger and can reduce the trigger jitters to be smaller than 1 μs. In order to reduce the energy consumption, the wireless digitizers are waken only when the tokamak is about to discharge. After discharges, they turn to a periodic checking mode with current consumption smaller than 200 μA. Because of low duty cycle, the wireless digitizers have a battery life of up to four weeks. In general, the wireless digitizers have better performance than normal isolation amplifiers and can greatly simplify the cable connections. They are very suitable for the data acquisition of dangerous and/or susceptible analog signals in tokamaks. PMID:26233380

  1. Correction of the axial asymmetry of the poloidal magnetic field in the Globus-M spherical tokamak

    SciTech Connect

    Petrov, Yu. V.; Patrov, M. I.; Varfolomeev, V. I.; Gusev, V. K.; Lamzin, E. A.; Sakharov, N. V.; Sychevskii, S. E.

    2010-06-15

    The toroidal inhomogeneity of the poloidal magnetic field-the so-called error fields that arise due to imperfections in manufacturing and assembling of the electromagnetic system-was measured in the Globus-M spherical tokamak. A substantial inhomogeneity corresponding to the n = 1 mode, which gave rise to a locked mode and led to discharge disruption, was revealed. After compensation of this inhomogeneity with the help of special correction coils, the discharge duration increased and the global plasma parameters improved substantially. A technique for determining and compensating the n = 1 mode inhomogeneity is described, the measured dependences of the penetration threshold of the m = 2/n = 1 mode on the plasma parameters are given, and results of experiments in which record parameters for the Globus-M tokamak were achieved after correction of the poloidal magnetic field are presented.

  2. Characteristics of High Poloidal Beta (βp) Plasma Formed by Electron Cyclotron Waves in Spherical Tokamak QUEST

    NASA Astrophysics Data System (ADS)

    Mishra, Kishore; Idei, Hiroshi; Zushi, Hideki; Tashima, Saya; Banerjee, Santanu; Hasegawa, Makoto; Hanada, Kazuaki; Nakamura, Kazuo; Fujisawa, Akihide; Matsuoka, Keisuke; Nagashima, Yoshihiko; Kawasaki, S.; Higashijima, A.; Nakashima, H.

    In spherical tokamak QUEST, plasma with high poloidal beta (βp) is obtained by injecting Electron Cyclotron Waves (ECW) into the Ohmic target plasma. With high βp, the plasma shape is transformed from an inboard limiter configuration to a natural divertor with the appearance of an poloidal null at the inboard side. By applying high vertical magnetic field (Bz), this high βp plasma is sustained. With suitable control of the equilibrium, effective current drive is observed in the later part of the discharge along with strong recharging of the Ohmic circuit and the Ip is sustained for > 1 s.

  3. Transport and confinement in the Mega Ampère Spherical Tokamak (MAST) plasma

    NASA Astrophysics Data System (ADS)

    Akers, R. J.; Ahn, J. W.; Antar, G. Y.; Appel, L. C.; Applegate, D.; Brickley, C.; Bunting, C.; Carolan, P. G.; Challis, C. D.; Conway, N. J.; Counsell, G. F.; Dendy, R. O.; Dudson, B.; Field, A. R.; Kirk, A.; Lloyd, B.; Meyer, H. F.; Morris, A. W.; Patel, A.; Roach, C. M.; Rohzansky, V.; Sykes, A.; Taylor, D.; Tournianski, M. R.; Valovi, M.; Wilson, H. R.; Axon, K. B.; Buttery, R. J.; Ciric, D.; Cunningham, G.; Dowling, J.; Dunstan, M. R.; Gee, S. J.; Gryaznevich, M. P.; Helander, P.; Keeling, D. L.; Knight, P. J.; Lott, F.; Loughlin, M. J.; Manhood, S. J.; Martin, R.; McArdle, G. J.; Price, M. N.; Stammers, K.; Storrs, J.; Walsh, M. J.; MAST, the; NBI Team

    2003-12-01

    A combination of recently installed state-of-the-art imaging and profile diagnostics, together with established plasma simulation codes, are providing for the first time on Mega Ampère Spherical Tokamak (MAST) the tools required for studying confinement and transport, from the core through to the plasma edge and scrape-off-layer (SOL). The H-mode edge transport barrier is now routinely turned on and off using a combination of poloidally localized fuelling and fine balancing of the X-points. Theory, supported by experiment, indicates that the edge radial electric field and toroidal flow velocity (thought to play an important role in H-mode access) are largest if gas fuelling is concentrated at the inboard side. H-mode plasmas show predominantly type III ELM characteristics, with confinement HH factor (w.r.t. scaling law IPB98[y, 2]) around ~1.0. Combining MAST H-mode data with the International Tokamak Physics Activities (ITPA) analyses, results in an L H power threshold scaling proportional to plasma surface area (rather than PLH ~ R2). In addition, MAST favours an inverse aspect ratio scaling PLH ~ egr0.5. Similarly, the introduction of type III ELMing H-mode data to the pedestal energy regression analysis introduces a scaling Wped ~ egr-2.13 and modifies the exponents on R, BT and kgr. Preliminary TRANSP simulations indicate that ion and electron thermal diffusivities in ELMing H-mode approach the ion-neoclassical level in the half-radius region of the plasma with momentum diffusivity a few times lower. Linear flux-tube ITG and ETG microstability calculations using GS2 offer explanations for the near-neoclassical ion diffusivity and significantly anomalous electron diffusivity seen on MAST. To complement the baseline quasi-steady-state H-mode, newly developed advanced regimes are being explored. In particular, 'broad' internal transport barriers (ITBs) have been formed using techniques developed at conventional aspect ratio. Electron and ion energy diffusivities

  4. Experimental Study of Parametric Dependence of Electron-scale Turbulence in a Spherical Tokamak

    SciTech Connect

    Ren, Y; Kaye, S M; Mazzucato, E; Bell, R E; Diallo, A; Domier, C W; LeBlanc, B P; Lee, K C; Smith, D R

    2012-05-23

    Electron-scale turbulence is predicted to drive anomalous electron thermal transport. However, experimental study of its relation with transport is still in its early stage. On the National Spherical Tokamak eXperiment (NSTX), electron-scale density fluctuations are studied with a novel tangential microwave scattering system with high radial resolution of ±2 cm. Here, we report a study of parametric dependence of electron-scale turbulence in NSTX H-mode plasmas. The dependence on density gradient is studied through the observation of a large density gradient variation in the core induced by an ELM event, where we found the first clear experimental evidence of density gradient stabilization of electron-gyro scale turbulence in a fusion plasma. This observation, cou- pled with linear gyro-kinetic calculations, leads to the identification of the observed instability as toroidal Electron Temperature Gradient (ETG) modes. It is observed that longer wavelength ETG modes, k⊥ρs < 10 (ρs is the ion gyroradius at electron temperature and k⊥ is the wavenumber perpendicular to local equilibrium magnetic field), are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in electron thermal diffusivity. Comparisons with nonlinear ETG gyrokinetic simulations shows ETG turbulence may be able to explain the experimental electron heat flux observed before the ELM event. The collisionality dependence of electron-scale turbulence is also studied by systematically varying plasma current and toroidal fleld, so that electron gyroradius (ρe ), electron beta (βe ) and safety factor (q95 ) are kept approximately constant. More than a factor of two change in electron collisionality, ν *e, was achieved, and we found that the spectral power of electron-scale turbulence appears to increase as ν *e is decreased in this collisonality scan. However, both linear and nonlinear simulations show no or weak dependence with the electron

  5. Collective fast ion instability-induced losses in National Spherical Tokamak Experiment

    SciTech Connect

    Fredrickson, E.D.; Bell, R.E.; Darrow, D.S.; Fu, G.Y.; Gorelenkov, N.N.; LeBlanc, B.P.; Medley, S.S.; Menard, J.E.; Park, H.; Roquemore, A.L.; Heidbrink, W.W.; Sabbagh, S.A.; Stutman, D.; Tritz, K.; Crocker, N.A.; Kubota, S.; Peebles, W.; Lee, K.C.; Levinton, F.M.

    2006-05-15

    A wide variety of fast ion driven instabilities are excited during neutral beam injection (NBI) in the National Spherical Torus Experiment (NSTX) [Nucl. Fusion 40, 557 (2000)] due to the large ratio of fast ion velocity to Alfven velocity, V{sub fast}/V{sub Alfven}, and high fast ion beta. The ratio V{sub fast}/V{sub Alfven} in ITER [Nucl. Fusion 39, 2137 (1999)] and NSTX is comparable. The modes can be divided into three categories: chirping energetic particle modes (EPM) in the frequency range 0 to 120 kHz, the toroidal Alfven eigenmodes (TAE) with a frequency range of 50 kHz to 200 kHz, and the compressional and global Alfven eigenmodes (CAE and GAE, respectively) between 300 kHz and the ion cyclotron frequency. Fast ion driven modes are of particular interest because of their potential to cause substantial fast ion losses. In all regimes of NBI heated operation we see transient neutron rate drops, correlated with bursts of TAE or fishbone-like EPMs. The fast ion loss events are predominantly correlated with the EPMs, although losses are also seen with bursts of multiple, large amplitude TAE. The latter is of particular significance for ITER; the transport of fast ions from the expected resonance overlap in phase space of a 'sea' of large amplitude TAE is the kind of physics expected in ITER. The internal structure and amplitude of the TAE and EPMs has been measured with quadrature reflectometry and soft x-ray cameras. The TAE bursts have internal amplitudes of n-tilde/n=1% and toroidal mode numbers 2tokamaks. Unlike the fishbones, the EPMs can be present with q(0)>1 and can have a toroidal mode number n>1. The range of the frequency chirp can be quite large and the resonance can be through a fishbone-like precessional drift resonance, or through a bounce resonance.

  6. High Power Heating of Magnetic Reconnection in UTokyo Spherical Tokamak Merging Experiment: TS-U

    NASA Astrophysics Data System (ADS)

    Ono, Y.; Kawanami, M.; Kimura, K.; Nakai, R.; Nishida, K.; Ishida, R.; Yamanaka, H.; Kuwahata, A.; Tanabe, H.; Inomoto, M.; Cheng, C. Z.; TS; UTST Team

    2015-11-01

    Significant ion heating of magnetic reconnection up to 0.2keV and 1.2keV were documented in two tokamak merging experiments: TS-3 and MAST, leading us to a new high-field merging experiment: TS-U in University of Tokyo. 1D and 2D contours of ion and electron temperatures measured in TS-3 already revealed clear energy-conversion of magnetic reconnection: huge outflow heating of ions in the downstream and electron heating localized at the X-point. It is noted that the ion heating energy is proportional to square of the reconnecting (poloidal) magnetic field Brec. It is because the reconnection outflow accelerates ions up to the poloidal Alfven speed. The accelerated ions are thermalized by shock-like density pileups in the downstreams. These results agree qualitatively with recent solar satellite observations and PIC simulation results. Based on those results, our poster will show the design of upscaled high-field tokamak merging experiment: TS-U. The high-power heating of tokamak merging is useful not only for laboratory study of reconnection heating mechanisms but also for economical startup and heating of tokamak plasmas. The tokamak merging with Brec>0.3T will enables us to heat the tokamak plasma to the burning regime: Ti>5keV without using any additional heating facility.

  7. Pioneering experiments on atomic-beam-assisted generation of drag currents in the Globus-M spherical tokamak

    NASA Astrophysics Data System (ADS)

    Shchegolev, P. B.; Bakharev, N. N.; Gusev, V. K.; Kurskiev, G. S.; Minaev, V. B.; Patrov, M. I.; Petrov, Yu. V.; Sakharov, N. V.

    2015-09-01

    Research data for drag currents in the Globus-M spherical tokamak are presented. The currents are generated by injecting atomic beams of hydrogen and deuterium. Experiments were carried out in the hydrogen and deuterium plasma of the tokamak. It has a divertor configuration with a lower X-point, a displacement along the larger radius from-1.0 to-2.5 cm, and a toroidal field of 0.4 T at a plasma current of 0.17-0.23 MA. The beam is injected into the tokamak in the equatorial plane tangentially to the magnetic axis of the plasma filament with an impact diameter of 32 cm. To provide a 28-keV 0.5-MW atomic beam with geometrical sizes of 4 × 20 cm (at a power level of 1/ e), an IPM-2 ion source is used. The generation of noninductive currents is detected from a rise in the loop current and a simultaneous dip of the loop voltage. The injection of the hydrogen and deuterium atomic beams into the deuterium plasma results in a noticeable and reproducible dip of the loop voltage (up to 0.5 V). Using the ASTRA transport code, a model is constructed that allows rapid calculation of noninductive currents. Calculations performed for a specific discharge confirm that the model adequately describes the effect of drag current generation.

  8. Two-fluid and magnetohydrodynamic modelling of magnetic reconnection in the MAST spherical tokamak and the solar corona

    NASA Astrophysics Data System (ADS)

    Browning, P. K.; Cardnell, S.; Evans, M.; Arese Lucini, F.; Lukin, V. S.; McClements, K. G.; Stanier, A.

    2016-01-01

    Twisted magnetic flux ropes are ubiquitous in laboratory and astrophysical plasmas, and the merging of such flux ropes through magnetic reconnection is an important mechanism for restructuring magnetic fields and releasing free magnetic energy. The merging-compression scenario is one possible start-up scheme for spherical tokamaks, which has been used on the Mega Amp Spherical Tokamak (MAST). Two current-carrying plasma rings or flux ropes approach each due to mutual attraction, forming a current sheet and subsequently merge through magnetic reconnection into a single plasma torus, with substantial plasma heating. Two-dimensional resistive and Hall-magnetohydrodynamic simulations of this process are reported, including a strong guide field. A model of the merging based on helicity-conserving relaxation to a minimum energy state is also presented, extending previous work to tight-aspect-ratio toroidal geometry. This model leads to a prediction of the final state of the merging, in good agreement with simulations and experiment, as well as the average temperature rise. A relaxation model of reconnection between two or more flux ropes in the solar corona is also described, allowing for different senses of twist, and the implications for heating of the solar corona are discussed.

  9. Plasma-Material Interface Development for Future Spherical Tokamak-based Devices in NSTX.

    SciTech Connect

    et. al, V

    2011-09-24

    The divertor plasma-material interface (PMI) must be able to withstand steady-state heat fluxes up to 10 MW/m{sup 2} (a limit imposed by the present day divertor material and engineering constraints) with minimal material erosion, as well as to provide impurity control and ion density pumping capabilities. In spherical tokamaks (STs), the compact divertor geometry and the requirement of low core electron collisionality n*{sub e} at n{sub e} < 0.5-0.7 n{sub G} (where n{sub G} is the Greenwald density) for increased neutral beam current drive efficiency impose much greater demands on divertor and first-wall particle and heat flux mitigation solutions. In NSTX, divertor heat flux mitigation and impurity control with an innovative 'snowflake' divertor configuration and ion density pumping by evaporated lithium wall and divertor coatings are studied. Lithium coatings have enabled ion density reduction up to 50% in NSTX through the reduction of wall and divertor recycling rates. The 'snowflake' divertor configuration was obtained in NSTX in 0.8-1 MA 4-6 MW NBI-heated H-mode lithium-assisted discharges using three divertor coils. The snowflake divertor formation was always accompanied by a partial detachment of the outer strike point with an up to 50% increase in divertor radiation from intrinsic carbon, the peak divertor heat flux reduction from 3-6 MW/m{sup 2} to 0.5-1 MW/m{sup 2}, and a significant increase in divertor volume recombination. High core confinement was maintained with the snowflake divertor, evidenced by the t{sub E}, W{sub MHD} and the H98(y,2) factors similar to those of the standard divertor discharges. Core carbon concentration and radiated power were reduced by 30-70%, apparently as a result of reduced divertor physical and chemical sputtering in the snowflake divertor and ELMs. In the SFD discharges, the MHD stability of the H-mode pedestal region was altered leading to the re-appearance of medium size (DW/W = 5-10%), Type I, ELMs otherwise

  10. A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks

    SciTech Connect

    R.D. Woolley

    2009-05-29

    A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress.

  11. A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks

    SciTech Connect

    Robert D. Woolley

    2009-06-11

    A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress.

  12. Investigating fusion plasma instabilities in the Mega Amp Spherical Tokamak using mega electron volt proton emissions (invited).

    PubMed

    Perez, R V; Boeglin, W U; Darrow, D S; Cecconello, M; Klimek, I; Allan, S Y; Akers, R J; Keeling, D L; McClements, K G; Scannell, R; Turnyanskiy, M; Angulo, A; Avila, P; Leon, O; Lopez, C; Jones, O M; Conway, N J; Michael, C A

    2014-11-01

    The proton detector (PD) measures 3 MeV proton yield distributions from deuterium-deuterium fusion reactions within the Mega Amp Spherical Tokamak (MAST). The PD's compact four-channel system of collimated and individually oriented silicon detectors probes different regions of the plasma, detecting protons (with gyro radii large enough to be unconfined) leaving the plasma on curved trajectories during neutral beam injection. From first PD data obtained during plasma operation in 2013, proton production rates (up to several hundred kHz and 1 ms time resolution) during sawtooth events were compared to the corresponding MAST neutron camera data. Fitted proton emission profiles in the poloidal plane demonstrate the capabilities of this new system. PMID:25430211

  13. Investigating fusion plasma instabilities in the Mega Amp Spherical Tokamak using mega electron volt proton emissions (invited)

    SciTech Connect

    Perez, R. V. Boeglin, W. U.; Angulo, A.; Avila, P.; Leon, O.; Lopez, C.; Darrow, D. S.; Cecconello, M.; Klimek, I.; Allan, S. Y.; Akers, R. J.; Keeling, D. L.; McClements, K. G.; Scannell, R.; Conway, N. J.; Turnyanskiy, M.; Jones, O. M.; Michael, C. A.

    2014-11-15

    The proton detector (PD) measures 3 MeV proton yield distributions from deuterium-deuterium fusion reactions within the Mega Amp Spherical Tokamak (MAST). The PD’s compact four-channel system of collimated and individually oriented silicon detectors probes different regions of the plasma, detecting protons (with gyro radii large enough to be unconfined) leaving the plasma on curved trajectories during neutral beam injection. From first PD data obtained during plasma operation in 2013, proton production rates (up to several hundred kHz and 1 ms time resolution) during sawtooth events were compared to the corresponding MAST neutron camera data. Fitted proton emission profiles in the poloidal plane demonstrate the capabilities of this new system.

  14. Coherence imaging of scrape-off-layer and divertor impurity flows in the Mega Amp Spherical Tokamak (invited)

    SciTech Connect

    Silburn, S. A. Sharples, R. M.; Harrison, J. R.; Meyer, H.; Michael, C. A.; Howard, J.; Gibson, K. J.

    2014-11-15

    A new coherence imaging Doppler spectroscopy diagnostic has been deployed on the UK’s Mega Amp Spherical Tokamak for scrape-off-layer and divertor impurity flow measurements. The system has successfully obtained 2D images of C III, C II, and He II line-of-sight flows, in both the lower divertor and main scrape-off-layer. Flow imaging has been obtained at frame rates up to 1 kHz, with flow resolution of around 1 km/s and spatial resolution better than 1 cm, over a 40° field of view. C III data have been tomographically inverted to obtain poloidal profiles of the parallel impurity flow in the divertor under various conditions. In this paper we present the details of the instrument design, operation, calibration, and data analysis as well as a selection of flow imaging results which demonstrate the diagnostic's capabilities.

  15. Note: Multi-pass Thomson scattering measurement on the TST-2 spherical tokamak.

    PubMed

    Togashi, H; Ejiri, A; Hiratsuka, J; Nakamura, K; Takase, Y; Yamaguchi, T; Furui, H; Imamura, K; Inada, T; Kakuda, H; Nakanishi, A; Oosako, T; Shinya, T; Sonehara, M; Tsuda, S; Tsujii, N; Wakatsuki, T; Hasegawa, M; Nagashima, Y; Narihara, K; Yamada, I; Tojo, H

    2014-05-01

    In multi-pass Thomson scattering (TS) scheme, a laser pulse makes multiple round trips through the plasma, and the effective laser energy is enhanced, and we can increase the signal-to-noise ratio as a result. We have developed a coaxial optical cavity in which a laser pulse is confined, and we performed TS measurements using the coaxial cavity in tokamak plasmas for the first time. In the optical cavity, the laser energy attenuation was approximately 30% in each round trip, and we achieved a photon number gain of about 3 compared with that obtained in the first round trip. In addition, the temperature measurement accuracy was improved by accumulating the first three round trip waveforms. PMID:24880428

  16. Note: Multi-pass Thomson scattering measurement on the TST-2 spherical tokamak

    SciTech Connect

    Togashi, H. Ejiri, A.; Hiratsuka, J.; Nakamura, K.; Takase, Y.; Yamaguchi, T.; Furui, H.; Imamura, K.; Inada, T.; Kakuda, H.; Nakanishi, A.; Oosako, T.; Shinya, T.; Sonehara, M.; Tsuda, S.; Tsujii, N.; Wakatsuki, T.; Hasegawa, M.; Nagashima, Y.; Narihara, K.; and others

    2014-05-15

    In multi-pass Thomson scattering (TS) scheme, a laser pulse makes multiple round trips through the plasma, and the effective laser energy is enhanced, and we can increase the signal-to-noise ratio as a result. We have developed a coaxial optical cavity in which a laser pulse is confined, and we performed TS measurements using the coaxial cavity in tokamak plasmas for the first time. In the optical cavity, the laser energy attenuation was approximately 30% in each round trip, and we achieved a photon number gain of about 3 compared with that obtained in the first round trip. In addition, the temperature measurement accuracy was improved by accumulating the first three round trip waveforms.

  17. Conceptual design study of a superconducting spherical tokamak reactor with a self-consistent system analysis code

    NASA Astrophysics Data System (ADS)

    Hong, B. G.; Hwang, Y. S.; Kang, J. S.; Lee, D. W.; Joo, H. G.; Ono, M.

    2011-11-01

    In a spherical tokamak (ST) reactor, the radial build of toroidal field coil and the shield play a key role in determining the size of the reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with a one-dimensional radiation transport code. A conceptual design study of a compact superconducting ST reactor with an aspect ratio of up to 2.0 is conducted and the optimum radial build is identified. It is shown that the use of an improved shielding material and high-temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at a low aspect ratio, and that by using an inboard neutron reflector instead of a breeding blanket, tritium self-sufficiency is possible with an outboard blanket only and thus a compact-sized all superconducting coil ST reactor is viable.

  18. Divertor impurity injection using high voltage arcs for impurity transport studies on the Mega Amp Spherical Tokamak.

    PubMed

    Leggate, H J; Lisgo, S W; Harrison, J R; Elmore, S; Allan, S Y; Gaffka, R C; Stephen, R C; Turner, M M

    2014-12-01

    The operation of next-generation fusion reactors will be significantly affected by impurity transport in the scrape-off layer (SOL). Current modelling efforts are restricted by a lack of detailed data on impurity transport in the SOL. In order to address this, a carbon injector has been designed and installed on the Mega Amp Spherical Tokamak (MAST). The injector creates short lived carbon plumes originating at the MAST divertor lasting less than 50 μs. High voltage capacitor banks are used to create a discharge across concentric carbon electrodes located in a probe mounted on the Divertor Science Facility in the MAST lower divertor. This results in a very short plume duration allowing observation of the evolution of the plume and precise localisation of the plume relative to the X-point on MAST. The emission from the carbon plume was imaged using fast visible cameras filtered in order to isolate the carbon II and carbon III emission lines centered around 514 nm and 465 nm. PMID:25554289

  19. Effect of asymmetrical eddy currents on magnetic diagnosis signals for equilibrium reconstruction in the Sino-UNIted Spherical Tokamak.

    PubMed

    Jiang, Y Z; Tan, Y; Gao, Z; Wang, L

    2014-11-01

    The vacuum vessel of Sino-UNIted Spherical Tokamak was split into two insulated hemispheres, both of which were insulated from the central cylinder. The eddy currents flowing in the vacuum vessel would become asymmetrical due to discontinuity. A 3D finite elements model was applied in order to study the eddy currents. The modeling results indicated that when the Poloidal Field (PF) was applied, the induced eddy currents would flow in the toroidal direction in the center of the hemispheres and would be forced to turn to the poloidal and radial directions due to the insulated slit. Since the eddy currents converged on the top and bottom of the vessel, the current densities there tended to be much higher than those in the equatorial plane were. Moreover, the eddy currents on the top and bottom of vacuum vessel had the same direction when the current flowed in the PF coils. These features resulted in the leading phases of signals on the top and bottom flux loops when compared with the PF waveforms. PMID:25430380

  20. Divertor impurity injection using high voltage arcs for impurity transport studies on the Mega Amp Spherical Tokamak

    SciTech Connect

    Leggate, H. J. Turner, M. M.; Lisgo, S. W.; Harrison, J. R.; Elmore, S.; Allan, S. Y.; Gaffka, R. C.; Stephen, R. C.

    2014-12-15

    The operation of next-generation fusion reactors will be significantly affected by impurity transport in the scrape-off layer (SOL). Current modelling efforts are restricted by a lack of detailed data on impurity transport in the SOL. In order to address this, a carbon injector has been designed and installed on the Mega Amp Spherical Tokamak (MAST). The injector creates short lived carbon plumes originating at the MAST divertor lasting less than 50 μs. High voltage capacitor banks are used to create a discharge across concentric carbon electrodes located in a probe mounted on the Divertor Science Facility in the MAST lower divertor. This results in a very short plume duration allowing observation of the evolution of the plume and precise localisation of the plume relative to the X-point on MAST. The emission from the carbon plume was imaged using fast visible cameras filtered in order to isolate the carbon II and carbon III emission lines centered around 514 nm and 465 nm.

  1. The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    Harrison, J. R.; Fishpool, G. M.; Thornton, A. J.; Walkden, N. R.

    2015-09-01

    The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFR of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1-2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5-1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.

  2. The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak

    SciTech Connect

    Harrison, J. R.; Fishpool, G. M.; Thornton, A. J.; Walkden, N. R.

    2015-09-15

    The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFR of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1–2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5–1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.

  3. Plasma start-up in a spherical tokamak with close-fitting conducting walls

    NASA Astrophysics Data System (ADS)

    Berzak, Laura Florence

    2010-12-01

    Conducting structures closely coupled to the plasma intensify the challenge of designing the magnetic fields necessary for successful plasma start-up. The need for close-fitting, conducting structures will continue to grow in importance, particularly as more experiments investigate alternative first walls and as experiments such as ITER incorporate large, highly conductive structures close to the plasma boundary. The Lithium Tokamak eXperiment (LTX) is designed to investigate the novel, low-recycling lithium wall operating regime for magnetically confined plasmas. LTX reaches this regime through a heated copper shell coated with liquid lithium internal to the vacuum vessel and conformal to the plasma last-closed-flux surface. The three-dimensional nature of the shell causes the eddy currents and resultant magnetic fields to be three-dimensional as well. An extensive array of unique magnetic diagnostics has been designed to yield data on non-axisymmetric, three-dimensional fields and implemented to quantify the temporal history and magnitude of fields due to the large eddy currents in the LTX shell. With data from the magnetic diagnostics, primary sources and paths of eddy currents have been elucidated. To further quantify eddy currents and their effect on in-vessel magnetic field, a two-dimensional electromagnetic code has been developed by utilizing data from the magnetic diagnostics. By developing an understanding of the effect of eddy currents on in-vessel fields through measurements and simulations, plasma start-up has been achieved in LTX. Highly-constrained plasma reconstructions have been calculated and used to predict the field evolution required for the desired discharge development. To capture full three-dimensional effects, a three-dimensional electromagnetic code set is under development. Using this code set, a new calibration algorithm for determining the actual versus model locations of field coils and sensors has been developed, as well as a novel

  4. Using Bayesian analysis and Gaussian processes to infer electron temperature and density profiles on the Mega-Ampere Spherical Tokamak experiment

    SciTech Connect

    Nessi, G. T. von; Hole, M. J.

    2013-06-15

    A unified, Bayesian inference of midplane electron temperature and density profiles using both Thomson scattering (TS) and interferometric data is presented. Beyond the Bayesian nature of the analysis, novel features of the inference are the use of a Gaussian process prior to infer a mollification length-scale of inferred profiles and the use of Gauss-Laguerre quadratures to directly calculate the depolarisation term associated with the TS forward model. Results are presented from an application of the method to data from the high resolution TS system on the Mega-Ampere Spherical Tokamak, along with a comparison to profiles coming from the standard analysis carried out on that system.

  5. Using Bayesian analysis and Gaussian processes to infer electron temperature and density profiles on the Mega-Ampere Spherical Tokamak experiment.

    PubMed

    von Nessi, G T; Hole, M J

    2013-06-01

    A unified, Bayesian inference of midplane electron temperature and density profiles using both Thomson scattering (TS) and interferometric data is presented. Beyond the Bayesian nature of the analysis, novel features of the inference are the use of a Gaussian process prior to infer a mollification length-scale of inferred profiles and the use of Gauss-Laguerre quadratures to directly calculate the depolarisation term associated with the TS forward model. Results are presented from an application of the method to data from the high resolution TS system on the Mega-Ampere Spherical Tokamak, along with a comparison to profiles coming from the standard analysis carried out on that system. PMID:23822343

  6. The effect of off-axis neutral beam injection on sawtooth stability in ASDEX Upgrade and Mega-Ampere Spherical Tokamak

    SciTech Connect

    Chapman, I. T.; de Bock, M. F.; Pinches, S. D.; Turnyanskiy, M. R.

    2009-07-15

    Sawtooth behavior has been investigated in plasmas heated with off-axis neutral beam injection in ASDEX Upgrade [A. Herrmann and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)] and the Mega-Ampere Spherical Tokamak (MAST) [A. Sykes et al., Nucl. Fusion 41, 1423 (2001)]. Provided that the fast ions are well confined, the sawtooth period is found to decrease as the neutral beam is injected further off-axis. Drift kinetic modeling of such discharges qualitatively shows that the passing fast ions born outside the q=1 rational surface can destabilize the n=1 internal kink mode, thought to be related to the sawtooth instability. This effect can be enhanced by optimizing the deposition of the off-axis beam energetic particle population with respect to the mode location.

  7. High βp plasma formation using off-axis ECCD in Ohmic heated plasma in the spherical tokamak QUEST

    NASA Astrophysics Data System (ADS)

    Mishra, Kishore; Zushi, H.; Idei, H.; Hasegawa, M.; Hanada, K.

    2015-03-01

    High poloidal beta (ɛβp ~ 1) operation in steady state condition in tokamaks is of great interest and has previously been demonstrated using NBI, LHCD and low current (Ip) plasma for a short time (<0.5 s). A very few experiments however, have been performed towards the investigation of highest obtainable βp in tokamak plasma. In this work we report the first result of high βp production and its sustainment though an off axis ECCD at two different frequencies (fundamental and second harmonic) in Ohmic (OH) target plasma. With application of ECCD, plasma βp increased to encounter an equilibrium limit and the standard limiter configuration is transformed to an Inboard Poloidal field Null (IPN) configuration. Both off-axis and on-axis ECCD is studied and found to have some distinctive features, which are discussed in this paper.

  8. Electron kinetics inferred from observations of microwave bursts during edge localized modes in the mega-amp spherical tokamak.

    PubMed

    Freethy, S J; McClements, K G; Chapman, S C; Dendy, R O; Lai, W N; Pamela, S J P; Shevchenko, V F; Vann, R G L

    2015-03-27

    Recent measurements of microwave and x-ray emission during edge localized mode (ELM) activity in tokamak plasmas provide a fresh perspective on ELM physics. It is evident that electron kinetics, which are not incorporated in standard (fluid) models for the instability that drives ELMs, play a key role in the new observations. These effects should be included in future models for ELMs and the ELM cycle. The observed radiative effects paradoxically imply acceleration of electrons parallel to the magnetic field combined with rapid acquisition of perpendicular momentum. It is shown that this paradox can be resolved by the action of the anomalous Doppler instability which enables fast collective radiative relaxation, in the perpendicular direction, of electrons accelerated in the parallel direction by inductive electric fields generated by the initial ELM instability. PMID:25860751

  9. Heat flux and plasma flow in the scrape off layer on the spherical tokamak QUEST with inboard poloidal field null configuration

    NASA Astrophysics Data System (ADS)

    Onchi, Takumi; Zushi, Hideki; Mishra, Kishore; Hanada, Kazuaki; Idei, Hiroshi; Nakamura, Kazuo; Fujisawa, Akihide; Nagashima, Yoshihiko; Hasegawa, Makoto; Kuzmin, Arseny; Nagaoka, Kenichi; QUEST Team

    2014-10-01

    Heat flux and plasma flow in the scrape off layer (SOL) are examined in the inboard poloidal null (IPN) configuration on the spherical tokamak (ST) QUEST. In the ST, trapped energetic electrons on the low field side are widely excursed from the last closed flux surface to SOL so that significant heat loss occurs. Interestingly, plasma flows in the core and the SOL are also observed in IPN though no inductive force like ohmic heating is applied. High heat flux (>1 MW/m2) and sonic flow (M > 1) in far-SOL arise in current ramp-up phase. In quasi-steady state, sawtooth-like oscillation of plasma current with 20 Hz has been observed. Heat flux and subsonic plasma flow in far-SOL are well correlated to plasma current oscillation. The toroidal Mach number largely increases from Mφ ~ 0.1 to ~ 0.5 and drops although the amplitude of plasma current is about 10% of that. Note that such flow modification occurs before plasma current crash, there may be some possibility that phenomena in the SOL or the edge trigger reactions in the core plasma. This work is supported by Grants-in-aid for Scientific Research (S24226020), NIFS Collaboration Research Program (NIFS12KUTR081), and the Collaborative Research Program of Research Institute for Applied Mechanics, Kyushu University.

  10. The influence of toroidal Alfvén modes on the confinement of fast particles in the Globus-M spherical tokamak

    NASA Astrophysics Data System (ADS)

    Petrov, Yu. V.; Bakharev, N. N.; Gusev, V. K.; Minaev, V. B.; Kornev, V. A.; Mel'nik, A. D.; Patrov, M. I.; Sakharov, N. V.; Tolstyakov, S. Yu.; Kurskiev, G. S.; Chernyshev, F. V.; Shchegolev, P. B.

    2014-12-01

    Neutral beam injection into the Globus-M spherical tokamak at the early stage of discharge leads to the development of instabilities in a frequency range of 50-200 kHz, which have been identified as toroidal Alfvén eigenmodes (TAEs) [1]. The influence of these modes on the confinement of fast particles has been studied with the aid of a neutral particle analyzer (NPA) and a neutron detector. The isotope effect was studied using hydrogen and deuterium both in the injected beam and in the target plasma. A correlation analysis of signals from magnetic probes showed that the observed modes in most cases contain a single harmonic with toroidal number n = 1. Upon the injection of deuterium into deuterium plasma, the development of TAEs led to a decrease in the neutron flux by 25%, whereas the fluxes of high-energy recharge atoms decreased by 75%. After the injection of hydrogen, a decrease in the flux measured by NPA did not exceed 25%.

  11. Non-inductive plasma start-up experiments on the TST-2 spherical tokamak using waves in the lower-hybrid frequency range

    NASA Astrophysics Data System (ADS)

    Shinya, T.; Takase, Y.; Wakatsuki, T.; Ejiri, A.; Furui, H.; Hiratsuka, J.; Imamura, K.; Inada, T.; Kakuda, H.; Kasahara, H.; Kumazawa, R.; Moeller, C.; Mutoh, T.; Nagashima, Y.; Nakamura, K.; Nakanishi, A.; Oosako, T.; Saito, K.; Seki, T.; Sonehara, M.; Togashi, H.; Tsuda, S.; Tsujii, N.; Yamaguchi, T.

    2015-07-01

    Non-inductive plasma current start-up and sustainment by waves in the lower-hybrid frequency range (200 MHz) have been studied on the TST-2 spherical tokamak (R0 ⩽ 0.38 m, a ⩽ 0.25 m, Bt0 ⩽ 0.3 T, Ip ⩽ 0.14 MA) using three types of antenna: the 11-element inductively-coupled combline antenna, the dielectric loaded 4-waveguide array antenna, and the 13-element capacitively-coupled combline (CCC) antenna. The maximum plasma currents of 15 kA, 10 kA and 16 kA were achieved, respectively. The highest current drive figure of merit η_CD \\equiv \\overline{n}e Ip R / P_RF was achieved by the CCC antenna. The efficiency of current drive should improve by reducing prompt orbit losses of high energy electrons by operating at higher plasma current (to improve orbit confinement) and higher toroidal magnetic field (to improve wave accessibility to the plasma core), while keeping the density high enough (to avoid excessive acceleration of electrons), but under the ‘density limit’.

  12. Investigations of the radial propagation of blob-like structure in a non-confined electron cyclotron resonance heated plasma on Q-shu University Experiment with a Steady-State Spherical Tokamak

    SciTech Connect

    Ogata, R.; Liu, H. Q.; Ishiguro, M.; Ikeda, T.; Hanada, K.; Zushi, H.; Nakamura, K.; Fujisawa, A.; Idei, H.; Hasegawa, M.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Nishino, N.; Collaboration: QUEST Group

    2011-09-15

    A study of radial propagation and electric fields induced by charge separation in blob-like structures has been performed in a non-confined cylindrical electron cyclotron resonance heating plasma on Q-shu University Experiment with a Steady-State Spherical Tokamak using a fast-speed camera and a Langmuir probe. The radial propagation of the blob-like structures is found to be driven by E x B drift. Moreover, these blob-like structures were found to have been accelerated, and the property of the measured radial velocities agrees with the previously proposed model [C. Theiler et al., Phys. Rev. Lett. 103, 065001 (2009)]. Although the dependence of the radial velocity on the connection length of the magnetic field appeared to be different, a plausible explanation based on enhanced short-circuiting of the current path can be proposed.

  13. Plasma Physics Regimes in Tokamaks with Li Walls

    SciTech Connect

    L.E. Zakharo; N.N. Gorelenkov; R.B. White; S.I. Krasheninnikov; G.V. Pereverzev

    2003-08-21

    Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors.

  14. Simulation of the conditions of wave excitation for the optimization of the lower hybrid current drive in the Globus-M spherical tokamak

    SciTech Connect

    Dyachenko, V. V.; Irzak, M. A.; Cherotchenko, E. D.; Shcherbinin, O. N.

    2013-02-15

    At present, the method of current drive by means of lower hybrid waves is not applied to low-aspect-ratio tokamaks, because, in the traditional approach, it would be necessary to use waves with a very high slowing-down factor. However, studies of new transparency regions for waves in a nonuniform magnetized plasma, performed earlier at the Ioffe Physical Technical Institute, Russian Academy of Sciences, made it possible to develop an approach in which slow waves are excited in the poloidal (rather than toroidal) direction. In this approach, moderately slowed-down waves first propagate in the poloidal direction, but then turn in the toroidalal direction and get into the dense plasma. In this work, this approach is further developed using numerical methods. In particular, the influence of the density profile in the edge plasma on the efficiency of wave excitation for given antenna parameters is studied in detail.

  15. Temperature and density characteristics of the Helicity Injected Torus-II spherical tokamak indicating closed flux sustainment using coaxial helicity injection

    SciTech Connect

    Hamp, W. T.; Jarboe, T. R.; Nelson, B. A.; O'Neill, R. G.; Raman, R.; Redd, A. J.; Stewart, B. T.; Mueller, D.

    2008-08-15

    The electron temperature and density profiles of plasmas in the Helicity Injected Torus [HIT-II: T. R. Jarboe et al., Phys. Plasmas 5, 1807 (1998)] experiment are measured by multipoint Thomson scattering (MPTS). The HIT-II device is a small low-aspect-ratio tokamak (major radius 0.3 m, minor radius 0.2 m, toroidal field of up to 0.5 T), capable of inductive ohmic (OH) current drive, Coaxial Helicity Injection (CHI) current drive, or combinations of both. The temperature and density characteristics have been characterized by a ruby laser MPTS diagnostic at up to six locations within the plasma for a single diagnostic time per discharge. Observed hollow temperature profiles of CHI discharges are inconsistent with open flux only predictions for CHI and indicate a closed flux region during CHI current drive.

  16. Simulation of the conditions of wave excitation for the optimization of the lower hybrid current drive in the Globus-M spherical tokamak

    NASA Astrophysics Data System (ADS)

    Dyachenko, V. V.; Irzak, M. A.; Cherotchenko, E. D.; Shcherbinin, O. N.

    2013-02-01

    At present, the method of current drive by means of lower hybrid waves is not applied to low-aspect-ratio tokamaks, because, in the traditional approach, it would be necessary to use waves with a very high slowing-down factor. However, studies of new transparency regions for waves in a nonuniform magnetized plasma, performed earlier at the Ioffe Physical Technical Institute, Russian Academy of Sciences, made it possible to develop an approach in which slow waves are excited in the poloidal (rather than toroidal) direction. In this approach, moderately slowed-down waves first propagate in the poloidal direction, but then turn in the toroidalal direction and get into the dense plasma. In this work, this approach is further developed using numerical methods. In particular, the influence of the density profile in the edge plasma on the efficiency of wave excitation for given antenna parameters is studied in detail.

  17. Recent Progress on Spherical Torus Research

    SciTech Connect

    Ono, Masayuki; Kaita, Robert

    2014-01-01

    The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ~ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ~ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of attractive fusion energy power source. Since the start of the two megaampere class ST facilities in 2000, National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all of fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.

  18. Spherical tokamak (ST) transmutation of nuclear wastes

    SciTech Connect

    Peng, Y.-K.M.; Galambos, J.D.; Cheng, E.T.; Cerbone, R.J.

    1995-12-31

    The concept for an ST fusion core that drives a He-cooled, actinide-bearing, molten-salt blanket of moderate power density to generate electricity is examined for the first time. The results show that the fusion core is suited for this purpose and require a level of plasma, power density, engineering, and material performances moderate in comparison with what has been considered desirable for fusion-only power plants. The low aspect ratio of ST introduces a relatively thick, diverted scrape-off layer which leads to reduced heat fluxes at the limiter and divertor tiles. The use of a demountable, water-cooled, single-turn copper center leg for the toroidal field coils enables simplifications of the fusion core configuration and improves overall practicality for future power applications. These result in much reduced size and cost of the fusion core for the transmutation power plant relative to an optimized fusion-only fusion core. Surrounded by a separate tritium-breeding zone, the molten-salt blanket concept is in principle less complex and costly than the thermal breeding blankets for fusion. These combine to effect major reductions in the cost and weight of the power core equipment for the transmutation power plant. The minimum cost of electricity for such a power plant is thus reduced from the best fusion-only counterpart by more than 30%, based on consistent but approximate modeling. The key issues, development steps, and the potential value inherent in the ST fusion core in addressing the world needs for nuclear waste energy production are discussed.

  19. ECH tokamak

    SciTech Connect

    Firestone, M.A.; Mau, T.K.; Conn, R.W.

    1985-04-01

    A small steady-state tokamak capable of producing power in the 100 to 300 MWe range and relying on electron cyclotron RF heating (ECH) for both heating and current drive is described. Working in the first MHD stability regime for tokamaks, the approach adheres to the recently discovered maximum beta limit. An appropriate figure of merit is the ratio of the fusion power to absorbed RF power. Efficient devices are feasible at both small and large values of fusion power, thereby pointing to a development path for an attractive commercial fusion reactor.

  20. Recent progress on spherical torus research

    NASA Astrophysics Data System (ADS)

    Ono, Masayuki; Kaita, Robert

    2015-04-01

    The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ˜ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ˜ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment in the United States and Mega Ampere Spherical Tokamak in UK, active ST research has been conducted worldwide. More than 16 ST research facilities operating during this period have achieved remarkable advances in all fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.

  1. Recent progress on spherical torus research

    SciTech Connect

    Ono, Masayuki; Kaita, Robert

    2015-04-15

    The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R{sub 0}/a) reduced to A ∼ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ∼ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment in the United States and Mega Ampere Spherical Tokamak in UK, active ST research has been conducted worldwide. More than 16 ST research facilities operating during this period have achieved remarkable advances in all fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.

  2. Physics of Spherical Torus Plasmas

    SciTech Connect

    Peng, Yueng Kay Martin

    2000-01-01

    Broad and important progress in plasma tests, theory, new experiments, and future visions of the spherical torus (ST, or very low aspect ratio tokamaks) have recently emerged. These have substantially improved our understanding of the potential properties of the ST plasmas, since the preliminary calculation of the ST magnetohydrodynamic equilibria more than a decade ago. Exciting data have been obtained from concept exploration level ST experiments of modest capabilities (with major radii up to 35 cm), making important scientific contributions to toroidal confinement in general. The results have helped approval and construction of new and/or more powerful ST experiments, and stimulated an increasing number of theoretical calculations of interest to magnetic fusion energy. Utilizing the broad knowledge base from the successful tokamak and advanced tokamak research, a wide range of new ST physics features has been suggested. These properties of the ST plasma will be tested at the 1 MA level with major radius up to similar to 80 cm in the new proof of principle devices National Spherical Torus Experiment (NSTX, U.S.) [M. Peng , European Conf. Abst. 22C, 451 (1998); S. M. Kaye , Fusion Technol. 36, 16 (1999); M. Ono , "Exploration of Spherical Torus Physics in the NSTX Device," 17th IAEA Fusion Energy Conf., paper IAEA-CN-69/ICP/01 (R), Yokohama, Japan (1998)], Mega Ampere Spherical Tokamak (MAST, U.K.) [A. C. Darke , Fusion Technol. 1, 799 (1995); Q. W. Morris , Proc. Int. Workshop on ST (Ioffe Inst., St. Petersburg, 1997), Vol. 1, p. 290], and Globus-M (R.F.) [V. K. Gusev , European Conf. Abst. 22C, 576 (1998)], which have just started full experimental operation. New concept exploration experiments, such as Pegasus (University of Wisconsin) [R. Fonck and the PEGASUS Team, Bull. Am. Phys. Soc. 44, 267 (1999)], Helicity Injected Tokamak-II (HIT-II, University of Washington) [T. R. Jarboe , Phys. Plasmas 5, 1807 (1998)], and Current Drive Experiment-Upgrade (CDX

  3. Turbulent equipartition pinch of toroidal momentum in spherical torus

    NASA Astrophysics Data System (ADS)

    Hahm, T. S.; Lee, J.; Wang, W. X.; Diamond, P. H.; Choi, G. J.; Na, D. H.; Na, Y. S.; Chung, K. J.; Hwang, Y. S.

    2014-12-01

    We present a new analytic expression for turbulent equipartition (TEP) pinch of toroidal angular momentum originating from magnetic field inhomogeneity of spherical torus (ST) plasmas. Starting from a conservative modern nonlinear gyrokinetic equation (Hahm et al 1988 Phys. Fluids 31 2670), we derive an expression for pinch to momentum diffusivity ratio without using a usual tokamak approximation of B ∝ 1/R which has been previously employed for TEP momentum pinch derivation in tokamaks (Hahm et al 2007 Phys. Plasmas 14 072302). Our new formula is evaluated for model equilibria of National Spherical Torus eXperiment (NSTX) (Ono et al 2001 Nucl. Fusion 41 1435) and Versatile Experiment Spherical Torus (VEST) (Chung et al 2013 Plasma Sci. Technol. 15 244) plasmas. Our result predicts stronger inward pinch for both cases, as compared to the prediction based on the tokamak formula.

  4. The physics of tokamak start-upa)

    NASA Astrophysics Data System (ADS)

    Mueller, D.

    2013-05-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  5. The Physics of Tokamak Start-up

    SciTech Connect

    D. Mueller

    2012-11-13

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. ITER, the National Spherical Torus eXperiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  6. The physics of tokamak start-up

    SciTech Connect

    Mueller, D.

    2013-05-15

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  7. Bifurcated helical core equilibrium states in tokamaks

    NASA Astrophysics Data System (ADS)

    Cooper, W. A.; Chapman, I. T.; Schmitz, O.; Turnbull, A. D.; Tobias, B. J.; Lazarus, E. A.; Turco, F.; Lanctot, M. J.; Evans, T. E.; Graves, J. P.; Brunetti, D.; Pfefferlé, D.; Reimerdes, H.; Sauter, O.; Halpern, F. D.; Tran, T. M.; Coda, S.; Duval, B. P.; Labit, B.; Pochelon, A.; Turnyanskiy, M. R.; Lao, L.; Luce, T. C.; Buttery, R.; Ferron, J. R.; Hollmann, E. M.; Petty, C. C.; van Zeeland, M.; Fenstermacher, M. E.; Hanson, J. M.; Lütjens, H.

    2013-07-01

    Tokamaks with weak to moderate reversed central shear in which the minimum inverse rotational transform (safety factor) qmin is in the neighbourhood of unity can trigger bifurcated magnetohydrodynamic equilibrium states, one of which is similar to a saturated ideal internal kink mode. Peaked prescribed pressure profiles reproduce the ‘snake’ structures observed in many tokamaks which has led to a novel explanation of the snake as a bifurcated equilibrium state. Snake equilibrium structures are computed in simulations of the tokamak à configuration variable (TCV), DIII-D and mega amp spherical torus (MAST) tokamaks. The internal helical deformations only weakly modulate the plasma-vacuum interface which is more sensitive to ripple and resonant magnetic perturbations. On the other hand, the external perturbations do not alter the helical core deformation in a significant manner. The confinement of fast particles in MAST simulations deteriorate with the amplitude of the helical core distortion. These three-dimensional bifurcated solutions constitute a paradigm shift that motivates the applications of tools developed for stellarator research in tokamak physics investigations.

  8. Measurements of Prompt and MHD-Induced Fast Ion Loss from National Spherical Torus Experiment Plasmas

    SciTech Connect

    D.S. Darrow; S.S. Medley; A.L. Roquemore; W.W. Heidbrink; A. Alekseyev; F.E. Cecil; J. Egedal; V.Ya. Goloborod'ko; N.N. Gorelenkov; M. Isobe; S. Kaye; M. Miah; F. Paoletti; M.H. Redi; S.N. Reznik; A. Rosenberg; R. White; D. Wyatt; V.A. Yavorskij

    2002-10-15

    A range of effects may make fast ion confinement in spherical tokamaks worse than in conventional aspect ratio tokamaks. Data from neutron detectors, a neutral particle analyzer, and a fast ion loss diagnostic on the National Spherical Torus Experiment (NSTX) indicate that neutral beam ion confinement is consistent with classical expectations in quiescent plasmas, within the {approx}25% errors of measurement. However, fast ion confinement in NSTX is frequently affected by magnetohydrodynamic (MHD) activity, and the effect of MHD can be quite strong.

  9. Tokamak Systems Code

    SciTech Connect

    Reid, R.L.; Barrett, R.J.; Brown, T.G.; Gorker, G.E.; Hooper, R.J.; Kalsi, S.S.; Metzler, D.H.; Peng, Y.K.M.; Roth, K.E.; Spampinato, P.T.

    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.

  10. Spherical Camera

    NASA Technical Reports Server (NTRS)

    1997-01-01

    Developed largely through a Small Business Innovation Research contract through Langley Research Center, Interactive Picture Corporation's IPIX technology provides spherical photography, a panoramic 360-degrees. NASA found the technology appropriate for use in guiding space robots, in the space shuttle and space station programs, as well as research in cryogenic wind tunnels and for remote docking of spacecraft. Images of any location are captured in their entirety in a 360-degree immersive digital representation. The viewer can navigate to any desired direction within the image. Several car manufacturers already use IPIX to give viewers a look at their latest line-up of automobiles. Another application is for non-invasive surgeries. By using OmniScope, surgeons can look more closely at various parts of an organ with medical viewing instruments now in use. Potential applications of IPIX technology include viewing of homes for sale, hotel accommodations, museum sites, news events, and sports stadiums.

  11. SOL Width Scaling in the MAST Tokamak

    NASA Astrophysics Data System (ADS)

    Ahn, Joon-Wook; Counsell, Glenn; Connor, Jack; Kirk, Andrew

    2002-11-01

    Target heat loads are determined in large part by the upstream SOL heat flux width, Δ_h. Considerable effort has been made in the past to develop analytical and empirical scalings for Δh to allow reliable estimates to be made for the next-step device. The development of scalings for a large spherical tokamak (ST) such as MAST is particularly important both for development of the ST concept and for improving the robustness of scalings derived for conventional tokamaks. A first such scaling has been developed in MAST DND plasmas. The scaling was developed by flux-mapping data from the target Langmuir probe arrays to the mid-plane and fitting to key upstream parameters such as P_SOL, bar ne and q_95. In order to minimise the effects of co-linearity, dedicated campaigns were undertaken to explore the widest possible range of each parameter while keeping the remainder as fixed as possible. Initial results indicate a weak inverse dependence on P_SOL and approximately linear dependence on bar n_e. Scalings derived from consideration of theoretical edge transport models and integration with data from conventional devices is under way. The established scaling laws could be used for the extrapolations to the future machine such as Spherical Tokamak Power Plant (STPP). This work is jointly funded by Euratom and UK Department of Trade and Industry. J-W. Ahn would like to recognise the support of a grant from the British Foreign & Commonwealth Office.

  12. Nonlinear Simulation Studies of Tokamaks and STs

    SciTech Connect

    W. Park; J. Breslau; J. Chen; G.Y. Fu; S.C. Jardin; S. Klasky; J. Menard; A. Pletzer; B.C. Stratton; D. Stutman; H.R. Strauss; L.E. Sugiyama

    2003-07-07

    The multilevel physics, massively parallel plasma simulation code, M3D, has been used to study spherical tori (STs) and tokamaks. The magnitude of outboard shift of density profiles relative to electron temperature profiles seen in NSTX [National Spherical Torus Experiment] under strong toroidal flow is explained. Internal reconnection events in ST discharges can be classified depending on the crash mechanism, just as in tokamak discharges; a sawtooth crash, disruption due to stochasticity, or high-beta disruption. Toroidal shear flow can reduce linear growth of internal kink. It has a strong stabilizing effect nonlinearly and causes mode saturation if its profile is maintained, e.g., through a fast momentum source. Normally, however, the flow profile itself flattens during the reconnection process, allowing a complete reconnection to occur. In some cases, the maximum density and pressure spontaneously occur inside the island and cause mode saturation. Gyrokinetic hot particle/MHD hybrid studies of NSTX show the effects of fluid compression on a fast-ion-driven n = 1 mode. MHD studies of recent tokamak experiments with a central current hole indicate that the current clamping is due to sawtooth-like crashes, but with n = 0.

  13. The physics of spherical torus plasmas

    NASA Astrophysics Data System (ADS)

    Peng, Y.-K. M.

    2000-05-01

    Broad and important progress in plasma tests, theory, new experiments, and future visions of the spherical torus (ST, or very low aspect ratio tokamaks) have recently emerged. These have substantially improved our understanding of the potential properties of the ST plasmas, since the preliminary calculation of the ST magnetohydrodynamic equilibria more than a decade ago. Exciting data have been obtained from concept exploration level ST experiments of modest capabilities (with major radii up to 35 cm), making important scientific contributions to toroidal confinement in general. The results have helped approval and construction of new and/or more powerful ST experiments, and stimulated an increasing number of theoretical calculations of interest to magnetic fusion energy. Utilizing the broad knowledge base from the successful tokamak and advanced tokamak research, a wide range of new ST physics features has been suggested. These properties of the ST plasma will be tested at the 1 MA level with major radius up to ˜80 cm in the new proof of principle devices National Spherical Torus Experiment (NSTX, U.S.) [M. Peng et al., European Conf. Abst. 22C, 451 (1998); S. M. Kaye et al., Fusion Technol. 36, 16 (1999); M. Ono et al., "Exploration of Spherical Torus Physics in the NSTX Device," 17th IAEA Fusion Energy Conf., paper IAEA-CN-69/ICP/01 (R), Yokohama, Japan (1998)], Mega Ampere Spherical Tokamak (MAST, U.K.) [A. C. Darke et al., Fusion Technol. 1, 799 (1995); Q. W. Morris et al., Proc. Int. Workshop on ST (Ioffe Inst., St. Petersburg, 1997), Vol. 1, p. 290], and Globus-M (R.F.) [V. K. Gusev et al., European Conf. Abst. 22C, 576 (1998)], which have just started full experimental operation. New concept exploration experiments, such as Pegasus (University of Wisconsin) [R. Fonck and the PEGASUS Team, Bull. Am. Phys. Soc. 44, 267 (1999)], Helicity Injected Tokamak-II (HIT-II, University of Washington) [T. R. Jarboe et al., Phys. Plasmas 5, 1807 (1998)], and Current

  14. A feasibility study for the spherical torus experiment

    SciTech Connect

    Lazarus, E; Peng, Yueng Kay Martin

    1985-10-01

    Oak Ridge National Laboratory (ORNL) proposes to build the Spherical Torus Experiment (STX), a very low aspect ratio toroidal confinement device. This proposal concentrates on tokamak operation of the experiment; however, it can in principle be operated as a pinch or reversed-field pinch as well. As a tokamak, the spherical torus confines a plasma that is characterized by high toroidal beta, low poloidal beta, large natural elongation, high plasma current for a given edge q, and strong paramagnetism. These features combine to offer the possibility of a compact, low-field fusion device. The figure below shows that when compared to a conventional tokamak the spherical torus represents a major change in geometry. The primary goals of the experiment will be to demonstrate a capability for high beta (20%) in the first stability regime, to extend our knowledge of tokamak confinement scaling, and to test oscillating-field current drive. The experiment will operate in the high-beta, collisionless regime, which is achieved in STX at low temperatures because of the geometry. At a minimum, operation of STX will help to resolve fundamental questions regarding the scaling of beta and confinement in tokamaks. Complete success in this program would have a significant impact on toroidal fusion research in that it would demonstrate solutions to the problems of beta and steady-state operation in the tokamak. The proposed device has a major radius of 0.45 m, a toroidai field of 0.5 T, a plasma current of 900 kA, and heating by neutral beam injection. We estimate 30 months for design, construction, and assembly. The budget estimate, including contingency and escalation, is $6.8 million.

  15. Modular tokamak magnetic system

    DOEpatents

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  16. Electron Bernstein waves in spherical torus plasmas

    SciTech Connect

    Saveliev, A. N.

    2006-11-30

    Propagation and absorption of the electron Bernstein waves (EBWs) in spherical tokamaks (STs) have been intensively discussed in recent years because the EBWs coupled with an externally launched electromagnetic beam seem to be the only opportunity for microwave plasma heating and current drive in the electron cyclotron (EC) frequency range in the STs. The whole problem of the electron Bernstein heating and current drive (EBWHCD) in spherical plasmas is naturally divided into three major parts: coupling of incident electromagnetic waves (EMWs) to the EBWs near the upper hybrid resonance (UHR) surface, propagation and absorption of the EBWs in the plasma interior and generation of noninductive current driven by the EBWs. The present paper is a brief survey of the most important theoretical and numerical results on the issue of EBWs.

  17. Imaging with Spherically Bent Crystals or Reflectors

    SciTech Connect

    Bitter, M; Hill, K W; Scott, S; Ince-Cushman, A; Reinke, M; Podpaly, Y; Rice, J E; Beiersdorfer, P

    2010-06-01

    This paper consists of two parts: Part I describes the working principle of a recently developed x-ray imaging crystal spectrometer, where the astigmatism of spherically bent crystals is being used with advantage to record spatially resolved spectra of highly charged ions for Doppler measurements of the ion-temperature and toroidal plasmarotation- velocity profiles in tokamak plasmas. This type of spectrometer was thoroughly tested on NSTX and Alcator C-Mod, and its concept was recently adopted for the design of the ITER crystal spectrometers. Part II describes imaging schemes, where the astigmatism has been eliminated by the use of matched pairs of spherically bent crystals or reflectors. These imaging schemes are applicable over a wide range of the electromagnetic radiation, which includes microwaves, visible light, EUV radiation, and x-rays. Potential applications with EUV radiation and x-rays are the diagnosis of laserproduced plasmas, imaging of biological samples with synchrotron radiation, and lithography.

  18. Edge Physics Studies on the NSTX Spherical Tokamak

    SciTech Connect

    Boedo, J.

    2013-01-29

    In this funding period, activities were focused on processing data for publication, writing code to facilitate processing of data in the future as programmatic needs arise, and closing down grant activities.

  19. Physics Basis for a Spherical Tokamak Power Plant

    NASA Astrophysics Data System (ADS)

    Jardin, S. C.; Kessel, C. E.; Menard, J.; Mau, T. K.; Lin-Liu, Y. R.; Miller, R. L.; Stambaugh, R. D.; Turnbull, A. D.

    1998-11-01

    We present the results of physics optimization studies done as part of the ARIES-ST power plant study. The baseline configuration has the following parameters: β = 54%, β N = 7.5, elongation κ = 3.4, triangularity δ = .65, Aspect Ratio A = 1.6. Calculations using the PEST-II, GATO, BALLOON, and BALOO codes show this is stable to ballooning and kink modes up to n=6 with an ideally conducting wall with a separation of 0.165 a. Neutral beams (40 MW at 120 kV) will provide 5% of the plasma current and the rotation needed to stabilize the resistive-wall mode. We also address the vertical control and plasma initiation issues. Vertical stability is provided by a vertically stabilizing wall segment behind the breeding blanket, but within a separation of 0.45 minor radii from the plasma boundary. Plasma startup is facilitated by a combination of bootstrap current and external heating and current drive. Requirements on heating and current-drive systems and the timescales for current rampup are discussed.

  20. Neoclassical Tearing Mode Analysis in Spherical Tokamak Burning Plasmas

    NASA Astrophysics Data System (ADS)

    Kurita, Daiki; Yamazaki, Kozo; Arimoto, Hideki; Oishi, Tetsutarou; Shoji, Tatsuo

    For stabilization of neoclassical tearing mode (NTM), non-resonant helical field (NRHF) is investigated. The time variation of magnetic island is described by modified Rutherford equation. In this work, plasma parameter change due to NTM is analyzed using 1.5-dimensional transport code TOTAL. In ST plasma, magnetic island at 3/2 mode grows by bootstrap current and the central temperature decreases. If NRHF is added, the effect of bootstrap current decreases and NTM is stabilized.

  1. Scrape Off Layer Flow Characteristics in the Spherical Tokamak QUEST

    NASA Astrophysics Data System (ADS)

    Banerjee, Santanu; Zushi, H.; Nishino, N.; Mahira, Y.; Nagaoka, K.; Mishra, K.; Tashima, S.; Nagashima, Y.; Hanada, K.; Nakamura, K.; Idei, H.; Hasegawa, M.; Fujisawa, A.; Matsuoka, K.

    Poloidal component of the plasma flow along the field lines in the scrape off layer (SOL) is measured with fast visible imaging in QUEST. With the injection of electron cyclotron waves (ECW) a distinct mode appears at 0.75-0.8 kHz. Fluctuation structures feature long range correlation and propagate both poloidally and radially. Strong cross-field transport associated with the change in plasma configuration from inboard limiter towards inboard poloidal null-like, are most likely the reasons behind this ECW induced flow.

  2. Key Aspects of EBW Heating and Current Drive in Tokamaks

    NASA Astrophysics Data System (ADS)

    Urban, Jakub; Decker, Joan; Preinhaelter, Josef; Taylor, Gary; Vahala, Linda; Vahala, George

    2010-11-01

    Electron Bernstein wave (EBW) heating and current drive is modeled by coupled mode conversion, ray-tracing (AMR) and Fokker-Planck (LUKE) codes. Deposition and current drive profiles are determined for EBW with various injection parameters under realistic spherical tokamak conditions. There parameters are varied to investigate the robustness of the applied scenarios. The importance of relativistic corrections to EBW absorption is considered. The differences between various relativistic models are explored.

  3. Transport of Energetic Ions during Relaxation Oscillations in Plasmas of Spherical Tori

    SciTech Connect

    Kolesnichenko, Ya.I.; Lutsenko, V.V.; White, R.B.; Yakovenko, Yu.V.

    2001-02-12

    It is shown that the confinement of trapped energetic ions in spherical tokamaks during MHD events, such as sawtooth oscillations and internal reconnection events (IRE), essentially depends on beta (beta is the ratio of plasma pressure to the magnetic field pressure). Namely, when beta is high, the energetic ions are expelled from the plasma core, whereas they are only weakly influenced by the MHD activity at relatively low beta. Because of this, moderate rather than the highest possible beta may be preferable in spherical tokamaks.

  4. Texas Experimental Tokamak

    SciTech Connect

    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.

  5. Completely bootstrapped tokamak

    SciTech Connect

    Weening, R.H. ); Boozer, A.H. )

    1992-01-01

    Numerical simulations of the evolution of large-scale magnetic fields have been developed using a mean-field Ohm's law. The Ohm's law is coupled to a {Delta}{prime} stabilty analysis and a magnetic island growth equation in order to simulate the behavior of tokamak plasmas that are subject to tearing modes. In one set of calculations, the magnetohydrodynamic (MHD)-stable regime of the tokamak is examined via the construction of an {ital l}{sub {ital i}} -{ital q}{sub {ital a}} diagram. The results confirm previous calculations that show that tearing modes introduce a stability boundary into the {ital l}{sub {ital i}} -{ital q}{sub {ital a}} space. In another series of simulations, the interaction between tearing modes and the bootstrap current is investigated. The results indicate that a completely bootstrapped tokamak may be possible, even in the absence of any externally applied loop voltage or current drive.

  6. Recent Progress on Spherical Torus Research and Implications for Fusion Energy Development Path

    NASA Astrophysics Data System (ADS)

    Ono, Masayuki

    2014-10-01

    The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A =R0 / a) reduced to A near 1.5, well below the normal tokamak operating range of A equal to 2.5 or greater. As the aspect ratio is reduced, the ideal tokamak beta (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural plasma elongation which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to the longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in the UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all areas of fusion research, including fundamental fusion energy science as well as technological innovation. These results suggest exciting future prospects for ST research in both the near and longer term. The talk will summarize the key physics results from worldwide ST experiments, and describe ST community plans to provide the database for FNSF design while improving predictive capabilities for ITER and beyond. This work supported by DoE Contract No. DE-AC02-09CH11466.

  7. Tearing Modes in Tokamaks

    NASA Astrophysics Data System (ADS)

    White, R. B.

    2008-05-01

    This lecture gives a basic introduction to magnetic £elds, magnetic surface destruction, toroidal equilibrium and tearing modes in a tokamak, including the linear and nonlinear development of these modes and their modi£cation by current drive and bootstrap current, and sawtooth oscillations and disruptions.

  8. Transport in gyrokinetic tokamaks

    SciTech Connect

    Mynick, H.E.; Parker, S.E.

    1995-01-01

    A comprehensive study of transport in full-volume gyrokinetic (gk) simulations of ion temperature gradient driven turbulence in core tokamak plasmas is presented. Though this ``gyrokinetic tokamak`` is much simpler than experimental tokamaks, such simplicity is an asset, because a dependable nonlinear transport theory for such systems should be more attainable. Toward this end, we pursue two related lines of inquiry. (1) We study the scalings of gk tokamaks with respect to important system parameters. In contrast to real machines, the scalings of larger gk systems (a/{rho}{sub s} {approx_gt} 64) with minor radius, with current, and with a/{rho}{sub s} are roughly consistent with the approximate theoretical expectations for electrostatic turbulent transport which exist as yet. Smaller systems manifest quite different scalings, which aids in interpreting differing mass-scaling results in other work. (2) With the goal of developing a first-principles theory of gk transport, we use the gk data to infer the underlying transport physics. The data indicate that, of the many modes k present in the simulation, only a modest number (N{sub k} {approximately} 10) of k dominate the transport, and for each, only a handful (N{sub p} {approximately} 5) of couplings to other modes p appear to be significant, implying that the essential transport physics may be described by a far simpler system than would have been expected on the basis of earlier nonlinear theory alone. Part of this analysis is the inference of the coupling coefficients M{sub kpq} governing the nonlinear mode interactions, whose measurement from tokamak simulation data is presented here for the first time.

  9. High Beta Tokamaks

    SciTech Connect

    Cowley, S.

    1998-11-14

    Perhaps the ideal tokamak would have high {beta} ({beta} {approx}> 1) and classical confinement. Such a tokamak has not been found, and we do not know if one does exist. We have searched for such a possibility, so far without success. In 1990, we obtained analytic equilibrium solutions for large aspect ratio tokamaks at {beta} {approx} {Omicron}(1) [1]. These solutions and the extension at high {beta} poloidal to finite aspect ratio [2] provided a basis for the study of high {beta} tokamaks. We have shown that these configurations can be stable to short scale MHD modes [3], and that they have reduced neoclassical transport [4]. Microinstabilities (such as the {del}T{sub i} mode) seem to be stabilized at high {beta} [5] - this is due to the large local shear [3] and the magnetic well. We have some concerns about modes associated with the compressional branch which may appear at high {beta}. Bill Dorland and Mike Kotschenreuther have studied this issue and our concerns may be unfounded. It is certainly tantalizing, especially given the lowered neoclassical transport values, that these configurations could have no microinstabilities and, one could assume, no anomalous transport. Unfortunately, while this work is encouraging, the key question for high {beta} tokamaks is the stability to large scale kink modes. The MHD {beta} limit (Troyon limit) for kink modes at large aspect ratio is problematically low. There is ample evidence from computations that the limit exists. However, it is not known if stable equilibria exist at much higher {beta}--none have been found. We have explored this question in the asymptotic high {beta} poloidal limit. Unfortunately, we are unable to find stable equilibrium and also unable to show that they don't exist. The results of these calculations will be published when a more definitive answer is found.

  10. Next Step Spherical Torus Design Studies

    SciTech Connect

    C. Neumeyer; P. Heitzenroeder; C. Kessel; M. Ono; M. Peng; J. Schmidt; R. Woolley; I. Zatz

    2002-11-08

    Studies are underway to identify and characterize a design point for a Next Step Spherical Torus (NSST) experiment. This would be a ''Proof of Performance'' device which would follow and build upon the successes of the National Spherical Torus Experiment (NSTX) a ''Proof of Principle'' device which has operated at PPPL since 1999. With the Decontamination and Decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR) nearly completed, the TFTR test cell and facility will soon be available for a device such as NSST. By utilizing the TFTR test cell, NSST can be constructed for a relatively low cost on a short time scale. In addition, while furthering spherical torus (ST) research, this device could achieve modest fusion power gain for short-pulse lengths, a significant step toward future large burning plasma devices now under discussion in the fusion community. The selected design point is Q=2 at HH=1.4, P subscript ''fusion''=60 MW, 5 second pulse, with R subscript ''0''=1.5 m, A=1.6, I subscript ''p''=10vMA, B subscript ''t''=2.6 T, CS flux=16 weber. Most of the research would be conducted in D-D, with a limited D-T campaign during the last years of the program.

  11. Next-Step Spherical Torus Experiment and Spherical Torus Strategy in the Fusion Energy Development Path

    SciTech Connect

    M. Ono; M. Peng; C. Kessel; C. Neumeyer; J. Schmidt; J. Chrzanowski; D. Darrow; L. Grisham; P. Heitzenroeder; T. Jarboe; C. Jun; S. Kaye; J. Menard; R. Raman; T. Stevenson; M. Viola; J. Wilson; R. Woolley; I. Zatz

    2003-10-27

    A spherical torus (ST) fusion energy development path which is complementary to proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact Component Test Facility (CTF) and higher performance advanced regimes leading to more attractive DEMO and Power Plant scale reactors. To provide the physics basis for the CTF an intermediate step needs to be taken which we refer to as the ''Next Step Spherical Torus'' (NSST) device and examine in some detail herein. NSST is a ''performance extension'' (PE) stage ST with the plasma current of 5-10 MA, R = 1.5 m, and Beta(sub)T less than or equal to 2.7 T with flexible physics capability. The mission of NSST is to: (1) provide a sufficient physics basis for the design of CTF, (2) explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, and (3) contribute to the general plasma/fusion science of high beta toroidal plasmas. The NSST facility is designed to utilize the Tokamak Fusion Test Reactor (or similar) site to minimize the cost and time required for the design and construction.

  12. Wave Driven Fast Ion Loss in the National Spherical Torus Experiment

    SciTech Connect

    E.D. Fredrickson; C.Z. Cheng; D. Darrow; G. Fu; N.N. Gorelenkov; G. Kramer; S.S. Medley; J. Menard; L. Roquemore; D. Stutman; R.B. White

    2003-01-28

    Spherical tokamaks, with their relatively low toroidal field, extend fast-ion-driven instability physics to parameter ranges not normally accessed in conventional tokamaks. The low field means that both the fast-ion Larmor radius normalized to the plasma minor radius and the ratio of the fast-ion velocity to the Alfven speed are relatively large. The large Larmor radius of the ions enhances their interaction with instability modes, influencing the structure of the unstable mode spectrum. The relatively large fast-ion velocity allows for a larger population of fast ions to be in resonance with the mode, increasing the drive. It is therefore an important goal of the present proof-of-principle spherical tokamaks to evaluate the role of fast-ion-driven instabilities in fast-ion confinement. This paper presents the first observations of fast-ion losses resulting from toroidal Alfven eigenmodes and a new, fishbone-like, energetic particle mode.

  13. Spherical neutron generator

    DOEpatents

    Leung, Ka-Ngo

    2006-11-21

    A spherical neutron generator is formed with a small spherical target and a spherical shell RF-driven plasma ion source surrounding the target. A deuterium (or deuterium and tritium) ion plasma is produced by RF excitation in the plasma ion source using an RF antenna. The plasma generation region is a spherical shell between an outer chamber and an inner extraction electrode. A spherical neutron generating target is at the center of the chamber and is biased negatively with respect to the extraction electrode which contains many holes. Ions passing through the holes in the extraction electrode are focused onto the target which produces neutrons by D-D or D-T reactions.

  14. Tritium catalyzed deuterium tokamaks

    SciTech Connect

    Greenspan, E.; Miley, G.H.; Jung, J.; Gilligan, J.

    1984-04-01

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the /sup 3/He from the D(D,n)/sup 3/He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general).

  15. Equilibrium reconstruction in the START tokamak

    NASA Astrophysics Data System (ADS)

    Appel, L. C.; Bevir, M. K.; Walsh, M. J.

    2001-02-01

    The computation of magnetic equilibria in the START spherical tokamak is more difficult than those in more conventional large aspect ratio tokamaks. This difficulty arises partly as a result of the use of induction compression to generate high current plasma, as this precludes the positioning of magnetic diagnostics close to the outboard side of the plasma. In addition, the effect of a conducting wall with a high, but finite, conductivity must be included. A method is presented for obtaining plasma equilibrium reconstructions based on the EFIT code. New constraints are used to relate isoflux surface locations deduced from radial profile measurements of electron temperature. A model of flux diffusion through the vessel wall is developed. It is shown that neglecting flux diffusion in the vessel wall can lead to a significant underestimate in the calculation of the plasma βt. Using a relatively sparse set of magnetic signals, βt can be obtained to within a fractional error of +/-10%. Using constraints to relate isoflux surface locations, the principle involved in determining the internal q profile is demonstrated.

  16. Magnetic diagnostics for the lithium tokamak experiment.

    PubMed

    Berzak, L; Kaita, R; Kozub, T; Majeski, R; Zakharov, L

    2008-10-01

    The lithium tokamak experiment (LTX) is a spherical tokamak with R(0)=0.4 m, a=0.26 m, B(TF) approximately 3.4 kG, I(P) approximately 400 kA, and pulse length approximately 0.25 s. The focus of LTX is to investigate the novel low-recycling lithium wall operating regime for magnetically confined plasmas. This regime is reached by placing an in-vessel shell conformal to the plasma last closed flux surface. The shell is heated and then coated with liquid lithium. An extensive array of magnetic diagnostics is available to characterize the experiment, including 80 Mirnov coils (single and double axis, internal and external to the shell), 34 flux loops, 3 Rogowskii coils, and a diamagnetic loop. Diagnostics are specifically located to account for the presence of a secondary conducting surface and engineered to withstand both high temperatures and incidental contact with liquid lithium. The diagnostic set is therefore fabricated from robust materials with heat and lithium resistance and is designed for electrical isolation from the shell and to provide the data required for highly constrained equilibrium reconstructions. PMID:19044600

  17. Energy confinement in tokamaks

    SciTech Connect

    Sugihara, M.; Singer, C.

    1986-08-01

    A straightforward generalization is made of the ohmic heating energy confinement scalings of Pfeiffer and Waltz and Blackwell et. al. The resulting model is systematically calibrated to published data from limiter tokamaks with ohmic, electron cyclotron, and neutral beam heating. With considerably fewer explicitly adjustable free parameters, this model appears to give a better fit to the available data for limiter discharges than the combined ohmic/auxiliary heating model of Goldston.

  18. Mechanisms of Stochastic Diffusion of Energetic Ions in Spherical Tori

    SciTech Connect

    Ya.I. Kolesnichenko; R.B. White; Yu.V. Yakovenko

    2001-01-18

    Stochastic diffusion of the energetic ions in spherical tori is considered. The following issues are addressed: (I) Goldston-White-Boozer diffusion in a rippled field; (ii) cyclotron-resonance-induced diffusion caused by the ripple; (iii) effects of non-conservation of the magnetic moment in an axisymmetric field. It is found that the stochastic diffusion in spherical tori with a weak magnetic field has a number of peculiarities in comparison with conventional tokamaks; in particular, it is characterized by an increased role of mechanisms associated with non-conservation of the particle magnetic moment. It is concluded that in current experiments on National Spherical Torus eXperiment (NSTX) the stochastic diffusion does not have a considerable influence on the confinement of energetic ions.

  19. Wide scanning spherical antenna

    NASA Technical Reports Server (NTRS)

    Shen, Bing (Inventor); Stutzman, Warren L. (Inventor)

    1995-01-01

    A novel method for calculating the surface shapes for subreflectors in a suboptic assembly of a tri-reflector spherical antenna system is introduced, modeled from a generalization of Galindo-Israel's method of solving partial differential equations to correct for spherical aberration and provide uniform feed to aperture mapping. In a first embodiment, the suboptic assembly moves as a single unit to achieve scan while the main reflector remains stationary. A feed horn is tilted during scan to maintain the illuminated area on the main spherical reflector fixed throughout the scan thereby eliminating the need to oversize the main spherical reflector. In an alternate embodiment, both the main spherical reflector and the suboptic assembly are fixed. A flat mirror is used to create a virtual image of the suboptic assembly. Scan is achieved by rotating the mirror about the spherical center of the main reflector. The feed horn is tilted during scan to maintain the illuminated area on the main spherical reflector fixed throughout the scan.

  20. Polarization spectroscopy of tokamak plasmas

    SciTech Connect

    Wroblewski, D.

    1991-09-01

    Measurements of polarization of spectral lines emitted by tokamak plasmas provide information about the plasma internal magnetic field and the current density profile. The methods of polarization spectroscopy, as applied to the tokamak diagnostic, are reviewed with emphasis on the polarimetry of motional Stark effect in hydrogenic neutral beam emissions. 25 refs., 7 figs.

  1. Gyrokinetics for high-frequency modes in tokamaks

    NASA Astrophysics Data System (ADS)

    Wang, Z. T.; Wang, L.; Long, L. X.; Dong, J. Q.; He, Zhixiong; Liu, Y.; Tang, C. J.

    2012-07-01

    Gyrokinetics for high-frequency modes in tokamaks is developed. It is found that the breakdown of the invariants by perturbed electromagnetic fields drives microinstability. The obtained diamagnetic frequency, ω∗, is proportional to only the toroidal mode number rather than transverse mode numbers. Therefore, there is no nonadiabatic drive for axisymmetrical modes in gyrokinetics. Meanwhile, the conventional eikonal Ansatz breaks down for the axisymmetrical modes. The ion drift-cyclotron instability discovered in a mirror machine is found for the first time in the toroidal system. The growth rates are proportional to ρi/Ln, and the slope changes with magnetic curvature. In spherical torus, where magnetic curvature is greater than that of traditional tokamaks, instability poses a potential danger to such devices.

  2. Gyrokinetics for high-frequency modes in tokamaks

    SciTech Connect

    Wang, Z. T.; Long, L. X.; Dong, J. Q.; He, Zhixiong; Wang, L.; Liu, Y.; Tang, C. J.

    2012-07-15

    Gyrokinetics for high-frequency modes in tokamaks is developed. It is found that the breakdown of the invariants by perturbed electromagnetic fields drives microinstability. The obtained diamagnetic frequency, {omega}{sup *}, is proportional to only the toroidal mode number rather than transverse mode numbers. Therefore, there is no nonadiabatic drive for axisymmetrical modes in gyrokinetics. Meanwhile, the conventional eikonal Ansatz breaks down for the axisymmetrical modes. The ion drift-cyclotron instability discovered in a mirror machine is found for the first time in the toroidal system. The growth rates are proportional to {rho}{sub i}/L{sub n}, and the slope changes with magnetic curvature. In spherical torus, where magnetic curvature is greater than that of traditional tokamaks, instability poses a potential danger to such devices.

  3. Large displacement spherical joint

    DOEpatents

    Bieg, Lothar F.; Benavides, Gilbert L.

    2002-01-01

    A new class of spherical joints has a very large accessible full cone angle, a property which is beneficial for a wide range of applications. Despite the large cone angles, these joints move freely without singularities.

  4. Excitation, propagation, and damping of electron Bernstein waves in tokamaks

    NASA Astrophysics Data System (ADS)

    Ram, A. K.; Schultz, S. D.

    2000-10-01

    The conventional ordinary O-mode and the extraordinary X-mode in the electron cyclotron range of frequencies are not suitable for core heating in high-β spherical tokamak plasmas, like the National Spherical Torus Experiment [M. Ono, S. Kaye, M. Peng et al., in Proceedings of the 17th International Atomic Energy Agency Fusion Energy Conference (International Atomic Energy Agency, Vienna, 1999), Vol. 3, p. 1135], as they are weakly damped at high harmonics of the electron cyclotron frequency. However, electron Bernstein waves (EBW) can be effective for heating and driving currents in spherical tokamak plasmas. Power can be coupled to EBWs via mode conversion of either the X-mode or the O-mode. The two mode conversions are optimized in different regions of the parameter space spanned by the parallel wavelength and wave frequency. The conditions for optimized mode conversion to EBWs are evaluated analytically and numerically using a cold plasma model and an approximate kinetic model. From geometric optics ray tracing it is found that the EBWs damp strongly near the Doppler-broadened resonance at harmonics of the electron cyclotron frequency.

  5. The tokamak as a neutron source

    SciTech Connect

    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.

  6. Physics Basis for a Spherical Torus Power Plant

    SciTech Connect

    C.E. Kessel; J. Menard; S.C. Jardin; T.K. Mau; et al

    1999-11-01

    The spherical torus, or low-aspect-ratio tokamak, is considered as the basis for a fusion power plant. A special class of wall-stabilized high-beta high-bootstrap fraction low-aspect-ratio tokamak equilibrium are analyzed with respect to MHD stability, bootstrap current and external current drive, poloidal field system requirements, power and particle exhaust and plasma operating regime. Overall systems optimization leads to a choice of aspect ratio A = 1:6, plasma elongation kappa = 3:4, and triangularity delta = 0:64. The design value for the plasma toroidal beta is 50%, corresponding to beta N = 7:4, which is 10% below the ideal stability limit. The bootstrap fraction of 99% greatly alleviates the current drive requirements, which are met by tangential neutral beam injection. The design is such that 45% of the thermal power is radiated in the plasma by Bremsstrahlung and trace Krypton, with Neon in the scrapeoff layer radiating the remainder.

  7. High power heating of magnetic reconnection in merging tokamak experimentsa)

    NASA Astrophysics Data System (ADS)

    Ono, Y.; Tanabe, H.; Yamada, T.; Gi, K.; Watanabe, T.; , T., Ii; Gryaznevich, M.; Scannell, R.; Conway, N.; Crowley, B.; Michael, C.

    2015-05-01

    Significant ion/electron heating of magnetic reconnection up to 1.2 keV was documented in two spherical tokamak plasma merging experiment on MAST with the significantly large Reynolds number R˜105. Measured 1D/2D contours of ion and electron temperatures reveal clearly energy-conversion mechanisms of magnetic reconnection: huge outflow heating of ions in the downstream and localized heating of electrons at the X-point. Ions are accelerated up to the order of poloidal Alfven speed in the reconnection outflow region and are thermalized by fast shock-like density pileups formed in the downstreams, in agreement with recent solar satellite observations and PIC simulation results. The magnetic reconnection efficiently converts the reconnecting (poloidal) magnetic energy mostly into ion thermal energy through the outflow, causing the reconnection heating energy proportional to square of the reconnecting (poloidal) magnetic field Brec2 ˜ Bp2. The guide toroidal field Bt does not affect the bulk heating of ions and electrons, probably because the reconnection/outflow speeds are determined mostly by the external driven inflow by the help of another fast reconnection mechanism: intermittent sheet ejection. The localized electron heating at the X-point increases sharply with the guide toroidal field Bt, probably because the toroidal field increases electron confinement and acceleration length along the X-line. 2D measurements of magnetic field and temperatures in the TS-3 tokamak merging experiment also reveal the detailed reconnection heating mechanisms mentioned above. The high-power heating of tokamak merging is useful not only for laboratory study of reconnection but also for economical startup and heating of tokamak plasmas. The MAST/TS-3 tokamak merging with Bp > 0.4 T will enables us to heat the plasma to the alpha heating regime: Ti > 5 keV without using any additional heating facility.

  8. On steady poloidal and toroidal flows in tokamak plasmas

    SciTech Connect

    McClements, K. G.

    2010-08-15

    The effects of poloidal and toroidal flows on tokamak plasma equilibria are examined in the magnetohydrodynamic limit. ''Transonic'' poloidal flows of the order of the sound speed multiplied by the ratio of poloidal magnetic field to total field B{sub {theta}/}B can cause the (normally elliptic) Grad-Shafranov (GS) equation to become hyperbolic in part of the solution domain. It is pointed out that the range of poloidal flows for which the GS equation is hyperbolic increases with plasma beta and B{sub {theta}/}B, thereby complicating the problem of determining spherical tokamak plasma equilibria with transonic poloidal flows. It is demonstrated that the calculation of the hyperbolicity criterion can be easily modified when the assumption of isentropic flux surfaces is replaced with the more tokamak-relevant one of isothermal flux surfaces. On the basis of the latter assumption, a simple expression is obtained for the variation of density on a flux surface when poloidal and toroidal flows are simultaneously present. Combined with Thomson scattering measurements of density and temperature, this expression could be used to infer information on poloidal and toroidal flows on the high field side of a tokamak plasma, where direct measurements of flows are not generally possible. It is demonstrated that there are four possible solutions of the Bernoulli relation for the plasma density when the flux surfaces are assumed to be isothermal, corresponding to four distinct poloidal flow regimes. Finally, observations and first principles-based theoretical modeling of poloidal flows in tokamak plasmas are briefly reviewed and it is concluded that there is no clear evidence for the occurrence of supersonic poloidal flows.

  9. Recent developments in Bayesian inference of tokamak plasma equilibria and high-dimensional stochastic quadratures

    NASA Astrophysics Data System (ADS)

    von Nessi, G. T.; Hole, M. J.; The MAST Team

    2014-11-01

    We present recent results and technical breakthroughs for the Bayesian inference of tokamak equilibria using force-balance as a prior constraint. Issues surrounding model parameter representation and posterior analysis are discussed and addressed. These points motivate the recent advancements embodied in the Bayesian Equilibrium Analysis and Simulation Tool (BEAST) software being presently utilized to study equilibria on the Mega-Ampere Spherical Tokamak (MAST) experiment in the UK (von Nessi et al 2012 J. Phys. A 46 185501). State-of-the-art results of using BEAST to study MAST equilibria are reviewed, with recent code advancements being systematically presented though out the manuscript.

  10. Dust Measurements in Tokamaks

    SciTech Connect

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-04-23

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  11. ITER tokamak device

    NASA Astrophysics Data System (ADS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-07-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER, a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fueling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  12. Sawtooth oscillation in tokamaks

    SciTech Connect

    Park, W.; Monticello, D.A.

    1989-03-01

    A three-dimensional nonlinear toroidal full MHD code, MH3D, has been used to study sawtooth oscillations in tokamaks. The profile evolution during the sawtooth crash phase compares well with experiment, but only if neoclassical resistivity is used in the rise phase. (Classical resistivity has been used in most of the previous theoretical sawtooth studies.) With neoclassical resistivity, the q value at the axis drops from 1 to about 0.8 before the crash phase, and then resets to 1 through a Kadomtsev-type complete reconnection process. This ..delta..q/sub 0/ approx. = 0.2 is much larger than ..delta..q/sub o/ approx. = 0.01, which is obtained if classical resistivity is used. The current profile is strongly peaked at the axis with a flat region around the singular surface, and is similar to the Textor profile. To understand this behavior, approximate formulas for the time behavior of current and q values are derived. A functional dependence of sawtooth period scaling is also derived. A semi-empirical scaling is found which fits the experimental data from various tokamaks. Some evidence is presented which indicates that the fast crash time is due to enhanced effective resistivity inside the singular current sheet, generated by, e.g., microinstability and electron parallel viscosity with stochastic fields at the x-point. 16 refs., 5 figs.

  13. Experimental setup for tungsten transport studies at the NSTX tokamak

    SciTech Connect

    Clementson, J.; Beiersdorfer, P.; Roquemore, A. L.; Skinner, C. H.; Mansfield, D. K.; Hartzfeld, K.; Lepson, J. K.

    2010-10-15

    Tungsten particles have been introduced into the National Spherical Torus Experiment (NSTX) in Princeton with the purpose to investigate the effects of tungsten injection on subsequent plasma discharges. An experimental setup for the study of tungsten particle transport is described where the particles are introduced into the tokamak using a modified particle dropper, otherwise used for lithium-powder injection. An initial test employing a grazing-incidence extreme ultraviolet spectrometer demonstrates that the tungsten-transport setup could serve to infer particle transport from the edge to the hot central plasmas of NSTX.

  14. Spherical torus fusion reactor

    DOEpatents

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  15. Spherical geodesic mesh generation

    SciTech Connect

    Fung, Jimmy; Kenamond, Mark Andrew; Burton, Donald E.; Shashkov, Mikhail Jurievich

    2015-02-27

    In ALE simulations with moving meshes, mesh topology has a direct influence on feature representation and code robustness. In three-dimensional simulations, modeling spherical volumes and features is particularly challenging for a hydrodynamics code. Calculations on traditional spherical meshes (such as spin meshes) often lead to errors and symmetry breaking. Although the underlying differencing scheme may be modified to rectify this, the differencing scheme may not be accessible. This work documents the use of spherical geodesic meshes to mitigate solution-mesh coupling. These meshes are generated notionally by connecting geodesic surface meshes to produce triangular-prismatic volume meshes. This mesh topology is fundamentally different from traditional mesh topologies and displays superior qualities such as topological symmetry. This work describes the geodesic mesh topology as well as motivating demonstrations with the FLAG hydrocode.

  16. Causes of major tokamak disruptions

    SciTech Connect

    White, R.B.; Monticello, D.A.

    1980-07-01

    The nonlinear saturation theory of the tearing mode is used to examine the necessary conditions for the occurrence of a major tokamak disruption. The results are compared with full three-dimensional numerical simulations, and with experimental data.

  17. The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    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{sup 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.

  18. Retroreflector spherical satellite

    NASA Astrophysics Data System (ADS)

    Akentyev, A. S.; Vasiliev, V. P.; Sadovnikov, M. A.; Sokolov, A. L.; Shargorodskiy, V. D.

    2015-10-01

    Specific features of spherical retroreflector arrays for high-precision laser ranging are considered, and errors in distance measurements are analyzed. A version of a glass retroreflector satellite with a submillimeter "target error" is proposed. Its corner cube reflectors are located in depressions to reduce the working angular aperture, and their faces have a dielectric interference coating.

  19. Hollow spherical supramolecular dendrimers.

    PubMed

    Percec, Virgil; Peterca, Mihai; Dulcey, Andrés E; Imam, Mohammad R; Hudson, Steven D; Nummelin, Sami; Adelman, Peter; Heiney, Paul A

    2008-10-01

    The synthesis of a library containing 12 conical dendrons that self-assemble into hollow spherical supramolecular dendrimers is reported. The design principles for this library were accessed by development of a method that allows the identification of hollow spheres, followed by structural and retrostructural analysis of their Pm3n cubic lattice. The first hollow spherical supramolecular dendrimer was made by replacing the tapered dendron, from the previously reported tapered dendritic dipeptide that self-assembled into helical pores, with its constitutional isomeric conical dendron. This strategy generated a conical dendritic dipeptide that self-assembled into a hollow spherical supramolecular dendrimer that self-organizes in a Pm3n cubic lattice. Other examples of hollow spheres were assembled from conical dendrons without a dipeptide at their apex. These are conical dendrons originated from tapered dendrons containing additional benzyl ether groups at their apex. The inner part of the hollow sphere assembled from the dipeptide resembles the path of a spherical helix or loxodrome and, therefore, is chiral. The spheres assembled from other conical dendrons are nonhelical, even when they contain stereocenters on the alkyl groups from their periphery. Functionalization of the apex of the conical dendrons with diethylene glycol allowed the encapsulation of LiOTf and RbOTf in the center of the hollow sphere. These experiments showed that hollow spheres function as supramolecular dendritic capsules and therefore are expected to display functions complementary to those of other related molecular and supramolecular structures. PMID:18771261

  20. Spherical ion source

    NASA Technical Reports Server (NTRS)

    Hall, L. G.

    1969-01-01

    Radial focusing of electrons in ion source produces greater ion densities, resulting in higher resolution and focus capability for a given source volume. Electron beam is focused near exit aperture by spherical fields. High density ions allow focusing ion beam to high density at echo, allowing high current through small aperture.

  1. Spherical mirror mount

    NASA Technical Reports Server (NTRS)

    Meyer, Jay L. (Inventor); Messick, Glenn C. (Inventor); Nardell, Carl A. (Inventor); Hendlin, Martin J. (Inventor)

    2011-01-01

    A spherical mounting assembly for mounting an optical element allows for rotational motion of an optical surface of the optical element only. In that regard, an optical surface of the optical element does not translate in any of the three perpendicular translational axes. More importantly, the assembly provides adjustment that may be independently controlled for each of the three mutually perpendicular rotational axes.

  2. Spherical colloidal photonic crystals.

    PubMed

    Zhao, Yuanjin; Shang, Luoran; Cheng, Yao; Gu, Zhongze

    2014-12-16

    CONSPECTUS: Colloidal photonic crystals (PhCs), periodically arranged monodisperse nanoparticles, have emerged as one of the most promising materials for light manipulation because of their photonic band gaps (PBGs), which affect photons in a manner similar to the effect of semiconductor energy band gaps on electrons. The PBGs arise due to the periodic modulation of the refractive index between the building nanoparticles and the surrounding medium in space with subwavelength period. This leads to light with certain wavelengths or frequencies located in the PBG being prohibited from propagating. Because of this special property, the fabrication and application of colloidal PhCs have attracted increasing interest from researchers. The most simple and economical method for fabrication of colloidal PhCs is the bottom-up approach of nanoparticle self-assembly. Common colloidal PhCs from this approach in nature are gem opals, which are made from the ordered assembly and deposition of spherical silica nanoparticles after years of siliceous sedimentation and compression. Besides naturally occurring opals, a variety of manmade colloidal PhCs with thin film or bulk morphology have also been developed. In principle, because of the effect of Bragg diffraction, these PhC materials show different structural colors when observed from different angles, resulting in brilliant colors and important applications. However, this angle dependence is disadvantageous for the construction of some optical materials and devices in which wide viewing angles are desired. Recently, a series of colloidal PhC materials with spherical macroscopic morphology have been created. Because of their spherical symmetry, the PBGs of spherical colloidal PhCs are independent of rotation under illumination of the surface at a fixed incident angle of the light, broadening the perspective of their applications. Based on droplet templates containing colloidal nanoparticles, these spherical colloidal PhCs can be

  3. Magnetized plasma flow injection into tokamak and high-beta compact torus plasmas

    NASA Astrophysics Data System (ADS)

    Matsunaga, Hiroyuki; Komoriya, Yuuki; Tazawa, Hiroyasu; Asai, Tomohiko; Takahashi, Tsutomu; Steinhauer, Loren; Itagaki, Hirotomo; Onchi, Takumi; Hirose, Akira

    2010-11-01

    As an application of a magnetized coaxial plasma gun (MCPG), magnetic helicity injection via injection of a highly elongated compact torus (magnetized plasma flow: MPF) has been conducted on both tokamak and field-reversed configuration (FRC) plasmas. The injected plasmoid has significant amounts of helicity and particle contents and has been proposed as a fueling and a current drive method for various torus systems. In the FRC, MPF is expected to generate partially spherical tokamak like FRC equilibrium by injecting a significant amount of magnetic helicity. As a circumstantial evidence of the modified equilibrium, suppressed rotational instability with toroidal mode number n = 2. MPF injection experiments have also been applied to the STOR-M tokamak as a start-up and current drive method. Differences in the responses of targets especially relation with beta value and the self-organization feature will be studied.

  4. Plasma current start-up by the outer ohmic heating coils in the Saskatchewan TORus Modified (STOR-M) iron core tokamak.

    PubMed

    Mitarai, O; Xiao, C; McColl, D; Dreval, M; Hirose, A; Peng, M

    2015-03-01

    A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. This result suggests a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. The effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments in the STOR-M tokamak. PMID:25832230

  5. Wave Driven Fast Ion Loss in the National Spherical Torus Experiment

    SciTech Connect

    E.D. Fredrickson; C.Z. Cheng; D. Darrow; G. Fu; N.N. Gorelenkov; G. Kramer; S.S. Medley; J. Menard; L. Roquemore; D. Stutman; R.B. White

    2003-08-05

    The study of fast ion instabilities in conventional aspect ratio tokamaks is motivated in large part by their potential to negatively impact the ignition threshold in fusion reactors by causing fast ion losses. Spherical tokamak's (ST), with intrinsically low magnetic fields, are particularly susceptible to fast ion driven instabilities. The 3.5 MeV alpha's from the D-T [deuterium-tritium] fusion reaction in proposed ST reactors will have velocities much higher than the Alfven speed. The Larmor radius of the fusion alphas, normalized to the plasma size, will also be larger than for conventional aspect ratio tokamak reactors. The resulting longer wavelengths of the *AE instabilities will be more effective in driving fast ion loss. The change in magnetic topology also influences the mode structure, as in the case of the Compressional Alfven Eigenmodes (CAE) seen on NSTX.

  6. Hollow spherical shell manufacture

    DOEpatents

    O'Holleran, T.P.

    1991-11-26

    A process is disclosed for making a hollow spherical shell of silicate glass composition in which an aqueous suspension of silicate glass particles and an immiscible liquid blowing agent is placed within the hollow spherical cavity of a porous mold. The mold is spun to reduce effective gravity to zero and to center the blowing agent, while being heated so as to vaporize the immiscible liquid and urge the water carrier of the aqueous suspension to migrate into the body of the mold, leaving a green shell compact deposited around the mold cavity. The green shell compact is then removed from the cavity, and is sintered for a time and a temperature sufficient to form a silicate glass shell of substantially homogeneous composition and uniform geometry. 3 figures.

  7. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-04-04

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  8. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  9. Hollow spherical shell manufacture

    DOEpatents

    O'Holleran, Thomas P.

    1991-01-01

    A process for making a hollow spherical shell of silicate glass composition in which an aqueous suspension of silicate glass particles and an immiscible liquid blowing agent is placed within the hollow spherical cavity of a porous mold. The mold is spun to reduce effective gravity to zero and to center the blowing agent, while being heated so as to vaporize the immiscible liquid and urge the water carrier of the aqueous suspension to migrate into the body of the mold, leaving a green shell compact deposited around the mold cavity. The green shell compact is then removed from the cavity, and is sintered for a time and a temperature sufficient to form a silicate glass shell of substantially homogeneous composition and uniform geometry.

  10. Spherical nitroguandine process

    SciTech Connect

    Sanchez, J.A.; Roemer, E.L.; Stretz, L.A.

    1990-10-30

    A process of preparing spherical high bulk density nitroguanidine by dissolving low bulk density nitroguanidine in N-methyl pyrrolidone at elevated temperatures and then cooling the solution to lower temperatures as a liquid characterized as a nonsolvent for the nitroguanidine is presented. The process is enhanced by inclusion in the solution of from about 1 ppm up to about 250 ppm of a metal salt such as nickel nitrate, zinc nitrate or chromium nitrate, preferably from about 20 to about 50 ppm.

  11. Spherical nitroguanidine process

    DOEpatents

    Sanchez, John A.; Roemer, Edward L.; Stretz, Lawrence A.

    1990-01-01

    A process of preparing spherical high bulk density nitroguanidine by dissing low bulk density nitroguanidine in N-methyl pyrrolidone at elevated temperatures and then cooling the solution to lower temperatures as a liquid characterized as a nonsolvent for the nitroguanidine is provided. The process is enhanced by inclusion in the solution of from about 1 ppm up to about 250 ppm of a metal salt such as nickel nitrate, zinc nitrate or chromium nitrate, preferably from about 20 to about 50 ppm.

  12. Spherical electrostatic electron spectrometer

    NASA Astrophysics Data System (ADS)

    Yang, T.-S.; Kolk, B.; Kachnowski, T.; Trooster, J.; Benczer-Koller, N.

    1982-06-01

    A high transmission, low energy spherical electrostatic electron spectrometer particularly suited to the geometry required for Mössbauer-conversion electron spectroscopy was built. A transmission of 13% at an energy resolution of 2% was obtained with an 0.5 cm diameter source of 13.6 keV electrons. Applications to the study of hyperfine interactions of surfaces and interfaces are discussed.

  13. Moving Divertor Plates in a Tokamak

    SciTech Connect

    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.

  14. Design Features and Commissioning of the Versatile Experiment Spherical Torus (VEST) at Seoul National University

    NASA Astrophysics Data System (ADS)

    J. Chung, K.; H. An, Y.; K. Jung, B.; Y. Lee, H.; C., Sung; S. Na, Y.; S. Hahm, T.; S. Hwang, Y.

    2013-03-01

    A new spherical torus called VEST (Versatile Experiment Spherical Torus) is designed, constructed and successfully commissioned at Seoul National University. A unique design feature of the VEST is two partial solenoid coils installed at both vertical ends of a center stack, which can provide sufficient magnetic fluxes to initiate tokamak plasmas while keeping a low aspect ratio configuration in the central region. According to initial double null merging start-up scenario using the partial solenoid coils, appropriate power supplies for driving a toroidal field coil, outer poloidal field coils, and the partial solenoid coils are fabricated and successfully commissioned. For reliable start-up, a pre-ionization system with two cost-effective homemade magnetron power supplies is also prepared. In addition, magnetic and spectroscopic diagnostics with appropriate data acquisition and control systems are well prepared for initial operation of the device. The VEST is ready for tokamak plasma operation by completing and commissioning most of the designed components.

  15. Ohmic Flux Consumption During Initial Operation of the NSTX Spherical Torus

    SciTech Connect

    J. Menard; B. LeBlanc; S.A. Sabbagh; M. Bell; R. Bell; E. Fredrickson; D. Gates; S. Jardin; S. Kaye; H. Kugel; R. Maingi; R. Maqueda; D. Mueller; M. Ono; S.Paul; C.H. Skinner; D. Stutman; and the NSTX Research Team.

    2000-10-05

    The spherical tokamak (ST), because of its slender central column, has very limited volt-second capability relative to a standard aspect ratio tokamak of similar plasma cross-section. Recent experiments on the National Spherical Torus Experiment (NSTX) have begun to quantify and optimize the ohmic current drive efficiency in a MA-class ST device. Sustainable ramp-rates in excess of 5MA/sec during the current rise phase have been achieved on NSTX, while faster ramps generate significant MHD activity. Discharges with Ip exceeding 1MA have been achieved in NSTX with nominal parameters: aspect ratio A=1.3--1.4, elongation k=2--2.2, triangularity d=0.4, internal inductance li=0.6, and Ejima coefficient CE=0.35. Flux consumption efficiency results, performance improvements associated with first boronization, and comparisons to neoclassical resistivity are described.

  16. Status of tokamak experiments

    SciTech Connect

    Wolf, G.H.

    1996-03-01

    Plasma-wall interaction, heat removal and ash exhaust have emerged as the dominant problems still to be solved in order to achieve ignition and - even more difficult - to maintain a state of self-sustained thermo-nuclear burn. This is of particular delicacy, since those operational regimes which yield the best energy confinement correspond to an even better particle confinement and confinement of impurities, which then tend to accumulate in the plasma core and to result in disruption or degradation of the tokamak discharge. Therefore, plasma-wall interaction, heat removal and particle exhaust will determine not only the structure and configuration of the plasma edge region, of the wall system and of the materials facing the plasma, but also the final choice of useful confinement regimes. Moreover, the potential effect of powerful {alpha}-particle heating on plasma stability and confinement has to be kept below critical values. For the latter requirement, a final answer can only be obtained in an ITER-type device where ignition and burn will become accessible. 72 refs., 12 figs.

  17. Resistive instabilities in tokamaks

    SciTech Connect

    Rutherford, P.H.

    1985-10-01

    Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed.

  18. Impedance of an intense plasma-cathode electron source for tokamak startup

    NASA Astrophysics Data System (ADS)

    Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Perry, J. M.

    2016-05-01

    An impedance model is formulated and tested for the ˜1 kV , 1 kA/cm2 , arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma ( narc≈1021 m-3 ) within the electron source, and the less dense external tokamak edge plasma ( nedge≈1018 m-3 ) into which current is injected at the applied injector voltage, Vinj . Experiments on the Pegasus spherical tokamak show that the injected current, Iinj , increases with Vinj according to the standard double layer scaling Iinj˜Vinj3 /2 at low current and transitions to Iinj˜Vinj1 /2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb˜Iinj/Vinj1 /2 . For low tokamak edge density nedge and high Iinj , the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb˜narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.

  19. Impedance of an intense plasma-cathode electron source for tokamak startup

    DOE PAGESBeta

    Hinson, Edward Thomas; Barr, Jayson L.; Bongard, Michael W.; Burke, Marcus Galen; Fonck, Raymond J.; Perry, Justin M.

    2016-05-31

    In this study, an impedance model is formulated and tested for the ~1kV, ~1kA/cm2, arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma (narc ≈ 1021 m-3) within the electron source, and the less dense external tokamak edge plasma (nedge ≈ 1018 m-3) into which current is injected at the applied injector voltage, Vinj. Experiments on the Pegasus spherical tokamak show the injected current, Iinj, increases with Vinj according to the standard double layer scaling Iinj ~ Vinj3/2 at low current and transitions to Iinj ~ Vinj1/2more » at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb ~ Iinj/Vinj1/2. For low tokamak edge density nedge and high Iinj, the inferred beam density nb is consistent with the requirement nb ≤ nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb ~ narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.« less

  20. Progress Towards High Performance, Steady-state Spherical Torus

    SciTech Connect

    M. Ono; M.G. Bell; R.E. Bell; T. Bigelow; M. Bitter; W. Blanchard; J. Boedo; C. Bourdelle; C. Bush; W. Choe; J. Chrzanowski; D.S. Darrow; S.J. Diem; R. Doerner; P.C. Efthimion; J.R. Ferron; R.J. Fonck; E.D. Fredrickson; G.D. Garstka; D.A. Gates; T. Gray; L.R. Grisham; W. Heidbrink; K.W. Hill; D. Hoffman; T.R. Jarboe; D.W. Johnson; R. Kaita; S.M. Kaye; C. Kessel; J.H. Kim; M.W. Kissick; S. Kubota; H.W. Kugel; B.P. LeBlanc; K. Lee; S.G. Lee; B.T. Lewicki; S. Luckhardt; R. Maingi; R. Majeski; J. Manickam; R. Maqueda; T.K. Mau; E. Mazzucato; S.S. Medley; J. Menard; D. Mueller; B.A. Nelson; C. Neumeyer; N. Nishino; C.N. Ostrander; D. Pacella; F. Paoletti; H.K. Park; W. Park; S.F. Paul; Y.-K. M. Peng; C.K. Phillips; R. Pinsker; P.H. Probert; S. Ramakrishnan; R. Raman; M. Redi; A.L. Roquemore; A. Rosenberg; P.M. Ryan; S.A. Sabbagh; M. Schaffer; R.J. Schooff; R. Seraydarian; C.H. Skinner; A.C. Sontag; V. Soukhanovskii; J. Spaleta; T. Stevenson; D. Stutman; D.W. Swain; E. Synakowski; Y. Takase; X. Tang; G. Taylor; J. Timberlake; K.L. Tritz; E.A. Unterberg; A. Von Halle; J. Wilgen; M. Williams; J.R. Wilson; X. Xu; S.J. Zweben; R. Akers; R.E. Barry; P. Beiersdorfer; J.M. Bialek; B. Blagojevic; P.T. Bonoli; M.D. Carter; W. Davis; B. Deng; L. Dudek; J. Egedal; R. Ellis; M. Finkenthal; J. Foley; E. Fredd; A. Glasser; T. Gibney; M. Gilmore; R.J. Goldston; R.E. Hatcher; R.J. Hawryluk; W. Houlberg; R. Harvey; S.C. Jardin; J.C. Hosea; H. Ji; M. Kalish; J. Lowrance; L.L. Lao; F.M. Levinton; N.C. Luhmann; R. Marsala; D. Mastravito; M.M. Menon; O. Mitarai; M. Nagata; G. Oliaro; R. Parsells; T. Peebles; B. Peneflor; D. Piglowski; G.D. Porter; A.K. Ram; M. Rensink; G. Rewoldt; P. Roney; K. Shaing; S. Shiraiwa; P. Sichta; D. Stotler; B.C. Stratton; R. Vero; W.R. Wampler; G.A. Wurden

    2003-10-02

    Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been

  1. Asymptotics for spherical needlets

    NASA Astrophysics Data System (ADS)

    Baldi, P.; Kerkyacharian, G.; Marinucci, D.; Picard, D.

    We investigate invariant random fields on the sphere using a new type of spherical wavelets, called needlets. These are compactly supported in frequency and enjoy excellent localization properties in real space, with quasi-exponentially decaying tails. We show that, for random fields on the sphere, the needlet coefficients are asymptotically uncorrelated for any fixed angular distance. This property is used to derive CLT and functional CLT convergence results for polynomial functionals of the needlet coefficients: here the asymptotic theory is considered in the high-frequency sense. Our proposals emerge from strong empirical motivations, especially in connection with the analysis of cosmological data sets.

  2. Tokamak coordinate conventions: COCOS

    NASA Astrophysics Data System (ADS)

    Sauter, O.; Medvedev, S. Yu.

    2013-02-01

    Dealing with electromagnetic fields, in particular current and related magnetic fields, yields "natural" physical vector relations in 3-D. However, when it comes to choosing local coordinate systems, the "usual" right-handed systems are not necessarily the best choices, which means that there are several options being chosen. In the magnetic fusion community such a difficulty exists for the choices of the cylindrical and of the toroidal coordinate systems. In addition many codes depend on knowledge of an equilibrium. In particular, the Grad-Shafranov axisymmetric equilibrium solution for tokamak plasmas, ψ, does not depend on the sign of the plasma current Ip nor that of the magnetic field B0. This often results in ill-defined conventions. Moreover the sign, amplitude and offset of ψ are of less importance, since the free sources in the equation depend on the normalized radial coordinate. The signs of the free sources, dp/dψ and dF2/dψ (p being the pressure, ψ the poloidal magnetic flux and F=RBφ), must be consistent to generate the current density profile. For example, RF and CD calculations (Radio Frequency heating and Current Drive) require an exact sign convention in order to calculate a co- or counter-CD component. It is shown that there are over 16 different coordinate conventions. This paper proposes a unique identifier, the COCOS convention, to distinguish between the 16 most-commonly used options. Given the present worldwide efforts towards code integration, the proposed new index COCOS defining uniquely the COordinate COnventionS required as input by a given code or module is particularly useful. As codes use different conventions, it is useful to allow different sign conventions for equilibrium code input and output, equilibrium being at the core of any calculations in magnetic fusion. Additionally, given two different COCOS conventions, it becomes simple to transform between them. The relevant transformations are described in detail.

  3. Tokamak and RFP ignition requirements

    SciTech Connect

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (n{tau}{sub 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{sub RAD} {approximately} 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 n{tau}{sub 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.

  4. Prospects for Tokamak Fusion Reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.

    1995-04-01

    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.

  5. High power heating of magnetic reconnection in merging tokamak experiments

    SciTech Connect

    Ono, Y.; Tanabe, H.; Gi, K.; Watanabe, T.; Ii, T.; Yamada, T.; Gryaznevich, M.; Scannell, R.; Conway, N.; Crowley, B.; Michael, C.

    2015-05-15

    Significant ion/electron heating of magnetic reconnection up to 1.2 keV was documented in two spherical tokamak plasma merging experiment on MAST with the significantly large Reynolds number R∼10{sup 5}. Measured 1D/2D contours of ion and electron temperatures reveal clearly energy-conversion mechanisms of magnetic reconnection: huge outflow heating of ions in the downstream and localized heating of electrons at the X-point. Ions are accelerated up to the order of poloidal Alfven speed in the reconnection outflow region and are thermalized by fast shock-like density pileups formed in the downstreams, in agreement with recent solar satellite observations and PIC simulation results. The magnetic reconnection efficiently converts the reconnecting (poloidal) magnetic energy mostly into ion thermal energy through the outflow, causing the reconnection heating energy proportional to square of the reconnecting (poloidal) magnetic field B{sub rec}{sup 2}  ∼  B{sub p}{sup 2}. The guide toroidal field B{sub t} does not affect the bulk heating of ions and electrons, probably because the reconnection/outflow speeds are determined mostly by the external driven inflow by the help of another fast reconnection mechanism: intermittent sheet ejection. The localized electron heating at the X-point increases sharply with the guide toroidal field B{sub t}, probably because the toroidal field increases electron confinement and acceleration length along the X-line. 2D measurements of magnetic field and temperatures in the TS-3 tokamak merging experiment also reveal the detailed reconnection heating mechanisms mentioned above. The high-power heating of tokamak merging is useful not only for laboratory study of reconnection but also for economical startup and heating of tokamak plasmas. The MAST/TS-3 tokamak merging with B{sub p} > 0.4 T will enables us to heat the plasma to the alpha heating regime: T{sub i} > 5 keV without using any additional heating facility.

  6. Spherically Symmetric Gravitational Fields

    NASA Astrophysics Data System (ADS)

    Vargas Moniz, P.

    The purpose of this paper is to investigate the quantum vacua directly implied by the wave function of a gravitational configuration characterized by the presence of an apparent horizon, namely the Vaidya space-time solution. Spherical symmetry is a main feature of this configuration, with a scalar field constituting a source [a Klein-Gordon geon or Berger-Chitre-Moncrief-Nutku (BCMN) type model]. The subsequent analysis requires solving a Wheeler-DeWitt equation near the apparent horizon (following the guidelinesintroduced by A. Tomimatsu,18; M. Pollock, 19 and developed by A. Hosoya and I. Oda20,21) with the scalar field herein expanded in terms of S2 spherical harmonics: midisuperspace quantization. The main results present in this paper are as follows. It is found that the mass function characteristic of the Vaidya metric is positive definite within this quantum approach. Furthermore, the inhomogeneous matter sector determines a descrip-tion in terms of open quantum (sub)systems, namely in the form of an harmonic oscillator whose frequency depends on the mass function. For this open (sub)system, a twofold approach is employed. On the one hand, an exact invariant observable is obtained from the effective Hamiltonian for the inhomogeneous matter modes. It is shown that this invariant admits a set of discrete eigenvalues which depend on the mass function. The corresponding set of eigenstates is constructed from a particular vacuum state. On the other hand, exact solutions are found for the Schrädinger equation associated with the inhomogeneous matter modes. This paper is concluded with a discussion, where two other issues are raised: (i) the possible application to realistic black hole dynamics of the results obtained for a simplified (BCMN) model and (ii) whether such vacuum states could be related with others defined instead within scalar field theories constructed in classical backgrounds.

  7. Spherical grating spectrometers

    NASA Astrophysics Data System (ADS)

    O'Donoghue, Darragh; Clemens, J. Christopher

    2014-07-01

    We describe designs for spectrometers employing convex dispersers. The Offner spectrometer was the first such instrument; it has almost exclusively been employed on satellite platforms, and has had little impact on ground-based instruments. We have learned how to fabricate curved Volume Phase Holographic (VPH) gratings and, in contrast to the planar gratings of traditional spectrometers, describe how such devices can be used in optical/infrared spectrometers designed specifically for curved diffraction gratings. Volume Phase Holographic gratings are highly efficient compared to conventional surface relief gratings; they have become the disperser of choice in optical / NIR spectrometers. The advantage of spectrometers with curved VPH dispersers is the very small number of optical elements used (the simplest comprising a grating and a spherical mirror), as well as illumination of mirrors off axis, resulting in greater efficiency and reduction in size. We describe a "Half Offner" spectrometer, an even simpler version of the Offner spectrometer. We present an entirely novel design, the Spherical Transmission Grating Spectrometer (STGS), and discuss exemplary applications, including a design for a double-beam spectrometer without any requirement for a dichroic. This paradigm change in spectrometer design offers an alternative to all-refractive astronomical spectrometer designs, using expensive, fragile lens elements fabricated from CaF2 or even more exotic materials. The unobscured mirror layout avoids a major drawback of the previous generation of catadioptric spectrometer designs. We describe laboratory measurements of the efficiency and image quality of a curved VPH grating in a STGS design, demonstrating, simultaneously, efficiency comparable to planar VPH gratings along with good image quality. The stage is now set for construction of a prototype instrument with impressive performance.

  8. Progress towards high-performance, steady-state spherical torus

    NASA Astrophysics Data System (ADS)

    Ono, M.; Bell, M. G.; Bell, R. E.; Bigelow, T.; Bitter, M.; Blanchard, W.; Boedo, J.; Bourdelle, C.; Bush, C.; Choe, W.; Chrzanowski, J.; Darrow, D. S.; Diem, S. J.; Doerner, R.; Efthimion, P. C.; Ferron, J. R.; Fonck, R. J.; Fredrickson, E. D.; Garstka, G. D.; Gates, D. A.; Gray, T.; Grisham, L. R.; Heidbrink, W.; Hill, K. W.; Hoffman, D.; Jarboe, T. R.; Johnson, D. W.; Kaita, R.; Kaye, S. M.; Kessel, C.; Kim, J. H.; Kissick, M. W.; Kubota, S.; Kugel, H. W.; LeBlanc, B. P.; Lee, K.; Lee, S. G.; Lewicki, B. T.; Luckhardt, S.; Maingi, R.; Majeski, R.; Manickam, J.; Maqueda, R.; Mau, T. K.; Mazzucato, E.; Medley, S. S.; Menard, J.; Mueller, D.; Nelson, B. A.; Neumeyer, C.; Nishino, N.; Ostrander, C. N.; Pacella, D.; Paoletti, F.; Park, H. K.; Park, W.; Paul, S. F.; Peng, Y.-K. M.; Phillips, C. K.; Pinsker, R.; Probert, P. H.; Ramakrishnan, S.; Raman, R.; Redi, M.; Roquemore, A. L.; Rosenberg, A.; Ryan, P. M.; Sabbagh, S. A.; Schaffer, M.; Schooff, R. J.; Seraydarian, R.; Skinner, C. H.; Sontag, A. C.; Soukhanovskii, V.; Spaleta, J.; Stevenson, T.; Stutman, D.; Swain, D. W.; Synakowski, E.; Takase, Y.; Tang, X.; Taylor, G.; Timberlake, J.; Tritz, K. L.; Unterberg, E. A.; Von Halle, A.; Wilgen, J.; Williams, M.; Wilson, J. R.; Xu, X.; Zweben, S. J.; Akers, R.; Barry, R. E.; Beiersdorfer, P.; Bialek, J. M.; Blagojevic, B.; Bonoli, P. T.; Carter, M. D.; Davis, W.; Deng, B.; Dudek, L.; Egedal, J.; Ellis, R.; Finkenthal, M.; Foley, J.; Fredd, E.; Glasser, A.; Gibney, T.; Gilmore, M.; Goldston, R. J.; Hatcher, R. E.; Hawryluk, R. J.; Houlberg, W.; Harvey, R.; Jardin, S. C.; Hosea, J. C.; Ji, H.; Kalish, M.; Lowrance, J.; Lao, L. L.; Levinton, F. M.; Luhmann, N. C.; Marsala, R.; Mastravito, D.; Menon, M. M.; Mitarai, O.; Nagata, M.; Oliaro, G.; Parsells, R.; Peebles, T.; Peneflor, B.; Piglowski, D.; Porter, G. D.; Ram, A. K.; Rensink, M.; Rewoldt, G.; Robinson, J.; Roney, P.; Shaing, K.; Shiraiwa, S.; Sichta, P.; Stotler, D.; Stratton, B. C.; Vero, R.; Wampler, W. R.; Wurden, G. A.

    2003-12-01

    Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (bgr), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bgrT of up to 35% with a near unity central bgrT have been obtained. NSTX will be exploring advanced regimes where bgrT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500 k

  9. Progress towards high-performance, steady-state spherical torus

    SciTech Connect

    Ono, M.; Bell, M. G.; Bell, R. E.; Bigelow, T.; Bitter, M.; Blanchard, W.; Boedo, J.; Bourdelle, C.; Bush, C.; Choe, W.; Chrzanowski, J.; Darrow, D. S.; Diem, S. J.; Doerner, R.; Efthimion, P. C.; Ferron, J. R.; Fonck, R. J.; Fredrickson, E. D.; Garstka, G. D.; Gates, D A; Gray, T.; Grisham, L. R.; Heidbrink, W.; Hill, K. W.; Hoffman, D.; Jarboe, T. R.; Johnson, D. W.; Kaita, R.; Kaye, S. M.; Kessel, C.; Kim, J. H.; Kissick, M. W.; Kubota, S.; Kugel, H. W.; LeBlanc, B. P.; Lee, K.; Lee, S. G.; Lewicki, B. T.; Luckhardt, S.; Maingi, R.; Majeski, R.; Manickam, J.; Maqueda, R.; Mau, T. K.; Mazzucato, E.; Medley, S. S.; Menard, J.; Mueller, D.; Nelson, B. A.; Neumeyer, C.; Nishino, N.; Ostrander, C. N.; Pacella, D.; Paoletti, F.; Park, H. K.; Park, W.; Paul, S. F.; Peng, Y-K M.; Phillips, C. K.; Pinsker, R.; Probert, P. H.; Ramakrishnan, S.; Raman, R.; Redi, M.; Roquemore, A. L.; Rosenberg, A.; Ryan, P. M.; Sabbagh, S. A.; Schaffer, M.; Schooff, R. J.; Seraydarian, R.; Skinner, C. H.; Sontag, A. C.; Soukhanovskii, V.; Spaleta, J.; Stevenson, T.; Stutman, D.; Swain, D. W.; Synakowski, E.; Takase, Y.; Tang, X.; Taylor, G.; Timberlake, J.; Tritz, K. L.; Unterberg, E. A.; Halle, A. Von.; Wilgen, J.; Williams, M.; Wilson, J. R.; Xu, X.; Zweben, S. J.; Akers, R.; Barry, R. E.; Beiersdorfer, P.; Bialek, J. M.; Blagojevic, B.; Bonoli, P. T.; Carter, M. D.; Davis, W.; Deng, B.; Dudek, L.; Egedal, J.; Ellis, R.; Finkenthal, M.; Foley, J.; Fredd, E.; Glasser, A.; Gibney, T.; Gilmore, M.; Goldston, R. J.; Hatcher, R. E.; Hawryluk, R. J.; Houlberg, W.; Harvey, R.; Jardin, S. C.; Hosea, J. C.; Ji, H.; Kalish, M.; Lowrance, J.; Lao, L. L.; Levinton, F. M.; Luhmann, N. C.; Marsala, R.; Mastravito, D.; Menon, M. M.; Mitarai, O.; Nagata, M.; Oliaro, G.; Parsells, R.; Peebles, T.; Peneflor, B.; Piglowski, D.; Porter, G. D.; Ram, A. K.; Rensink, M.; Rewoldt, G.; Robinson, J.; Roney, P.; Shaing, K.; Shiraiwa, S.; Sichta, P.; Stotler, D.; Stratton, B. C.; Vero, R.; Wampler, W. R.; Wurden, G. A.

    2003-12-01

    Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (β), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values βT of up to 35% with a near unity central βT have been obtained. NSTX will be exploring advanced regimes where βT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~ 60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to

  10. Progress Towards High-Performance, Steady-State Spherical Torus

    SciTech Connect

    Lawrence Livermore National Laboratory

    2004-01-04

    Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta ({beta}), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values {beta}{sub T} of up to 35% with a near unity central {beta}{sub T} have been obtained. NSTX will be exploring advanced regimes where {beta}{sub T} up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction ({approx}60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fastwave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX

  11. Progress towards high-performance, steady-state spherical torus.

    SciTech Connect

    Lee, S.G; Kugel, W.; Efthimion, P. C.; Kissick, M. W.; Bourdelle, C.; Kim, J.H; Gray, T.; Garstka, G. D.; Fonck, R. J.; Doerner, R.; Diem, S.J.; Pacella, D.; Nishino, N.; Ferron, J. R.; Skinner, C. H.; Stutman, D.; Soukhanovskii, V.; Choe, W.; Chrzanowski, J.; Mau, T.K.; Bell, Michael G.; Raman, R.; Peng, Y-K. M.; Ono, M.; Park, W.; Hoffman, D.; Maqueda, R.; Kaye, S. M.; Kaita, R.; Jarboe, T.R.; Hill, K.W.; Heidbrink, W.; Spaleta, J.; Sontag, A.C; Seraydarian, R.; Schooff, R.J.; Sabbagh, S.A.; Menard, J.; Mazzucato, E.; Lee, K.; LeBlanc, B.; Probert, P. H.; Blanchard, W.; Wampler, William R.; Swain, D. W.; Ryan, P.M.; Rosenberg, A.; Ramakrishnan, S.; Phillips, C.K.; Park, H.K.; Roquemore, A. L.; Paoletti, F.; Medley, S. S.; Fredrickson, E. D.; Kessel, C. E.; Stevenson, T.; Darrow, D. S.; Majeski, R.; Bitter, M.; Neumeyer, C.; Nelson, B.A.; Paul, S. F.; Manickam, J.; Ostrander, C. N.; Mueller, D.; Lewicki, B.T; Luckhardt, S.; Johnson, D.W.; Grisham, L.R.; Kubota, Shigeru; Gates, D.A.; Bush, C.; Synakowski, E.J.; Schaffer, M.; Boedo, J.; Maingi, R.; Redi, M.; Pinsker, R.; Bigelow, T.; Bell, R. E.

    2004-06-01

    Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta ({beta}), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values {beta}{sub T} of up to 35% with a near unity central {beta}{sub T} have been obtained. NSTX will be exploring advanced regimes where {beta}{sub T} up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction ({approx}60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX

  12. Bootstrapped tokamak with oscillating field current drive

    SciTech Connect

    Weening, R.H. )

    1993-07-01

    A magnetic helicity conserving mean-field Ohm's law is used to study bootstrapped tokamaks with oscillating field current drive. The Ohm's law leads to the conclusion that the tokamak bootstrap effect can convert the largely alternating current of oscillating field current drive into a direct toroidal plasma current. This plasma current rectification is due to the intrinsically nonlinear nature of the tokamak bootstrap effect, and suggests that it may be possible to maintain the toroidal current of a tokamak reactor by supplementing the bootstrap current with oscillating field current drive. Steady-state tokamak fusion reactors operating with oscillating field current drive could provide an alternative to tokamak reactors operating with external current drive.

  13. Double slotted socket spherical joint

    DOEpatents

    Bieg, Lothar F.; Benavides, Gilbert L.

    2001-05-22

    A new class of spherical joints is disclosed. These spherical joints are capable of extremely large angular displacements (full cone angles in excess of 270.degree.), while exhibiting no singularities or dead spots in their range of motion. These joints can improve or simplify a wide range of mechanical devices.

  14. Microwave Tokamak Experiment: Overview and status

    SciTech Connect

    Not Available

    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.

  15. Composition And Electrical Properties Of Dust From Tokamak Compass

    SciTech Connect

    Vaverka, J.; Beranek, M.; Pavlu, J.; Richterova, I.; Vysinka, M.; Safrankova, J.; Nemecek, Z.

    2011-11-29

    In spite of the fact that fusion is a subject of the study for many years, there are still a lot of open questions. One of the interesting topics in fusion research is a presence of dust grains in reactors. In the paper, dust grains born in tokamak Compass are studied and compared with samples of a spherical geometry and well known composition. A unique experimental setup was used for investigations of charging properties of such grains and the SEM and EDX spectroscopy was applied for a study of grain composition. We focus on the secondary emission because this process plays a prominent role when a portion of energetic electrons is present in surroundings of a particular grain. It was shown that depending on the grain size and material, energetic electrons charge the grains to positive potentials comparable with the energy of impinging electrons.

  16. Validation of Tokamak Equilibria: Reconciling Theory and Observation Using BEAST

    NASA Astrophysics Data System (ADS)

    von Nessi, Gregory; Hole, Matthew; Svensson, Jakob

    2011-10-01

    We present a new technique for reconciling force-balance models with diagnostic observations via the statistical theory of Bayesian analysis. This method forms the backbone of a new data analysis code called BEAST (Bayesian Equilibrium Analysis and Simulation Technique) and is based on refactoring the force-balance relation into two different forward models, each associated with a 'fractional' observation, which are subsequently used in the Bayesian inference of the plasma equilibrium. By using a variant of the nested sampling algorithm, the evidence of the inferred posterior distribution is calculated and provides a relative quantification of how much the inferred equilibrium differs from a force-balance solution. Results are presented for discharges on the Mega-Ampere Spherical Tokamak (MAST), which are calculated using pickup coil, flux loop and Motional-Stark Effect (MSE) diagnostic data.

  17. Comprehensive numerical modelling of tokamaks

    SciTech Connect

    Cohen, R.H.; Cohen, B.I.; Dubois, P.F.

    1991-01-03

    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.

  18. Burn Control Mechanisms in Tokamaks

    NASA Astrophysics Data System (ADS)

    Hill, Maxwell; Stacey, Weston

    2013-10-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamaks, especially those used as a neutron source for fusion-fission hybrid reactors, such as the Subcritical Advanced Burner Reactor (SABR) concept. At Georgia Tech, we are developing a new burning plasma dynamics code to investigate passive safety mechanisms that could prevent power excursions in tokamak reactors. This code solves the coupled set of balance equations governing burning plasmas in conjunction with a two-point SOL-divertor model. Predictions have been benchmarked against data from DIII-D. We are examining several potential negative feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instabilities, iii) the degradation of alpha-particle confinement resulting from ripples in the toroidal field, iv) modifications to the radial current profile, v) ``divertor choking'' and vi) Type 1 ELMs.

  19. MHD stability of tokamak plasmas

    SciTech Connect

    Chance, M.S. Sun, Y.C.; Jardin, S.C.; Kessel, C.E.; Okabayashi, M.

    1992-08-01

    This paper will give an overview of the some of the methods which are used to simulate the ideal MHD properties of tokamak plasmas. A great deal of the research in this field is necessarily numerical and the substantial progress made during the past several years has roughly paralleled the continuing availability of more advanced supercomputers. These have become essential to accurately model the complex configurations necessary for achieving MHD stable reactor grade conditions. Appropriate tokamak MHD equilibria will be described. Then the stability properties is discussed in some detail, emphasizing the difficulties of obtaining stable high {beta} discharges in plasmas in which the current is mainly ohmically driven and thus demonstrating the need for tailoring the current and pressure profiles of the plasma away from the ohmic state. The outline of this paper will roughly follow the physics development to attain the second region of stability in the PBX-M device at The Princeton Plasmas Physics Laboratory.

  20. Density limit disruptions in tokamaks

    NASA Astrophysics Data System (ADS)

    Kleva, Robert G.; Drake, J. F.

    1991-02-01

    Magnetohydrodynamic simulations are presented which reproduce the rapid drop in the central temperature observed during density limit disruptions in tokamaks. The loss of central confinement is triggered by edge radiation which destabilizes a q=1 kink mode. A bubble of cold plasma is injected from the edge into the center by the q=1 kink. This bubble bears a striking resemblance to the cold plasma that is observed to move from the edge into the center during density limit disruptions on the JET tokamak [Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. 1, p. 433], initiating the loss of central confinement. The bubble profile produced by the q=1 kink is unstable to a broad spectrum of modes which progressively reduce the magnetic shear between the q=2 surface and the center. The q=2 mode then grows across the center, broadening the current and throwing the hot plasma to the wall.

  1. Magnetic Diagnostics for the Lithium Tokamak eXperiment

    SciTech Connect

    Berzak, L.; Kaita, R.; Kozub, T.; Majeski, R.; Zakharov, L.

    2008-06-20

    The Lithium Tokamak eXperiment (LTX) is a spherical tokamak with R0 = 0.4m, a = 0.26m, BTF ~ 3.4kG, IP ~ 400kA, and pulse length ~ 0.25s. The focus of LTX is to investigate the novel, low-recycling Lithium Wall operating regime for magnetically confined plasmas. This regime is reached by placing an in-vessel shell conformal to the plasma last closed flux surface. The shell is heated and then coated with liquid lithium. An extensive array of magnetic diagnostics is available to characterize the experiment, including 80 Mirnov coils (single and double-axis, internal and external to the shell), 34 flux loops, 3 Rogowskii coils, and a diamagnetic loop. Diagnostics are specifically located to account for the presence of a secondary conducting surface and engineered to withstand both high temperatures and incidental contact with liquid lithium. The diagnostic set is therefore fabricated from robust materials with heat and lithium resistance and is designed for electrical isolation from the shell and to provide the data required for highly constrained equilibrium reconstructions.

  2. ADVANCES IN DUST DETECTION AND REMOVAL FOR TOKAMAKS

    SciTech Connect

    Campos, A.; Skinner, C.H.

    2009-01-01

    Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. In the tokamak environment, large particles or fi bers can fall on the electrostatic detector potentially causing a permanent short. An electrostatic dust detector developed in the laboratory is being applied to the National Spherical Torus Experiment (NSTX). We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments at atmospheric pressure with varying nozzle designs, backing pressures, puff durations and exit fl ow orientations have given an optimal confi guration that effectively removes particles from a 25 cm² area. Similar removal effi ciencies were observed under a vacuum base pressure of 1 mTorr. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tri-polar grid of fi ne interdigitated traces has been designed that generates an electrostatic traveling wave for conveying dust particles to a “drain.” First trials with only two working electrodes have shown particle motion in optical microscope images.

  3. Buckling of spherical capsules.

    PubMed

    Knoche, Sebastian; Kierfeld, Jan

    2011-10-01

    We investigate buckling of soft elastic capsules under negative pressure or for reduced capsule volume. Based on nonlinear shell theory and the assumption of a hyperelastic capsule membrane, shape equations for axisymmetric and initially spherical capsules are derived and solved numerically. A rich bifurcation behavior is found, which is presented in terms of bifurcation diagrams. The energetically preferred stable configuration is deduced from a least-energy principle both for prescribed volume and prescribed pressure. We find that buckled shapes are energetically favorable already at smaller negative pressures and larger critical volumes than predicted by the classical buckling instability. By preventing self-intersection for strongly reduced volume, we obtain a complete picture of the buckling process and can follow the shape from the initial undeformed state through the buckling instability into the fully collapsed state. Interestingly, the sequences of bifurcations and stable capsule shapes differ for prescribed volume and prescribed pressure. In the buckled state, we find a relation between curvatures at the indentation rim and the bending modulus, which can be used to determine elastic moduli from experimental shape analysis. PMID:22181297

  4. Consider a spherical cow

    SciTech Connect

    Harte, J.

    1985-01-01

    Consider a Spherical Cow describes relatively simple mathematical methods for developing quantitative answers to often complex environmental problems. Early chapters provide systematic insights into problem solving and identifying mathematical tools and models that lead to back of the envelope answers. Subsequent chapters treat increasingly complex problems. Solutions are sought at different levels, e.g., informed guesses, quantitative solutions based on detailed analytical models, and ultimately, critical evaluation of the consequences of removing simplifying assumptions from the models. The vehicle employed is a collection of 44 challenging problems, with clearly worked out solutions, plus ample exercises. The book, though directed at environmentalists, should appeal to chemists. Many of the problems are rooted in chemistry, including acid rain, the CO/sub 2/ greenhouse effect, chemical contamination, and the disturbing of cyclical chemical balances. Readers feeling a civic responsibility to think and speak more clearly on environmental issues will find the essential modeling and quantitative approaches valuable assets beyond the help provided by the usual courses in science and mathematics. In fact, the techniques of problem solving have broad applicability beyond the specific environmental examples covered in this text.

  5. A relativistic spherical vortex

    PubMed Central

    Pekeris, C. L.

    1976-01-01

    This investigation is concerned with stationary relativistic flows of an inviscid and incompressible fluid. In choosing a density-pressure relation to represent relativistic “incompressibility,” it is found that a fluid in which the velocity of sound equals the velocity of light is to be preferred for reasons of mathematical simplicity. In the case of axially symmetric flows, the velocity field can be derived from a stream function obeying a partial differential equation which is nonlinear. A transformation of variables is found which makes the relativistic differential equation linear. An exact solution is obtained for the case of a vortex confined to a stationary sphere. One can make all three of the components of velocity vanish on the surface of the sphere, as in the nonrelativistic Hicks spherical vortex. In the case of an isolated vortex on whose surface the pressure is made to vanish, it is found that the pressure at the center of the sphere becomes negative, as in the nonrelativistic case. A solution is also obtained for a relativistic vortex advancing in a fluid. The sphere is distorted into an oblate spheroid. The maximum possible velocity of advance of the vortex is (2/3) c. PMID:16578745

  6. Immunomodulatory spherical nucleic acids

    PubMed Central

    Radovic-Moreno, Aleksandar F.; Chernyak, Natalia; Mader, Christopher C.; Nallagatla, Subbarao; Kang, Richard S.; Hao, Liangliang; Walker, David A.; Halo, Tiffany L.; Merkel, Timothy J.; Rische, Clayton H.; Anantatmula, Sagar; Burkhart, Merideth; Mirkin, Chad A.; Gryaznov, Sergei M.

    2015-01-01

    Immunomodulatory nucleic acids have extraordinary promise for treating disease, yet clinical progress has been limited by a lack of tools to safely increase activity in patients. Immunomodulatory nucleic acids act by agonizing or antagonizing endosomal toll-like receptors (TLR3, TLR7/8, and TLR9), proteins involved in innate immune signaling. Immunomodulatory spherical nucleic acids (SNAs) that stimulate (immunostimulatory, IS-SNA) or regulate (immunoregulatory, IR-SNA) immunity by engaging TLRs have been designed, synthesized, and characterized. Compared with free oligonucleotides, IS-SNAs exhibit up to 80-fold increases in potency, 700-fold higher antibody titers, 400-fold higher cellular responses to a model antigen, and improved treatment of mice with lymphomas. IR-SNAs exhibit up to eightfold increases in potency and 30% greater reduction in fibrosis score in mice with nonalcoholic steatohepatitis (NASH). Given the clinical potential of SNAs due to their potency, defined chemical nature, and good tolerability, SNAs are attractive new modalities for developing immunotherapies. PMID:25775582

  7. Gyrosheath near the tokamak edge

    SciTech Connect

    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.

  8. Neoclassical magnetic microislands in tokamaks

    SciTech Connect

    Kovalishen, E.A.; Mikhailovskii, A.B.; Botov, P.V.; Shirokov, M.S.; Konovalov, S.V.; Tsypin, V.S.; Galvao, R.M.O.

    2005-09-15

    Possibility of existence of neoclassical magnetic microislands (island width smaller than the ion Larmor radius) in a tokamak in the banana regime is shown. The rotation frequency of such islands is found. It is shown that for the case of positive electron temperature gradient, the bootstrap current destabilizes the microislands while the polarization current leads to their stabilization. Maximally possible neoclassical microisland width is estimated.

  9. Magnetic island formation in tokamaks

    SciTech Connect

    Yoshikawa, S.

    1989-04-01

    The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs.

  10. Magnetic confinement experiment. I: Tokamaks

    SciTech Connect

    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 nT{tau}`s {approximately} 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.

  11. Engineering design of the National Spherical Torus Experiment

    SciTech Connect

    C. Neumeyer; P. Heitzenroeder; J. Spitzer, J. Chrzanowski; et al

    2000-05-11

    NSTX is a proof-of-principle experiment aimed at exploring the physics of the ``spherical torus'' (ST) configuration, which is predicted to exhibit more efficient magnetic confinement than conventional large aspect ratio tokamaks, amongst other advantages. The low aspect ratio (R/a, typically 1.2--2 in ST designs compared to 4--5 in conventional tokamaks) decreases the available cross sectional area through the center of the torus for toroidal and poloidal field coil conductors, vacuum vessel wall, plasma facing components, etc., thus increasing the need to deploy all components within the so-called ``center stack'' in the most efficient manner possible. Several unique design features have been developed for the NSTX center stack, and careful engineering of this region of the machine, utilizing materials up to their engineering allowables, has been key to meeting the desired objectives. The design and construction of the machine has been accomplished in a rapid and cost effective manner thanks to the availability of extensive facilities, a strong experience base from the TFTR era, and good cooperation between institutions.

  12. Numerical tokamak turbulence project (OFES grand challenge)

    SciTech Connect

    Beer, M; Cohen, B I; Crotinger, J; Dawson, J; Decyk, V; Dimits, A M; Dorland, W D; Hammett, G W; Kerbel, G D; Leboeuf, J N; Lee, W W; Lin, Z; Nevins, W M; Reynders, J; Shumaker, D E; Smith, S; Sydora, R; Waltz, R E; Williams, T

    1999-08-27

    The primary research objective of the Numerical Tokamak Turbulence Project (NTTP) is to develop a predictive ability in modeling turbulent transport due to drift-type instabilities in the core of tokamak fusion experiments, through the use of three-dimensional kinetic and fluid simulations and the derivation of reduced models.

  13. Steady State Tokamak Equilibria without Current Drive

    SciTech Connect

    Shaing, K.C.; Aydemir, A.Y.; Lin-Liu, Y.R.; Miller, R.L.

    1997-11-01

    Steady state tokamak equilibria without current drive are found. This is made possible by including the potato bootstrap current close to the magnetic axis. Tokamaks with this class of equilibria do not need seed current or current drive, and are intrinsically steady state. {copyright} {ital 1997} {ital The American Physical Society}

  14. Physics of compact ignition tokamak designs

    SciTech Connect

    Singer, C.E.; Ku, L.P.; Bateman, G.; Seidl, F.; Sugihara, M.

    1986-03-01

    Models for predicting plasma performance in compact ignition experiments are constructed on the basis of theoretical and empirical constraints and data from tokamak experiments. Emphasis is placed on finding transport and confinement models which reproduce results of both ohmically and auxiliary heated tokamak data. Illustrations of the application of the models to compact ignition designs are given.

  15. Advanced tokamak research on the DIII-D tokamak

    SciTech Connect

    Chan, V.S.

    1994-01-01

    The objective of the planned research in advanced tokamak development on DIII-D at General Atomics, San Diego, USA. is to establish improved tokamak operation through significant improvements in the stability factor, confinement quality, and bootstrap current fraction using localized radio frequency (rf) current profile control, rf and neutral beam heating for pressure profile control, as well as control of plasma rotation and optimization of the plasma boundary conditions. Recent research results in H-mode confinement, modifications of current profiles to achieve higher confinement and higher {beta}, a new regime of improved confinement (VH-mode), and rf noninductive current drive are encouraging. In this talk, arguments will be presented supporting the need for improved performance in tokamak reactors. Experimentally observed advanced performance regimes on DIII-D will be discussed. Confinement improvement up to H = 4, where H is the ratio of energy confinement time to the ITER89-P scaling H{triple_bond} {Tau}{sub E}/{Tau}{sub E-ITER89-P}, has been achieved. In other discharges {beta}{sub N} = {beta}/(I/aB),[%-m{center_dot}{Tau}/MA] {approx_gt} 6 has been obtained. These values have so far been achieved transiently and independently. Techniques, will be described which can extend the high performance to quasi-steady-state and sustain the high H and {beta}{sub N} values simultaneously. Two high performance regimes, one in first stable regime and the other in second stable regime, have been simulated br self-consistently evolving a magnetohydrodynamic (MHD) equilibrium-transport code. Finally, experimental program plans and outstanding important physics issues will be discussed.

  16. Results and future plans of the Lithium Tokamak eXperiment (LTX)

    NASA Astrophysics Data System (ADS)

    Schmitt, J. C.; Abrams, T.; Baylor, L. R.; Berzak Hopkins, L.; Biewer, T.; Bohler, D.; Boyle, D.; Granstedt, E.; Gray, T.; Hare, J.; Jacobson, C. M.; Jaworski, M.; Kaita, R.; Kozub, T.; LeBlanc, B.; Lundberg, D. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Ryou, A.; Shi, E.; Squire, J.; Stotler, D.; Thomas, C. E.; Tritz, K.; Zakharov, L.

    2013-07-01

    The Lithium Tokamak eXperiment (LTX) is a spherical tokamak with the unique capability of studying the low-recycling regime by coating nearly 90% of the first wall with lithium in either solid or liquid form. Several grams of lithium are evaporated onto the plasma-facing side of the first wall. Without lithium coatings, the plasma discharge is limited to less than 5 ms and only 10 kA of plasma current, and the first wall acts as a particle source. With cold lithium coatings, plasma discharges last up to 20 ms with plasma currents up to 70 kA. The lithium coating provides a low-recycling first wall condition for the plasma and higher fueling rates are required to realize plasma densities similar to that of pre-lithium walls. Traditional puff fueling, supersonic gas injection, and molecular cluster injection (MCI) are used. Liquid lithium experiments will begin in 2012.

  17. Magnetic diagnostics for equilibrium reconstructions with eddy currents on the Lithium Tokamak eXperiment.

    PubMed

    Schmitt, J C; Bialek, J; Lazerson, S; Majeski, R

    2014-11-01

    The Lithium Tokamak eXperiment is a spherical tokamak with a close-fitting low-recycling wall composed of thin lithium layers evaporated onto a stainless steel-lined copper shell. Long-lived non-axisymmetric eddy currents are induced in the shell and vacuum vessel by transient plasma and coil currents and these eddy currents influence both the plasma and the magnetic diagnostic signals that are used as constraints for equilibrium reconstruction. A newly installed set of re-entrant magnetic diagnostics and internal saddle flux loops, compatible with high-temperatures and lithium environments, is discussed. Details of the axisymmetric (2D) and non-axisymmetric (3D) treatments of the eddy currents and the equilibrium reconstruction are presented. PMID:25430382

  18. Magnetic diagnostics for equilibrium reconstructions with eddy currents on the lithium tokamak experiment

    SciTech Connect

    Schmitt, J. C. Lazerson, S.; Majeski, R.; Bialek, J.

    2014-11-15

    The Lithium Tokamak eXperiment is a spherical tokamak with a close-fitting low-recycling wall composed of thin lithium layers evaporated onto a stainless steel-lined copper shell. Long-lived non-axisymmetric eddy currents are induced in the shell and vacuum vessel by transient plasma and coil currents and these eddy currents influence both the plasma and the magnetic diagnostic signals that are used as constraints for equilibrium reconstruction. A newly installed set of re-entrant magnetic diagnostics and internal saddle flux loops, compatible with high-temperatures and lithium environments, is discussed. Details of the axisymmetric (2D) and non-axisymmetric (3D) treatments of the eddy currents and the equilibrium reconstruction are presented.

  19. Magnetic diagnostics for equilibrium reconstructions with eddy currents on the lithium tokamak experimenta)

    SciTech Connect

    Schmitt, J. C.; Bialek, J.; Lazerson, S.; Majeski, R.

    2014-11-01

    The Lithium Tokamak eXperiment is a spherical tokamak with a close-fitting low-recycling wall composed of thin lithium layers evaporated onto a stainless steel-lined copper shell. Long-lived non-axisymmetric eddy currents are induced in the shell and vacuum vessel by transient plasma and coil currents and these eddy currents influence both the plasma and the magnetic diagnositc signals that are used as constraints for equilibrium reconstruction. A newly installed set of re-entrant magnetic diagnostics and internal saddle flux loops, compatible with high-temperatures and lithium environments, is discussed. Details of the axisymmetric (2D) and non-axisymmetric (3D) treatments of the eddy currents and the equilibrium reconstruction are presented.

  20. Linear optimal control of tokamak fusion devices

    SciTech Connect

    Kessel, C.E.; Firestone, M.A.; Conn, R.W.

    1989-05-01

    The control of plasma position, shape and current in a tokamak fusion reactor is examined using linear optimal control. These advanced tokamaks are characterized by non up-down symmetric coils and structure, thick structure surrounding the plasma, eddy currents, shaped plasmas, superconducting coils, vertically unstable plasmas, and hybrid function coils providing ohmic heating, vertical field, radial field, and shaping field. Models of the electromagnetic environment in a tokamak are derived and used to construct control gains that are tested in nonlinear simulations with initial perturbations. The issues of applying linear optimal control to advanced tokamaks are addressed, including complex equilibrium control, choice of cost functional weights, the coil voltage limit, discrete control, and order reduction. Results indicate that the linear optimal control is a feasible technique for controlling advanced tokamaks where the more common classical control will be severely strained or will not work. 28 refs., 13 figs.

  1. Observation of quasi-coherent edge fluctuations in Ohmic plasmas on National Spherical Torus Experiment

    NASA Astrophysics Data System (ADS)

    Banerjee, Santanu; Diallo, A.; Zweben, S. J.

    2016-04-01

    A quasi-coherent edge density mode with frequency fmode ˜ 40 kHz is observed in Ohmic plasmas in National Spherical Torus Experiment using the gas puff imaging diagnostic. This mode is located predominantly just inside the separatrix, with a maximum fluctuation amplitude significantly higher than that of the broadband turbulence in the same frequency range. The quasi-coherent mode has a poloidal wavelength λpol ˜ 16 cm and a poloidal phase velocity of Vpol ˜ 4.9 ± 0.3 km s-1 in the electron diamagnetic direction, which are similar to the characteristics expected from a linear drift-wave-like mode in the edge. This is the first observation of a quasi-coherent edge mode in an Ohmic diverted tokamak, and so may be useful for validating tokamak edge turbulence codes.

  2. Ohmic flux consumption during initial operation of the NSTX spherical torus

    NASA Astrophysics Data System (ADS)

    Menard, J. E.; LeBlanc, B. P.; Sabbagh, S. A.; Bell, M. G.; Bell, R. E.; Fredrickson, E. D.; Gates, D. A.; Jardin, S. C.; Johnson, D. W.; Kaye, S. M.; Kugel, H. W.; Maingi, R.; Maqueda, R. J.; Mueller, D.; Ono, M.; Paoletti, F.; Paul, S. F.; Skinner, C. H.; Stutman, D.; NSTX Research Team

    2001-09-01

    Spherical tokamaks, because of their slender central columns, have very limited volt second capability relative to that of standard aspect ratio tokamaks of similar plasma cross-section. Recent experiments on NSTX have begun to quantify and optimize the ohmic current drive efficiency in an MA class ST device. Sustainable ramp rates in excess of 5 MA/s during the current rise phase have been achieved on NSTX, while faster ramps generate significant MHD activity. Discharges with Ip exceeding 1 MA have been achieved in NSTX with nominal parameters: aspect ratio A = 1.3-1.4, elongation κ = 2-2.2, triangularity δ = 0.4, internal inductance li = 0.6, Ejima coefficient CE = 0.35. Flux consumption efficiency results, performance improvements associated with first boronization and comparisons with neoclassical resistivity are described.

  3. Transport equations in tokamak plasmas

    SciTech Connect

    Callen, J. D.; Hegna, C. C.; Cole, A. J.

    2010-05-15

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for neoclassical effects on the parallel Ohm's law, fluctuation-induced transport, heating, current-drive and flow sources and sinks, small magnetic field nonaxisymmetries, magnetic field transients, etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed using a kinetic-based approach. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales and constraints they impose are considered sequentially: compressional Alfven waves (Grad-Shafranov equilibrium, ion radial force balance), sound waves (pressure constant along field lines, incompressible flows within a flux surface), and collisions (electrons, parallel Ohm's law; ions, damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on a plasma species: seven ambipolar collision-based ones (classical, neoclassical, etc.) and eight nonambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients, etc.). The plasma toroidal rotation equation results from setting to zero the net radial current induced by the nonambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the nonambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The 'mean field' effects of microturbulence on the parallel Ohm's law, poloidal ion flow, particle fluxes, and toroidal momentum and energy transport are all included self-consistently. The

  4. The Microwave Tokamak Experiment (MTX)

    SciTech Connect

    Thomassen, K.I.; Cohen, B.I.; Hooper, E.B.; Lang, D.D.; Nevins, W.M.

    1987-10-02

    A new experimental facility is being assembled at the Lawrence Livermore National Laboratory (LLNL) for studying microwave propagation and absorption in high density plasmas. A unique feature of the facility is the free electron laser (FEL) used to generate high peak power microwaves at 250 GHz, at a repetition rate so as to produce up to 2 MW of average power for up to 30 s. Called the Microwave Tokamak Experiment (MTX), the facility will be used for studies of plasma heating, current drive, and confinement.

  5. Breakdown in the pretext tokamak

    SciTech Connect

    Benesch, J.F.

    1981-06-01

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges.

  6. Milking the spherical cow - on aspherical dynamics in spherical coordinates

    NASA Astrophysics Data System (ADS)

    Pontzen, Andrew; Read, Justin I.; Teyssier, Romain; Governato, Fabio; Gualandris, Alessia; Roth, Nina; Devriendt, Julien

    2015-08-01

    Galaxies and the dark matter haloes that host them are not spherically symmetric, yet spherical symmetry is a helpful simplifying approximation for idealized calculations and analysis of observational data. The assumption leads to an exact conservation of angular momentum for every particle, making the dynamics unrealistic. But how much does that inaccuracy matter in practice for analyses of stellar distribution functions, collisionless relaxation, or dark matter core-creation? We provide a general answer to this question for a wide class of aspherical systems; specifically, we consider distribution functions that are `maximally stable', i.e. that do not evolve at first order when external potentials (which arise from baryons, large-scale tidal fields or infalling substructure) are applied. We show that a spherically symmetric analysis of such systems gives rise to the false conclusion that the density of particles in phase space is ergodic (a function of energy alone). Using this idea we are able to demonstrate that: (a) observational analyses that falsely assume spherical symmetry are made more accurate by imposing a strong prior preference for near-isotropic velocity dispersions in the centre of spheroids; (b) numerical simulations that use an idealized spherically symmetric setup can yield misleading results and should be avoided where possible; and (c) triaxial dark matter haloes (formed in collisionless cosmological simulations) nearly attain our maximally stable limit, but their evolution freezes out before reaching it.

  7. Simulation of microtearing turbulence in national spherical torus experiment

    SciTech Connect

    Guttenfelder, W.; Kaye, S. M.; Bell, R. E.; Hammett, G. W.; LeBlanc, B. P.; Mikkelsen, D. R.; Ren, Y.; Candy, J.; Nevins, W. M.; Wang, E.; Zhang, J.; Crocker, N. A.; Yuh, H.

    2012-05-15

    Thermal energy confinement times in National Spherical Torus Experiment (NSTX) dimensionless parameter scans increase with decreasing collisionality. While ion thermal transport is neoclassical, the source of anomalous electron thermal transport in these discharges remains unclear, leading to considerable uncertainty when extrapolating to future spherical tokamak (ST) devices at much lower collisionality. Linear gyrokinetic simulations find microtearing modes to be unstable in high collisionality discharges. First non-linear gyrokinetic simulations of microtearing turbulence in NSTX show they can yield experimental levels of transport. Magnetic flutter is responsible for almost all the transport ({approx}98%), perturbed field line trajectories are globally stochastic, and a test particle stochastic transport model agrees to within 25% of the simulated transport. Most significantly, microtearing transport is predicted to increase with electron collisionality, consistent with the observed NSTX confinement scaling. While this suggests microtearing modes may be the source of electron thermal transport, the predictions are also very sensitive to electron temperature gradient, indicating the scaling of the instability threshold is important. In addition, microtearing turbulence is susceptible to suppression via sheared E Multiplication-Sign B flows as experimental values of E Multiplication-Sign B shear (comparable to the linear growth rates) dramatically reduce the transport below experimental values. Refinements in numerical resolution and physics model assumptions are expected to minimize the apparent discrepancy. In cases where the predicted transport is strong, calculations suggest that a proposed polarimetry diagnostic may be sensitive to the magnetic perturbations associated with the unique structure of microtearing turbulence.

  8. Predictive Modeling of Tokamak Configurations*

    NASA Astrophysics Data System (ADS)

    Casper, T. A.; Lodestro, L. L.; Pearlstein, L. D.; Bulmer, R. H.; Jong, R. A.; Kaiser, T. B.; Moller, J. M.

    2001-10-01

    The Corsica code provides comprehensive toroidal plasma simulation and design capabilities with current applications [1] to tokamak, reversed field pinch (RFP) and spheromak configurations. It calculates fixed and free boundary equilibria coupled to Ohm's law, sources, transport models and MHD stability modules. We are exploring operations scenarios for both the DIII-D and KSTAR tokamaks. We will present simulations of the effects of electron cyclotron heating (ECH) and current drive (ECCD) relevant to the Quiescent Double Barrier (QDB) regime on DIII-D exploring long pulse operation issues. KSTAR simulations using ECH/ECCD in negative central shear configurations explore evolution to steady state while shape evolution studies during current ramp up using a hyper-resistivity model investigate startup scenarios and limitations. Studies of high bootstrap fraction operation stimulated by recent ECH/ECCD experiments on DIIID will also be presented. [1] Pearlstein, L.D., et al, Predictive Modeling of Axisymmetric Toroidal Configurations, 28th EPS Conference on Controlled Fusion and Plasma Physics, Madeira, Portugal, June 18-22, 2001. * Work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

  9. Superconducting magnet system for the TPX Tokamak

    SciTech Connect

    Hassenzahl, W.V.; Chaplin, M.R.; Heim, J.R.

    1993-09-15

    The Tokamak Physics Experiment (TPX) will be the first Tokamak using superconducting magnets for both the poloidal and toroidal field. It is designed for advanced Tokamak physics experiments in steady-state and long-pulse operation. The TPX superconducting magnets use an advanced cable-in-conduit conductor (CICC) design similar to that developed in support of the International Thermonuclear Experimental Reactor (ITER). The toroidal field magnets provide 4.0 T at 2.25 m with a stored energy of 1.05 GJ. The poloidal field magnets provide 18.0 V-s to ohmically start and control long burns of a 2.0 MA plasma.

  10. Next-step spherical torus experiment and spherical torus strategy in the course of development of fusion energy

    NASA Astrophysics Data System (ADS)

    Ono, M.; Peng, M.; Kessel, C.; Neumeyer, C.; Schmidt, J.; Chrzanowski, J.; Darrow, D.; Grisham, L.; Heitzenroeder, P.; Jarboe, T.; Jun, C.; Kaye, S.; Menard, J.; Raman, R.; Stevenson, T.; Viola, M.; Wilson, J.; Woolley, R.; Zatz, I.

    2004-03-01

    A spherical torus (ST) fusion energy development path which is complementary to the proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact component test facility (CTF) and high performance advanced regimes leading to more attractive Demo and power plant scale reactors. To provide the physical basis for the CTF an intermediate step needs to be taken, which we refer to as the 'next-step spherical torus' (NSST) device and which we examine in some detail herein. NSST is a 'performance extension' stage ST with a plasma current of 5-10 MA, R = 1.5 m, BT les 2.6 T and the possibility of varying physical parameters. The mission of NSST is to (1) provide a sufficient physical basis for the design of a CTF; (2) explore advanced operating scenarios with high bootstrap current fraction and high performance which can be utilized by CTF, Demo, and power plants; and (3) contribute to the general science of high bgr toroidal plasmas. The NSST is designed to utilize a TFTR-like site to minimize the cost and time required for design and construction.