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.
A Research Program of Spherical Tokamak in China
NASA Astrophysics Data System (ADS)
He, Ye-xi
2002-08-01
The mission of this program is to explore the spherical torus plasma with a SUNIST spherical tokamak. Main experiments in the start phase will be involved with breakdown and plasma current set-up with a mode of saving volt-second and without ohmic heating system, equilibrium and instability, current driving, heating and profile modification. The SUNIST is a university-scale conceptual spherical tokamak, with R = 0.3 m, A 1.3, Ip ~ 50 kA, BT < 0.15 T, and PRF = 100 kW. The only peculiarity of SUNIST is that there is a toroidal insulating break along the outer wall of vacuum vessel. The expected that advantages of this arrangement are helpful not only for saving flux swing, but also for having a deep understanding of what will influence the discharge startup and globe performances of plasma under different conditions of strong vessel eddy and ECR power assistance. Of course, the vessel structure of cross seal will be at a great risk of controlling vacuum quality, although we have achieved positive results on simulation test and vacuum vessel test.
The Reconstruction of the Plasma Boundary in the Sino-United Spherical Tokamak Experiments
NASA Astrophysics Data System (ADS)
Liu, Jian; Feng, Chunhua; Yang, Xuanzong; Wang, Long
2010-04-01
A method for the reconstruction of the plasma boundary in the sino-united spherical tokamak (SUNIST) based on the outer plasma magnetic diagnostics is reported. In SUNIST, the magnetic flux loop integral signals were measured recently and the plasma boundary could be reconstructed well with a current filament (CF) model by setting 2 to 8 current filaments. There are three additional filament positional parameters in addition to the filament current to minimize the square root error in the CF model. The plasma configuration obtained with the CF method is consistent with the visible plasma image from the CCD camera. The average difference in the minor radii for the plasma boundary, by applying the CF model and EFIT code, is below 6 mm.
Zhong, H. Tan, Y.; Liu, Y. Q.; Xie, H. Q.; Gao, Z.
2016-11-15
A single-channel 3 mm interferometer has been developed for plasma density diagnostics in the Sino-UNIted Spherical Tokamak (SUNIST). The extremely compact microwave interferometer utilizes one corrugated feed horn antenna for both emitting and receiving the microwave. The beam path lies on the equatorial plane so the system would not suffer from beam path deflection problems due to the symmetry of the cross section. A focusing lens group and an oblique vacuum window are carefully designed to boost the signal to noise ratio, which allows this system to show good performance even with the small-diameter central column itself as a reflector, without a concave mirror. The whole system discards the reference leg for maximum compactness, which is particularly suitable for the small-sized tokamak. An auto-correcting algorithm is developed to calculate the phase evolution, and the result displays good phase stability of the whole system. The intermediate frequency is adjustable and can reach its full potential of 2 MHz for best temporal resolution. Multiple measurements during ohmic discharges proved the interferometer’s capability to track typical density fluctuations in SUNIST, which enables this system to be utilized in the study of MHD activities.
NASA Astrophysics Data System (ADS)
Zhong, H.; Tan, Y.; Liu, Y. Q.; Xie, H. Q.; Gao, Z.
2016-11-01
A single-channel 3 mm interferometer has been developed for plasma density diagnostics in the Sino-UNIted Spherical Tokamak (SUNIST). The extremely compact microwave interferometer utilizes one corrugated feed horn antenna for both emitting and receiving the microwave. The beam path lies on the equatorial plane so the system would not suffer from beam path deflection problems due to the symmetry of the cross section. A focusing lens group and an oblique vacuum window are carefully designed to boost the signal to noise ratio, which allows this system to show good performance even with the small-diameter central column itself as a reflector, without a concave mirror. The whole system discards the reference leg for maximum compactness, which is particularly suitable for the small-sized tokamak. An auto-correcting algorithm is developed to calculate the phase evolution, and the result displays good phase stability of the whole system. The intermediate frequency is adjustable and can reach its full potential of 2 MHz for best temporal resolution. Multiple measurements during ohmic discharges proved the interferometer's capability to track typical density fluctuations in SUNIST, which enables this system to be utilized in the study of MHD activities.
Zhong, H; Tan, Y; Liu, Y Q; Xie, H Q; Gao, Z
2016-11-01
A single-channel 3 mm interferometer has been developed for plasma density diagnostics in the Sino-UNIted Spherical Tokamak (SUNIST). The extremely compact microwave interferometer utilizes one corrugated feed horn antenna for both emitting and receiving the microwave. The beam path lies on the equatorial plane so the system would not suffer from beam path deflection problems due to the symmetry of the cross section. A focusing lens group and an oblique vacuum window are carefully designed to boost the signal to noise ratio, which allows this system to show good performance even with the small-diameter central column itself as a reflector, without a concave mirror. The whole system discards the reference leg for maximum compactness, which is particularly suitable for the small-sized tokamak. An auto-correcting algorithm is developed to calculate the phase evolution, and the result displays good phase stability of the whole system. The intermediate frequency is adjustable and can reach its full potential of 2 MHz for best temporal resolution. Multiple measurements during ohmic discharges proved the interferometer's capability to track typical density fluctuations in SUNIST, which enables this system to be utilized in the study of MHD activities.
Do spherical tokamaks have a thermonuclear future?
NASA Astrophysics Data System (ADS)
Mirnov, S. V.
2012-12-01
This work has been initiated by the publication of a review by B.V.Kuteev et al., "Intense Fusion Neutron Sources" [Plasma Physics Reports 36, 281 (2010)]. It is stated that the key thesis of the above review that a spherical tokamak can be recommended for research neutron sources and for demonstration hybrid systems as an alternative to expensive "classical" tokamaks of the JET and ITER type is inconsistent. The analysis of the experimental material obtained during the last 10 years in the course of studies on the existing spherical tokamaks shows that the TIN-ST fusion neutron source spherical tokamak proposed by the authors of the review and intended, according to the authors' opinion, to replace "monsters" in view of its table-top dimensions (2 m3) and laboratory-level energetics cannot be transformed into any noticeable stationary megawatt-power neutron source competing with the existing classical tokamaks (in particular, with JET with its quasi-steady DT fusion power at a level of 5 MW). Namely, the maximum plasma current in the proposed tokamak will be not 3 MA, as the authors suppose erroneously, but, according to the present-day practice of spherical tokamaks, within 0.6-0.7 MA, which will lead to a reduction on the neutron flux by two to three orders of magnitude from the expected 5 MW. The possibility of the maintenance of the stationary process itself even in such a "weakened" spherical tokamak is very doubtful. The experience of the largest existing devices of this type (such as NSTX and MAST) has shown that they are incapable of operating even in a quasi-steady operating mode, because the discharge in them is spontaneously interrupted about 1 s after the beginning of the current pulse, although its expected duration is of up to 5 s. The nature of this phenomenon is the subject of further study of the physics of spherical tokamaks. This work deals with a critical analysis of the available experimental data concerning such tokamaks and a discussion of
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.
Assessment of the beam path deflection for a vertically installed microwave interferometer in SUNIST
Zhong, H.; Tan, Y.; Gao, Z.; Ling, B. L.
2016-08-15
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 n{sub e0}/n{sub c} = 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{sup 19} m{sup −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.
Overview of spherical tokamak research in Japan
NASA Astrophysics Data System (ADS)
Takase, Y.; Ejiri, A.; Fujita, T.; Fukumoto, N.; Fukuyama, A.; Hanada, K.; Idei, H.; Nagata, M.; Ono, Y.; Tanaka, H.; Uchida, M.; Horiuchi, R.; Kamada, Y.; Kasahara, H.; Masuzaki, S.; Nagayama, Y.; Oishi, T.; Saito, K.; Takeiri, Y.; Tsuji-Iio, S.
2017-10-01
Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.
LHCD Scenarios for Spherical Tokamak Plasmas
Takase, Y.; Ejiri, A.; Oosako, T.; Bonoli, P. T.; Wright, J. C.
2007-09-28
Noninductive plasma current start-up and sustainment are crucial issues for spherical tokamak reactors and other applications such as component test facility. It is widely recognized that the lower hybrid wave (the slow wave), which is most efficient in driving current, is not accessible to the core of a fully developed spherical tokamak plasma with very high dielectric constant. However, it may be useful in the initial plasma current ramp-up phase while the density is still low, where it is not practical to use other methods of noninductive current drive. Such a possibility is investigated theoretically for planned experiments on the TST-2 spherical tokamak at the University of Tokyo. The transmitters previously used for FWCD experiments on JFT-2M (200 MHz) are being prepared for this experiment. The combline antenna used for JFT-2M has been modified for use in TST-2. This antenna will be used to excite a unidirectional fast wave traveling in the toroidal direction with a toroidal mode number of 12 (corresponding to an initial parallel index of refraction of about 5). The fast wave can mode convert to the lower hybrid wave and drive current under some conditions. Examination of the dispersion relation indicates that there may be a suitable regime at relatively high field (0.3 T) and low density (<1x10{sup 19} m{sup -3})
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.
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.
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.
Energetic particles in spherical tokamak plasmas
NASA Astrophysics Data System (ADS)
McClements, K. G.; Fredrickson, E. D.
2017-05-01
Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together with higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion-ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of beam ion
Energetic particles in spherical tokamak plasmas
McClements, K. G.; Fredrickson, E. D.
2017-03-21
Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together withmore » higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion–ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of
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
Nonlinear fishbone dynamics in spherical tokamaks
NASA Astrophysics Data System (ADS)
Wang, Feng; Fu, G. Y.; Shen, Wei
2017-01-01
Linear and nonlinear kinetic-MHD hybrid simulations have been carried out to investigate linear stability and nonlinear dynamics of beam-driven fishbone instability in spherical tokamak plasmas. Realistic NSTX parameters with finite toroidal rotation were used. The results show that the fishbone is driven by both trapped and passing particles. The instability drive of passing particles is comparable to that of trapped particles in the linear regime. The effects of rotation are destabilizing and a new region of instability appears at higher q min (>1.5) values, q min being the minimum of safety factor profile. In the nonlinear regime, the mode saturates due to flattening of beam ion distribution, and this persists after initial saturation while mode frequency chirps down in such a way that the resonant trapped particles move out radially and keep in resonance with the mode. Correspondingly, the flattening region of beam ion distribution expands radially outward. A substantial fraction of initially non-resonant trapped particles become resonant around the time of mode saturation and keep in resonance with the mode as frequency chirps down. On the other hand, the fraction of resonant passing particles is significantly smaller than that of trapped particles. Our analysis shows that trapped particles provide the main drive to the mode in the nonlinear regime.
Nonlinear fishbone dynamics in spherical tokamaks
Wang, Feng; Fu, G. Y.; Shen, Wei
2016-11-22
Linear and nonlinear kinetic-MHD hybrid simulations have been carried out to investigate linear stability and nonlinear dynamics of beam-driven fishbone instability in spherical tokamak plasmas. Realistic NSTX parameters with finite toroidal rotation were used. Our results show that the fishbone is driven by both trapped and passing particles. The instability drive of passing particles is comparable to that of trapped particles in the linear regime. The effects of rotation are destabilizing and a new region of instability appears at higher q _{min} (>1.5) values, q _{min} being the minimum of safety factor profile. In the nonlinear regime, the mode saturates due to flattening of beam ion distribution, and this persists after initial saturation while mode frequency chirps down in such a way that the resonant trapped particles move out radially and keep in resonance with the mode. Correspondingly, the flattening region of beam ion distribution expands radially outward. Furthermore, a substantial fraction of initially non-resonant trapped particles become resonant around the time of mode saturation and keep in resonance with the mode as frequency chirps down. On the other hand, the fraction of resonant passing particles is significantly smaller than that of trapped particles. Finally, our analysis shows that trapped particles provide the main drive to the mode in the nonlinear regime.
Nonlinear fishbone dynamics in spherical tokamaks
Wang, Feng; Fu, G.Y.; Shen, Wei
2017-01-01
Linear and nonlinear kinetic-MHD hybrid simulations have been carried out to investigate linear stability and nonlinear dynamics of beam-driven fishbone instability in spherical tokamak plasmas. Realistic NSTX parameters with finite toroidal rotation were used. The results show that the fishbone is driven by both trapped and passing particles. The instability drive of passing particles is comparable to that of trapped particles in the linear regime. The effects of rotation are destabilizing and a new region of instability appears at higher q min (>1.5) values, q min being the minimum of safety factor profile. In the nonlinear regime, the mode saturates due to flattening of beam ion distribution, and this persists after initial saturation while mode frequency chirps down in such a way that the resonant trapped particles move out radially and keep in resonance with the mode. Correspondingly, the flattening region of beam ion distribution expands radially outward. A substantial fraction of initially non-resonant trapped particles become resonant around the time of mode saturation and keep in resonance with the mode as frequency chirps down. On the other hand, the fraction of resonant passing particles is significantly smaller than that of trapped particles. Our analysis shows that trapped particles provide the main drive to the mode in the nonlinear regime.
Nonlinear fishbone dynamics in spherical tokamaks
Wang, Feng; Fu, G. Y.; Shen, Wei
2016-11-22
Linear and nonlinear kinetic-MHD hybrid simulations have been carried out to investigate linear stability and nonlinear dynamics of beam-driven fishbone instability in spherical tokamak plasmas. Realistic NSTX parameters with finite toroidal rotation were used. Our results show that the fishbone is driven by both trapped and passing particles. The instability drive of passing particles is comparable to that of trapped particles in the linear regime. The effects of rotation are destabilizing and a new region of instability appears at higher q min (>1.5) values, q min being the minimum of safety factor profile. In the nonlinear regime, the mode saturatesmore » due to flattening of beam ion distribution, and this persists after initial saturation while mode frequency chirps down in such a way that the resonant trapped particles move out radially and keep in resonance with the mode. Correspondingly, the flattening region of beam ion distribution expands radially outward. Furthermore, a substantial fraction of initially non-resonant trapped particles become resonant around the time of mode saturation and keep in resonance with the mode as frequency chirps down. On the other hand, the fraction of resonant passing particles is significantly smaller than that of trapped particles. Finally, our analysis shows that trapped particles provide the main drive to the mode in the nonlinear regime.« less
The role of spherical torus in clarifying tokamak physics
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.
Thermal Fluid Multiphysics Optimization of Spherical Tokamak
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.
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.
Shape reconstruction of merging spherical tokamak plasma in UTST device
NASA Astrophysics Data System (ADS)
Ushiki, Tomohiko; Itagaki, Masafumi; Inomoto, Michiaki
2016-10-01
Spherical tokamak (ST) merging method is one of the ST start-up methods which heats the plasma through magnetic reconnection. In the present study reconstruction of eddy current profile and plasma shape was performed during spherical tokamak merging only using external sensor signals by the Cauchy condition surface (CCS) method. CCS method have been implemented for JT-60 (QST), QUEST (Kyushu University), KSTAR (NFRI), RELAX (KIT), and LHD (Nifs). In this method, CCS was assumed inside each plasmas, where both flux function and its normal derivative are unknown. Effect of plasma current was replaced by the boundary condition of CCS, assuming vacuum field everywhere. Also, the nodal points for the boundary integrals of eddy current density were set using quadratic elements in order to express the complicated vacuum vessel shape. Reconstructed profiles of the eddy current and the magnetic flux were well coincided with the reference in each phase of merging process. Magnetic sensor installation plan for UTST was determined from these calculation results. This work was supported by the JSPS A3 Foresight Program ``Innovative Tokamak Plasma Startup and Current Drive in Spherical Torus''.
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
Magneto-hydro-dynamic limits in spherical tokamaks
NASA Astrophysics Data System (ADS)
Hender, T. C.; Allfrey, S. J.; Akers, R.; Appel, L. C.; Bevir, M. K.; Buttery, R. J.; Gryaznevich, M.; Jenkins, I.; Kwon, O. J.; McClements, K. G.; Martin, R.; Medvedev, S.; Nightingale, M. P. S.; Ribeiro, C.; Roach, C. M.; Robinson, D. C.; Sharapov, S. E.; Sykes, A.; Villard, L.; Walsh, M. J.
1999-05-01
The operational limits observed in spherical tokamaks, notably the small tight aspect ratio tokamak (START) device [A. Sykes et al., Nucl. Fusion 32, 694 (1992)], are consistent with those found in conventional aspect ratio tokamaks. In particular the highest β achieved (˜40%) is consistent with an ideal magneto-hydro-dynamic (MHD) Troyon type limit, the upper limit on density is well described by the Greenwald density (πa2n¯e/Ip˜1) and the normalized current (Ip/aBt) is limited such that q95≳2. Stability calculations indicate scope for increasing both normalized β and normalized current beyond the values so far achieved, although wall stabilization is generally needed for low-n modes. In double null configurations current terminating disruptions occur at each of the operational boundaries, though the current quench tends to be slow at the density limit and disruptions at high β may be due to the low q. In early limiter START discharges, before the divertor coils were installed, disruptions rarely occurred. Instead internal reconnection events which have all the characteristics of a disruption except the current quench occurred. These various disruptive behaviors are explained in terms of a model in which helicity is conserved during the disruption. Due to the low toroidal field beam ions in START, and α particles in a ST power plant, are super-Alfvénic. This gives the possibility for toroidal Alfvén eigenmodes (TAEs) to occur and such modes are frequently observed in START neutral beam injection (NBI) discharges, but seem to be benign. The features of these observed TAEs are shown to be in agreement with MHD calculations.
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.
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.
Resolving electron scale turbulence in spherical tokamaks with flow shear
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.
Spherical tokamak Globus-M2: design, integration, construction
NASA Astrophysics Data System (ADS)
Minaev, V. B.; Gusev, V. K.; Sakharov, N. V.; Varfolomeev, V. I.; Bakharev, N. N.; Belyakov, V. A.; Bondarchuk, E. N.; Brunkov, P. N.; Chernyshev, F. V.; Davydenko, V. I.; Dyachenko, V. V.; Kavin, A. A.; Khitrov, S. A.; Khromov, N. A.; Kiselev, E. O.; Konovalov, A. N.; Kornev, V. A.; Kurskiev, G. S.; Labusov, A. N.; Melnik, A. D.; Mineev, A. B.; Mironov, M. I.; Miroshnikov, I. V.; Patrov, M. I.; Petrov, Yu. V.; Rozhansky, V. A.; Saveliev, A. N.; Senichenkov, I. Yu.; Shchegolev, P. B.; Shcherbinin, O. N.; Shikhovtsev, I. V.; Sladkomedova, A. D.; Solokha, V. V.; Tanchuk, V. N.; Telnova, A. Yu.; Tokarev, V. A.; Tolstyakov, S. Yu.; Zhilin, E. G.
2017-06-01
The Globus-M spherical tokamak has demonstrated practically all of the project objectives during the 15-year period of operation. The main factor limiting further progress in plasma performance is a relatively low toroidal magnetic field. The maximum toroidal magnetic field achieved on Globus-M was 0.4 T with the exception of a limited number of shots with 0.55 T, which led to damage of the toroidal field coil in 2002. The increase of the magnetic field up to 1.0 T together with the plasma current up to 0.5 MA will result in the significant enhancement of the operating parameters in the upgraded Globus-M2 machine. The experimental program will be focused on plasma heating and non-inductive current drive and will contribute to the creation of a physical and technological base for the compact fusion neutron source development. In the article a brief overview of the physical background for the machine upgrade is outlined. The current status of the project implementation is described. First experimental results on moderate magnetic field increase from 0.4 T up to 0.5 T in the existing Globus-M machine are discussed. The improvement of plasma confinement as well as enhancement of efficiency of the beam driven current is observed.
Optimization of magnetic field system for glass spherical tokamak GLAST-III
NASA Astrophysics Data System (ADS)
Ahmad, Zahoor; Ahmad, S.; Naveed, M. A.; Deeba, F.; Aqib Javeed, M.; Batool, S.; Hussain, S.; Vorobyov, G. M.
2017-04-01
GLAST-III (Glass Spherical Tokamak) is a spherical tokamak with aspect ratio A = 2. The mapping of its magnetic system is performed to optimize the GLAST-III tokamak for plasma initiation using a Hall probe. Magnetic field from toroidal coils shows 1/R dependence which is typical with spherical tokamaks. Toroidal field (TF) coils can produce 875 Gauss field, an essential requirement for electron cyclotron resonance assisted discharge. The central solenoid (CS) of GLAST-III is an air core solenoid and requires compensation coils to reduce unnecessary magnetic flux inside the vessel region. The vertical component of magnetic field from the CS in the vacuum vessel region is reduced to 1.15 Gauss kA-1 with the help of a differential loop. The CS of GLAST can produce flux change up to 68 mVs. Theoretical and experimental results are compared for the current waveform of TF coils using a combination of fast and slow capacitor banks. Also the magnetic field produced by poloidal field (PF) coils is compared with theoretically predicted values. It is found that calculated results are in good agreement with experimental measurement. Consequently magnetic field measurements are validated. A tokamak discharge with 2 kA plasma current and pulse length 1 ms is successfully produced using different sets of coils.
Formation of Spherical Tokamak by ECH without Center Solenoid in the LATE Device
Tanaka, H.; Uchida, M.; Yoshinaga, T.; Yamada, J.; Maekawa, T.; Yamaguchi, S.
2005-09-26
In the LATE device, Spherical Tokamak (ST) plasmas are formed by electron cyclotron heating (ECH) without center solenoid. Two types of experiments are described: (1) slow formation of ST plasmas and (2) spontaneous formation of ST plasmas with rapid-current-rise. A ST formation scenario is discussed including the current-drive mechanism, particle confinement and MHD equilibrium.
Nonlinear Fishbone Dynamics in Spherical Tokamaks with Toroidal Rotation
NASA Astrophysics Data System (ADS)
Wang, Feng; Fu, G. Y.
2015-11-01
Fishbone is ubiquitous in tokamak plasmas with fast ions. A numerical study of nonlinear dynamics of fishbone has been carried out in this work. Realistic parameters of NSTX are used to understand instability and nonlinear frequency chirping in tokamak plasmas. First, the effects of shear toroidal rotation are considered for fishbone instability. It's shown that with low qmin, it has small effects on the mode; while with high qmin, a new unstable region with a strong ballooning feature in mode structure appears. Second, a detailed study of nonlinear frequency chirping and energetic particles' dynamics is carried out. Linearly, the mode is driven by both trapped and passing particles, with dresonance condition ωd ~= ω for trapped particles and ωϕ +ωθ ~= ω for passing particles. As the mode grows, resonance particles oscillate and move outward in Pϕ space, which reduces particles' frequency. We believe that this is the main reason for the mode frequency chirping down. Finally, as the mode frequency chirping down, particles with lower orbit frequencies, which are non-resonant linearly, can turn into resonant particles in the nonlinear regime. This effect can sustain a quasi-steady state mode amplitude.
Plasma Current Start-up by ECW and Vertical Field in the TST-2 Spherical Tokamak
NASA Astrophysics Data System (ADS)
Mitarai, Osamu; Takase, Yuichi; Ejiri, Akira; Shiraiwa, Syunichi; Kasahara, Hiroshi; Yamada, Takuma; Ohara, Shinya; TST-2 Team; Nakamura, Kazuo; Iyomasa, Atsuhiro; Hasegawa, Makoto; Idei, Hiroshi; Sakamoto, Mizuki; Hanada, Kazuaki; Satoh, Kohnosuke; Zushi, Hideki; TRIAM Group; Nishino, Nobuhiro
Plasma current start-up and ramp-up to 10 kA have been demonstrated in the TST-2 spherical tokamak without the use of the central solenoid. Only the electron cyclotron wave (ECW) and the outer equilibrium field coils are used. The plasma current evolution depends on the poloidal coil arrangement. It is also demonstrated that the plasma current start-up can take place without the field null.
Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations
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.
Microstability in a ``MAST-like'' high confinement mode spherical tokamak equilibrium
NASA Astrophysics Data System (ADS)
Applegate, D. J.; Roach, C. M.; Cowley, S. C.; Dorland, W. D.; Joiner, N.; Akers, R. J.; Conway, N. J.; Field, A. R.; Patel, A.; Valovic, M.; Walsh, M. J.
2004-11-01
Gyrokinetic microstability analyses, with and without electromagnetic effects, are presented for a spherical tokamak plasma equilibrium closely resembling that from a high confinement mode (H mode) discharge in the mega-ampere spherical tokamak (MAST) [A. Sykes et al., Nucl. Fusion 41, 1423 (2001)]. Electrostatic ion temperature gradient driven modes (ITG modes) were found to be unstable on all surfaces, though they are likely to be substantially stabilized by equilibrium E×B flow shear. Electron temperature gradient driven modes (ETG modes) have stronger growth rates that substantially exceed the equilibrium flow shearing rates. Mixing length arguments suggest that ITG modes would give rise to significant transport if they are not stabilized by sheared flows, and predict weak transport from ETG turbulence. Significant plasma flows have been neglected in this first analysis, and are probably important in the delicate balance between ITG growth rates and flow shear, and in the formation of internal transport barriers on MAST. Electromagnetic effects are found to be important even in this low β discharge, especially for longer length-scale modes with k⊥ρi
Bounce Precession Fishbones in the National Spherical Tokamak Experiment
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.
NASA Astrophysics Data System (ADS)
Sykes, Alan
1997-05-01
The world's first high-power auxiliary heating experiments in a tight aspect ratio (or spherical) tokamak have been performed on the Small Tight Aspect Ratio Tokomak (START) device [Sykes et al., Nucl. Fusion 32, 694 (1992)] at Culham Laboratory, using the 40 keV, 0.5 MW Neutral Beam Injector loaned by the Oak Ridge National Laboratory. Injection has been mainly of hydrogen into hydrogen or deuterium target plasmas, with a one-day campaign to explore D→D operation. In each case injection provides a combination of higher density operation and effective heating of both ions and electrons. The highest β values achieved to date in START are volume average βT˜11.5% and central beta βO˜50%. Already high, these values are expected to increase further with the use of higher beam power.
Nonlocal neoclassical transport in tokamak and spherical torus experiments
Wang, W. X.; Rewoldt, G.; Tang, W. M.; Hinton, F. L.; Manickam, J.; Zakharov, L. E.; White, R. B.; Kaye, S.
2006-08-15
Large ion orbits can produce nonlocal neoclassical effects on ion heat transport, the ambipolar radial electric field, and the bootstrap current in realistic toroidal plasmas. Using a global {delta}f particle simulation, it is found that the conventional local, linear gradient-flux relation is broken for the ion thermal transport near the magnetic axis. With regard to the transport level, it is found that details of the ion temperature profile determine whether the transport is higher or lower when compared with the predictions of standard neoclassical theory. Particularly, this nonlocal feature is suggested to exist in the National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, Y.-K. M. Peng et al., Nucl. Fusion 40, 557 (2000)], being consistent with NSTX experimental evidence. It is also shown that a large ion temperature gradient can increase the bootstrap current. When the plasma rotation is taken into account, the toroidal rotation gradient can drive an additional parallel flow for the ions and then additional bootstrap current, either positive or negative, depending on the gradient direction. Compared with the carbon radial force balance estimate for the neoclassical poloidal flow, our nonlocal simulation predicts a significantly deeper radial electric field well at the location of an internal transport barrier of an NSTX discharge.
Next-step-targeted experiments on the Mega-Amp Spherical Tokamak
NASA Astrophysics Data System (ADS)
Gryaznevich, M.; Akers, R. J.; Counsell, G. F.; Cunningham, G.; Dnestrovskij, A.; Field, A. R.; Hender, T. C.; Kirk, A.; Lloyd, B.; Meyer, H.; Morris, A. W.; Sykes, A.; Tabasso, A.; Valovic, M.; Voss, G. M.; Wilson, H. R.
2003-05-01
Since its first physics campaign, the principal parameters on MAST (Mega-Amp Spherical Tokamak) [A. Sykes et al., Nuclear Fusion 41, 1423 (2001)] have been brought up towards their design values. Considerable advances have been made in a range of physics areas of direct relevance to the International Thermonuclear Experimental Reactor (ITER) [ITER Physics Basis, Nuclear Fusion 39, 2175 (1999)]. In this paper, results on H-mode access, global confinement and pedestal studies are presented and compared with conventional aspect ratio tokamak scalings. Physics and engineering requirements relevant to next step spherical tokamak devices are discussed, in particular the plasma formation, current ramp-up and sustainment, and plasma exhaust. Results of first experiments directly targeting these issues are presented: Plasma current up to 0.5 MA has been produced without use of the central solenoid flux, and current ramp-up and sustainment without use of the central solenoid flux has been demonstrated. Experiments on neutral beam heating and current drive (CD) demonstrate up to 50% bootstrap current fraction and good CD efficiency, and divertor power loading has been found to be tolerable and have a favorable outboard asymmetry.
Fast scanning probe for the NSTX spherical tokamak
Boedo, J. A.; Crocker, N.; Chousal, L.; Hernandez, R.; Chalfant, J.; Kugel, H.; Roney, P.; Wertenbaker, J.; Collaboration: NSTX Team
2009-12-15
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 {approx}7 m/s and {approx}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/m{sup 2}. The National Spherical Torus Experiment SOL features electron temperature, T{sub e}{approx}10-30 eV, and electron density, n{sub e}{approx}0.1-5x10{sup 12} cm{sup -3} while the pedestal features n{sub e}{approx}0.5-1.5x10{sup 13} cm{sup -3} and T{sub e}{approx}30-150 eV. The probe described here has ten tips which obtain a wide spectrum of plasma parameters: electron temperature profile T{sub e}(r), electron density profile n{sub e}(r) and Mach number profile M(r), floating potential V{sub f}(r), poloidal and radial electric field profiles E{sub {theta}}(r) and E{sub {rho}}(r), saturation current profile I{sub sat}(r), and their fluctuations up to 3 MHz. We describe the probe and show representative radial profiles of various parameters.
Fast scanning probe for the NSTX spherical tokamak.
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.
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.
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.
Characteristics of major plasma discharge disruption in the Globus-M spherical tokamak
NASA Astrophysics Data System (ADS)
Sakharov, N. V.; Gusev, V. K.; Iblyaminova, A. D.; Kavin, A. A.; Kamenshchikov, S. N.; Kurskiev, G. S.; Lobanov, K. M.; Mineev, A. B.; Patrov, M. I.; Petrov, Yu. V.; Tolstyakov, S. Yu.
2017-04-01
The characteristics of the major disruption of plasma discharges in the Globus-M spherical tokamak are analyzed. The process of current quench is accompanied by the loss of the vertical stability of the plasma column. The plasma boundary during the disruption is reconstructed using the algorithm of movable filaments. The plasma current decay is preceded by thermal quench, during which the profiles of the temperature and electron density were measured. The data on the time of disruption, the plasma current quench rate, and the toroidal current induced in the tokamak vessel are compared for hydrogen and deuterium plasmas. It is shown that the disruption characteristics depend weakly on the ion mass and the current induced in the vessel increases with the disruption time. The decay rate of the plasma toroidal magnetic flux during the disruption is determined using diamagnetic measurements. Such a decay is a source of the poloidal current induced in the vessel; it may also cause poloidal halo currents.
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.
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.
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.
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.
Scaling of energy confinement time in the Globus-M spherical tokamak
NASA Astrophysics Data System (ADS)
Kurskiev, G. S.; Gusev, V. K.; Sakharov, N. V.; Bakharev, N. N.; Iblyaminova, A. D.; Shchegolev, P. B.; Avdeeva, G. F.; Kiselev, E. O.; Minaev, V. B.; Mukhin, E. E.; Patrov, M. I.; Petrov, Yu V.; Telnova, A. Yu; Tolstyakov, S. Yu
2017-04-01
The paper is devoted to an energy confinement study at the Globus-M spherical tokamak (ST). Experiments were performed in single null divertor configuration with elongation as high as 1.8–1.9 for variable plasma current and fixed toroidal magnetic field. The confinement time (τ E) dependence on density for ohmic-heated (OH) deuterium plasma is presented. It was found that τ E rises linearly with plasma current in H-mode with pure ohmic heating. Pronounced electron and ion heating was achieved in discharges with neutral beam injection at a moderate density level. The dependence of τ E on absorbed power was weak.
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.
Inductive plasma current start-up by the outer vertical field coil in a spherical tokamak
NASA Astrophysics Data System (ADS)
Mitarai, Osamu
1999-12-01
Plasma current-start up induced by an outer vertical field coil is studied during the ignition access phase in a spherical tokamak reactor. We have illustrated the concept that the plasma current of ~50 MA could be induced by the outer vertical field coil in the proposed spherical tokamak with the help of the small central solenoid flux of +/-5 V s and the strong heating power less than 100 MW for the internal inductance of icons/Journals/Common/ell" ALT="ell" ALIGN="TOP"/>i~0.4-0.8 without the help of bootstrap current and non-inductive current drive power. The required condition to achieve this operation scenario is that the flux produced by the equilibrium vertical field is larger than the inductive flux. Current start-up operation is achieved by adding the small ohmic heating solenoid flux for the flux waveform adjustment because the flux from the outer vertical field coil cannot solely induce the desired plasma current waveform in the case of the preprogramming of the heating power.
Wang, W. X.; Ethier, S.; Ren, Y.; ...
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
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
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 k_{θρs} ~ 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
Sykes, A.
1997-05-01
The world{close_quote}s first high-power auxiliary heating experiments in a tight aspect ratio (or spherical) tokamak have been performed on the Small Tight Aspect Ratio Tokomak (START) device [Sykes {ital et al.}, Nucl. Fusion {bold 32}, 694 (1992)] at Culham Laboratory, using the 40 keV, 0.5 MW Neutral Beam Injector loaned by the Oak Ridge National Laboratory. Injection has been mainly of hydrogen into hydrogen or deuterium target plasmas, with a one-day campaign to explore D{r_arrow}D operation. In each case injection provides a combination of higher density operation and effective heating of both ions and electrons. The highest {beta} values achieved to date in START are volume average {beta}{sub T}{approximately}11.5{percent} and central beta {beta}{sub O}{approximately}50{percent}. Already high, these values are expected to increase further with the use of higher beam power. {copyright} {ital 1997 American Institute of Physics.}
Simulation of current-filament dynamics and relaxation in the Pegasus Spherical Tokamak
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.
Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe
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.
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.
Far infrared tangential interferometry/polarimetry on the National Spherical Tokamak Experiment
NASA Astrophysics Data System (ADS)
Park, H. K.; Domier, C. W.; Geck, W. R.; Luhmann, N. C.
1999-01-01
Measurement of the core BT(r,t) value is essential in the National Spherical Tokamak Experiment (NSTX), since the effects of paramagnetism and diamagnetism in the NSTX are expected to be considerably greater than that in higher aspect ratio tokamaks. Therefore, without independent BT(r,t) measurement, plasma parameters dependent upon BT such as the q profile and local β value cannot be evaluated. Tangential interferometer/polarimeter systems (eight channels) [H. Park, L. Guttadora, C. Domier, W. R. Geck, and N. C. Luhman, Jr., First and Second NSTX Research Forums, Princeton, NJ, 1997 (unpublished)] for the NSTX will provide temporally and radially resolved toroidal field profile [BT(r,t)] and two-dimensional electron density profile [ne(r,t)] data. The outcome of the proposed system is extremely important to the study of confinement, heating, and stability of the NSTX plasmas. The research task is largely based on utilizing existing hardware from the TFTR multichannel infrared interferometer system [D. K. Mansfield, H. K. Park, L. C. Johnson, H. Anderson, S. Foote, B. Clifton, and C. H. Ma, Appl. Opt. 26, 4469 (1987) and H. K. Park, D. K. Mansfield, and C. L. Johnson, Proceedings of the 3rd International Symposium on Laser-Aided Plasma Diagnostic, Los Angeles, CA, 28-30 Oct. 1987 (unpublished), pp. 96-104] which will be reconfigured into a tangential system for NSTX, and to develop the additional hardware required to complete the system.
Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe.
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.
Modelling the power deposition into a spherical tokamak fusion power plant
NASA Astrophysics Data System (ADS)
Windsor, C. G.; Morgan, J. G.; Buxton, P. F.; Costley, A. E.; Smith, G. D. W.; Sykes, A.
2017-03-01
Numerical studies have been made to improve the performance of the central column of a superconducting spherical tokamak fusion pilot plant. The assumed neutron shield includes concentric layers of tungsten carbide and water. The relative thickness of the water layers was varied and a minimum power deposition was found at about 17% of water. It was found advantageous to have an approximately 1.7 times thicker water layer next to the core and a similarly thinner layer next to the plasma. The use of tungsten boride instead of tungsten carbide was shown to make an improvement especially if placed close to the central superconducting core, the inner layer alone reducing the power deposition by 29%. Engineering features such as a central steel tie-bar, an insulating thermal vacuum gap, a wall gap next to the plasma and knowledge of the vertical energy distribution are essential to a successful design and their effects on the power deposition are shown in an appendix. The results have been fitted to model distributions and incorporated into the Tokamak Energy System Code, which can then give predictions of the power deposition as a function of other parameters such as the plasma major radius and the maximum magnetic field permitted on the superconductors.
Diamagnetic Fishbone Mode Associated with Circulating Fast Ions in Spherical Tokamaks
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.
Simulation of High-Harmonic Fast-Wave Heating on the National Spherical Tokamak Experiment
Green, David L; Jaeger, Erwin Frederick; Chen, Guangye; Berry, Lee A; Pugmire, Dave; Canik, John; Ryan, Philip Michael
2011-01-01
Images associated with radio-frequency heating of low-confinement mode plasmas in the National Spherical Tokamak Experiment, as calculated by computer simulation, are presented. The AORSA code has been extended to simulate the whole antenna-to-plasma heating system by including both the kinetic physics of the well-confined core plasma and a poorly confined scrape-off plasma and vacuum vessel structure. The images presented show the 3-D electric wave field amplitude for various antenna phasings. Visualization of the simulation results in 3-D makes clear that -30 degrees phasing excites kilo-volt per meter coaxial standing modes in the scrape-off plasma and shows magnetic-field-aligned whispering-gallery type modes localized to the plasma edge.
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.
Core Plasma Characteristics of a Spherical Tokamak D-3He Fusion Reactor
NASA Astrophysics Data System (ADS)
Shi, Bingren
2005-04-01
The magnetic fusion reactor using the advanced D-3He fuels has the advantage of much less-neutron productions so that the consequent damages to the first wall are less serious. If the establishment of this kind of reactor becomes realistic, the exploration of 3He on the moon will be largely motivated. Based on recent progresses in the spherical torus (ST) research, we have physically designed a D-3He fusion reactor using the extrapolated results from the ST experiments and also the present-day tokamak scaling. It is found that the reactor size significantly depends on the wall reflection coefficient of the synchrotron radiation and of the impurity contaminations. The secondary reaction between D-D that promptly leads to the D-T reaction producing 14 MeV neutrons is also estimated. Comparison of this D-3He ST reactor with the D-T reactor is made.
Reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak
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.
Ignition and burn criteria for D/sup 3/He tokamak and spherical torus reactors
Galambos, J.D.; Peng, Y-K. M.
1989-01-01
D-/sup 3/He ignition and burn criteria for tokamaks and spherical torus reactors are examined in a global analysis with profile corrections. Particle confinement and ash buildup effects are included with the power balance, which results in an increased sensitivity of the ignition criteria to losses via brehmsstrahlung and synchrotron radiation. Plasma beta scaling via an /epsilon//beta//sub p/ limit provides the needed aspect ratio (A) dependence, and permits an analysis in all A values of the first and second stability regimes. Energy confinement time (/tau//sub E/) associated with particle diffusion (/tau//sub p/) and energy conduction (/tau//sub c/) are used; parabolic profile are assumed with exponents /alpha//sub n/ = 1.0 and /alpha//sub T/ = 1.5; and we set /tau//sub p/ = 2/tau//sub c/. The ignition condition for minimum n/tau//sub E/ is found to be sensitive to /beta/ but not to the magnetic field. Steady state burn in second stability tokamaks (/epsilon//beta//sub p/ /ge/ 0.6) at high A (>4) with average synchrotron wall reflectivities below 95% requires n/tau//sub E/ above 5 /times/ 10/sup 21/ m/sup /minus/3/s or strong plasma elongation (/kappa/ > 3). Ignition in a spherical tori can be achieved with wall reflectivities below 80% and at n/tau//sub E/ /le/ 10/sup 21/ m/sup /minus/3/ s, without requiring strong plasma shaping or /epsilon//beta//sub p/ > 0.6. The need to minimize n/tau//sub E/ for ignition and burn strongly limits the synchrotron radiation loss to less than 20% of the fusion power for all values of A. Synchrotron power fractions can be increased to 40% by increasing n/tau//sub E/ to its upperbound of ignition. Further increases of this fraction can be obtained only by assuming preferential ash removal. 19 refs., 8 figs.
Anomalous ion heating during helicity injection in the Pegasus spherical tokamak
NASA Astrophysics Data System (ADS)
Burke, Marcus Galen
Plasmas in the Pegasus spherical tokamak are initiated and grown by local helicity injection (LHI) current drive, resulting in toroidal plasma current Ip > 180 kA with 5 kA of injected current. The LHI system consists of 3 adjacent electron current sources that inject helical current streams into the plasma edge and generate toroidal current through magnetic reconnection of adjacent passes of the injected current helix. Anomalously high impurity ion temperatures are observed during LHI with the OV Ti,OV≤ 650 eV, which is in contrast to T i,OV≤ 70 eV from ohmic heating alone. Spatial profiles of Ti,OV indicate an edge localized heating source with Ti,OVapprox 650 eV near the outboard major radius of the injectors dropping to Ti,OVapprox 150 eV near the plasma magnetic axis. The location of this anomalously high Ti is in agreement with the predicted location of reconnection activity by simulations of LHI current drive. Experiments without a background tokamak plasma indicate the ion heating results from magnetic reconnection between adjacent current filaments of the multi-injector set. In these filaments-only experiments, the HeII Ti perpendicular to the magnetic field is found to scale with the reconnecting field strength, local density, and guide field, while Ti,∥ experiences little change, in agreement with two-fluid reconnection theory. In addition, the non-thermal HeII ion velocity distributions that are observed near the injectors with a tokamak plasma present are attributed to the collisional isotropization of the reconnection-driven Ti,⊥. Finally, the Ti evolution is broadly correlated with the amplitude of high frequency MHD activity and is not temporally correlated with the dominant, intermittent, n=1 mode that is postulated to be related to the LHI current drive mechanism. The ion heating described in this work does not significantly impact the LHI plasma performance as it doesn't contribute significantly to the electron heating. The power
Study of plasma heating in discharges with neutral beam injection in the Globus-M spherical tokamak
Ayushin, B. B.; Barsukov, A. G.; Gusev, V. K.; Esipov, L. A.; Zhilin, E. G.; Kurskiev, G. S.; Levin, R. G.; Leonov, V. M.; Minaev, V. B.; Patrov, M. I.; Petrov, Yu. V.; Sakharov, N. V.; Tilinin, G. N.; Tolstyakov, S. Yu.; Chernyshev, F. V.
2008-02-15
Results from experimental studies on the injection of high-energy neutral hydrogen beams into the plasma of the Globus-M spherical tokamak are reviewed. In the Introduction, the importance of these studies for implementing the controlled fusion research program and constructing the ITER tokamak is proved. Some problems related to the use of neutral beam injection in small and low-aspect-ratio tokamaks is analyzed. Results are presented from numerical simulations of the experiment by using the ASTRA transport code. It is shown that the use of neutral beam injection in the Globus-M tokamak ensures efficient ion heating and increases the plasma stored energy. The greater part of the review is devoted to the survey of experiments on the injection of 22-to 30-keV hydrogen and deuterium beams with a power of 0.4-0.8 MW into the plasma of the Globus-M spherical tokamak in a wide range of plasma currents and densities. The experimental results are analyzed and compared with the results of numerical simulations. The achievement of top plasma parameters is highlighted.
Drift kinetic effects on the plasma response in high beta spherical tokamak experiments
Wang, Zhirui; Park, Jong-Kyu; Menard, Jonathan E.; ...
2017-09-21
Highmore » $$\\beta$$ plasma response to the rotating n=1 external magnetic perturbations is numerically studied and compared with National Spherical Torus eXperiment (NSTX). The hybrid magnetohydrodynamic(MHD)-kinetic modeling shows the drift kinetic effects are important to resolve the disagreement of plasma response between the ideal MHD prediction and the NSTX experimental observation when plasma pressure reaches and exceeds the no-wall limit [F. Troyon et al., Plasma Phys. Control. Fusion 26, 209 (1984)]. Thus, since the external rotating fields and high plasma rotation are presented in NSTX experiments, the importance of resistive wall effect and plasma rotation on determining the plasma response is also identified, where the resistive wall suppresses the plasma response through the wall eddy current. The inertial energy, due to plasma rotation, destabilizes the plasma. Finally, the complexity of plasma response, in this study, indicates that MHD modeling, including comprehensive physics e.g. the drift kinetic effects, resistive wall and plasma rotation, is essential to reliably predict the plasma behavior in high beta spherical tokamak device.« less
Distinct turbulence sources and confinement features in the spherical tokamak plasma regime
Wang, W. X.; Ethier, S.; Ren, Y.; ...
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
Distinct turbulence sources and confinement features in the spherical tokamak plasma regime
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.
Measurement of eddy-current distribution in the vacuum vessel of the Sino-UNIted Spherical Tokamak.
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.
NASA Astrophysics Data System (ADS)
Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Mano, Junji
We applied the TOTAL (toroidal transport analysis linkage) simulation code for the analysis of the operational scenario of D-3He spherical tokamak reactors with high beta values and high bootstrap current fractions. Several technical elements, such as the control of the fuel ratio or selective exhaust of the α particle, need to be developed to establish steady-state burning. Negative magnetic shear configuration is a candidate for the high bootstrap current fraction operation.
Plasma diagnostics in spherical tokamaks with silicon charged-particle detectors
NASA Astrophysics Data System (ADS)
Netepenko, A.; Boeglin, W. U.; Darrow, D. S.; Ellis, R.; Sibilia, M. J.
2016-11-01
Detection of charged fusion products, such as protons and tritons resulting from D(d, p) t reactions, can be used to determine the position and time dependent fusion reaction rate profile in spherical tokamak plasmas with neutral beam heating. We have developed a prototype instrument consisting of 6 ion-implanted-silicon surface barrier detectors combined with collimators in such a way that each detector can accept 3 MeV protons and 1 MeV tritons and thus provides a curved view across the plasma cross section. The combination of the results from all six detectors will provide information on the spatial distribution of the fusion reaction rate. The expected time resolution of about 1 ms makes it possible to study changes in the reaction rate due to slow variations in the neutral beam density profile, as well as rapid changes resulting from MHD instabilities. Details of the new instrument, its data acquisition system, simulation results, and electrical noise testing results are discussed in this paper. First experimental data are expected to be taken during the current experimental campaign at NSTX-U.
Toroidal ripple transport of beam ions in the mega-ampère spherical tokamak
NASA Astrophysics Data System (ADS)
McClements, K. G.; Hole, M. J.
2012-07-01
The transport of injected beam ions due to toroidal magnetic field ripple in the mega-ampère 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 m2 s-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.
Characteristics of a novel lower hybrid wave antenna for the TST-2 spherical tokamak
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.
Plasma diagnostics in spherical tokamaks with silicon charged-particle detectors
Netepenko, A. Boeglin, W. U.; Darrow, D. S.; Ellis, R.; Sibilia, M. J.
2016-11-15
Detection of charged fusion products, such as protons and tritons resulting from D(d, p) t reactions, can be used to determine the position and time dependent fusion reaction rate profile in spherical tokamak plasmas with neutral beam heating. We have developed a prototype instrument consisting of 6 ion-implanted-silicon surface barrier detectors combined with collimators in such a way that each detector can accept 3 MeV protons and 1 MeV tritons and thus provides a curved view across the plasma cross section. The combination of the results from all six detectors will provide information on the spatial distribution of the fusion reaction rate. The expected time resolution of about 1 ms makes it possible to study changes in the reaction rate due to slow variations in the neutral beam density profile, as well as rapid changes resulting from MHD instabilities. Details of the new instrument, its data acquisition system, simulation results, and electrical noise testing results are discussed in this paper. First experimental data are expected to be taken during the current experimental campaign at NSTX-U.
Toroidal ripple transport of beam ions in the mega-ampere spherical tokamak
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.
Heat deposition into the superconducting central column of a spherical tokamak fusion plant
NASA Astrophysics Data System (ADS)
Windsor, C. G.; Morgan, J. G.; Buxton, P. F.
2015-02-01
A key challenge in designing a fusion power plant is to manage the heat deposition into the central core containing superconducting toroidal field coils. Spherical tokamaks have limited space for shielding the central core from fast neutrons produced by fusion and the resulting gamma rays. This paper reports a series of three-dimensional computations using the Monte Carlo N-particle code to calculate the heat deposition into the superconducting core. For a given fusion power, this is considered as a function of plasma major radius R0, core radius rsc and shield thickness d. Computations over the ranges 0.6 m ⩽ R0 ⩽ 1.6 m, 0.15 m ⩽ rsc ⩽ 0.25 m and 0.15 m ⩽ d ⩽ 0.4 m are presented. The deposited power shows an exponential dependence on all three variables to within around 2%. The additional effects of source profile, the outer shield and shield material are all considered. The results can be interpolated to 2% accuracy and have been successfully incorporated into a system code. A possible pilot plant with 174 MW of fusion is shown to lead to a heat deposition into the superconducting core of order 30 kW. An estimate of 1.7 MW is made for the cryogenic plant power necessary for heat removal, and of 88 s running time for an adiabatic experiment where the heat deposition is absorbed by a 10 K temperature rise.
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.
A high resolution Mirnov array for the Mega Ampere Spherical Tokamak.
Hole, M J; Appel, L C; Martin, R
2009-12-01
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.
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.
Compressional Alfvén eigenmodes in rotating spherical tokamak plasmas
Smith, H. M.; Fredrickson, E. D.
2017-02-07
Spherical tokamaks often have a considerable toroidal plasma rotation of several tens of kHz. Compressional Alfvén eigenmodes in such devices therefore experience a frequency shift, which if the plasma were rotating as a rigid body, would be a simple Doppler shift. However, since the rotation frequency depends on minor radius, the eigenmodes are affected in a more complicated way. The eigenmode solver CAE3B (Smith et al 2009 Plasma Phys. Control. Fusion 51 075001) has been extended to account for toroidal plasma rotation. The results show that the eigenfrequency shift due to rotation can be approximated by a rigid body rotationmore » with a frequency computed from a spatial average of the real rotation profile weighted with the eigenmode amplitude. To investigate the effect of extending the computational domain to the vessel wall, a simplified eigenmode equation, yet retaining plasma rotation, is solved by a modified version of the CAE code used in Fredrickson et al (2013 Phys. Plasmas 20 042112). Lastly, both solving the full eigenmode equation, as in the CAE3B code, and placing the boundary at the vessel wall, as in the CAE code, significantly influences the calculated eigenfrequencies.« less
Linear stability and nonlinear dynamics of the fishbone mode in spherical tokamaks
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.
Integrated predictive modeling simulations of the Mega-Amp Spherical Tokamak
NASA Astrophysics Data System (ADS)
Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.; Akers, Robert; Byrom, Calum; Sykes, Alan
2002-09-01
Integrated predictive modeling simulations are carried out using the BALDUR transport code [Singer et al., Comput. Phys. Commun. 49, 275 (1982)] for high confinement mode (H-mode) and low confinement mode (L-mode) discharges in the Mega-Amp Spherical Tokamak (MAST) [Sykes et al., Phys. Plasmas 8, 2101 (2001)]. Simulation results, obtained using either the Multi-Mode transport model (MMM95) or, alternatively, the mixed-Bohm/gyro-Bohm transport model, are compared with experimental data. In addition to the anomalous transport, neoclassical transport is included in the simulations and the ion thermal diffusivity in the inner third of the plasma is found to be predominantly neoclassical. The sawtooth oscillations in the simulations radially spread the neutral beam injection heating profiles across a broad sawtooth mixing region. The broad sawtooth oscillations also flatten the central temperature and electron density profiles. Simulation results for the electron temperature and density profiles are compared with experimental data to test the applicability of these models and the BALDUR integrated modeling code in the limit of low aspect ratio toroidal plasmas.
Compressional Alfvén eigenmodes in rotating spherical tokamak plasmas
NASA Astrophysics Data System (ADS)
Smith, H. M.; Fredrickson, E. D.
2017-03-01
Spherical tokamaks often have a considerable toroidal plasma rotation of several tens of kHz. Compressional Alfvén eigenmodes in such devices therefore experience a frequency shift, which if the plasma were rotating as a rigid body, would be a simple Doppler shift. However, since the rotation frequency depends on minor radius, the eigenmodes are affected in a more complicated way. The eigenmode solver CAE3B (Smith et al 2009 Plasma Phys. Control. Fusion 51 075001) has been extended to account for toroidal plasma rotation. The results show that the eigenfrequency shift due to rotation can be approximated by a rigid body rotation with a frequency computed from a spatial average of the real rotation profile weighted with the eigenmode amplitude. To investigate the effect of extending the computational domain to the vessel wall, a simplified eigenmode equation, yet retaining plasma rotation, is solved by a modified version of the CAE code used in Fredrickson et al (2013 Phys. Plasmas 20 042112). In summary, both solving the full eigenmode equation, as in the CAE3B code, and placing the boundary at the vessel wall, as in the CAE code, significantly influences the calculated eigenfrequencies.
Electron and Ion Heating Characteristics during Magnetic Reconnection in the MAST Spherical Tokamak.
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.
Linear stability and nonlinear dynamics of the fishbone mode in spherical tokamaks
NASA Astrophysics Data System (ADS)
Wang, Feng; Fu, G. Y.; Breslau, J. A.; Liu, J. Y.
2013-10-01
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 qmin 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 qmin values, consistent with the observed appearance of the fishbone before the "long-lived mode" in MAST and NSTX experiments. In contrast, at lower qmin 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.
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.
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.
Relevant parameter space and stability of spherical tokamaks with a plasma center column
NASA Astrophysics Data System (ADS)
Lampugnani, L. G.; Garcia-Martinez, P. L.; Farengo, R.
2017-02-01
A spherical tokamak (ST) with a plasma center column (PCC) can be formed inside a simply connected chamber via driven magnetic relaxation. From a practical perspective, the ST-PCC could overcome many difficulties associated with the material center column of the standard ST reactor design. Besides, the ST-PCC concept can be regarded as an advanced helicity injected device that would enable novel experiments on the key physics of magnetic relaxation and reconnection. This is because the concept includes not only a PCC but also a coaxial helicity injector (CHI). This combination implies an improved level of flexibility in the helicity injection scheme required for the formation and sustainment phases. In this work, the parameter space determining the magnetic structure of the ST-PCC equilibria is studied under the assumption of fully relaxed plasmas. In particular, it is shown that the effect of the external bias field of the PCC and the CHI essentially depends on a single parameter that measures the relative amount of flux of these two entities. The effect of plasma elongation on the safety factor profile and the stability to the tilt mode are also analyzed. In the first part of this work, the stability of the system is explained in terms of the minimum energy principle, and relevant stability maps are constructed. While this picture provides an adequate insight into the underlying physics of the instability, it does not include the stabilizing effect of line-tying at the electrodes. In the second part, a dynamical stability analysis of the ST-PCC configurations, including the effect of line-tying, is performed by numerically solving the magnetohydrodynamic equations. A significant stability enhancement is observed when the PCC contains more than the 70% of the total external bias flux, and the elongation is not higher than two.
Recent progress of magnetic reconnection research in the MAST spherical tokamak
NASA Astrophysics Data System (ADS)
Tanabe, Hiroshi
2016-10-01
In the last three years, magnetic reconnection research in the MAST spherical tokamak achieved major progress by use of new 32 chord ion Doppler tomography, 130 channel YAG- and 300 channel Ruby-Thomson scattering diagnostics. In addition to the significant plasma heating up to 1 keV, detailed full temperature profile measurements including the diffusion region have been achieved for the first time. 2D imaging measurements of Ti and Te profiles have revealed that magnetic reconnection mostly heats ions globally in the downstream region of outflow jet and electrons locally at the X-point. The higher toroidal field in MAST (Bt > 0.3 T) strongly inhibits cross-field thermal transport scaling as 1 /Bt2 and the characteristic peaked Te profile at the X point is sustained on a millisecond time scale. In contrast, ions are mostly heated in the downstream region of outflow acceleration inside the current sheet width (c /ωpi 0.1 m) and around the stagnation point formed by reconnected flux mostly by viscosity dissipation and shock-like compressional damping of the outflow jet. Toroidal confinement also contributes to the characteristic Ti profile, forming a ring structure aligned with the closed flux surface. There is an effective confinement of the downstream thermal energy due to a thick layer of reconnected flux. The characteristic structure is sustained for longer than an ion-electron energy relaxation time (τeiE 4 - 11 ms) and the energy exchange between ions and electrons contributes to the bulk electron heating in the downstream region. The toroidal guide field mostly contributes to the formation of a localized electron heating structure at the X-point but not to bulk ion heating downstream. This work was supported by Grant-in-Aid for Scientific Research 15H05750, 15K14279 and 15K20921.
Recent progress of magnetic reconnection research in the MAST spherical tokamak
NASA Astrophysics Data System (ADS)
Tanabe, H.; Yamada, T.; Watanabe, T.; Gi, 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.
2017-05-01
In the last three years, magnetic reconnection research in the MAST spherical tokamak achieved major progress by the use of new 32 chord ion Doppler tomography and 130 channel YAG and 300 channel Ruby Thomson scattering diagnostics. In addition to the previously achieved high power plasma heating during merging, detailed full temperature profile measurements including the diffusion region have been achieved for the first time. 2D imaging measurements of ion and electron temperature profiles have revealed that magnetic reconnection mostly heats ions globally in the downstream region of outflow jet and electrons locally around the X-point. The toroidal field in MAST "over 0.3T" strongly inhibits cross-field thermal transport, and the characteristic peaked electron temperature profile around the X-point is sustained on a millisecond time scale. In contrast, ions are mostly heated in the downstream region of outflow acceleration and around the stagnation point formed by reconnected flux mostly by viscosity dissipation and shock-like compressional damping of the outflow jet. Toroidal confinement also contributes to the characteristic ion temperature profile, forming a ring structure aligned with the closed flux surface. There is an effective confinement of the downstream thermal energy due to a thick layer of reconnected flux. The characteristic structure is sustained for longer than an ion-electron energy relaxation time ( ˜4 ms), and the energy exchange between ions and electrons contributes to the bulk electron heating in the downstream region. The toroidal guide field mostly contributes to the formation of a localized electron heating structure around the X-point but not to bulk ion heating downstream.
The poloidal distribution of turbulent fluctuations in the Mega-Ampere Spherical Tokamak
Antar, G.Y.; Counsell, G.; Ahn, J.-W.; Yang, Y.; Price, M.; Tabasso, A.; Kirk, A.
2005-03-01
Recently, it was shown that intermittency observed in magnetic fusion devices is caused by large-scales events with high radial velocity reaching about 1/10th of the sound speed (called avaloids or blobs) [G. Antar et al., Phys. Rev. Lett. 87 065001 (2001)]. In the present paper, the poloidal distribution of turbulence is investigated on the Mega-Ampere Spherical Tokamak [A. Sykes et al., Phys. Plasmas 8 2101 (2001)]. To achieve our goal, target probes that span the divertor strike points are used and one reciprocating probe at the midplane. Moreover, a fast imaging camera that can reach 10 {mu}s exposure time looks tangentially at the plasma allowing us to view a poloidal cut of the plasma. The two diagnostics allow us to have a rather accurate description of the particle transport in the poloidal plane for L-mode discharges. Turbulence properties at the low-field midplane scrape-off layer are discussed and compared to other poloidal positions. On the low-field target divertor plates, avaloids bursty signature is not detected but still intermittency is observed far from the strike point. This is a consequence of the field line expansion which transforms a structure localized in the poloidal plane into a structure which expands over several tens of centimeters at the divertor target plates. Around the X point and in the high-field side, however, different phenomena enter into play suppressing the onset of convective transport generation. No signs of intermittency are observed in these regions. Accordingly, like 'normal' turbulence, the onset of convective transport is affected by the local magnetic curvature and shear.
The poloidal distribution of turbulent fluctuations in the Mega-Ampère Spherical Tokamak
NASA Astrophysics Data System (ADS)
Antar, G. Y.; Counsell, G.; Ahn, J.-W.; Yang, Y.; Price, M.; Tabasso, A.; Kirk, A.
2005-03-01
Recently, it was shown that intermittency observed in magnetic fusion devices is caused by large-scales events with high radial velocity reaching about 1/10th of the sound speed (called avaloids or blobs) [G. Antar et al., Phys. Rev. Lett. 87 065001 (2001)]. In the present paper, the poloidal distribution of turbulence is investigated on the Mega-Ampère Spherical Tokamak [A. Sykes et al., Phys. Plasmas 8 2101 (2001)]. To achieve our goal, target probes that span the divertor strike points are used and one reciprocating probe at the midplane. Moreover, a fast imaging camera that can reach 10μs exposure time looks tangentially at the plasma allowing us to view a poloidal cut of the plasma. The two diagnostics allow us to have a rather accurate description of the particle transport in the poloidal plane for L-mode discharges. Turbulence properties at the low-field midplane scrape-off layer are discussed and compared to other poloidal positions. On the low-field target divertor plates, avaloids bursty signature is not detected but still intermittency is observed far from the strike point. This is a consequence of the field line expansion which transforms a structure localized in the poloidal plane into a structure which expands over several tens of centimeters at the divertor target plates. Around the X point and in the high-field side, however, different phenomena enter into play suppressing the onset of convective transport generation. No signs of intermittency are observed in these regions. Accordingly, like "normal" turbulence, the onset of convective transport is affected by the local magnetic curvature and shear.
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
Fusion nuclear science facilities and pilot plants based on the spherical tokamak
Menard, J. E.; Brown, T.; El-Guebaly, L.; Boyer, M.; Canik, J.; Colling, B.; Raman, R.; Wang, Z.; Zhai, Y.; Buxton, P.; Covele, B.; D’Angelo, C.; Davis, A.; Gerhardt, S.; Gryaznevich, M.; Harb, M.; Hender, T. C.; Kaye, S.; Kingham, D.; Kotschenreuther, M.; Mahajan, S.; Maingi, R.; Marriott, E.; Meier, E. T.; Mynsberge, L.; Neumeyer, C.; Ono, M.; Park, J. -K.; Sabbagh, S. A.; Soukhanovskii, V.; Valanju, P.; Woolley, R.
2016-08-16
Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is ${{R}_{0}}\\geqslant 1.7$ m, and a smaller R _{0} = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high
Fusion nuclear science facilities and pilot plants based on the spherical tokamak
Menard, J. E.; Brown, T.; El-Guebaly, L.; ...
2016-08-16
Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 ismore » $${{R}_{0}}\\geqslant 1.7$$ m, and a smaller R 0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high
Fusion nuclear science facilities and pilot plants based on the spherical tokamak
NASA Astrophysics Data System (ADS)
Menard, J. E.; Brown, T.; El-Guebaly, L.; Boyer, M.; Canik, J.; Colling, B.; Raman, R.; Wang, Z.; Zhai, Y.; Buxton, P.; Covele, B.; D'Angelo, C.; Davis, A.; Gerhardt, S.; Gryaznevich, M.; Harb, M.; Hender, T. C.; Kaye, S.; Kingham, D.; Kotschenreuther, M.; Mahajan, S.; Maingi, R.; Marriott, E.; Meier, E. T.; Mynsberge, L.; Neumeyer, C.; Ono, M.; Park, J.-K.; Sabbagh, S. A.; Soukhanovskii, V.; Valanju, P.; Woolley, R.
2016-10-01
A fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m-2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is {{R}0}≥slant 1.7 m, and a smaller R 0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R 0
Fusion nuclear science facilities and pilot plants based on the spherical tokamak
Menard, J. E.; Brown, T.; El-Guebaly, L.; Boyer, M.; Canik, J.; Colling, B.; Raman, R.; Wang, Z.; Zhai, Y.; Buxton, P.; Covele, B.; D’Angelo, C.; Davis, A.; Gerhardt, S.; Gryaznevich, M.; Harb, M.; Hender, T. C.; Kaye, S.; Kingham, D.; Kotschenreuther, M.; Mahajan, S.; Maingi, R.; Marriott, E.; Meier, E. T.; Mynsberge, L.; Neumeyer, C.; Ono, M.; Park, J. -K.; Sabbagh, S. A.; Soukhanovskii, V.; Valanju, P.; Woolley, R.
2016-08-16
Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is ${{R}_{0}}\\geqslant 1.7$ m, and a smaller R _{0} = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high
Collective fast ion instability-induced losses in National Spherical Tokamak Experiment
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 2
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.
Experimental study of toroidal Alfven modes in the Globus-M spherical tokamak
Petrov, Yu. V.; Patrov, M. I.; Gusev, V. K.; Ivanov, A. E.; Minaev, V. B.; Sakharov, N. V.; Tolstyakov, S. Yu.; Kurskiev, G. S.
2011-12-15
In the experiments carried out on the Globus-M tokamak in regimes with injection of 26-keV neutral beams with a power of 0.75-0.85 MW, two branches of instabilities excited by fast ions were observed in the early stage of a discharge: a low-frequency energetic particle mode (EPM) in the frequency range of 5-30 kHz and a high-frequency mode in the range of 50-200 kHz, identified as a toroidal Alfven eigenmode (TAE). The TAE developed in the initial phase of the discharge at q(0) > 1 and terminated when sawtooth oscillations were excited at q(0) < 1. The spectrum and spatial localization of the mode agree with predictions of the linear theory. The modes observed in the Globus-M tokamak possess both properties common to other tokamaks and their own specific features.
A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks
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.
A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks
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.
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.
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.
NASA Astrophysics Data System (ADS)
Tanabe, H.; Yamada, T.; Watanabe, T.; Gi, K.; Inomoto, M.; Imazawa, R.; Gryaznevich, M.; Scannell, R.; Conway, N. J.; Michael, C.; Crowley, B.; Fitzgerald, I.; Meakins, A.; Hawkes, N.; McClements, K. G.; Harrison, J.; O'Gorman, T.; Cheng, C. Z.; Ono, Y.; The MAST Team
2017-05-01
We present results of recent studies of merging/reconnection heating during central solenoid (CS)-free plasma startup in the Mega Amp Spherical Tokamak (MAST). During this process, ions are heated globally in the downstream region of an outflow jet, and electrons locally around the X-point produced by the magnetic field of two internal P3 coils and of two plasma rings formed around these coils, the final temperature being proportional to the reconnecting field energy. There is an effective confinement of the downstream thermal energy, due to a thick layer of reconnected flux. The characteristic structure is sustained for longer than an ion-electron energy relaxation time, and the energy exchange between ions and electrons contributes to the bulk electron heating in the downstream region. The peak electron temperature around the X-point increases with toroidal field, but the downstream electron and ion temperatures do not change.
NASA Astrophysics Data System (ADS)
Chai, Song; Xu, Yuhong; Gao, Zhe; Wang, Wenhao; Liu, Yangqing; Tan, Yi
2017-03-01
The characteristics of the energy transfer and nonlinear coupling among edge electromagnetic turbulence have been dedicatedly studied in various discharge stages at the sino-united spherical tokamak using multiple Langmuir and magnetic probe arrays. The wavelet bispectral analysis and the modified Kim's method are applied to investigate turbulence properties and their linear growth/damping and nonlinear energy transfer rates, along with multi-field turbulence interactions. The results show diverse features in the linear growth and nonlinear energy transfer between multi-field fluctuations during the current ramp-up, stationary, and internal connection event discharge phases. The diversity implies the importance to develop more sophisticated multi-field models to directly estimate the energy transfer rate among multiple turbulent fields.
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.
NASA Astrophysics Data System (ADS)
Carolan, P. G.; Patel, A.; Conway, N. J.; Akers, R. J.; Bunting, C. A.; Counsell, G. F.; Dowling, J.; Dunstan, M. R.; Kirk, A.; Lott, F.; Price, M. N.; Tournianski, M. R.; Walsh, M. J.
2004-10-01
The Mega Amp Spherical Torus (MAST) diagnostic needs are strongly influenced by physics goals that often require diagnostic integration and cross-mapping, especially in fine-scale investigations, such as transport barriers. Conversely, the unrivalled viewing access to the edge, scrape-off layer (SOL) and divertor regions, provided by the MAST open geometry, impacts on the physics program priorities. A supporting suite diagnostics, such as the high definition Thomson scattering systems, provide considerable added value in detailed data interpretation (e.g., bremsstrahlung emissivity in terms of Zeff). Thus, to exploit these advantages, an extensive set of high-resolution imaging diagnostics have been installed, encompassing soft x-rays, visible bremsstrahlung, charge exchange recognisation radiation, Dα from NBI, and edge plasma neutrals, and infrared (IR) from the divertor and wall regions. Plasma light collection optics provide near parallel illumination of narrow bandpass interference filters to give monochromatic images. One adaptation provides multiwavelength images; another accommodates smooth variation of wavelength across an image (e.g., for a range of Doppler shifts beam fast neutrals). Diagnostic synergy is enhanced by combining such diagnostics to common viewing optics which allow exact-mapping. Soft x-ray tangential imaging has been achieved by using a two dimensional charge coupled device detector in a pinhole camera. Finally, a fast IR camera monitors the power deposition on the first wall and divertor plates, important in quantifying power losses (e.g., ELMs, disruptions), and complemented by visible viewing of the SOL, and linear Dα cameras.
Carolan, P.G.; Patel, A.; Conway, N.J.; Akers, R.J.; Bunting, C.A.; Counsell, G.F.; Dowling, J.; Dunstan, M.R.; Kirk, A.; Lott, F.; Price, M.N.; Tournianski, M.R.; Walsh, M.J.
2004-10-01
The Mega Amp Spherical Torus (MAST) diagnostic needs are strongly influenced by physics goals that often require diagnostic integration and cross-mapping, especially in fine-scale investigations, such as transport barriers. Conversely, the unrivalled viewing access to the edge, scrape-off layer (SOL) and divertor regions, provided by the MAST open geometry, impacts on the physics program priorities. A supporting suite diagnostics, such as the high definition Thomson scattering systems, provide considerable added value in detailed data interpretation (e.g., bremsstrahlung emissivity in terms of Z{sub eff}). Thus, to exploit these advantages, an extensive set of high-resolution imaging diagnostics have been installed, encompassing soft x-rays, visible bremsstrahlung, charge exchange recognisation radiation, D{sub {alpha}} from NBI, and edge plasma neutrals, and infrared (IR) from the divertor and wall regions. Plasma light collection optics provide near parallel illumination of narrow bandpass interference filters to give monochromatic images. One adaptation provides multiwavelength images; another accommodates smooth variation of wavelength across an image (e.g., for a range of Doppler shifts beam fast neutrals). Diagnostic synergy is enhanced by combining such diagnostics to common viewing optics which allow exact-mapping. Soft x-ray tangential imaging has been achieved by using a two dimensional charge coupled device detector in a pinhole camera. Finally, a fast IR camera monitors the power deposition on the first wall and divertor plates, important in quantifying power losses (e.g., ELMs, disruptions), and complemented by visible viewing of the SOL, and linear D{sub {alpha}} cameras.
Antar, G.Y.
2006-05-15
This article describes the poloidal plasma particle distribution of type-III edge localized modes (ELMs) in the Mega-Ampere spherical tokamak [R.-J. Akers et al., Phys. Plasmas 9, 3919 (2002)]. A fast imaging camera with 10 {mu}s exposure time is used to record the D{sub {alpha}} light coming from the entire poloidal cross section. Furthermore, three sets of probes, triggered at the same time, acquired at 1 MHz, and located at different poloidal, radial, and toroidal locations in the tokamak are used. ELMs are observed to affect the D{sub {alpha}} emission throughout the low-field scrape-off layer; on the high-field side, however, this effect is found to be small. The results obtained by imaging agree with the pointwise measurements using Langmuir probes. The radial propagation is shown to occur at a speed of 250 m/s, whereas the toroidal convection from the top to the bottom of the plasma is shown to be consistent with a transport at the local sound speed. Strong correlation amplitudes are reported among the probes that are poloidally and toroidally separated by several meters. The study of the cross-correlation coefficients as a function of the frequency indicates that this correlation is caused by the low-frequency component of the signal and that the high-frequency part is not correlated. Consequently, the filamentary structures are interpreted as caused by the onset of turbulence during an ELM and do not constitute the ELM itself.
Note: Multi-pass Thomson scattering measurement on the TST-2 spherical tokamak
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.
Full orbit simulation of collisional transport of impurity ions in the MAST spherical tokamak
NASA Astrophysics Data System (ADS)
Romanelli, M.; McClements, K. G.; Cross, J.; Knight, P. J.; Thyagaraja, A.; Callaghan, J.
2011-05-01
Transport analysis of MAST discharges indicates that collisions are an important loss mechanism in the core of a tight aspect ratio tokamak. In the strongly varying equilibrium fields of MAST many of the assumptions of drift kinetic and neoclassical theory (e.g. small plasma inverse aspect ratio and low ratio of toroidal Larmor radius to poloidal Larmor radius) are not met by all particle species and it becomes appropriate to use full orbit analysis to evaluate heat and particle fluxes. Collisional transport of impurity ions (C6+ and W20+) has been studied using a full orbit solver, CUEBIT, to integrate the test-particle dynamics. Electromagnetic fields in MAST plasma have been modelled using the cylindrical and toroidal two-fluid codes CUTIE and CENTORI. A detailed study of the scaling of the test-particle diffusivity with collisionality in the equilibrium field reveals deviations from the standard neoclassical theory, in both the Pfirsch-Schlüter and banana regimes, and difficulties in defining a local diffusivity at low collisionalities. The effect of electric and magnetic fluctuations is also briefly addressed. It is found that field fluctuations enhance the non-diffusive nature of transport. The full orbit analysis presented here predicts levels of transport and confinement times for the examined species broadly consistent with the experimental observations.
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.
NASA Astrophysics Data System (ADS)
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.
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.
NASA Astrophysics Data System (ADS)
Jones, O. M.; Michael, C. A.; McClements, K. G.; Conway, N. J.; Crowley, B.; Akers, R. J.; Lake, R. J.; Pinches, S. D.; the MAST Team
2013-08-01
Using the recently installed fast-ion deuterium alpha (FIDA) spectrometer, the effects of low-frequency (20-50 kHz) chirping energetic particle modes with toroidal mode number n ⩾ 1 on the neutral beam injection-driven fast-ion population in Mega-Ampere Spherical Tokamak (MAST) plasmas are considered. Results from the FIDA diagnostic are presented and discussed in the light of the present theoretical understanding of these modes, known as fishbones, in plasmas with reversed shear. Measurements of the fast-ion population reveal strong redistribution of fast ions in both real and velocity space as a result of the fishbones. Time-resolved measurements throughout the evolution of a fishbone show radial redistribution of fast ions with energies up to 95% of the primary beam injection energy. Correlations between changes in the FIDA signal and the peak time derivative of the magnetic field perturbation are observed in a limited range of operating scenarios. The transient reduction in signal caused by a fishbone may in some cases reach 50% of the signal intensity before mode onset.
Raman, R; Nelson, B A; Bell, M G; Jarboe, T R; Mueller, D; Bigelow, T; Leblanc, B; Maqueda, R; Menard, J; Ono, M; Wilson, R
2006-10-27
A method of coaxial helicity injection has successfully produced a closed flux current without the use of the central solenoid in the NSTX device, on a size scale closer to a spherical torus reactor, for a proof-of-principle demonstration of this concept. For the first time, a remarkable 60 times current multiplication factor was achieved. Grad-Shafranov plasma equilibrium reconstructions are used to verify the existence of closed flux current. In some discharges the generated current persists for a surprisingly long time approximately 400 ms.
Raman, R.
2006-10-01
A method of coaxial helicity injection has successfully produced a closed flux current without the use of the central solenoid in the NSTX device, on a size scale closer to a spherical torus reactor, for a proof-of-principle demonstration of this concept. For the first time, a remarkable 60 times current multiplication factor was achieved. Grad-Shafranov plasma equilibrium reconstructions are used to verify the existence of closed flux current. In some discharges the generated current persists for a surprisingly long time ~400 ms.
The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak
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.
NASA Astrophysics Data System (ADS)
Hillesheim, J. C.; Crocker, N. A.; Peebles, W. A.; Meyer, H.; Meakins, A.; Field, A. R.; Dunai, D.; Carr, M.; Hawkes, N.; the MAST Team
2015-07-01
The high-k (7≲ {{k}\\bot}{ρi}≲ 11 ) wavenumber spectrum of density fluctuations has been measured for the first time in MAST (Lloyd et al 2003 Nucl. Fusion 43 1665). This was accomplished with the first implementation of Doppler backscattering (DBS) for core measurements in a spherical tokamak. DBS has become a well-established and versatile diagnostic technique for the measurement of intermediate- k ({{k}\\bot}{ρi}˜ 1 , and higher) density fluctuations and flows in magnetically confined fusion experiments. Previous implementations of DBS for core measurements have been in standard, large aspect ratio tokamaks. A novel implementation with two-dimensional (2D) steering was necessary to enable DBS measurements in MAST, where the large variation of the magnetic field pitch angle presents a challenge. We report on the scattering considerations and ray tracing calculations used to optimize the design and present data demonstrating measurement capabilities. Initial results confirm the applicability of the design and implementation approaches, showing the strong dependence of scattering alignment on the toroidal launch angle and demonstrating that DBS is sensitive to the local magnetic field pitch angle. We also present comparisons of DBS plasma velocity measurements with charge exchange recombination and beam emission spectroscopy measurements, which show reasonable agreement over most of the minor radius, but imply large poloidal flows approaching the magnetic axis in a discharge with an internal transport barrier. The 2D steering is shown to enable high-k measurements with DBS, at {{k}\\bot}>20 cm-1 ({{k}\\bot}{ρi}>10 ) for launch frequencies less than 75 GHz; this capability is used to measure the wavenumber spectrum of turbulence and we find \\mid n≤ft({{k}\\bot}\\right){{\\mid}2}\\propto k\\bot-4.7+/- 0.2 for {{k}\\bot}{ρi}≈ 7 -11, which is similar to the expectation for the turbulent kinetic cascade of \\mid n≤ft({{k}\\bot}\\right){{\\mid}2}\\propto
Formation of spherical tokamak equilibria by ECH in the LATE device
NASA Astrophysics Data System (ADS)
Maekawa, T.; Terumichi, Y.; Tanaka, H.; Uchida, M.; Yoshinaga, T.; Yamaguchi, S.; Igami, H.; Konno, M.; Katsuura, K.; Hayashi, K.; Abe, Y.; Yamada, J.; Maebara, S.; Imai, T.
2005-11-01
The main objective of the Low Aspect Ratio Torus Experiment (LATE) device is to demonstrate the formation of spherical torus (ST) plasmas by electron cyclotron heating (ECH) alone without a centre solenoid and establish its physical bases. By injecting a 2.45 GHz microwave pulse for 4 s, a plasma current of 1.2 kA is spontaneously initiated by P = 5 kW under a weak steady vertical field of Bv = 12 G and then ramped up slowly with a slow ramp-up of Bv for the equilibrium of the plasma loop and finally reaches 6.3 kA by P = 30 kW at Bv = 70 G. This current amounts to 10% of the total coil currents of 60 kA flowing through the centre post for the toroidal field. Magnetic measurements show that an ST equilibrium, having the last closed flux surface with an aspect ratio of R0/a sime 20.4 cm/14.5 cm sime 1.4, an elongation of κ sime 1.5 and qedge sime 37, has been produced and maintained for 0.5 s at the final stage of discharge. Spontaneous formation of ST equilibria under steady Bv fields, where plasma current increases rapidly in the time scale of a few milliseconds, is also effective and a plasma current of 6.8 kA is spontaneously generated and maintained at Bv = 85 G by a 5 GHz microwave pulse (130 kW, 60 ms). In both cases, the plasma centre locates near the second or third harmonic EC resonance layer and the line averaged electron density significantly exceeds the plasma cutoff density, suggesting that the harmonic EC heating by the mode-converted electron Bernstein waves supports the plasma.
NASA Astrophysics Data System (ADS)
Tani, K.; Shinohara, K.; Oikawa, T.; Tsutsui, H.; McClements, K. G.; Akers, R. J.; Liu, Y. Q.; Suzuki, M.; Ide, S.; Kusama, Y.; Tsuji-Iio, S.
2016-11-01
As part of the verification and validation of a newly developed non-steady-state orbit-following Monte-Carlo code, application studies of time dependent neutron rates have been made for a specific shot in the Mega Amp Spherical Tokamak (MAST) using 3D fields representing vacuum resonant magnetic perturbations (RMPs) and toroidal field (TF) ripples. The time evolution of density, temperature and rotation rate in the application of the code to MAST are taken directly from experiment. The calculation results approximately agree with the experimental data. It is also found that a full orbit-following scheme is essential to reproduce the neutron rates in MAST.
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.
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.
NASA Astrophysics Data System (ADS)
Shinya, T.; Takase, Y.; Yajima, S.; Moeller, C.; Yamazaki, H.; Tsujii, N.; Yoshida, Y.; Ejiri, A.; Togashi, H.; Toida, K.; Furui, H.; Homma, H.; Nakamura, K.; Roidl, B.; Sonehara, M.; Takahashi, W.; Takeuchi, T.
2017-03-01
Non-inductive plasma current start-up experiments were performed using the lower hybrid wave (LHW) on the TST-2 spherical tokamak. The density limit, observed in previous experiments using the outboard-launch antenna, disappeared after changing the plasma condition in the scrape-off layer, and the plasma current reached about 20 kA. In order to improve the LHW power deposition in the plasma core through an up-shift of the parallel wavenumber during the first pass through the plasma, a new top-launch antenna was designed, fabricated and installed. The plasma current ramp-up to 12 kA was achieved using the top-launch antenna alone in a preliminary experiment. Ray-tracing calculations using the measured plasma parameters showed a large up-shift during the first pass, satisfying the strong electron Landau damping condition in the plasma core.
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.
NASA Astrophysics Data System (ADS)
Chapman, I. T.; de Bock, M. F.; Pinches, S. D.; Turnyanskiy, M. R.; Mast Team; Igochine, V. G.; Maraschek, M.; Tardini, G.; ASDEX Upgrade Team
2009-07-01
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.
Nagashima, Yoshihiko; Oosako, Takuya; Takase, Yuichi; Ejiri, Akira; Watanabe, Osamu; Kobayashi, Hiroaki; Adachi, Yuuki; Tojo, Hiroshi; Yamaguchi, Takashi; Kurashina, Hiroki; Yamada, Kotaro; An, Byung Il; Kasahara, Hiroshi; Shimpo, Fujio; Kumazawa, Ryuhei; Hayashi, Hiroyuki; Matsuzawa, Haduki; Hiratsuka, Junichi; Hanashima, Kentaro; Kakuda, Hidetoshi; Sakamoto, Takuya; Wakatsuki, Takuma
2010-06-18
We present an observation of beat oscillation generation by coupled modes associated with parametric decay instability (PDI) during radio frequency (rf) wave heating experiments on the Tokyo Spherical Tokamak-2. Nearly identical PDI spectra, which are characterized by the coexistence of the rf pump wave, the lower-sideband wave, and the low-frequency oscillation in the ion-cyclotron range of frequency, are observed at various locations in the edge plasma. A bispectral power analysis was used to experimentally discriminate beat oscillation from the resonant mode for the first time. The pump and lower-sideband waves have resonant mode components, while the low-frequency oscillation is exclusively excited by nonlinear coupling of the pump and lower-sideband waves. Newly discovered nonlocal transport channels in spectral space and in real space via PDI are described.
Effect of toroidal Alfvén eigenmodes on fast particle confinement in the spherical tokamak Globus-M
NASA Astrophysics Data System (ADS)
Petrov, Yu. V.; Bakharev, N. N.; Gusev, V. K.; Minaev, V. B.; Kornev, V. A.; Kurskiev, G. S.; Patrov, M. I.; Sakharov, N. V.; Tolstyakov, S. Yu.; Shchegolev, P. B.
2015-12-01
> In experiments with neutral beam injection at the early stage of a Globus-M discharge, instabilities were observed that were excited by fast ions in the frequency range of 50-200 kHz, which were identified as toroidal Alfvén eigenmodes (TAE) (Petrov et al., Plasma Phys. Rep., vol. 37, 2011, pp. 1001-1005). In contradiction with the NSTX and MAST tokamaks, a regime of TAE generation was realized with strongly developed single modes. Magnetic measurements with fast Mirnov probes have shown that most of the modes have toroidal number . The influence of the modes on the fast particle confinement was recorded by means of a tangentially directed neutral particle analyser (NPA) and neutron detector. Hydrogen and deuterium were used as target plasma and injected beam for study of the isotopic effect. At deuterium injection into the deuterium plasma, TAE led to the neutron rate dropping by 25 %, whereas NPA fluxes of high energy dropped by 75 %. At hydrogen injection, the drop in the measured NPA fluxes did not exceed 25 %.
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’.
NASA Astrophysics Data System (ADS)
Wakatsuki, Takuma; Ejiri, Akira; Takase, Yuichi; Furui, Hirokazu; Hashimoto, Takahiro; Hiratsuka, Junichi; Kakuda, Hidetoshi; Kato, Kunihiko; Nakanishi, Ayaka; Oosako, Takuya; Shinya, Takahiro; Sonehara, Masateru; Togashi, Hiro; Yamaguchi, Takashi; Kasahara, Hiroshi; Kumazawa, Ryuhei; Saito, Kenji; Seki, Tetsuo; Shimpo, Fujio; Nagashima, Yoshihiko
2012-10-01
Plasma current start-up experiments were performed on the TST-2 spherical tokamak (R= 0.38 m, a = 0.25 m, Bt = 0.3 T, Ip = 0.1 MA) using the lower hybrid wave (LHW) at f = 200 MHz. A waveguide array antenna consisting of four dielectric (alumina, ɛr = 10.0) loaded waveguides was used. The coupling characteristics of this antenna were investigated by low power experiments (PFWD< 5 kW). The measured characteristics were qualitatively consistent with those predicted by calculations using a finite element method solver package (COMSOL). The experimentally observed reflection coefficient is large (greater than 36 % averaged over four waveguides), and there are large differences in reflectivities in neighboring waveguides. It was necessary to take into account of the private limiter surrounding the antenna in order to reproduce these features. Non-inductive plasma current start-up to 6 kA has been demonstrated using 20 kW of LHW power. In this experiment, the reflection coefficient was very high because the initial plasma density was much lower than the predicted optimum plasma density.
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.
Raman, R; Mueller, D; Nelson, B A; Jarboe, T R; Gerhardt, S; Kugel, H W; Leblanc, B; Maingi, R; Menard, J; Ono, M; Paul, S; Roquemore, L; Sabbagh, S; Soukhanovskii, V
2010-03-05
Transient coaxial helicity injection (CHI) started discharges in the National Spherical Torus Experiment (NSTX) have attained peak currents up to 300 kA and when coupled to induction, it has produced up to 200 kA additional current over inductive-only operation. CHI in NSTX has shown to be energetically quite efficient, producing a plasma current of about 10 A/J of capacitor bank energy. In addition, for the first time, the CHI-produced toroidal current that couples to induction continues to increase with the energy supplied by the CHI power supply at otherwise similar values of the injector flux, indicating the potential for substantial current generation capability by CHI in NSTX and in future toroidal devices.
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.
Plasma Physics Regimes in Tokamaks with Li Walls
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.
Features of spherical torus plasmas
Peng, Y.K.M.; Strickler, D.J.
1985-12-01
The spherical torus is a very small aspect ratio (A < 2) confinement concept obtained by retaining only the indispensable components inboard to the plasma torus. MHD equilibrium calculations show that spherical torus plasmas with safety factor q > 2 are characterized by high toroidal beta (..beta../sub t/ > 0.2), low poloidal beta (..beta../sub p/ < 0.3), naturally large elongation (kappa greater than or equal to 2), large plasma current with I/sub p//(aB/sub t0/) up to about 7 MA/mT, strong paramagnetism (B/sub t//B/sub t0/ > 1.5), and strong plasma helicity (F comparable to THETA). A large near-omnigeneous region is seen at the large-major-radius, bad-curvature region of the plasma in comparison with the conventional tokamaks. These features combine to engender the spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost. Because of its strong paramagnetism and helicity, the spherical torus plasma shares some of the desirable features of spheromak and reversed-field pinch (RFP) plasmas, but with tokamak-like confinement and safety factor q. The general class of spherical tori, which includes the spherical tokamak (q > 1), the spherical pinch (1 > q > O), and the spherical RFP (q < O), have magnetic field configurations unique in comparison with conventional tokamaks and RFPs. 22 refs., 12 figs.
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.
Recent Progress on Spherical Torus Research
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.
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.
Cash boost for UK’s MAST tokamak
NASA Astrophysics Data System (ADS)
2017-06-01
The Mega Amp Spherical Tokamak (MAST) at the Culham Centre for Fusion Energy in Oxfordshire, UK, has received £21m for a series of upgrades to study the best way to extract waste fuel from the plasma it contains.
Dust in tokamaks: An overview of the physical model of the dust in tokamaks code
NASA Astrophysics Data System (ADS)
Bacharis, Minas; Coppins, Michael; Allen, John E.
2010-04-01
The dynamical behavior of dust produced in tokamaks is an important issue for fusion. In this work, the current status of the dust in tokamaks (DTOKS) [J. D. Martin et al., Europhys Lett. 83, 65001 (2008)] dust transport code will be presented. A detailed description of the various elements of its underlying physical model will be given together with representative simulation results for the mega amp spherical tokamak (MAST) [A. Sykes et al., Nucl. Fusion 41, 1423 (2001)]. Furthermore, a brief description of the various components of the dust transport (DUSTT) [R. D. Smirnov et al., Plasma Phys. Controlled Fusion 49, 347 (2007)] code will also be presented in comparison with DTOKS.
Spherical tokamak (ST) transmutation of nuclear wastes
Peng, Yueng Kay Martin; Cheng, E.T.; Galambos, John D; Cerbone, R. J.
1995-01-01
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 reduction and energy production are discussed.
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.
Recent progress on spherical torus research
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.
Furth, H.P.
1984-10-01
The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT.
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.
The physics of tokamak start-up
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.
The Physics of Tokamak Start-up
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.
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.
Physics of Spherical Torus Plasmas
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
Diagnosing transient plasma status: from solar atmosphere to tokamak divertor
NASA Astrophysics Data System (ADS)
Giunta, A. S.; Henderson, S.; O'Mullane, M.; Harrison, J.; Doyle, J. G.; Summers, H. P.
2016-09-01
This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.
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.
Options for commercial tokamaks
Dabiri, A.E.; Keeton, D.C.; Thomson, S.L.
1986-07-01
Systems studies have been performed at the Fusion Engineering Design Center (FEDC) to assess commercial tokamak options. One study investigates the economics of high-beta operation and determines an optimum operating range of 10 to 20% beta, with a corresponding neutron wall loading of 6 to 8 MW/m/sup 2/. A second study determines conditions under which small, low-power tokamaks can be economically combined into a 1200-MW(electric) multiplex power plant. The results of these studies have directed future efforts at the FEDC toward a high-beta, tokamak design using a modular maintenance configuration.
Baker, C.C.
1981-01-01
This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features.
Murray, J.G.; Gorker, G.E.
1985-01-01
Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.
Modular tokamak magnetic system
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.
Steady state self-induced current in tokamak
Gott, Yu. V.; Yurchenko, E. I.
2009-11-15
A model, which may make it possible to self-consistently calculate the self-driven current in tokamaks taking into account asymmetry and bootstrap currents, is presented. It is shown that the described self-driven current can provide steady-state tokamak operation without the seed current produced with the help of additional methods. The total self-consistent, self-driven current does not depend on magnetic field magnitude and is proportional to the square root from plasma pressure. The experimental data obtained in the National Spherical Torus Experiment are satisfactorily described by this model.
Advanced commercial tokamak study
Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.
1985-12-01
Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs.
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.
Completely bootstrapped tokamak
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.
Steiner, D.; Embrechts, M.
1990-07-01
This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein.
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.
White, R. B.
2008-05-14
This lecture gives a basic introduction to magnetic pound elds, magnetic surface destruction, toroidal equilibrium and tearing modes in a tokamak, including the linear and nonlinear development of these modes and their modi pound cation by current drive and bootstrap current, and sawtooth oscillations and disruptions.
Thomson, S.L.
1985-01-01
This report is concerned with the modular tokamak configuration, and presents information on the following topics: modularity; external vacuum boundary; vertical maintenance; combined reactor building/biological shield with totally remote maintenance; independent TF coils; minimum TF coil bore; saddle PF coils; and heat transport system in bore.
Engineering Overview of the National Spherical Torus Experiment (NSTX)
Neumeyer, C. Author
1997-01-01
The National Spherical Torus Experiment (NSTX) Project will provide a national facility for the study of plasma confinement, heating, and current drive in a low-aspect-ratio, spherical torus (ST) configuration. The ST configuration is an alternate confinement concept which is characterized by high beta, high elongation, high bootstrap fraction, and low toroidal magnetic field compared to conventional tokamaks. The NSTX is the next-step ST experiment following smaller experiments such as the Princeton Plasma Physics Laboratory CDX-U (Current Drive Experiment-Upgrade), the START (Small Tight Aspect Ratio Tokamak) at Culham Laboratory, UK, and the HIT (Helicity Injected Tokamak) at the University of Washington, and it is smaller in scale to the MAST (Meg-Amp Spherical Tokamak) machine now under construction at Culham.This paper provides a description of the NSTX mission and gives an overview of the main engineering features of the design of the machine and facility and discusses some of the key design solutions.
Sheffield, J.
1981-06-01
During the past two decades, the tokamak program has made significant advances. As a result of these advances, the program has expanded, and construction is nearly completed of large facilities, including the US Tokamak Fusion Test Reactor (TFTR), that are capable of achieving energy break-even in the plasma. Design work is under way on the next application phase, exemplified by the US Fusion Engineering Device (FED) and the IAEA INTOR programs. There are many excellent reviews that give a broad history of the evolution of the tokamak program. In addition, reviews are available that give detailed descriptions of individual tokamak elements. This paper reviews the status of the tokamak program, concentrating on recent advances and areas important to the further advancement of the tokamak concept.
HHFW Experiments on the TST-2 Spherical Tokamak
NASA Astrophysics Data System (ADS)
Takase, Y.; Ejiri, A.; Iijima, D.; Kasahara, H.; Kasuya, N.; Kobori, Y.; Nagashima, Y.; Nishi, T.; Shiraiwa, S.; Taniguchi, T.; Ushigome, M.; Wada, H.; Yamada, T.; Yamagishi, K.
2001-10-01
A 6-element combline antenna was used in TST-2 to excite a unidirectional high-harmonic fast wave (HHFW) at low-power (1 kW level). This antenna behaves as a band-pass filter with a passband of 22--28 MHz. At 25 MHz the excited toroidal wavenumber is 13 m-1 at R = 0.57 m, which corresponds to a toroidal refractive index of 25 and a toroidal mode number of 7.4. Antenna current measurements indicated nearly complete radiation into the plasma from the first two radiating elements. Such high plasma loading is advantageous for conventional loop antennas for heating, but in order to excite a wave spectrum with high directivity required for current drive, the radiation resistance must be reduced. RF magnetic fields were measured by magnetic probes located on the low field side of the torus near the midplane. The measured phase shifts were consistent with the excited toroidal mode numbers of 7--8. A broadening of the frequency spectrum was observed when the plasma was present, which suggests scattering of the incident HHFW by low frequency density fluctuations.
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.
Disruption mitigation studies on the Mega Amp Spherical Tokamak (MAST)
NASA Astrophysics Data System (ADS)
Mast Team Thornton, A. J.; Gibson, K. J.; Harrison, J. R.; Kirk, A.; Lisgo, S. W.; Lehnen, M.; Martin, R.; Naylor, G.; Scannell, R.; Cullen, A.
2011-08-01
Disruptions pose a significant challenge in future devices where the increased stored energy can lead to unacceptably large transient heat loads on plasma facing components (PFCs). One means of mitigating disruptions is that of massive gas injection (MGI), which produces a radiative collapse of the plasma discharge through the injection of impurity gases. The MAST disruption mitigation system is capable of injecting up to 1.95 bar litres into the MAST vacuum vessel over a timescale of 1-2 ms, corresponding to a particle inventory of 5 × 1022, around 100 times the plasma particle inventory. High speed infrared thermography, offering full divertor coverage, has shown a 60-70% reduction in divertor power loads during mitigation. A combination of high temporal (0.2 ms) and spatial resolution (1 cm) Thomson scattering and soft X-ray camera array data show evidence for a cooling front associated with the inward propagation of the injected impurities.
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.
Edge Physics Studies on the NSTX Spherical Tokamak
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.
Energy confinement in tokamaks
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.
TPX tokamak construction management
Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.
1995-12-31
A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly.
Magnetic diagnostics for the lithium tokamak experiment.
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.
Rotation driven by fast ions in tokamaks
Thyagaraja, A.; Schwander, F.; McClements, K. G.
2007-11-15
Collective fast ion effects on flows in tokamaks are investigated analytically and numerically. A general analysis of noncollisional electrodynamic momentum transfer from fast ions to bulk plasma is presented, with polarization effects and dissipation in the bulk plasma taken into account. The analysis is illustrated using idealized simulations of fast ion orbits and radial electric fields in the Mega-Ampere Spherical Tokamak (MAST) [A. Sykes, R. J. Akers, L. C. Appel et al., Nucl. Fusion 41, 1423 (2001)], the Joint European Torus (JET) [P. H. Rebut et al., Nucl. Fusion 25, 1011 (1985)], and ITER [R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Controlled Fusion 44, 519 (2002)]. In the MAST simulation, prompt losses of beam ions injected counter to the plasma current drive up a radial electric field that saturates at a level such that beam ions subsequently injected are confined electrostatically. Although the actual radial electric fields in counterinjected MAST discharges are lower than this, the scenario explored in the simulation would be approached in MAST plasmas with sufficiently low collisionality. The JET simulation, although unrealistic, shows that a similar process could be driven by losses of fusion {alpha}-particles from a burning plasma. Test-particle simulations of {alpha}-particles in ITER suggest that performance-limiting instabilities such as neoclassical tearing modes and resistive wall modes could be affected significantly by flows associated with radial fast particle currents.
Rotation driven by fast ions in tokamaks
NASA Astrophysics Data System (ADS)
Thyagaraja, A.; Schwander, F.; McClements, K. G.
2007-11-01
Collective fast ion effects on flows in tokamaks are investigated analytically and numerically. A general analysis of noncollisional electrodynamic momentum transfer from fast ions to bulk plasma is presented, with polarization effects and dissipation in the bulk plasma taken into account. The analysis is illustrated using idealized simulations of fast ion orbits and radial electric fields in the Mega-Ampère Spherical Tokamak (MAST) [A. Sykes, R. J. Akers, L. C. Appel et al., Nucl. Fusion 41, 1423 (2001)], the Joint European Torus (JET) [P. H. Rebut et al., Nucl. Fusion 25, 1011 (1985)], and ITER [R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Controlled Fusion 44, 519 (2002)]. In the MAST simulation, prompt losses of beam ions injected counter to the plasma current drive up a radial electric field that saturates at a level such that beam ions subsequently injected are confined electrostatically. Although the actual radial electric fields in counterinjected MAST discharges are lower than this, the scenario explored in the simulation would be approached in MAST plasmas with sufficiently low collisionality. The JET simulation, although unrealistic, shows that a similar process could be driven by losses of fusion α-particles from a burning plasma. Test-particle simulations of α-particles in ITER suggest that performance-limiting instabilities such as neoclassical tearing modes and resistive wall modes could be affected significantly by flows associated with radial fast particle currents.
Tokamak Simulation Code modeling of NSTX
S.C. Jardin; S. Kaye; J. Menard; C. Kessel; A.H. Glasser
2000-07-20
The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption.
Magnetic confinement experiment -- 1: Tokamaks
Goldston, R.J.
1994-12-31
This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization.
Imaging with Spherically Bent Crystals or Reflectors
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.
Sager, G.T.
1988-06-01
Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs.
On the bootstrap current in stellarators and tokamaks
Helander, P.; Geiger, J.; Maassberg, H.
2011-09-15
The expression for the long-mean-free-path limit of the bootstrap current in stellarators is rederived in such a way that the expansion procedure is identical to that used in the corresponding calculation for a tokamak. In addition, the first correction due to finite collisionality is calculated and shown to vanish in quasi-isodynamic configurations without net current. This correction, which is proportional to the square root of the collisionality, is found to compare well with a numerical solution of the first-order drift kinetic equation in spherical tokamak geometry. Numerically, it appears that there is a similar correction in general stellarator geometry, which however depends on the strength of the radial electric field.
The Tokamak Physics Experiment
Davidson, R.C.; Goldston, R.J.; Neilson, G.H.; Thomassen, K.I.
1995-06-01
The mission of the Tokamak Physics Experiment (TPX) [Nevins {ital et} {ital al}., {ital Plasma} {ital Physics} {ital and} {ital Controlled} {ital Nuclear} {ital Fusion}, Wuerzburg (International Atomic Energy Agency, Vienna, 1992), Vol. 3, p. 279] is to develop the scientific basis for an economically competitive and continuously operating tokamak fusion power source. This complements the primary mission of the International Thermonuclear Experimental Reactor (ITER) [ITER Document Ser. No. 18 (International Atomic Energy Agency, Vienna, 1991)], the demonstration of ignition and long-pulse burn, and the integration of nuclear technologies. The TPX program is focused on making the demonstration power plant that follows ITER as compact and attractive as possible, and on permitting ITER to achieve its ultimate goal of steady-state operation. This mission of TPX requires the development of steady-state regimes with high beta, good confinement, and a high fraction of a self-driven bootstrap current. These regimes must be compatible with plasma stability, strong heat-flux dispersion in the divertor region, and effective particle control.
The Tokamak Physics Experiment
NASA Astrophysics Data System (ADS)
Davidson, Ronald C.; Goldston, Robert J.; Neilson, George H.; Thomassen, Keith I.
1995-06-01
The mission of the Tokamak Physics Experiment (TPX) [Nevins et al., Plasma Physics and Controlled Nuclear Fusion, Würzburg (International Atomic Energy Agency, Vienna, 1992), Vol. 3, p. 279] is to develop the scientific basis for an economically competitive and continuously operating tokamak fusion power source. This complements the primary mission of the International Thermonuclear Experimental Reactor (ITER) [ITER Document Ser. No. 18 (International Atomic Energy Agency, Vienna, 1991)], the demonstration of ignition and long-pulse burn, and the integration of nuclear technologies. The TPX program is focused on making the demonstration power plant that follows ITER as compact and attractive as possible, and on permitting ITER to achieve its ultimate goal of steady-state operation. This mission of TPX requires the development of steady-state regimes with high beta, good confinement, and a high fraction of a self-driven bootstrap current. These regimes must be compatible with plasma stability, strong heat-flux dispersion in the divertor region, and effective particle control.
NASA Astrophysics Data System (ADS)
Nedospasov, A. V.
1992-12-01
Edge turbulence is of decisive importance for the distribution of particle and energy fluxes to the walls of tokamaks. Despite the availability of extensive experimental data on the turbulence properties, its nature still remains a subject for discussion. This paper contains a review of the most recent theoretical and experimental studies in the field, including mainly the studies to which Wootton (A.J. Wooton, J. Nucl. Mater. 176 & 177 (1990) 77) referred to most in his review at PSI-9 and those published later. The available theoretical models of edge turbulence with volume dissipation due to collisions fail to fully interpret the entire combination of experimental facts. In the scrape-off layer of a tokamak the dissipation prevails due to the flow of current through potential shifts near the surface of limiters of divertor plates. The different origins of turbulence at the edge and in the core plasma due to such dissipation are discussed in this paper. Recent data on the electron temperature fluctuations enabled one to evaluate the electric probe measurements of turbulent flows of particles and heat critically. The latest data on the suppression of turbulence in the case of L-H transitions are given. In doing so, the possibility of exciting current instabilities in biasing experiments (rather than only to the suppression of existing turbulence) is given some attention. Possible objectives of further studies are also discussed.
The Aneutronic Rodless Ultra Low Aspect Ratio Tokamak
NASA Astrophysics Data System (ADS)
Ribeiro, Celso
2016-10-01
The replacement of the metal centre-post in spherical tokamaks (STs) by a plasma centre-post (PCP, the TF current carrier) is the ideal scenario for a ST reactor. A simple rodless ultra low aspect-ratio tokamak (RULART) using a screw-pinch PCP ECR-assisted with an external solenoid has been proposed in the most compact RULART [Ribeiro C, SOFE-15]. There the solenoid provided the stabilizing field for the PCP and the toroidal electrical field for the tokamak start-up, which will stabilize further the PCP, acting as stabilizing closed conducting surface. Relative low TF will be required. The compactness (high ratio of plasma-spherical vessel volume) may provide passive stabilization and easier access to L-H mode transition. It is presented here: 1) stability analysis of the PCP (initially MHD stable due to the hollow J profile); 2) tokamak equilibrium simulations, and 3) potential use for aneutronic reactions studies via pairs of proton p and boron 11B ion beams in He plasmas. The beams' line-of-sights sufficiently miss the sources of each other, thus allowing a near maximum relative velocities and reactivity. The reactions should occur close to the PCP mid-plane. Some born alphas should cross the PCP and be dragged by the ion flow (higher momentum exchange) towards the anode but escape directly to a direct electricity converter. Others will reach evenly the vessel directly or via thermal diffusion (favourable heating by the large excursion 2a), leading to the lowest power wall load possible. This might be a potential hybrid direct-steam cycle conversion reactor scheme, nearly aneutronic, and with no ash or particle retention problems, as opposed to the D-T thermal reaction proposals.
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.
Tokamak building-design considerations for a large tokamak device
Barrett, R.J.; Thomson, S.L.
1981-01-01
Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release.
Causes of major tokamak disruptions
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.
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.
Bootstrap current in a tokamak
Kessel, C.E.
1994-03-01
The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and {beta}{sub p} must be kept below a critical value.
NASA Astrophysics Data System (ADS)
Argenti, L.; Bonizzoni, G.; Cirant, S.; Corti, S.; Grosso, G.; Lampis, G.; Rossi, L.; Carretta, U.; Jacchia, A.; de Luca, F.
1981-06-01
The principle characteristics of plasma discharges produced in Thor tokamak experiments are discussed. The equilibrium and stability characteristics of the plasma produced are considered, with attention given to the density limits and critical streaming parameter for stable operation. The temporal evolution of the main plasma parameters, including electron density, electron temperature distribution, hard X-ray emission from suprathermal electrons, neutral gas influx, plasma density and Ohmic heating efficiency, is then examined, with particular emphasis on means used to control the electron runaway. The results achieved are noted to have demonstrated the possibility of controlling both plasma equilibrium and discharge regime, and further improvements expected by the use of more efficient preionization, gas puffing and feedback poloidal control of column position are indicated.
Cardozo, N.J.; Barth, C.J.; Chu, C.C.; Lok, J.; Montvai, A.; Oomens, A.A.; Peters, M.; Pijper, F.J.; de Rover, M.; Schueller, F.C.; Steenbakkers, M.F.; RTP team
1995-09-01
The relevance of a nest of toroidal flux surfaces as a paradigm of the magnetic topology of a tokamak plasma is challenged. High resolution Thomson scattering measurements of electron temperature and density in RTP show several hot filaments in the plasma center and sharp gradients near the sawtooth inversion radius and structures outside the sawtooth region under central ECH. In ohmic plasmas, too, the pressure and temperature profiles show significant bumps. These measurements give evidence of a complex magnetic topology. Transport in a medium with spatially strongly varying diffusivity is considered. It is shown that macroscopic transport is determined by the microscopic structure: a transport theory must predict this structure and the diffusivity in the insulating regions, while the {open_quote}turbulent{close_quote} diffusivity is irrelevant. A numerical approach to equilibria with broken surfaces is presented. {copyright} {ital 1995 American Institute of Physics.}
Tokamak Transmutation of (nuclear) Waste (TTW): Parametric studies
NASA Astrophysics Data System (ADS)
Cheng, E. T.; Krakowski, R. A.; Peng, Y. K. M.
Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses.
Tokamak transmutation of (nuclear) waste (TTW): Parametric studies
Cheng, E.T.; Krakowski, R.A.; Peng, Y.K.M.
1994-06-01
Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low-aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses.
The ARIES tokamak reactor study
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.
Wave Driven Fast Ion Loss in the National Spherical Torus Experiment
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.
Bibliography of fusion product physics in tokamaks
Hively, L. M.; Sigmar, D. J.
1989-09-01
Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category.
Connections between physics and economics for Tokamak fusion power plants
NASA Astrophysics Data System (ADS)
Krakowski, R. A.; Delene, J. G.
1988-03-01
tokamaks based on (a) operation in the second-stability region ( β=0.2, increased aspect ratio, reduced elongation), (b) super high-field but low-beta operation, (c) very low aspect ratio and highly elongated spherical torus, and (d) a direct application of the present database using a long-pulsed, low-beta tokamak. The economic impact of a range of base-case parameters and operating variables is examined, including current-drive efficiency, beta, stability limits, advanced magnets, economy of scale, blanket/shield lifetime, blanket thickness, and plant lead time. It is found that a range of tokamak options, relative to the optimistic base case selected for this study, may provide economically competitive power plants. Areas where physics and technology advances are needed to achieve this attractive end product are quantitively elucidated for all tokamak options considered.
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.
Transition to subcritical turbulence in a tokamak plasma
NASA Astrophysics Data System (ADS)
van Wyk, F.; Highcock, E. G.; Schekochihin, A. A.; Roach, C. M.; Field, A. R.; Dorland, W.
2016-12-01
Tokamak turbulence, driven by the ion-temperature gradient and occurring in the presence of flow shear, is investigated by means of local, ion-scale, electrostatic gyrokinetic simulations (with both kinetic ions and electrons) of the conditions in the outer core of the Mega-Ampere Spherical Tokamak (MAST). A parameter scan in the local values of the ion-temperature gradient and flow shear is performed. It is demonstrated that the experimentally observed state is near the stability threshold and that this stability threshold is nonlinear: sheared turbulence is subcritical, i.e. the system is formally stable to small perturbations, but, given a large enough initial perturbation, it transitions to a turbulent state. A scenario for such a transition is proposed and supported by numerical results: close to threshold, the nonlinear saturated state and the associated anomalous heat transport are dominated by long-lived coherent structures, which drift across the domain, have finite amplitudes, but are not volume filling; as the system is taken away from the threshold into the more unstable regime, the number of these structures increases until they overlap and a more conventional chaotic state emerges. Whereas this appears to represent a new scenario for transition to turbulence in tokamak plasmas, it is reminiscent of the behaviour of other subcritically turbulent systems, e.g. pipe flows and Keplerian magnetorotational accretion flows.
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.
Moving Divertor Plates in a Tokamak
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.
Tokamak Physics Experiment (TPX) design
Schmidt, J.A.
1995-12-31
TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995.
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.
Resistive instabilities in tokamaks
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.
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.
Mechanisms of Stochastic Diffusion of Energetic Ions in Spherical Tori
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.
Summary discussion: An integrated advanced tokamak reactor
Sauthoff, N.R.
1994-12-31
The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ``figures of merit`` for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept.
The spherical birdcage resonator
NASA Astrophysics Data System (ADS)
Harpen, Michael D.
A description of the operation of a spherical resonator capable of producing a uniform magnetic induction throughout a spherical volume is presented. Simple closed-form expressions for the spectrum of resonant frequencies are derived for both the low-pass and the high-pass configuration of the resonator and are shown to compare favorably with observation in an experimental coil system. It is shown that the spherical resonator produces a uniform spherical field of view when used as a magnetic resonance imaging radiofrequency coil.
Impedance of an intense plasma-cathode electron source for tokamak startup
Hinson, Edward Thomas; Barr, Jayson L.; Bongard, Michael W.; ...
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
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.
Understanding disruptions in tokamaks
NASA Astrophysics Data System (ADS)
Zakharov, Leonid
2011-10-01
Disruptions in tokamaks are known since 1963 but even now some aspects of them remain a mystery. This talk describes progress made recently in understanding disruptions. A major step forward occurred in 2007 when the importance of galvanic contact of the plasma with the wall in plasma dynamics was pointed out. The toroidal asymmetry of plasma current, observed in JET vertical disruptions, was explained by the theory of the wall touching kink mode. The currents shared by the plasma with the wall and responsible for the asymmetry were identified as generated by the kink mode. Such currents are referred to as Hiro currents. They have shown exceptional consistency with the entire JET disruption data base (more than 5500 cases) and ruled out the long lasting interpretation based on ``halo currents,'' which contradict experiments even in the sign of the measured asymmetry. Accordingly, the sideways forces are understood and their scaling from JET to ITER was justified. Hiro currents provide also a plausible explanation of the current spike at the beginning of the disruptions. The important role of the plasma edge and its interaction with the wall was revealed. Based on this new understanding of disruptions, dedicated experiments on the current spike (J-TEXT, Wuhan, China) and runaway prevention by the repetitive triggering of kink modes (T-10, AUG, Tore Supra) were motivated and are in progress. Accordingly, the need for new, adaptive grid approaches to numerical simulations of disruptions became evident. In addition to the core MHD, simulations of realistic wall geometry, disruption specific plasma edge physics, plasma-wall interaction, and energetic particles need be developed. The first results of simulations of the fast MHD regime, Hiro current generation, and slower plasma decay due to a wall touching kink mode made with the new DSC code are presented. This work is supported by US DoE contract No. DE-AC02-09-CH11466.
Physics Basis for a Spherical Torus Power Plant
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.
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.
An enhanced tokamak startup model
NASA Astrophysics Data System (ADS)
Goswami, Rajiv; Artaud, Jean-François
2017-01-01
The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.
Prospects for Tokamak Fusion Reactors
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.
Tokamak and RFP ignition requirements
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.
High Energy Particles in Tokamaks
White, R. B.
2008-05-14
This lecture covers the derivation of guiding center equations in a tokamak, orbit classification, the effect of magnetic perturbations and ripple, the interaction of particles with magnetohydrodynamic modes, including passing particle resonance, toroidal Alfven mode drive and saturation, the fishbone mode, and sawtooth stabilization.
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.
High power heating of magnetic reconnection in merging tokamak experiments
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.
Bootstrapped tokamak with oscillating field current drive
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.
Large displacement spherical joint
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.
The Precessing Spherical Pendulum.
ERIC Educational Resources Information Center
Olsson, M. G.
1978-01-01
Explains how the spherical pendulum could be used to observe nonreentrant orbits, and shows, using theoretical analysis, that for small displacements the elliptical orbit will precess at a rate proportional to its area. (GA)
Wave Driven Fast Ion Loss in the National Spherical Torus Experiment
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.
Cryogenic needs for future tokamaks
NASA Astrophysics Data System (ADS)
Katheder, H.
The ITER tokamak is a machine using superconducting magnets. The windings of these magnets will be subjected to high heat loads resulting from a combination of nuclear energy absorption and AC-losses. It is estimated that about 100 kW at 4.5 K are needed. The total cooling mass flow rate will be around 10 - 15 kg/s. In addition to the large cryogenic power required for the superconducting magnets cryogenic power is also needed for refrigerated radiation shield, various cryopumps, fuel processing and test beds. A general description of the overall layout and the envisaged refrigerator cycle, necessary cold pumps and ancillary equipment is given. The basic cryogenic layout for the ITER tokakmak design, as developed during the conceptual design phase and a short overview about existing tokamak designs using superconducting magnets is given.
Options for an ignited tokamak
Sheffield, J.
1984-02-01
It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon ..beta../sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed.
Folded waveguide designs for tokamaks
NASA Astrophysics Data System (ADS)
Hoffman, D. J.; Bigelow, T. S.; Fogelman, C. H.; Yugo, J. J.; Caughman, J. B. O.; Gardner, W. L.; Carter, M. D.; Probert, P. H.; Barbato, E.
The folded waveguide (FWG) has been tested to the megawatt level in RFTF and shows great promise for tokamak use. It has three primary advantages: low electric field (anywhere) per unit power coupled to the plasma, strong structural capabilities, and better spectral content than loops. A tokamak test is now needed. Potential candidates include C-Mod at 80 MHz and FTU at 433 MHz. The waveguide test on the first machine will be directed at conventional ion cyclotron heating, while the test on the latter will be directed at direct electron heating. In addition, a variation of the folded waveguide is proposed to be tested on Phaedrus-T. In this paper, we discuss the advantages of the waveguide, the design layout, some of the potential physics programs, and how these programs may have an impact on its potential use in ITER.
Spherical torus fusion reactor
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.
Neoclassical magnetic microislands in tokamaks
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.
Transport Equations In Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Callen, J. D.
2009-11-01
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 (trapped particle effects on resistivity, bootstrap current); fluctuation-induced transport; heating, current-drive and flow sources and sinks; small B field non-axisymmetries; 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 recently using a kinetic-based framework. 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 Alfv'en waves (Grad-Shafranov equilibrium, ion radial force balance); sound waves (pressure constant along field lines, incompressible flows within a flux surface); and ion collisions (damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on the plasma fluid: 7 ambipolar collision-based ones (classical, neoclassical, etc.) and 8 non-ambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients etc.). The plasma toroidal rotation equation [1] results from setting to zero the net radial current induced by the non-ambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the non-ambipolar 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 resultant transport equations will be presented and contrasted with the usual ones. [4pt] [1] J.D. Callen, A.J. Cole, C.C. Hegna, ``Toroidal Rotation In
Magnetic island formation in tokamaks
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.
Equilibrium Reconstruction in EAST Tokamak
NASA Astrophysics Data System (ADS)
Qian, Jinping; Wan, Baonian; L. Lao, L.; Shen, Biao; A. Sabbagh, S.; Sun, Youwen; Liu, Dongmei; Xiao, Bingjia; Ren, Qilong; Gong, Xianzu; Li, Jiangang
2009-04-01
Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.
Magnetic confinement experiment. I: Tokamaks
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.
NASA Technical Reports Server (NTRS)
Peeples, Steven
2015-01-01
A three degree of freedom (DOF) spherical actuator is proposed that will replace functions requiring three single DOF actuators in robotic manipulators providing space and weight savings while reducing the overall failure rate. Exploration satellites, Space Station payload manipulators, and rovers requiring pan, tilt, and rotate movements need an actuator for each function. Not only does each actuator introduce additional failure modes and require bulky mechanical gimbals, each contains many moving parts, decreasing mean time to failure. A conventional robotic manipulator is shown in figure 1. Spherical motors perform all three actuation functions, i.e., three DOF, with only one moving part. Given a standard three actuator system whose actuators have a given failure rate compared to a spherical motor with an equal failure rate, the three actuator system is three times as likely to fail over the latter. The Jet Propulsion Laboratory reliability studies of NASA robotic spacecraft have shown that mechanical hardware/mechanism failures are more frequent and more likely to significantly affect mission success than are electronic failures. Unfortunately, previously designed spherical motors have been unable to provide the performance needed by space missions. This inadequacy is also why they are unavailable commercially. An improved patentable spherically actuated motor (SAM) is proposed to provide the performance and versatility required by NASA missions.
Overview of the Pegasus Extremely Low-Aspect Ratio Tokamak
NASA Astrophysics Data System (ADS)
Fonck, R.; Garstka, G.; Intrator, T.; Lewicki, B.; Thorson, T.; Toonen, R.; Tritz, K. L.; White, B.; Winz, G.
1996-11-01
Pegasus is a new experiment designed to explore the potential of Extremely Low Aspect Ratio Tokamaks (ELART) at very high toroidal β. Ohmic induction for plasma startup will be followed by ohmic sustainment initially and noninductive RF current drive in the future. Plasma parameters are projected to be Ip <= 0.3 MA, Bt < 0.2T, <β> ≈ 5-40 % or higher, A=1.1-2, R=0.2-0.4 m, and P_RF <= 2MW. Goals of the program include: demonstrate high-β spherical tokamak operation in the near term; examine the stability, n=0 stability properties at high elongation and low- A, confinement and scaling characteristics at A <= 1.25; and extend high power ST operation to the extrema of A <= 1.1. Hollow current profiles should be accessible in Pegasus using a fast current ramp during formation plus off-axis FWCD in the longer term. Recent changes to the design include: increased vacuum vessel height to allow for divertor operation with an internal X-point plus increased accessible elongations (i.,e., κ <= 3.7 at A = 1.25); additional coils for X-point control; and elimination of toroidal gaps in favor of a resistive vacuum vessel. Initial operation will emphasize ohmic access to high- β, followed by high power RF heating.
ADVANCES IN DUST DETECTION AND REMOVAL FOR TOKAMAKS
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.
Numerical study of tokamak equilibria with arbitrary flow
NASA Astrophysics Data System (ADS)
Guazzotto, L.; Betti, R.; Manickam, J.; Kaye, S.
2004-02-01
The effects of toroidal and poloidal flows on the equilibrium of tokamak plasmas are numerically investigated using the code FLOW. The code is used to determine the changes in the profiles induced by large toroidal flows on NSTX-like equilibria [with NSTX being the National Spherical Torus Experiment, M. Ono, S.M. Kaye, Y.-K.M. Peng et al., Nucl. Fusion 40, 557 (2000)] where flows exceeding the sound speed lead to a considerable outward shift of the plasma. The code is also used to study the effects of poloidal flow when the flow velocity profile varies from subsonic to supersonic with respect to the poloidal sound speed. It is found that pressure and density profiles develop a pedestal structure characterized by radial discontinuities at the transonic surface where the poloidal velocity abruptly jumps from subsonic to supersonic values. These results confirm the conclusions of the analytic theory of R. Betti and J. P. Freidberg [Phys. Plasmas 7, 2439 (2000)], derived for a low-β, large aspect ratio tokamak with a circular cross section.
Magnetic Diagnostics for the Lithium Tokamak eXperiment
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.
Composition And Electrical Properties Of Dust From Tokamak Compass
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.
Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas.
Green, D L; Berry, L A; Chen, G; Ryan, P M; Canik, J M; Jaeger, E F
2011-09-30
Observations of improved radio frequency (rf) heating efficiency in ITER relevant high-confinement (H-)mode plasmas on the National Spherical Tokamak Experiment are investigated by whole-device linear simulation. The steady-state rf electric field is calculated for various antenna spectra and the results examined for characteristics that correlate with observations of improved or reduced rf heating efficiency. We find that launching toroidal wave numbers that give fast-wave propagation in the scrape-off plasma excites large amplitude (∼kV m(-1)) coaxial standing modes between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggests that these modes are a probable cause of degraded heating efficiency.
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.
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.
Ohmic Flux Consumption During Initial Operation of the NSTX Spherical Torus
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.
Leakage of runaway electrons from tokamaks
Wong, K.L.
1982-02-01
Runaway electron orbits are calculated in a tokamak magnetic field. It is shown that these electrons tend to drift towards a larger major radius with a velocity v Vector/sub R/ = qcE/B/sub 0/ R. This effect may be relevant to some recent experimental observations in tokamaks.
Numerical tokamak turbulence project (OFES grand challenge)
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.
High-performance discharges in the Small Tight Aspect Ratio Tokamak (START)
NASA Astrophysics Data System (ADS)
Gates, D. A.; Akers, R.; Appel, L.; Carolan, P. G.; Conway, N.; Dowling, J.; Gryaznevich, M.; Hender, T.; Kwon, O. J.; Martin, R.; Nightingale, M.; Price, M.; Roach, C.; Sykes, A.; Tournianski, M. R.; Walsh, M.; Warrick, C. D.
1998-05-01
The Small Tight Aspect Ratio Tokamak (START) [A. Sykes et al., Nucl. Fusion 32, 769 (1994)] spherical tokamak has recently achieved the record value of toroidal β˜30% in a tokamak-like configuration. The improvements that have made these results possible are presented along with a description of the global equilibrium parameters of the discharges. The ideal magnetohydrodynamic (MHD) stability of these discharges is analyzed, and they are found to be in close proximity to both the ballooning limit and the external current driven kink limit, but they are found to be far from the pressure driven external kink limit. Disruptivity for a range of shots is not correlated with the normalized β limit, but does correlate well with the empirical high-li disruption limit. The transport properties of these high-β equilibria are analyzed and compared to conventional tokamak scaling laws and transport models. The global transport is at least as good as that predicted by the ITER97-ELMy (edge-localized) scaling law. The local ion transport is in good agreement with that predicted by neoclassical models. The electron transport is anomalous, showing rough agreement with the Lackner-Gottardi transport model.
Spherical geodesic mesh generation
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.
Natural current profiles in a tokamak
Taylor, J.B.
1990-08-01
In this paper I show how one may arrive at a universal, or natural, family of Tokamak profiles using only accepted physical principles. These particular profiles are similar to ones proposed previously on the basis of ad hoc variational principles and the point of the present paper is to provide a justification for them. However in addition, the present work provides an interesting view of Tokamak fluctuations and leads to a new result -- a relationship between the inward particle pinch velocity, the diffusion coefficient and the current profile. The basic Tokamak model is described in this paper. Then an analogy is developed between Tokamak profiles and the equilibrium of a realisable dynamical system. Then the equations governing the natural Tokamak profiles are derived by applying standard statistical mechanics to this analog. The profiles themselves are calculated and some other results of the theory are described.
Linear optimal control of tokamak fusion devices
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.
A Fusion Breeder Reactor Based on a Catalyzed D-D Spherical Torus.
1986-08-08
cooling and fissile breeding. The need for tritium breeding is eliminated by the use of’ a catalyzed 0-0 fuel cycle. Analysis of this novel reactor...in heavy water which flows through the first wall and blanket providing both cooling and fissile breeding. The need I for tritium breeding is...studies in that: (1) a deuterium fuel cycle is used to eliminate the : need to breed tritium ; (2) a compact tokamak (spherical torus) is used as a
Transport equations in tokamak plasmas
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
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.
The Microwave Tokamak Experiment (MTX)
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.
Alpha particle confinement in tokamaks
White, R.B.; Mynick, H.E.
1988-11-01
An assessment of diffusive tokamak transport mechanisms of concern for alpha particles indicates that the ''stochastic regime'' is the only one which appears to pose a real danger for adequate alpha confinement. This fact, in conjunction with the threshold character of that mechanism, allows one to decide whether an alpha born at a given location will be lost or confined, according to a very simple criterion. Implementing this criterion numerically results in a new code for the assessment of alpha confinement, which is orders of magnitude faster than earlier codes used for this purpose. 13 refs., 3 figs., 1 tab.
Breakdown in the pretext tokamak
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.
Noniterative reconstruction of tokamak equilibria
NASA Astrophysics Data System (ADS)
Rodrigues, Paulo; Bizarro, João P. S.
2009-02-01
Unlike iterative approaches, noniterative equilibria reconstruction schemes are designed to keep two measured internal profiles fixed along a given chord while solving a sequence of linear differential equations, providing a unique and asymptotic solution to the Grad-Shafranov (GS) equation directly in laboratory coordinates. A noniterative algorithm is extended to handle plasma configurations that are not symmetric with respect to the tokamak midplane and then used to compute an equilibrium solution from an actual experimental data set. A number of issues concerning how available experimental data can be handled and provided as input to the GS solver in practical situations are also discussed.
Spherical colloidal photonic crystals.
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
Integrated modeling of H-mode tokamak discharges with ASTRA and B2SOLPS numerical codes
NASA Astrophysics Data System (ADS)
Senichenkov, I. Yu; Kaveeva, E. G.; Rozhansky, V. A.; Voskoboynikov, S. P.; Molchanov, P. A.; Coster, D. P.; Pereverzev, G. V.; the ASDEX Upgrade Team; the Globus-M Team
2014-05-01
The numerical codes ASTRA and B2SOLPS5.2 are coupled to perform an integrated modeling of particle and energy transport and to obtain continuous self-consistent profiles of the main plasma parameters from the magnetic axis up to target plates. The unique distinguishing feature of the new coupling scheme is the presence of a region of overlap of the 1D and 2D computational domains, where the 1D solution coincides with the 2D one at the equatorial midplane. In the 2D transport equation system, all relevant drift flows and currents are taken into account, which allows us to calculate the poloidal variation of the density, temperatures and electrostatic potential, and obtain neoclassical radial fluxes in a self-consistent manner. Such an approach allows us to model tokamaks for which neoclassical effects give a significant contribution to the ion heat transport, and in particular, spherical tokamaks.
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.
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.
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.
Progress Towards High Performance, Steady-state Spherical Torus
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
Tokamak Physics Experiment divertor design
Anderson, P.M.
1995-12-31
The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m{sup 2}. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services.
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.
Theoretical Transport Model for Tokamaks
NASA Astrophysics Data System (ADS)
Ghanem, Elsayed Mohammad
In the present thesis work a theoretical transport model is suggested to study the anomalous transport of plasma particles and energy across the axisymmetric equilibrium toroidal magnetic flux surfaces in tokamaks. The model suggests a linear combination of two transport mechanisms; drift waves, which dominate the transport in the core region, and resistive ballooning modes, which dominate the transport in the edge region. The resulting unified model has been used in a predictive transport code to simulate the plasma transport in different tokamak experiments operating in both the ohmic heating phase and the low confinement mode (L-mode). For ohmic plasma, the model was used to study the saturation of energy confinement time at high plasma density. The effect of the resistive ballooning mode as a possible cause of the saturation phenomena has been investigated together with the effect of the ion temperature gradient mode. For the low confinement mode plasmas, the study has emphasized on using the model to obtain a scaling law for the energy confinement time with the various plasma parameters compared to the scaling laws that are derived based on fitting the experimental data.
Spherical coordinate descriptions of cylindrical and spherical Bessel beams.
Poletti, M A
2017-03-01
This paper derives a generalized spherical harmonic description of Bessel beams. The spherical harmonic description of the well-known cylindrical Bessel beams is reviewed and a family of spherical Bessel beams are introduced which can provide a number of azimuthal phase variations for a single beam radial amplitude. The results are verified by numerical simulations.
Superconducting magnet system for the TPX Tokamak
NASA Astrophysics Data System (ADS)
Hassenzahl, W. V.; Chaplin, M. R.; Heim, J. R.; Lang, D. D.; O'Connor, T. G.; Slack, D. S.; Wong, R. L.; Zbasnik, J. P.; Brown, T. G.; Citrolo, J. C.
1994-07-01
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.
Superconducting magnet system for the TPX Tokamak
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.
Superconducting magnet system for the TPX Tokamak
Hassenzahl, W.V.; Chaplin, M.R.; Heim, J.R.
1994-07-01
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 TPC 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.
Impedance of an intense plasma-cathode electron source for tokamak startup
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/cm^{2}, arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma (n_{arc} ≈ 10^{21} m^{-3}) within the electron source, and the less dense external tokamak edge plasma (n_{edge} ≈ 10^{18} m^{-3}) into which current is injected at the applied injector voltage, V_{inj}. Experiments on the Pegasus spherical tokamak show the injected current, I_{inj}, increases with V_{inj} according to the standard double layer scaling I_{inj} ~ V_{inj}^{3/2} at low current and transitions to I_{inj} ~ V_{inj}^{1/2} at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density n_{b} ~ I_{inj}/V_{inj}^{1/2}. For low tokamak edge density n_{edge} and high I_{inj}, the inferred beam density n_{b} is consistent with the requirement n_{b} ≤ n_{edge} imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, n_{b} ~ n_{arc} is observed, consistent with a limit to n_{b} imposed by expansion of the double layer sheath. These results suggest that n_{arc} is a viable control actuator for the source impedance.
Impedance of an intense plasma-cathode electron source for tokamak startup
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/cm^{2}, arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma (n_{arc} ≈ 10^{21} m^{-3}) within the electron source, and the less dense external tokamak edge plasma (n_{edge} ≈ 10^{18} m^{-3}) into which current is injected at the applied injector voltage, V_{inj}. Experiments on the Pegasus spherical tokamak show the injected current, I_{inj}, increases with V_{inj} according to the standard double layer scaling I_{inj} ~ V_{inj}^{3/2} at low current and transitions to I_{inj} ~ V_{inj}^{1/2} at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density n_{b} ~ I_{inj}/V_{inj}^{1/2}. For low tokamak edge density n_{edge} and high I_{inj}, the inferred beam density n_{b} is consistent with the requirement n_{b} ≤ n_{edge} imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, n_{b} ~ n_{arc} is observed, consistent with a limit to n_{b} imposed by expansion of the double layer sheath. These results suggest that n_{arc} is a viable control actuator for the source impedance.
Spherical torus fusion reactor
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.
Spherical torus fusion reactor
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.
NASA Technical Reports Server (NTRS)
Lee, M. C.; Kendall, J. M., Jr.; Bahrami, P. A.; Wang, T. G.
1986-01-01
Fluid-dynamic and capillary forces can be used to form nearly perfect, very small spherical shells when a liquid that can solidify is passed through an annular die to form an annular jet. Gravity and certain properties of even the most ideal materials, however, can cause slight asymmetries. The primary objective of the present work is the control of this shell formation process in earth laboratories rather than space microgravity, through the development of facilities and methods that minimize the deleterious effects of gravity, aerodynamic drag, and uncontrolled cooling. The spherical shells thus produced can be used in insulation, recyclable filter materials, fire retardants, explosives, heat transport slurries, shock-absorbing armor, and solid rocket motors.
Noncommuting spherical coordinates
Bander, Myron
2004-10-15
Restricting the states of a charged particle to the lowest Landau level introduces a noncommutativity between Cartesian coordinate operators. This idea is extended to the motion of a charged particle on a sphere in the presence of a magnetic monopole. Restricting the dynamics to the lowest energy level results in noncommutativity for angular variables and to a definition of a noncommuting spherical product. The values of the commutators of various angular variables are not arbitrary but are restricted by the discrete magnitude of the magnetic monopole charge. An algebra, isomorphic to angular momentum, appears. This algebra is used to define a spherical star product. Solutions are obtained for dynamics in the presence of additional angular dependent potentials.
Hollow spherical shell manufacture
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.
NASA Technical Reports Server (NTRS)
Lee, M. C.; Kendall, J. M., Jr.; Bahrami, P. A.; Wang, T. G.
1986-01-01
Fluid-dynamic and capillary forces can be used to form nearly perfect, very small spherical shells when a liquid that can solidify is passed through an annular die to form an annular jet. Gravity and certain properties of even the most ideal materials, however, can cause slight asymmetries. The primary objective of the present work is the control of this shell formation process in earth laboratories rather than space microgravity, through the development of facilities and methods that minimize the deleterious effects of gravity, aerodynamic drag, and uncontrolled cooling. The spherical shells thus produced can be used in insulation, recyclable filter materials, fire retardants, explosives, heat transport slurries, shock-absorbing armor, and solid rocket motors.
Hollow spherical shell manufacture
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.
Spherical nitroguanidine process
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.
Stability of thick spherical shells
NASA Astrophysics Data System (ADS)
Liu, I.-Shih
1995-06-01
The pressure-radius relation of spherical rubber balloons has been derived and its stability behavior investigated before. In this work, we show that similar results remain valid for thick spherical shells of Mooney-Rivlin materials. In addition, we show that eversion of a spherical shell is possible for any incompressible isotropic materials if the shell is not too thick.
OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS
LIN-LIU,YR; STAMBAUGH,RD
2002-11-01
OAK A271 OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS. The dependence of the ideal ballooning {beta} limit on aspect ratio, A, and elongation {kappa} is systematically explored for nearly 100% bootstrap current driven tokamak equilibria in a wide range of the shape parameters (A = 1.2-7.0, {kappa} = 1.5-6.0 with triangularity {delta} = 0.5). The critical {beta}{sub N} is shown to be optimal at {kappa} = 3.0-4.0 for all A studied and increases as A decreases with a dependence close to A{sup -0.5}. The results obtained can be used as a theoretical basis for the choice of optimum aspect ratio and elongation of next step burning plasma tokamaks or tokamak reactors.
Control of Dust Inventory in Tokamaks
Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Pitcher, C. S.; Taylor, N.; Furlan, J.
2008-09-07
Particles with sizes ranging from 100 nm to 100 {mu}m are produced in tokamaks by the interaction of the plasma with the first wall materials and divertor. Dust has not yet been of a major concern in existing tokamaks mainly because their quantities are small and these devices are not nuclear facilities. However, in ITER and in future reactors, they could represent operational and potential safety issues. The aim of this paper is thus to describe the dust creation processes in the tokamak environment. The diagnostics and removal techniques that are needed to be implemented to measure and minimise the dust inventory are also presented. The integration of these techniques into a tokamak environment is also discussed.
NASA Astrophysics Data System (ADS)
Timokhin, V. M.; Rykachevskii, A. I.; Miroshnikov, I. V.; Sergeev, V. Yu.; Kochergin, M. M.; Koval', A. N.; Mukhin, E. E.; Tolstyakov, S. Yu.; Voronin, A. V.
2016-08-01
A diagnostic technique that is based on measuring the ratios of neutral-helium line strengths has been developed for peripheral distributions of electron temperature and density of tokamak plasma. The main components of the technique are a four-channel filter-lens imaging polychromator (FLIP-4) and a Phantom Miro M110 high-speed camera for recording the images. The polychromator has been assembled, adjusted, and tested on an optical test bench. The optical system was installed on the spherical Globus-M tokamak. Some preliminary experiments were carried out. Images of the plasma-gun jet were obtained at neutralhelium lines.
Active tokamak limiters; symmetrizing the edge plasma
Motley, R.W.
1981-02-01
The surface layers of tokamak plasmas are strongly unstable to low frequency oscillations. The limiter, by imposing a non-axisymmetric state in the plasma scrape-off layer, may be the source of the free energy driving the instabilities. An active, two-faced sandwich limiter is proposed to symmetrize the scrape-off layer, thereby minimizing the mismatch between the inner and outer tokamak states.
Evans, K.E. Jr.; Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Finn, P.A.; Jung, J.; Mattas, R.F.; Misra, B.; Smith, D.L.; Stevens, H.C.
1980-11-01
A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated.
Burn Control Mechanisms in Tokamaks
NASA Astrophysics Data System (ADS)
Hill, M. A.; Stacey, W. M.
2015-11-01
Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.
Final Report: Spectral Analysis of L-shell Data in the Extreme Ultraviolet from Tokamak Plasmas
Lepson, J.; Jernigan, J. Garrett; Beiersdorfer, P.
2016-02-05
We performed detailed analyses of extreme ultraviolet spectra taken by Lawrence Livermore National Laboratory on the National Spherical Torus Experiment at Princeton Plasma Physics Laboratory and on the Alcator CKmod tokamak at the M.I.T. Plasma Science and Fusion Center. We focused on the emission of iron, carbon, and other elements in several spectral band pass regions covered by the Atmospheric Imaging Assembly on the Solar Dynamics Observatory. We documented emission lines of carbon not found in currently used solar databases and demonstrated that this emission was due to charge exchange.
Progress Towards High-Performance, Steady-State Spherical Torus
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
Science and Technology of the 10-MA Spherical Tori
Peng, Y-K.M.
1999-11-14
The Spherical Torus (ST) configuration has recently emerged as an example of confinement concept innovation that enables attractive steps in the development of fusion energy. The scientific potential for the ST has been indicated by recent encouraging results from START,2 CDX-U, and HIT. The scientific principles for the D-fueled ST will soon be tested by NSTX (National Spherical Torus Experiment3) in the U.S. and MAST (Mega-Amp Spherical Tokamak4) in the U.K. at the level of l-2 MA in plasma current. More recently, interest has grown in the U.S. in the possibility of near-term ST fusion burn devices at the level of 10 MA in plasma current. The missions for these devices would be to test burning plasma performance in a small, pulsed D-T-fueled ST (i.e., DTST) and to develop fusion energy technologies in a small steady state ST-based Volume Neutron Source (VNS). This paper reports the results of analysis of the key science and technology issues for these devices.
Holographic Spherically Symmetric Metrics
NASA Astrophysics Data System (ADS)
Petri, Michael
The holographic principle (HP) conjectures, that the maximum number of degrees of freedom of any realistic physical system is proportional to the system's boundary area. The HP has its roots in the study of black holes. It has recently been applied to cosmological solutions. In this article we apply the HP to spherically symmetric static space-times. We find that any regular spherically symmetric object saturating the HP is subject to tight constraints on the (interior) metric, energy-density, temperature and entropy-density. Whenever gravity can be described by a metric theory, gravity is macroscopically scale invariant and the laws of thermodynamics hold locally and globally, the (interior) metric of a regular holographic object is uniquely determined up to a constant factor and the interior matter-state must follow well defined scaling relations. When the metric theory of gravity is general relativity, the interior matter has an overall string equation of state (EOS) and a unique total energy-density. Thus the holographic metric derived in this article can serve as simple interior 4D realization of Mathur's string fuzzball proposal. Some properties of the holographic metric and its possible experimental verification are discussed. The geodesics of the holographic metric describe an isotropically expanding (or contracting) universe with a nearly homogeneous matter-distribution within the local Hubble volume. Due to the overall string EOS the active gravitational mass-density is zero, resulting in a coasting expansion with Ht = 1, which is compatible with the recent GRB-data.
A Fusion Proton Diagnostic for Low Field Tokamaks
NASA Astrophysics Data System (ADS)
Lampson, Alexander
The use of a pinhole-type detector to image the high-energy protons produced by interactions of fast particles within the plasma of low field tokamaks such as the Mega Amp Spherical Tokamak is explored. The minimum number of detector elements in a pinhole-type detector needed to produce accurate images of the plasma was found to be 144, giving a 5cm by 5cm resolution of the plasma in the R, Z plane. The image is recreated via a combination of singular value decomposition and maximum entropy methods. The proton production rate is a function of the spatial distribution of fast particles, and so images of the proton production distribution can be used to directly diagnose the fast particle dis-tribution within the plasma. These fast particles are mostly produced by neutral beam injection, and so the methods presented here can be used to more accurately control the neutral beam injection energy to ensure that the heating and current drive is applied to the desired regions of the plasma. As the orbits of the fusion born protons are influenced by the magnetic field within the plasma, a transformation matrix is created to link the detector image with the intial proton distribution within the plasma. To do this, large numbers of test particles are needed. The calculation of the orbits of the test particles is performed using CUDA GPUs which can calculate the orbits of up to 100,000 particles per second. This takes advantage of the embarrassingly parallel nature of the problem, which is ideally suited to calculation on the GPU.
Integrated modelling of the Globus-M tokamak plasma and a comparison with SOL width scaling
NASA Astrophysics Data System (ADS)
Senichenkov, I. Yu.; Kaveeva, E. G.; Gogoleva, A. V.; Vekshina, E. O.; Zadvitskiy, G. V.; Molchanov, P. A.; Rozhansky, V. A.; Voskoboynikov, S. P.; Khromov, N. A.; Lepikhov, S. A.; Gusev, V. K.; The Globus-M Team
2015-05-01
Recently a scheme for the coupling of the one-dimensional core transport code ASTRA and the two-dimensional edge transport code B2SOLPS was developed, thus providing the integrated modelling of tokamak discharge. Here, this scheme is improved by taking impurities into account and by considering a real flux surface shape using the equilibrium code SPIDER. This integrated modelling is applied to discharges of the spherical tokamak Globus-M to study the dependence of the scrape-off layer (SOL) width and divertor heat loads on the discharge power and the plasma current. Since these values, together with the magnetic field, are relatively small in Globus-M, this study can test the existing scaling against data in a wider range of tokamak operational parameters. The modelling results agree reasonably with Thomson scattering and Langmuir probe measurements and allow, in principle, the determination of the physical mechanisms responsible for the SOL structure formation. It is found that the SOL width is approximately inversely proportional to the plasma current, in agreement with existing experimental scaling, while its dependence on discharge power is found to be quite weak.
What we should do for transition from current tokamaks to fusion-fission reactor
NASA Astrophysics Data System (ADS)
Mirnov, S.
2012-06-01
The Russian fission community places several heavy demands to quality of fusion neutron source for the first step of investigation of minority transmutations ("burning") and breading of nuclear fuel. They are: the steady state regime of neutron production (not rare 80% of main operation time), the total power on neutron flux should be not lower than 20MW with surface neutron load not lower than 0.2MW/m2. Between the current fusion devices: mirror traps, reverse field pinches, stellarators, spherical torus and tokamaks only lasts have today the some probability to fulfill in the near future these hard demands. Two well known DT-tokamaks with neutron power production higher 10MW - TFTR and JET-had maximal neutron load approximately 0.1MW/m2 only in transient (with time scale lower 1s) regimes. The quasi steady state neutron emission regime (˜5MW, 5sec) was performed in JET with mean surface neutron load lower than 0.025MW/m2 only. In this communication it will be discussed the main needs of JET scale tokamak improvement for increase on neutron load up to 0.2MW/m2. They are: decrease of Zeff by ECRH and lithium use as plasma facing components, the increase of energy of steady state neutral injectors up to 150-170keV (tritium), the He removal and creation of closed loop of DT fuel circulation.
Collective electric field effects on the confinement of fast ions in tokamaks
McClements, K.G.; Thyagaraja, A.
2006-04-15
The injection of neutral particle beams counter to the plasma current direction in the Mega-Ampere Spherical Tokamak (MAST) [A. Sykes, R. J. Akers, L. C. Appel et al., Nucl. Fusion, 41, 1423 (2001)] leads to substantial losses of energetic beam ions and also rapid toroidal rotation. The electrodynamic consequences of energetic ion loss on tokamak plasmas are explored in light of results from the MAST counterinjection experiments and test particle calculations of the current density due to escaping ions. Previous authors have noted that there are two possible consequences of such a current: either a compensating bulk plasma return current is set up, or the plasma behaves as an insulator, with the energetic ion current balanced by a displacement current rather than a conduction current. Radial electric fields and hence toroidal flows occur in both cases, but higher fields are predicted in the insulating case. Such fields are important because they can confine both fast ions and bulk plasma (via the suppression of turbulent transport). The return current scenario, which appears to be operative during counterinjection in MAST, is shown to be applicable if there is a sufficiently high level of momentum transport in the bulk ions; electrons cannot carry the return current, although they contribute to an ambipolar particle flux on the plasma confinement time scale. The insulating scenario may be applicable to high confinement regimes in burning tokamak plasmas.
Collective electric field effects on the confinement of fast ions in tokamaks
NASA Astrophysics Data System (ADS)
McClements, K. G.; Thyagaraja, A.
2006-04-01
The injection of neutral particle beams counter to the plasma current direction in the Mega-Ampère Spherical Tokamak (MAST) [A. Sykes, R. J. Akers, L. C. Appel et al., Nucl. Fusion, 41, 1423 (2001)] leads to substantial losses of energetic beam ions and also rapid toroidal rotation. The electrodynamic consequences of energetic ion loss on tokamak plasmas are explored in light of results from the MAST counterinjection experiments and test particle calculations of the current density due to escaping ions. Previous authors have noted that there are two possible consequences of such a current: either a compensating bulk plasma return current is set up, or the plasma behaves as an insulator, with the energetic ion current balanced by a displacement current rather than a conduction current. Radial electric fields and hence toroidal flows occur in both cases, but higher fields are predicted in the insulating case. Such fields are important because they can confine both fast ions and bulk plasma (via the suppression of turbulent transport). The return current scenario, which appears to be operative during counterinjection in MAST, is shown to be applicable if there is a sufficiently high level of momentum transport in the bulk ions; electrons cannot carry the return current, although they contribute to an ambipolar particle flux on the plasma confinement time scale. The insulating scenario may be applicable to high confinement regimes in burning tokamak plasmas.
Mission and physics design of the Tokamak Physics Experiment
Neilson, G.H.; Batchelor, D.B.; Mioduszewski, P.K.; Strickler, D.J.; Bonoli, P.T.; Porkolab, M.; Goldston, R.J.; Jardin, S.C.; Bialek, J.M.; Kessel, C.E.
1994-11-01
Improvements in the confinement, stability limits, current-drive efficiency and divertor performance, combined with steady-state operation, can lead to a more economical tokamak fusion reactor than one based on the present physics data base. The Tokamak Physics Experiment (TPX) is planned to extend the recent advances in these areas, achieved in pulsed tokamaks, to the steady-state regime. In so doing, it will develop a data base needed for the design of an economically attractive tokamak reactor.
Engineering design of the National Spherical Torus Experiment
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.
Study of spherical torus based volume neutron source
Cheng, E.T.; Peng, Yueng Kay Martin
1998-01-01
With the worldwide development of fusion power focusing on the design of the International Thermonuclear Experimental Reactor (ITER), developmental strategies for the demonstration fusion power plant (DEMO) are being discussed. A relatively prudent strategy is to construct and operate a small deuterium tritium fuelled volumetric neutron source (VNS) in parallel with ITER. The VNS is to provide, over a period less than 20 years, a relatively high fusion neutron fluence of 6 MW year m2 and wall loading of 1 MW m2 or more, over an accessible blanket test area of more than 10 m2. Such a VNS would complement ITER in testing, developing, and qualifying nuclear technology components, materials, and their combinations for DEMO and future commercial power plants. The effort of this study has established the potential of the spherical tokamak as a credible VNS concept that satisfies the above requirements.
Spherical artifacts on ferrograms
NASA Technical Reports Server (NTRS)
Jones, W. R., Jr.
1976-01-01
In the past, hollow spheres detected on ferrograms have been interpreted as being due to fretting, abrasion, cavitation erosion, and fatigue-related processes. Here it is reported that such spheres were found to result from the fact that a routine grinding operation on a steel plate was carried out about 20 feet away from the ferrograph. A similar grinding operation was performed on a piece of low carbon steel a few feet from the ferrograph, and after a few minutes of grinding, the resulting ferrogram contained thousands of particles of which more than 90% were spherical. Because of the widespread occurrence of ordinary grinding operations, it seems prudent that those utilizing the ferrograph be cognizant of this type of artifact.
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.
Tokamak x ray diagnostic instrumentation
Hill, K.W.; Beiersdorfer, P.; Bitter, M.; Fredrickson, E.; Von Goeler, S.; Hsuan, H.; Johnson, L.C.; Liew, S.L.; McGuire, K.; Pare, V.
1987-01-01
Three classes of x-ray diagnostic instruments enable measurement of a variety of tokamak physics parameters from different features of the x-ray emission spectrum. (1) The soft x-ray (1 to 50 keV) pulse-height-analysis (PHA) diagnostic measures impurity concentrations from characteristic line intensities and the continuum enhancement, and measures the electron temperature from the continuum slope. (2) The Bragg x-ray crystal spectrometer (XCS) measures the ion temperature and neutral-beam-induced toroidal rotation velocity from the Doppler broadening and wavelength shift, respectively, of spectral lines of medium-Z impurity ions. Impurity charge state distributions, precise wavelengths, and inner-shell excitation and recombination rates can also be studied. X rays are diffracted and focused by a bent crystal onto a position-sensitive detector. The spectral resolving power E/..delta..E is greater than 10/sup 4/ and time resolution is 10 ms. (3) The x-ray imaging system (XIS) measures the spatial structure of rapid fluctuations (0.1 to 100 kHZ) providing information on MHD phenomena, impurity transport rates, toroidal rotation velocity, plasma position, and the electron temperature profile. It uses an array of silicon surface-barrier diodes which view different chords of the plasma through a common slot aperture and operate in current (as opposed to counting) mode. The effectiveness of shields to protect detectors from fusion-neutron radiation effects has been studied both theoretically and experimentally.
Intrinsic rotation in tokamaks: theory
NASA Astrophysics Data System (ADS)
Parra, Felix I.; Barnes, Michael
2015-04-01
Self-consistent equations for intrinsic rotation in tokamaks with small poloidal magnetic field Bp compared to the total magnetic field B are derived. The model gives the momentum redistribution due to turbulence, collisional transport and energy injection. Intrinsic rotation is determined by the balance between the momentum redistribution and the turbulent diffusion and convection. Two different turbulence regimes are considered: turbulence with characteristic perpendicular lengths of the order of the ion gyroradius, ρi, and turbulence with characteristic lengths of the order of the poloidal gyroradius, (B/Bp)ρi. Intrinsic rotation driven by gyroradius scale turbulence is mainly due to the effect of neoclassical corrections and of finite orbit widths on turbulent momentum transport, whereas for the intrinsic rotation driven by poloidal gyroradius scale turbulence, the slow variation of turbulence characteristics in the radial and poloidal directions and the turbulent particle acceleration can be become as important as the neoclassical and finite orbit width effects. The magnetic drift is shown to be indispensable for the intrinsic rotation driven by the slow variation of turbulence characteristics and the turbulent particle acceleration. The equations are written in a form conducive to implementation in a flux tube code, and the effect of the radial variation of the turbulence is included in a novel way that does not require a global gyrokinetic formalism.
Microtearing modes in tokamak discharges
NASA Astrophysics Data System (ADS)
Rafiq, T.; Weiland, J.; Kritz, A. H.; Luo, L.; Pankin, A. Y.
2016-06-01
Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.
Microtearing modes in tokamak discharges
Rafiq, T.; Kritz, A. H.; Weiland, J.; Luo, L.; Pankin, A. Y.
2016-06-15
Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.
Toroidal Flow in Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Callen, J. D.; Cole, A. J.; Hegna, C. C.
2007-11-01
Many effects influence toroidal flow evolution in tokamak plasmas. Momentum sources and radial diffusion due to axisymmetric neoclassical, paleoclassical and anomalous transport are usually considered. In addition, the toroidal flow can be affected by field errors. Small, non-axisymmetric field errors arise from coil irregularities, active control coils and collective plasma magnetic distortions (e.g., NTMs, RWMs). Resonant field errors cause localized electromagnetic torques near rational surfaces in the plasma, which can lock the plasma to the wall leading to magnetic islands and reduced confinement or disruptions. Their penetration into the plasma is limited by flow-shielding effects; but they can be amplified by the plasma response at high beta. Non-resonant field errors cause magnetic pumping and radial banana drifts, and lead to toroidal flow damping over the entire plasma. Many of these processes can also produce momentum pinch and intrinsic flow effects. This poster will seek to present a coherent picture of all these effects and suggest ways they could be tested and distinguished experimentally.
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.
Activation analysis of the compact ignition tokamak
Selcow, E.C.
1986-01-01
The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak.
Helicity content and tokamak applications of helicity
Boozer, A.H.
1986-05-01
Magnetic helicity is approximately conserved by the turbulence associated with resistive instabilities of plasmas. To generalize the application of the concept of helicity, the helicity content of an arbitrary bounded region of space will be defined. The definition has the virtues that both the helicity content and its time derivative have simple expressions in terms of the poloidal and toroidal magnetic fluxes, the average toroidal loop voltage and the electric potential on the bounding surface, and the volume integral of E-B. The application of the helicity concept to tokamak plasmas is illustrated by a discussion of so-called MHD current drive, an example of a stable tokamak q profile with q less than one in the center, and a discussion of the possibility of a natural steady-state tokamak due to the bootstrap current coupling to tearing instabilities.
Double slotted socket spherical joint
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.
SPHERICAL SHOCK WAVES IN SOLIDS
Contents: Introduction-Reasons for Studying Spherical Shock Waves, Physics of Cavity Expansion due to Explosive Impact, General Nature of Shock Waves...Governing Differential Equation of Self-Similar Motion; Application of the Theory of Self-Similar Motion to the Problem of Expansion of a Spherical
Physics of Tokamak Plasma Start-up
NASA Astrophysics Data System (ADS)
Mueller, Dennis
2012-10-01
This tutorial describes and reviews the state-of-art in tokamak plasma start-up and its importance to next step devices such as ITER, a Fusion Nuclear Science Facility and a Tokamak/ST demo. Tokamak plasma start-up includes breakdown of the initial gas, ramp-up of the plasma current to its final value and the control of plasma parameters during those phases. Tokamaks rely on an inductive component, typically a central solenoid, which has enabled attainment of high performance levels that has enabled the construction of the ITER device. Optimizing the inductive start-up phase continues to be an area of active research, especially in regards to achieving ITER scenarios. A new generation of superconducting tokamaks, EAST and KSTAR, experiments on DIII-D and operation with JET's ITER-like wall are contributing towards this effort. Inductive start-up relies on transformer action to generate a toroidal loop voltage and successful start-up is determined by gas breakdown, avalanche physics and plasma-wall interaction. The goal of achieving steady-sate tokamak operation has motivated interest in other methods for start-up that do not rely on the central solenoid. These include Coaxial Helicity Injection, outer poloidal field coil start-up, and point source helicity injection, which have achieved 200, 150 and 100 kA respectively of toroidal current on closed flux surfaces. Other methods including merging reconnection startup and Electron Bernstein Wave (EBW) plasma start-up are being studied on various devices. EBW start-up generates a directed electron channel due to wave particle interaction physics while the other methods mentioned rely on magnetic helicity injection and magnetic reconnection which are being modeled and understood using NIMROD code simulations.
Current drive by spheromak injection into a tokamak
NASA Astrophysics Data System (ADS)
Brown, M. R.; Bellan, P. M.
1990-04-01
We report the first observation of current drive by injection of a spheromak plasma into a tokamak (Caltech ENCORE small reasearch tokamak) due to the process of helicity injection. After an abrupt 30% increase, the tokamak current decays by a factor of 3 due to plasma cooling caused by the merging of the relatively cold spheromak with the tokamak. The tokamak density profile peaks sharply due to the injected spheromak plasma (n¯3 increases by a factor of 6) then becomes hollow, suggestive of an interchange instability.
Optimal mollifiers for spherical deconvolution
NASA Astrophysics Data System (ADS)
Hielscher, Ralf; Quellmalz, Michael
2015-08-01
This paper deals with the inversion of the spherical Funk-Radon transform, and, more generally, with the inversion of spherical convolution operators from the point of view of statistical inverse problems. This means we consider discrete data perturbed by white noise and aim at estimators with optimal mean square error for functions out of a Sobolev ball. To this end we analyze a specific class of estimators built upon the spherical hyperinterpolation operator, spherical designs and the mollifier approach. Eventually, we determine optimal mollifier functions with respect to the noise level, the number of data points and the smoothness of the original function. We complete this paper by providing a fast algorithm for the numerical computation of the estimator, which is based on the fast spherical Fourier transform, and by illustrating our theoretical results with numerical experiments.
Electron cyclotron emission diagnostics on KSTAR tokamak
Jeong, S. H.; Lee, K. D.; Kwon, M.; Kogi, Y.; Kawahata, K.; Nagayama, Y.; Mase, A.
2010-10-15
A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.
Overview of the National Centralized Tokamak programme
NASA Astrophysics Data System (ADS)
Kikuchi, M.; Tamai, H.; Matsukawa, M.; Fujita, T.; Takase, Y.; Sakurai, S.; Kizu, K.; Tsuchiya, K.; Kurita, G.; Morioka, A.; Hayashi, N.; Miura, Y.; Itoh, S.; Bialek, J.; Navratil, G.; Ikeda, Y.; Fujii, T.; Kurihara, K.; Kubo, H.; Kamada, Y.; Miya, N.; Suzuki, T.; Hamamatsu, K.; Kawashima, H.; Kudo, Y.; Masaki, K.; Takahashi, H.; Takechi, M.; Akiba, M.; Okuno, K.; Ishida, S.; Ichimura, M.; Imai, T.; Hashizume; Miura, Y. M.; Horiike, H.; Kimura, A.; Tsutsui, H.; Matsuoka, M.; Uesugi, Y.; Sagara, A.; Nishimura, A.; Shimizu, A.; Sakamoto, M.; Nakamura, K.; Sato, K.; Okano, K.; Ida, K.; Shimada, H. R.; Kishimoto, Y.; Azechi, H.; Tanaka, S.; Yatsu, K.; Yoshida, N.; Inutake, M.; Fujiwara, M.; Inoue, N.; Hosogane, N.; Kuriyama, M.; Ninomiya, H.
2006-03-01
An overview is given of the National Centralized Tokamak (NCT) programme as a research programme for advanced tokamak research to succeed JT-60U. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility is pursued in aspect ratio and shape controllability for the demonstration of the high-β steady-state, feedback control of resistive wall modes, wide current and pressure profile control capability and also very long pulse steady-state operation. Existing JT-60 infrastructure such as the heating and current drive system, power supplies and cooling systems will be best utilized for this modification.
Tokamak power systems studies, FY 1985
Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.
1985-12-01
The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.
Tokamak Spectroscopy for X-Ray Astronomy
NASA Technical Reports Server (NTRS)
Fournier, Kevin B.; Finkenthal, M.; Pacella, D.; May, M. J.; Soukhanovskii, V.; Mattioli, M.; Leigheb, M.; Rice, J. E.
2000-01-01
This paper presents the measured x-ray and Extreme Ultraviolet (XUV) spectra of three astrophysically abundant elements (Fe, Ca and Ne) from three different tokamak plasmas. In every case, each spectrum touches on an issue of atomic physics that is important for simulation codes to be used in the analysis of high spectral resolution data from current and future x-ray telescopes. The utility of the tokamak as a laboratory test bed for astrophysical data is demonstrated. Simple models generated with the HULLAC suite of codes demonstrate how the atomic physics issues studied can affect the interpretation of astrophysical data.
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.
Immunomodulatory spherical nucleic acids.
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
2015-03-31
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.
Immunomodulatory spherical nucleic acids
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
A spherical electrostatic orrery
NASA Astrophysics Data System (ADS)
Smetana, Carole; Alexander, David; Robertson, Scott; Vilkaitis, Kim; Walch, Bob
1996-11-01
An electrostatic orrery for studying Keplerian orbits has been constructed in which one or more negatively charged hollow glass microparticles orbit a 9.5-mm-diam metal sphere at +8-kV potential in a vacuum. The device is similar to an earlier cylindrical orrery in which particles orbit a rod [Biewer et al., Am. J. Phys. 62(9), 821-827 (1994)]. Electrically biased cylinders covering the rod supporting the sphere give nearly spherical potential surfaces inside the trap. Additional electrodes at the boundary are used to reduce the perturbation of gravity and to prevent motion resulting in collisions with the supporting rod. Orbits last approximately 10 min or about 104 revolutions. The orbiters are illuminated with a slide projector and can be seen with the naked eye as well as videotaped. The trap has been used to observe orbital precession, interparticle collisions, and the effects of time-independent perturbations. This apparatus provides an opportunity for the study and demonstration of orbital motion in a laboratory.
Kinetic full wave analyses of O-X-B mode conversion of EC waves in tokamak plasmas
NASA Astrophysics Data System (ADS)
Fukuyama, Atsushi; Khan, Shabbir Ahmad; Igami, Hiroe; Idei, Hiroshi
2016-10-01
For heating and current drive in a high-density plasma of tokamak, especially spherical tokamak, the use of electron Bernstein waves and the O-X-B mode conversion were proposed and experimental observations have been reported. In order to evaluate the power deposition profile and the current drive efficiency, kinetic full wave analysis using an integral form of dielectric tensor has been developed. The incident angle dependence of wave structure and O-X-B mode conversion efficiency is examined using one-dimensional analysis in the major radius direction. Two-dimensional analyses on the horizontal plane and the poloidal plane are also conducted, and the wave structure and the power deposition profile are compared with those of previous analyses using ray tracing method and cold plasma approximation. This work is supported by JSPS KAKENHI Grant Number JP26630471.
Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas
Green, David L; Jaeger, E. F.; Berry, Lee A; Chen, Guangye; Ryan, Philip Michael; Canik, John
2011-01-01
Observations of improved radio frequency (RF) heating efficiency in high-confinement (H-) mode plasmas on the National Spherical Tokamak Experiment (NSTX) are investigated by whole-device linear simulation. We present the first full-wave simulation to couple kinetic physics of the well confined core plasma to the poorly confined scrape-off plasma. The new simulation is used to scan the launched fast-wave spectrum and examine the steady-state electric wave field structure for experimental scenarios corresponding to both reduced, and improved RF heating efficiency. We find that launching toroidal wave-numbers that required for fast-wave propagation excites large amplitude (kVm 1 ) coaxial standing modes in the wave electric field between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggest these modes are a probable cause of degraded heating efficiency. Also, the H-mode density pedestal and fast-wave cutoff within the confined plasma allow for the excitation of whispering gallery type eigenmodes localised to the plasma edge.
Spherical 3D isotropic wavelets
NASA Astrophysics Data System (ADS)
Lanusse, F.; Rassat, A.; Starck, J.-L.
2012-04-01
Context. Future cosmological surveys will provide 3D large scale structure maps with large sky coverage, for which a 3D spherical Fourier-Bessel (SFB) analysis in spherical coordinates is natural. Wavelets are particularly well-suited to the analysis and denoising of cosmological data, but a spherical 3D isotropic wavelet transform does not currently exist to analyse spherical 3D data. Aims: The aim of this paper is to present a new formalism for a spherical 3D isotropic wavelet, i.e. one based on the SFB decomposition of a 3D field and accompany the formalism with a public code to perform wavelet transforms. Methods: We describe a new 3D isotropic spherical wavelet decomposition based on the undecimated wavelet transform (UWT) described in Starck et al. (2006). We also present a new fast discrete spherical Fourier-Bessel transform (DSFBT) based on both a discrete Bessel transform and the HEALPIX angular pixelisation scheme. We test the 3D wavelet transform and as a toy-application, apply a denoising algorithm in wavelet space to the Virgo large box cosmological simulations and find we can successfully remove noise without much loss to the large scale structure. Results: We have described a new spherical 3D isotropic wavelet transform, ideally suited to analyse and denoise future 3D spherical cosmological surveys, which uses a novel DSFBT. We illustrate its potential use for denoising using a toy model. All the algorithms presented in this paper are available for download as a public code called MRS3D at http://jstarck.free.fr/mrs3d.html
Transformer Recharging with Alpha Channeling in Tokamaks
N.J. Fisch
2009-12-21
Transformer recharging with lower hybrid waves in tokamaks can give low average auxiliary power if the resistivity is kept high enough during the radio frequency (rf) recharging stage. At the same time, operation in the hot ion mode via alpha channeling increases the effective fusion reactivity. This paper will address the extent to which these two large cost saving steps are compatible. __________________________________________________
Spontaneous generation of rotation in tokamak plasmas
Parra Diaz, Felix
2013-12-24
Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.
INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS
HUMPHREYS,D.A; FERRON,J.R; JOHNSON,R.D; LEUER,J.A; PENAFLOR,B.G; WALKER,M.L; WELANDER,A.S; KHAYRUTDINOV,R.R; DOKOUKA,V; EDGELL,D.H; FRANSSON,C.M
2003-10-01
OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance.
Simulation of runaway electrons in tokamak
NASA Astrophysics Data System (ADS)
Guo, Zehua; Tang, Xianzhu; McDevitt, Chris
2015-11-01
Runaway electrons with relativisitc energy (>Mev) are generated in tokamaks when the acceleration by parallel electric field exceeds the drag due to Coulomb collisions with the bulk plasma. Carrying about 70% of the ITER thermal current (15MA), they can possibly cause severe damage to tokamak facing components. Here we report the development of a solver for computing the evolution of runaway electron distribution in tokamak geometries. Essential effects from Coulomb collisions, radiation losses, toroidal effects and the radial transport are included on the same footings. Numerical techniques (implicit-explicit time-stepping, KT/NT central schemes) to overcome the difficulties arising from the wide spread of time scales in runaway electron dynamics and the hyperbolic nature of the relativistic Fokker-Planck equation will be discussed. We will use the solver to study two important physics: 1) the presence of stable point in the phase space and its relation to the electric field threshold; 2) the radial transport of runaways in tokamak geometry and its effects on the distribution function. Work supported by DOE via LANL-LDRD.
Tokamak startup with electron cyclotron heating
Holly, D J; Prager, S C; Shepard, D A; Sprott, J C
1980-04-01
Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed.
Analysis of sawtooth relaxation oscillations in tokamaks
Yamazaki, K.; McGuire, K.; Okabayashi, M.
1982-07-01
Sawtooth relaxation oscillations are analyzed using the Kadomtsev's disruption model and a thermal relaxation model. The sawtooth period is found to be very sensitive to the thermal conduction loss. Qualitative agreement between these calculations and the sawtooth period observed in several tokamaks is demonstrated.
Banana drift transport in tokamaks with ripple
Linsker, R.; Boozer, A.H.
1982-01-01
Ripple transport in tokamaks is discussed for the ''banana drift'' collisionality regime, which lies below the ripple plateau regime treated earlier. The physical mechanisms that dominate banana drift transport are found to differ from those considered in previous work on this regime, and consequently the resulting transport coefficients can differ by several orders of magnitude.
Banana drift transport in tokamaks with ripple
Linsker, R.; Boozer, A.H.
1981-04-01
Ripple transport in tokamaks is discussed for the banana drift collisionality regime, which lies below the ripple plateau regime treated earlier. The physical mechanisms that dominate banana drift transport are found to differ from those considered in previous work on this regime, and the resulting transport coefficients can consequently differ by several orders of magnitude.
Stabilization of tokamak plasma by lithium streams
L.E. Zakharov
2000-08-07
The stabilization theory of free-boundary magnetohydrodynamic instabilities in tokamaks by liquid lithium streams driven by magnetic propulsion is formulated. While the conventional, wall-locked, resistive wall mode can be well suppressed by the flow, a new, stream-locked mode determines the limits of the flow stabilization.
UCLA Tokamak Program Close Out Report.
Taylor, Robert John
2014-02-04
The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.
Microinstabilities in weak density gradient tokamak systems
Tang, W.M.; Rewoldt, G.; Chen, L.
1986-04-01
A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient.
Fusion reactor design studies. [ARIES Tokamak
Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.
1990-10-12
This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources. (LSP)
Elementary Processes Underlying Alpha Channeling in Tokamaks
NM.J. Fisch
2012-06-15
Alpha channeling in tokamaks is speculative, but also extraordinarily attractive. Waves that can accomplish this effect have been identified. Key aspects of the theory now enjoy experimental confirmation. This paper will review the elementary processes of wave-particle interactions in plasma that underlie the alpha channeling effect
Diagnostics for neutral-beam-heated tokamaks
Goldston, R.J.
1982-12-01
Diagnostic techniques for neutral-beam-heated tokamak plasmas fall into three categories: (1) magnetic diagnostics for measurements of gross stored energy, (2) profile diagnostics for measurements of stored thermal and beam energy, impurity content and plasma rotation, and (3) fast time resolution diagnostics to study MHD fluctuations and micro-turbulence.
C.Z. Cheng; G.Y. Fu; M.V. Gorelenkova; N.N. Gorelenkov; R. White; S. Kaye
1999-12-10
Toroidicity induced Alfven Eigenmode (TAE) stability in National Spherical Torus Experiment (NSTX) is analyzed using the improved NOVA-K code, which includes finite orbit width and Larmor radius effects and is able to predicts the saturation amplitude for the mode using the quasilinear theory. Broad spectrum of unstable global TAEs with different toroidal mode numbers is predicted. Due to the strong poloidal field and the presence of the magnetic well in NSTX better particle confinement in the presence of TAEs in comparison with tokamaks is illustrated making use of the ORBIT code.
First Observation Of ELM Pacing With Vertical Jogs In A Spherical Torus
Gerhardt, S P; Canik, J M; Maingi, R; Bell, R; Gates, d; Goldston, R; Hawryluk, R; Le Blanc, B P; Menard, J; Sontag, A C; Sabbagh, S
2010-07-15
Experiments in a number of conventional aspect ratio tokamaks have been successful in pacing edge localized modes (ELMs) by rapid vertical jogging of the plasma. This paper demonstrates the first pacing of ELMs in a spherical torus plasma. Applied 30 Hz vertical jogs synchronized the ELMs with the upward motion of the plasma. 45 Hz jogs also lead to an increase in the ELM frequency, though the synchronization of the ELMs and jogs was unclear. A reduction in the ELM energy was observed at the higher driven ELM frequencies. __________________________________________________
Simulation of microtearing turbulence in national spherical torus experiment
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.
Gas Puff Imaging Studies of Tokamak Edge Physics in the National Spherical Torus Experiment
NASA Astrophysics Data System (ADS)
Sechrest, Yancey
In order to be viable, Next-step fusion devices must overcome two pressing problems: they must be able to achieve high levels of confinement while also handling potentially damaging heat loads on material surfaces. The study of plasma edge physics promises solutions to both problems because the plasma edge, being the boundary between confined and unconfined regions, plays a key role in determining the global confinement and the plasma interaction with material surfaces (e.g. edge transport barriers, pedestal evolution, and edge localized modes). However, the steep gradients in density and temperature in the plasma edge that drive strong fluctuations in plasma parameters require measurements of fluctuations with high spatial and temporal resolution. By measuring drift scale (kyrhos < 2) fluctuations for frequencies less than ˜ 200 kHz, Gas Puff Imaging (GPI) meets these requirements while providing two-dimensional coverage at a large number of measurement locations. This dissertation presents GPI studies of transitions from low to high confinement regimes (L-H transitions) and Edge Localized Modes (ELMs). In 2010, a study of L-H transitions with the GPI diagnostic revealed quasi-periodic reductions in the scrape-off-layer turbulence levels during the 30 ms preceding the transition. The two-dimensional flow fields for these "quiet-periods", estimated from the GPI data by a pattern-matching velocimetry technique, exhibit intriguing similarity with the Drift Wave - Zonal Flow paradigm, a leading candidate in explaining L-H transitions. Following this study, a survey of GPI data from RF heated H-mode plasmas near the L-H power threshold identified short-lived, coherent oscillations in edge emission preceding the ELM crash. These observations provide detailed two-dimensional dynamics of the growth, filamentation, and crash of the ELM event, which could improve our understanding through comparison with nonlinear simulation. Cross diagnostic comparisons of GPI and Beam Emission Spectroscopy measurements of edge fluctuations are also presented.
Bi-directional Alfvén cyclotron instabilities in the mega-amp spherical tokamak
Sharapov, S. E. Akers, R.; Ayed, N. Ben; Cunningham, G.; Lilley, M. K.; Cecconello, M.; Cook, J. W. C.; Verwichte, E.
2014-08-15
Alfvén cyclotron instabilities excited by velocity gradients of energetic beam ions were investigated in MAST experiments with super-Alfvénic neutral beam injection over a wide range of toroidal magnetic fields from ∼0.34 T to ∼0.585 T. In MAST discharges with high magnetic field, a discrete spectrum of modes in the sub-cyclotron frequency range is excited toroidally propagating counter to the beam and plasma current (toroidal mode numbers n < 0). At lower magnetic field ≤0.45 T, a discrete spectrum of Compressional Alfvén Eigenmodes (CAEs) with n > 0 arises, in addition to the modes with n < 0. At lowest magnetic fields, the CAEs with n > 0 become dominant, they are observed in frequency range from ∼250 kHz for n=1 to ∼3.5 MHz for n=15, well above the on-axis ion cyclotron frequency (∼2.5 MHz). The data is interpreted in terms of normal and anomalous Doppler resonances modified by magnetic drift terms due to inhomogeneity and curvature of the magnetic field. A Hall MHD model is applied for computing the eigenfrequencies and the spatial mode structure of CAEs and a good agreement with the experimental frequencies is found.
Kinetic effects in the conversion of fast waves in pre-heated, low aspect ratio tokamak plasmas
NASA Astrophysics Data System (ADS)
Kommoshvili, K.; Cuperman, S.; Bruma, C.
2003-03-01
Kinetic effects in the conversion of fast waves to Alfvèn waves and their subsequent deposition in low aspect ratio (spherical) tokamaks (LARTs) have been investigated theoretically. More specifically, we have considered the consequences of incorporation of kinetic effects in the electron parallel (to the ambient magnetic field) dynamics derived by following the drift-tearing mode analysis of Chen et al (Chen L, Rutherford P H and Tang W M 1977 Phys. Rev. Lett. 39 460), and particle-conserving Krook collision operator for the passing electrons involved (Mett R R and Mahajan S M 1992 Phys. Fluids B 4 2885). The perpendicular plasma dynamics is described by a quite general resistive two-fluid (2F) model based dielectric tensor-operator (Cuperman S, Bruma C and Komoshvili K 2002 Solution of the resistive 2F wave equations for Alfvènic modes in spherical tokamak plasmas J. Plasma Phys. accepted for publication). The full-wave electromagnetic equations, formulated in terms of the vector and scalar potentials, have been solved by the aid of an advanced finite elements numerical code (Sewell G 1993 Adv. Eng. Software 17 105). Detailed solutions of the full-wave equations are obtained and compared with those corresponding to a pure resistive 2F model, this, for the illustrative pre-heated START-type device (Sykes 1994). Our results quantitatively confirm the general theory of the conversion of fast waves with subsequent power dissipation for the conditions of spherical tokamaks thus providing the required auxilliary energy source for the succesful operation of LARTs. Moreover, these results indicate the absolute necessity of using a full model for the parallel electron dynamics, i.e. including both kinetic and collisional effects.
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.
Spherical harmonics in texture analysis
NASA Astrophysics Data System (ADS)
Schaeben, Helmut; van den Boogaart, K. Gerald
2003-07-01
The objective of this contribution is to emphasize the fundamental role of spherical harmonics in constructive approximation on the sphere in general and in texture analysis in particular. The specific purpose is to present some methods of texture analysis and pole-to-orientation probability density inversion in a unifying approach, i.e. to show that the classic harmonic method, the pole density component fit method initially introduced as a distinct alternative, and the spherical wavelet method for high-resolution texture analysis share a common mathematical basis provided by spherical harmonics. Since pole probability density functions and orientation probability density functions are probability density functions defined on the sphere Ω3⊂ R3 or hypersphere Ω4⊂ R4, respectively, they belong at least to the space of measurable and integrable functions L1( Ωd), d=3, 4, respectively. Therefore, first a basic and simplified method to derive real symmetrized spherical harmonics with the mathematical property of providing a representation of rotations or orientations, respectively, is presented. Then, standard orientation or pole probability density functions, respectively, are introduced by summation processes of harmonic series expansions of L1( Ωd) functions, thus avoiding resorting to intuition and heuristics. Eventually, it is shown how a rearrangement of the harmonics leads quite canonically to spherical wavelets, which provide a method for high-resolution texture analysis. This unified point of view clarifies how these methods, e.g. standard functions, apply to texture analysis of EBSD orientation measurements.
Neural net prediction of tokamak plasma disruptions
NASA Astrophysics Data System (ADS)
Hernandez, J. V.; Lin, Z.; Horton, W.; Vannucci, A.; McCool, S. C.
1994-10-01
The computation based on neural net algorithms in predicting minor and major disruptions in TEXT tokamak discharges has been performed. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal, measured by a single Mirnov coil. The time step used (= 0.04ms) corresponds to the experimental data sampling rate. Two kinds of approaches are adopted for the task, the contiguous future prediction and the multi-timescale prediction. Results are shown for comparison. Both networks are trained through the back-propagation algorithm with inertial terms. The degree of this success indicates that the magnetic fluctuations associated with tokamak disruptions may be characterized by a relatively low-dimensional dynamical system.
Global migration of impurities in tokamaks
NASA Astrophysics Data System (ADS)
Hakola, A.; Airila, M. I.; Björkas, C.; Borodin, D.; Brezinsek, S.; Coad, J. P.; Groth, M.; Järvinen, A.; Kirschner, A.; Koivuranta, S.; Krieger, K.; Kurki-Suonio, T.; Likonen, J.; Lindholm, V.; Makkonen, T.; Mayer, M.; Miettunen, J.; Müller, H. W.; Neu, R.; Petersson, P.; Rohde, V.; Rubel, M.; Widdowson, A.; the ASDEX Upgrade Team; Contributors, JET-EFDA
2013-12-01
The migration of impurities in tokamaks has been studied with the help of tracer-injection (13C and 15N) experiments in JET and ASDEX Upgrade since 2001. We have identified a common pattern for the migrating particles: scrape-off layer flows drive impurities from the low-field side towards the high-field side of the vessel. Migration is also sensitive to the density and magnetic configuration of the plasma, and strong local variations in the resulting deposition patterns require 3D treatment of the migration process. Moreover, re-erosion of the deposited particles has to be taken into account to properly describe the migration process during steady-state operation of the tokamak.
Filamentary probe on the COMPASS tokamak
NASA Astrophysics Data System (ADS)
Kovarik, K.; Duran, I.; Stockel, J.; Seidl, J.; Adamek, J.; Spolaore, M.; Vianello, N.; Hacek, P.; Hron, M.; Panek, R.
2017-03-01
This paper describes a new filamentary probe recently introduced on the COMPASS tokamak. It allows the measurement of electrostatic and magnetic properties of the filaments and their changes in dependence on distance from the separatrix in the region between a divertor and midplane. The probe head is mounted on a manipulator moving the probe radially on a shot-to-shot basis. This configuration is suitable for the long term statistical measurement of the plasma filaments and the measurement of their evolution during their propagation from the separatrix to the wall. The basics of the filamentary probe construction, the evolution of the plasma parameters, and first conditional averages of the plasma filaments in the scrape-off layer of the COMPASS tokamak during the L-mode regime are presented.
Rapidly Moving Divertor Plates In A Tokamak
S. Zweben
2011-05-16
It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.
Filamentary probe on the COMPASS tokamak.
Kovarik, K; Duran, I; Stockel, J; Seidl, J; Adamek, J; Spolaore, M; Vianello, N; Hacek, P; Hron, M; Panek, R
2017-03-01
This paper describes a new filamentary probe recently introduced on the COMPASS tokamak. It allows the measurement of electrostatic and magnetic properties of the filaments and their changes in dependence on distance from the separatrix in the region between a divertor and midplane. The probe head is mounted on a manipulator moving the probe radially on a shot-to-shot basis. This configuration is suitable for the long term statistical measurement of the plasma filaments and the measurement of their evolution during their propagation from the separatrix to the wall. The basics of the filamentary probe construction, the evolution of the plasma parameters, and first conditional averages of the plasma filaments in the scrape-off layer of the COMPASS tokamak during the L-mode regime are presented.
Models for impurity effects in tokamaks
Hogan, J.T.
1980-03-01
Models for impurity effects in tokamaks are described with an emphasis on the relationship between attainment of high ..beta.. and impurity problems. We briefly describe the status of attempts to employ neutral beam heating to achieve high ..beta.. in tokamaks and propose a qualitative model for the mechanism by which heavy metal impurities may be produced in the startup phase of the discharge. We then describe paradoxes in impurity diffusion theory and discuss possible resolutions in terms of the effects of large-scale islands and sawtooth oscillations. Finally, we examine the prospects for the Zakharov-Shafranov catastrophe (long time scale disintegration of FCT equilibria) in the context of present and near-term experimental capability.
A low aspect ratio tokamak transmutation system
NASA Astrophysics Data System (ADS)
Qiu, L. J.; Wu, Y. C.; Xiao, B. J.; Xu, Q.; Huang, Q. Y.; Wu, B.; Chen, Y. X.; Xu, W. N.; Chen, Y. P.; Liu, X. P.
2000-03-01
A low aspect ratio tokamak transmutation system is proposed as an alternative application of fusion energy on the basis of a review of previous studies. This system includes: (1) a low aspect ratio tokamak as fusion neutron driver, (2) a radioactivity-clean nuclear power system as blanket, and (3) a novel concept of liquid metal centre conductor post as part of the toroidal field coils. In the conceptual design, a driver of 100 MW fusion power under 1 MW/m2 neutron wall loading can transmute the amount of high level waste (including minor actinides and fission products) produced by ten standard pressurized water reactors of 1 GW electrical power output. Meanwhile, the system can produce tritium on a self-sustaining basis and an output of about 2 GW of electrical energy. After 30 years of operation, the biological hazard potential level of the whole system will decrease by two orders of magnitude.
Boundary Plasma Turbulence Simulations for Tokamaks
Xu, X; Umansky, M; Dudson, B; Snyder, P
2008-05-15
The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T{sub e}; T{sub i}) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics.
Tritium Retention and Removal in Tokamaks
Skinner, Charles H.
2009-02-19
Management of tritium inventory remains one of the grand challenges in the development of fusion energy. Tritium is an important source term in safety assessments, it is expensive and in short supply. Tritium can be continuously retained in a tokamak by codeposition with eroded carbon or beryllium and JET and TFTR with carbon plasma facing components showed a tritium retention level that would be unacceptable in ITER or future fusion reactors. Asdex-U and Alcator C-mod have shown reduced hydrogenic retention with tungsten clad and molybdenum plasma facing components. Once the tritium inventory approaches the administrative limit, tritium must be removed to permit continued D-T plasma operations. Several candidate techniques are being considered and need to be proven at a relevant speed and efficiency in contemporary tokamaks. Projections for ITER are discussed.
Magnetohydrodynamic stability of tokamak edge plasmas
Connor, J.W.; Hastie, R.J.; Wilson, H.R.; Miller, R.L.
1998-07-01
A new formalism for analyzing the magnetohydrodynamic stability of a limiter tokamak edge plasma is developed. Two radially localized, high toroidal mode number n instabilities are studied in detail: a peeling mode and an edge ballooning mode. The peeling mode, driven by edge current density and stabilized by edge pressure gradient, has features which are consistent with several properties of tokamak behavior in the high confinement {open_quotes}H{close_quotes}-mode of operation, and edge localized modes (or ELMs) in particular. The edge ballooning mode, driven by the pressure gradient, is identified; this penetrates {approximately}n{sup 1/3} rational surfaces into the plasma (rather than {approximately}n{sup 1/2}, expected from conventional ballooning mode theory). Furthermore, there exists a coupling between these two modes and this coupling provides a picture of the ELM cycle.
X-ray spectroscopy on tokamaks
von Goeler, S.; Bitter, M.; Cohen, S.
1982-01-01
During the last decade, the x-ray spectroscopy of high temperature plasmas has witnessed a rapid development. Most of the impulses have come from astrophysics, in particular, from the research on solar flares. On the other hand, the attainment of well-diagnosed, high-temperature laboratory plasmas in laser-pellet implosions and in tokamaks, has precipitated a fertile exchange between theory and experiment. Agreement and very detailed understanding has been reached for a great number of spectra with the result, that x-ray spectroscopy represents today a powerful and reliable new plasma diagnostic with important applications for fusion plasmas as well as solar flares. This paper is a short review of the experimental results from tokamaks.
Snowflake divertor configuration studies in National Spherical Torus Experiment
Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R.; and others
2012-08-15
Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width {lambda}{sub q} was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m{sup 2} (from 4-7 MW/m{sup 2} in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P{sub rad} in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.
Self-Organized Stationary States of Tokamaks
Jardin, S. C.; Ferraro, N.; Krebs, I.
2015-11-01
We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."
Heating the Compact Ignition Tokamak (CIT)
Ignat, D.W.
1989-11-01
The proposed CIT starts operation in the late 1990's with 20 MW of rf heating power. The tokamak and facility are to be designed to accommodate 50 MW auxiliary heating. The heating methods new being considered are ion cyclotron heating (ICH) and electron cyclotron heating (ECH). Aspects of these systems are described, and the choice of power level and type is discussed. 18 refs.
Self-Organized Stationary States of Tokamaks.
Jardin, S C; Ferraro, N; Krebs, I
2015-11-20
We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."
High beta plasmas in the PBX tokamak
Bol, K.; Buchenauer, D.; Chance, M.; Couture, P.; Fishman, H.; Fonck, R.; Gammel, G.; Grek, B.; Ida, K.; Itami, K.
1986-04-01
Bean-shaped configurations favorable for high ..beta.. discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present ..beta.. limit.
Instrumentation and controls of an ignited Tokamak
NASA Astrophysics Data System (ADS)
Becraft, W. R.; Golzy, J.; Houlberg, W. A.; Kukielka, C. A.; Onega, R. J.; Raju, G. V. S.; Stone, R. S.
1980-10-01
The instrumentation and controls of an ignited plasma magnetically confined in a Tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. Effort in these and other areas are reported. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton ratio in the plasma, synchrotron radiation, and divertor control.
Tokamak with liquid metal toroidal field coil
Ohkawa, Tihiro; Schaffer, Michael J.
1981-01-01
Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof.
Mathematical modeling plasma transport in tokamaks
NASA Astrophysics Data System (ADS)
Qiang, Ji
1998-11-01
In this work, we have applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in the next generation machine, ITER. The ignition probability of ITER for engineering design activity (EDA) parameters can be formally as high as 99.9% in the present context. The same probability for conceptual design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%. This suggests that EDA parameters for ITER tokamak are very likely to achieve the self- sustained thermonuclear reaction, but CDA parameters are risky for the realization of ignition.
Neutral-beam current drive in tokamaks
Devoto, R.S.
1986-01-01
The theory of neutral-beam current drive in tokamaks is reviewed. Experiments are discussed where neutral beams have been used to drive current directly and also indirectly through neoclassical effects. Application of the theory to an experimental test reactor is described. It is shown that neutral beams formed from negative ions accelerated to 500 to 700 keV are needed for this device.
Confinement scaling and ignition in tokamaks
Perkins, F.W.; Sun, Y.C.
1985-10-01
A drift wave turbulence model is used to compute the scaling and magnitude of central electron temperature and confinement time of tokamak plasmas. The results are in accord with experiment. Application to ignition experiments shows that high density (1 to 2) . 10/sup 15/ cm/sup -3/, high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition.
Plasma filamentation in the Rijnhuizen tokamak RTP
Lopes Cardozo, N.J.; Schueller, F.C.; Barth, C.J.; Chu, C.C.; Pijper, F.J.; Lok, J.; Oomens, A.A.M. )
1994-07-11
Evidence for small scale magnetic structures in the Rijnhuizen tokamak RTP is presented. These are manifest through steps and peaks in the electron temperature and pressure, measured with multiposition Thomson scattering. During central electron cyclotron heating, several filaments of high pressure are found in the power deposition region. They live hundreds of microseconds. Near the sawtooth inversion radius a step'' in the temperature profile occurs. Further out, quasiperiodic structures are observed, in both Ohmic and heated discharges.
Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)
NASA Astrophysics Data System (ADS)
Azizov, E. A.
2012-02-01
The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined.
Development of a free-boundary tokamak equilibrium solver for advanced study of tokamak equilibria
NASA Astrophysics Data System (ADS)
Jeon, Young Mu
2015-09-01
A free-boundary Tokamak equilibrium solver (TES), developed for advanced study of tokamak equilibra, is described with two distinctive features. One is a generalized method to resolve the intrinsic axisymmetric instability, which is encountered in all equilibrium calculations with a freeboundary condition. The other is an extension to deal with a new divertor geometry such as snowflake or X divertors. For validations, the uniqueness of a solution is confirmed by the independence of variations in the computational domain, the mathematical correctness and accuracy of equilibrium profiles are checked by using a direct comparison with an analytic equilibrium known as a generalized Solov'ev equilibrium, and the governing force balance relation is tested by examining the intrinsic axisymmetric instabilities. As an application of an advanced equilibrium study, a snow-flake divertor configuration that requires a second-order zero of the poloidal magnetic flux is discussed in the circumstance of the Korea superconducting tokamak advanced research (KSTAR) coil system.
Vdovin, V.
2014-02-12
The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].
Spherical demons: fast surface registration.
Yeo, B T Thomas; Sabuncu, Mert; Vercauteren, Tom; Ayache, Nicholas; Fischl, Bruce; Golland, Polina
2008-01-01
We present the fast Spherical Demons algorithm for registering two spherical images. By exploiting spherical vector spline interpolation theory, we show that a large class of regularizers for the modified demons objective function can be efficiently implemented on the sphere using convolution. Based on the one parameter subgroups of diffeomorphisms, the resulting registration is diffeomorphic and fast - registration of two cortical mesh models with more than 100k nodes takes less than 5 minutes, comparable to the fastest surface registration algorithms. Moreover, the accuracy of our method compares favorably to the popular FreeSurfer registration algorithm. We validate the technique in two different settings: (1) parcellation in a set of in-vivo cortical surfaces and (2) Brodmann area localization in ex-vivo cortical surfaces.
Spherical Demons: Fast Surface Registration
Yeo, B.T. Thomas; Sabuncu, Mert; Vercauteren, Tom; Ayache, Nicholas; Fischl, Bruce; Golland, Polina
2009-01-01
We present the fast Spherical Demons algorithm for registering two spherical images. By exploiting spherical vector spline interpolation theory, we show that a large class of regularizers for the modified demons objective function can be efficiently implemented on the sphere using convolution. Based on the one parameter subgroups of diffeomorphisms, the resulting registration is diffeomorphic and fast – registration of two cortical mesh models with more than 100k nodes takes less than 5 minutes, comparable to the fastest surface registration algorithms. Moreover, the accuracy of our method compares favorably to the popular FreeSurfer registration algorithm. We validate the technique in two different settings: (1) parcellation in a set of in-vivo cortical surfaces and (2) Brodmann area localization in ex-vivo cortical surfaces. PMID:18979813
Nondiffusive plasma transport at tokamak edge
NASA Astrophysics Data System (ADS)
Krasheninnikov, S. I.
2000-10-01
Recent findings show that cross field edge plasma transport at tokamak edge does not necessarily obey a simple diffusive law [1], the only type of a transport model applied so far in the macroscopic modeling of edge plasma transport. Cross field edge transport is more likely due to plasma filamentation with a ballistic motion of the filaments towards the first wall. Moreover, it so fast that plasma recycles on the main chamber first wall rather than to flow into divertor as conventional picture of edge plasma fluxes suggests. Crudely speaking particle recycling wise diverted tokamak operates in a limiter regime due to fast anomalous non-diffusive cross field plasma transport. Obviously that this newly found feature of edge plasma anomalous transport can significantly alter a design of any future reactor relevant tokamaks. Here we present a simple model describing the motion of the filaments in the scrape off layer and discuss it implications for experimental observations. [1] M. Umansky, S. I. Krasheninnikov, B. LaBombard, B. Lipschultz, and J. L. Terry, Phys. Plasmas 6 (1999) 2791; M. Umansky, S. I. Krasheninnikov, B. LaBombard and J. L. Terry, Phys. Plasmas 5 (1998) 3373.
ADX - Advanced Divertor and RF Tokamak Experiment
NASA Astrophysics Data System (ADS)
Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl
2015-11-01
The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.
ECH on the MTX (Microwave Tokamak Experiment)
Stallard, B.W.; Byers, J.A.; Hooper, E.B.; Makowski, M.A.; Meassick, S.; Rice, B.W.; Rognlien, T.D.; Verboncoeur, J.
1989-04-01
The Microwave Tokamak Experiment (MTX) at LLNL is investigating the heating of high density Tokamak plasmas using an intense pulse FEL. Our first experiments, now beginning, will study the absorption and plasma heating of single FEL pulses (20 ns pulse length and peak power up to 2 GW) at a frequency of 140 GHz. A later phase of experiments also at 140 GHz will study FEL heating at 5 kHz rate for a pulse train up to 50 pulses (35 ns pulse length and peak power up to 4 GW). Future operations are planned at 250 GHz with an average power of 2 MW for a pulse train of 0.5 s. The microwave output of the FEL is transported quasi-optically to the tokamak through a window-less, evacuated pipe of 20 in. diameter, using a six mirror system. Computational modelling of the non-linear absorption for the MTX geometry predicts single-pass absorption of 40% at a density and temperature of 1.8 /times/ 10/sup 20/m/sup /minus/3/ and 1 keV, respectively. To measure plasma microwave absorption and backscatter, diagnostics are available to measure forward and reflected power (parallel wire grid beam-splitter and mirror directional couplers) and power transmitted through the plasma (segmented calorimeter and waveguide detector). Other fast diagnostics include ECE, Thompson scattering, soft x-rays, and fast magnetic probes. 8 refs., 2 figs.
Remote feedback stabilization of tokamak instabilities
Sen, A.K. )
1994-05-01
A novel remote suppressor consisting of an injected ion beam has been used for the stabilization of plasma instabilities. A collisionless curvature-driven trapped-particle instability, an [bold E][times][bold B] flute mode and an ion temperature gradient (ITG) instability have been successfully suppressed down to noise levels using this scheme. Furthermore, the first experimental demonstration of a multimode feedback stabilization with a single sensor--suppressor pair has been achieved. Two modes (an [bold E][times][bold B] flute and an ITG mode) were simultaneously stabilized with a simple state-feedback-type method where more state'' information was generated from a single-sensor Langmuir probe by appropriate signal processing. The above experiments may be considered as paradigms for controlling several important tokamak instabilities. First, feedback suppression of edge fluctuations in a tokamak with a suitable form of insulated segmented poloidal limiter sections used as Langmuir-probe-like suppressors is proposed. Other feedback control schemes are proposed for the suppression of electrostatic core fluctuations via appropriately phased ion density input from a modulated neutral beam. Most importantly, a scheme to control major disruptions in tokamaks via feedback suppression of kink (and possibly) tearing modes is discussed. This may be accomplished by using a modulated neutral beam suppressor in a feedback loop, which will supply a momentum input of appropriate phase and amplitude. Simple theoretical models predict modest levels of beam energy, current, and power.
Forced Magnetic Reconnection In A Tokamak Plasma
NASA Astrophysics Data System (ADS)
Callen, J. D.; Hegna, C. C.
2015-11-01
The theory of forced magnetic field reconnection induced by an externally imposed resonant magnetic perturbation usually uses a sheared slab or cylindrical magnetic field model and often focuses on the potential time-asymptotic induced magnetic island state. However, tokamak plasmas have significant magnetic geometry and dynamical plasma toroidal rotation screening effects. Also, finite ion Larmor radius (FLR) and banana width (FBW) effects can damp and thus limit the width of a nascent magnetic island. A theory that is more applicable for tokamak plasmas is being developed. This new model of the dynamics of forced magnetic reconnection considers a single helicity magnetic perturbation in the tokamak magnetic field geometry, uses a kinetically-derived collisional parallel electron flow response, and employs a comprehensive dynamical equation for the plasma toroidal rotation frequency. It is being used to explore the dynamics of bifurcation into a magnetically reconnected state in the thin singular layer around the rational surface, evolution into a generalized Rutherford regime where the island width exceeds the singular layer width, and assess the island width limiting effects of FLR and FBW polarization currents. Support by DoE grants DE-FG02-86ER53218, DE-FG02-92ER54139.
Edge-localized-modes in tokamaks
Leonard, A. W.
2014-09-15
Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively, rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heat flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. Encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.
Edge-localized-modes in tokamaks
Leonard, Anthony W.
2014-09-11
Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heatmore » flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. As a result, encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.« less
Edge-localized-modes in tokamaks
Leonard, Anthony W.
2014-09-11
Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heat flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. As a result, encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.
Interactive, multiobjective Bayesian optimization of tokamak scenarios
NASA Astrophysics Data System (ADS)
Urban, Jakub; Artaud, Jean-François
2016-10-01
Bayesian optimization is applied to tokamak scenario optimizations. The key advantages are 1) a reduced number of objective function evaluations, 2) no need for derivatives, and 3) the possibility to include a prior knowledge. This is of a great value for optimizing tokamak scenarios, where several (competing) objectives with often unknown magnitudes exist and the number of parameters is large (>10). The first two properties imply that Bayesian optimization is well suited for heavy, complex objective functions. Reusing previous iterations as priors for next optimization steps effectively enables interactive, multiobjective optimizations, regardless of whether a human decision maker is included or not. We show that these features make Bayesian optimization an outstanding tool for optimizing tokamak scenarios. Objective functions and constraints, targeting, e.g., fusion gain, flux consumption, coils currents limits or q-profile, can be assembled interactively. The optimized parameter vector may include actuators like plasma current or heating waveforms. We demonstrate the capabilities on optimizing ITER and DEMO-like scenarios, simulated by the METIS code.
Tokamak Physics Experiment diagnostic plans (invited)
NASA Astrophysics Data System (ADS)
Medley, S. S.
1995-01-01
A superconducting Tokamak Physics Experiment (TPX) whose mission is to develop the scientific basis for a compact and continuously operating tokamak fusion reactor is being designed by an integrated U.S. national team. Key physics features such as strong shaping, a double-null poloidal divertor, full noninductive current drive, and current profile control capability will be used to explore improvements in energy confinement and beta limit scaling in high-aspect-ratio plasmas with a high bootstrap current fraction. Steady-state operation of TPX permits these studies to be extended to time scales significantly exceeding the global current-relaxation time and the plasma-wall equilibrium time. The diagnostic requirements are determined by the TPX mission and supporting objectives, such as optimization of plasma performance through active control of the current profile and of the plasma-wall interactions. Diagnostic measurements are needed to characterize the plasma behavior over the full range of conventional tokamak plasma parameters with appropriate spatial and temporal resolution as well as for control and monitoring of aspects of the machine operation such as the plasma position and shape, plasma current, vacuum vessel currents, electron density and temperature, and the divertor and limiter temperatures. In addition, several diagnostic capabilities that are especially critical for the TPX project will be discussed.
Basketballs as spherical acoustic cavities
NASA Astrophysics Data System (ADS)
Russell, Daniel A.
2010-06-01
The sound field resulting from striking a basketball is found to be rich in frequency content, with over 50 partials in the frequency range of 0-12 kHz. The frequencies are found to closely match theoretical expectations for standing wave patterns inside a spherical cavity. Because of the degenerate nature of the mode shapes, explicit identification of the modes is not possible without internal investigation with a microphone probe. A basketball proves to be an interesting application of a boundary value problem involving spherical coordinates.
Radiance calibration of spherical integrators
NASA Technical Reports Server (NTRS)
Mclean, James T.; Guenther, Bruce W.
1989-01-01
Techniques for improving the knowledge of the radiance of large area spherical and hemispherical integrating energy sources have been investigated. Such sources are used to calibrate numerous aircraft and spacecraft remote sensing instruments. Comparisons are made between using a standard source based calibration method and a quantum efficient detector (QED) based calibration method. The uncertainty involved in transferring the calibrated values of the point source standard lamp to the extended source is estimated to be 5 to 10 percent. The use of the QED allows an improvement in the uncertainty to 1 to 2 percent for the measurement of absolute radiance from a spherical integrator source.
NASA Astrophysics Data System (ADS)
Nieto, M.; Allain, J. P.; Hassanein, A.; Titov, V.; Hendricks, M.; Gray, T.; Kaita, R.; Kugel, H.; Majeski, R.; Mansfield, D.; Spaleta, J.; Timberlake, J.
2006-12-01
The role of lithium on the modification of recycling regimes in fusion reactors has renewed interest of previous lithium supershot experiments carried out in TFTR. There is a need to understand the interaction between edge plasmas and lithiated plasma-facing components (PFCs), which have the potential of enabling fusion reactors to operate at low-recycling regimes. The Interaction of Materials with Particles and Components Testing (IMPACT) facility at Argonne National Laboratory is currently collaborating with Princeton Plasma Physics Laboratory (PPPL) to conduct lithiated surface studies for the National Spherical Tokamak Experiment (NSTX) and the Current Drive eXperiment — Upgrade (CDX-U). IMPACT has the necessary tools to perform experiments that diagnose the surface dynamics of lithium thin films on metallic and non-metallic substrates, and can be monitored with multiple in-situ techniques (LEISS, AES, QMS and XPS) capturing real-time surface dynamics. Therefore, these techniques are available during He+ and D+ irradiation. Surface sputtering measurements can be performed using a quartz crystal microbalance — dual crystal unit (QCM-DCU) with very high sensitivity. Initial results suggest that lithium intercalation into graphite occurs quite rapidly and only a fraction lithium can be kept on the surface. On metallic substrates this intercalation is absent. Additional results of Li/metal systems show lithium surface self-healing with temperature. It was also found that the presence of lithium seems to inhibit hydrocarbon formation during D+ bombardment of graphite. Experiments in CDX-U have tested the effect of both solid and liquid lithium PFCs on tokamak plasmas, and significant changes in tokamak operation are observed. These include a strong reduction in both recycling and impurity levels in the gas phase, lowered loop voltage during ohmic operation, and an increased electron temperature at the edge.
Burke, Marcus G.; Barr, Jayson L.; Bongard, Michael W.; ...
2017-05-16
Plasmas in the Pegasus spherical tokamak are initiated and grown by the non-solenoidal local helicity injection (LHI) current drive technique. The LHI system consists of three adjacent electron current sources that inject multiple helical current filaments that can reconnect with each other. Anomalously high impurity ion temperatures are observed during LHI with Ti,OV ≤ 650 eV, which is in contrast to Ti,OV ≤ 70 eV from Ohmic heating alone. Spatial profiles of Ti,OV indicate an edge localized heating source, with Ti,OV ~ 650 eV near the outboard major radius of the injectors and dropping to ~150 eV near the plasma magnetic axis. Experiments without a background tokamak plasma indicate the ion heating results from magnetic reconnection between adjacent injected current filaments. In these experiments, the HeII T i perpendicular to the magnetic field is found to scale with the reconnecting field strength, local density, and guide field, whilemore » $${{T}_{\\text{i},\\parallel}}$$ experiences little change, in agreement with two-fluid reconnection theory. In conclusion, this ion heating is not expected to significantly impact the LHI plasma performance in Pegasus, as it does not contribute significantly to the electron heating. However, estimates of the power transfer to the bulk ion are quite large, and thus LHI current drive provides an auxiliary ion heating mechanism to the tokamak plasma.« less
NASA Astrophysics Data System (ADS)
Burke, M. G.; Barr, J. L.; Bongard, M. W.; Fonck, R. J.; Hinson, E. T.; Perry, J. M.; Reusch, J. A.; Schlossberg, D. J.
2017-07-01
Plasmas in the Pegasus spherical tokamak are initiated and grown by the non-solenoidal local helicity injection (LHI) current drive technique. The LHI system consists of three adjacent electron current sources that inject multiple helical current filaments that can reconnect with each other. Anomalously high impurity ion temperatures are observed during LHI with T i,OV ⩽ 650 eV, which is in contrast to T i,OV ⩽ 70 eV from Ohmic heating alone. Spatial profiles of T i,OV indicate an edge localized heating source, with T i,OV ~ 650 eV near the outboard major radius of the injectors and dropping to ~150 eV near the plasma magnetic axis. Experiments without a background tokamak plasma indicate the ion heating results from magnetic reconnection between adjacent injected current filaments. In these experiments, the HeII T i perpendicular to the magnetic field is found to scale with the reconnecting field strength, local density, and guide field, while {{T}\\text{i,\\parallel}} experiences little change, in agreement with two-fluid reconnection theory. This ion heating is not expected to significantly impact the LHI plasma performance in Pegasus, as it does not contribute significantly to the electron heating. However, estimates of the power transfer to the bulk ion are quite large, and thus LHI current drive provides an auxiliary ion heating mechanism to the tokamak plasma.
Importance of Plasma Response to Non-axisymmetric Perturbations in Tokamaks
Jong-kyu Park, Allen H. Boozer, Jonathan E. Menard, Andrea M. Garofalo, Michael J. Schaffer, Richard J. Hawryluk, Stanley M. Kaye, Stefan P. Gerhardt, Steve A. Sabbagh, and the NSTX Team
2009-04-22
Tokamaks are sensitive to deviations from axisymmetry as small as δB=B0 ~ 10-4. These non-axisymmetric perturbations greatly modify plasma confinement and performance by either destroying magnetic surfaces with subsequent locking or deforming magnetic surfaces with associated non-ambipolar transport. The Ideal Perturbed Equilibrium Code (IPEC) calculates ideal perturbed equilibria and provides important basis for understanding the sensitivity of tokamak plasmas to perturbations. IPEC calculations indicate that the ideal plasma response, or equiva- lently the effect by ideally perturbed plasma currents, is essential to explain locking experiments on National Spherical Torus eXperiment (NSTX) and DIII-D. The ideal plasma response is also important for Neoclassical Toroidal Viscosity (NTV) in non-ambipolar transport. The consistency between NTV theory and magnetic braking experiments on NSTX and DIII-D can be improved when the variation in the field strength in IPEC is coupled with generalized NTV theory. These plasma response effects will be compared with the previous vacuum superpositions to illustrate the importance. However, plasma response based on ideal perturbed equilibria is still not suffciently accurate to predict the details of NTV transport, and can be inconsistent when currents associated with a toroidal torque become comparable to ideal perturbed currents.
Ion cyclotron and spin-flip emissions from fusion products in tokamaks
Arunasalam, V.; Greene, G.J.; Young, K.M.
1993-02-01
Power emission by fusion products of tokamak plasmas in their ion cyclotron range of frequencies (ICRF) and at their spin-flip resonance frequency is calculated for some specific model fusion product velocity-space distribution functions. The background plasma of say deuterium (D) is assumed to be in equilibrium with a Maxwellian distribution both for the electrons and ions. The fusion product velocity distributions analyzed here are: (1) A monoenergetic velocity space ring distribution. (2) A monoenergetic velocity space spherical shell distribution. (3) An anisotropic Maxwellian distribution with T [perpendicular] [ne] T[parallel]and with appreciable drift velocity along the confining magnetic field. Single dressed'' test particle spontaneous emission calculations are presented first and the radiation temperature for ion cyclotron emission (ICE) is analyzed both for black-body emission and nonequilibrium conditions. Thresholds for instability and overstability conditions are then examined and quasilinear and nonlinear theories of the electromagnetic ion cyclotron modes are discussed. Distinctions between kinetic or causal instabilities'' and hydrodynamic instabilities'' are drawn and some numerical estimates are presented for typical tokamak parameters. Semiquantitative remarks are offered on wave accessibility, mode conversion, and parametric decay instabilities as possible for spatially localized ICE. Calculations are carried out both for k[parallel] = 0 for k[parallel] [ne] 0. The effects of the temperature anisotropy and large drift velocities in the parallel direction are also examined. Finally, proton spin-flip resonance emission and absorption calculations are also presented both for thermal equilibrium conditions and for an inverted'' population of states.
Ion cyclotron and spin-flip emissions from fusion products in tokamaks
Arunasalam, V.; Greene, G.J.; Young, K.M.
1993-02-01
Power emission by fusion products of tokamak plasmas in their ion cyclotron range of frequencies (ICRF) and at their spin-flip resonance frequency is calculated for some specific model fusion product velocity-space distribution functions. The background plasma of say deuterium (D) is assumed to be in equilibrium with a Maxwellian distribution both for the electrons and ions. The fusion product velocity distributions analyzed here are: (1) A monoenergetic velocity space ring distribution. (2) A monoenergetic velocity space spherical shell distribution. (3) An anisotropic Maxwellian distribution with T {perpendicular} {ne} T{parallel}and with appreciable drift velocity along the confining magnetic field. Single ``dressed`` test particle spontaneous emission calculations are presented first and the radiation temperature for ion cyclotron emission (ICE) is analyzed both for black-body emission and nonequilibrium conditions. Thresholds for instability and overstability conditions are then examined and quasilinear and nonlinear theories of the electromagnetic ion cyclotron modes are discussed. Distinctions between ``kinetic or causal instabilities`` and ``hydrodynamic instabilities`` are drawn and some numerical estimates are presented for typical tokamak parameters. Semiquantitative remarks are offered on wave accessibility, mode conversion, and parametric decay instabilities as possible for spatially localized ICE. Calculations are carried out both for k{parallel} = 0 for k{parallel} {ne} 0. The effects of the temperature anisotropy and large drift velocities in the parallel direction are also examined. Finally, proton spin-flip resonance emission and absorption calculations are also presented both for thermal equilibrium conditions and for an ``inverted`` population of states.
Berzak, L; Jones, A D; Kaita, R; Kozub, T; Logan, N; Majeski, R; Menard, J; Zakharov, L
2010-10-01
The lithium tokamak experiment (LTX) is a modest-sized spherical tokamak (R(0)=0.4 m and a=0.26 m) designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 °C. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.
Drift-kinetic simulation of neoclassical transport with impurities in tokamaks
Kolesnikov, R. A.; Wang, W. X.; Rewoldt, G.; Tang, W. M.; Hinton, F. L.
2010-02-15
Plasmas in modern tokamak experiments contain a significant fraction of impurity ions in addition to the main deuterium background ions. A new multiple ion-species deltaf particle simulation capability has been developed to self-consistently study the nonlocal effects of impurities on neoclassical transport in toroidal plasmas. A new algorithm for an unlike-particle collision operator, including test-particle and conserving field-particle parts, is described. Effects of the carbon impurity on the main deuterium species heat flux as well as an ambipolar radial electric field in a National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, Y.-K. M. Peng et al., Nucl. Fusion 40, 557 (2000)] configuration were studied. A difference between carbon poloidal rotation found from simulation and from conventional theoretical estimates has been investigated and was identified to be a nonlocal finite orbit effect. In the case of large-aspect ratio tokamak configurations with steep toroidal flow profiles, we propose a theoretical model to describe this nonlocal effect. The dominant mechanisms captured by the model are associated with ion parallel velocity modification due to steep toroidal flow and radial electric field profiles. We present simulation results for carbon poloidal velocity in NSTX. Comparisons with neoclassical theory are discussed.
Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment
NASA Astrophysics Data System (ADS)
Boyle, D. P.; Bell, R. E.; Kaita, R.; Lucia, M.; Schmitt, J. C.; Scotti, F.; Kubota, S.; Hansen, C.; Biewer, T. M.; Gray, T. K.
2016-10-01
The Lithium Tokamak Experiment (LTX) is a modest-sized spherical tokamak with all-metal plasma facing components (PFCs), uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma. This work presents measurements of core plasma impurity concentrations and transport in LTX. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with 2 - 4 % Li, 0.6 - 2 % C, 0.4 - 0.7 % O, and Zeff < 1.2 . Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, and neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two. However, time-independent simulations with MIST indicated that neoclassical theory did not fully capture the impurity transport and anomalous transport likely played a significant role in determining impurity profiles. Progress on additional analysis, including time-dependent impurity transport simulations and impurity measurements with liquid lithium coatings, and plans for diagnostic upgrades and future experiments in LTX- β will also be presented. This work supported by US DOE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.
Buckling Analysis of TF Coil Inner Leg for Central Solenoidless Tokamak
NASA Astrophysics Data System (ADS)
Song, Yuntao; Yang, Qingxi; Nishio, Satoshi
2006-07-01
The central post is one of the critical components for the low aspect ratio tokamak, which endures not only a tremendous ohmic heating because it carries a rather high current, but also a large neutron heating and irradiation owing to the plasma operation. The DS copper alloy Glidcop AL-25[8] was chosen as the conductor material for its adequate mechanical properties and physics properties. The central post has a cylindrical structure with lots of cooling channels. The length of the central post for the next generation of nuclear fusion spherical tokamaks will be more than 10 m or 20 m. The structural stability is very crucial. When the applied load is larger than the structure critical buckling load, the device will lose its stability and collapse. In order to calculate the critical buckling load, a 1/6-segment finite element model was used and the force acting on the central post was simulated. The results showed that the vertical compressive stresses mainly affect the stability of the central post. The linear buckling analysis results with finite element method based on small deformation theory were given in this paper. The relation curves and functions for buckling factor, depending on the different lengths and the radius of the central post, the diameter of cooling channel and the maximum allowable current density, were also shown.
Kaita, R.; Kozub, T.; Logan, N.; Majeski, R.; Menard, J.; Zakharov, L.
2010-12-10
The lithium tokamak experiment LTX is a modest-sized spherical tokamak R0=0.4 m and a =0.26 m designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 oC. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.
Modeling the effect of toroidal plasma rotation on drift-magnetohydrodynamic modes in tokamaks
NASA Astrophysics Data System (ADS)
Chapman, I. T.; Sharapov, S. E.; Huysmans, G. T. A.; Mikhailovskii, A. B.
2006-06-01
A new code, MISHKA-F (Flow), has been developed as an extension of the ideal magneto-hydrodynamic (MHD) code MISHKA-1 [Mikhailovskii et al., Plasma Phys. Rep. 23, 844 (1997)] in order to investigate the linear MHD stability of ideal and resistive eigenmodes with respect to the effects of toroidal rotation in tokamaks in general toroidal geometry with the ion diamagnetic drift effect taken into account. Benchmark test results of the MISHKA-F code show good agreement with analytic theory [A. B. Mikhailovskii and S. E. Sharapov, Plasma Phys. Controlled Fusion 42, 57 (2000)] for the stability limits of the ideal n /m=1/1 internal kink mode. The combined stabilizing effects of the ion diamagnetic drift frequency, ω*i, and the toroidal flow shear are also studied. The ω*i stabilization of the internal kink mode is found to be more effective at finite flow shear. Finite-n ballooning modes are studied in plasmas with the toroidal flow shear effect included. The stabilization of the ballooning modes by toroidal rotation is found to agree well with earlier predictions [Webster et al., Phys. Plasmas 11, 2135 (2004)]. The effect of high flow shear is analyzed for a sawtoothing discharge typical in the Mega Ampère Spherical Tokamak (MAST) [Sykes et al., Nucl. Fusion 41, 1423 (2001)]. It is found that the ideal n =1 internal kink mode can be stabilized by toroidal rotation at values observed experimentally.
Modeling the effect of toroidal plasma rotation on drift-magnetohydrodynamic modes in tokamaks
Chapman, I.T.; Sharapov, S.E.; Huysmans, G.T.A.; Mikhailovskii, A. B.
2006-06-15
A new code, MISHKA-F (Flow), has been developed as an extension of the ideal magneto-hydrodynamic (MHD) code MISHKA-1 [Mikhailovskii et al., Plasma Phys. Rep. 23, 844 (1997)] in order to investigate the linear MHD stability of ideal and resistive eigenmodes with respect to the effects of toroidal rotation in tokamaks in general toroidal geometry with the ion diamagnetic drift effect taken into account. Benchmark test results of the MISHKA-F code show good agreement with analytic theory [A. B. Mikhailovskii and S. E. Sharapov, Plasma Phys. Controlled Fusion 42, 57 (2000)] for the stability limits of the ideal n/m=1/1 internal kink mode. The combined stabilizing effects of the ion diamagnetic drift frequency, {omega}{sub *i}, and the toroidal flow shear are also studied. The {omega}{sub *i} stabilization of the internal kink mode is found to be more effective at finite flow shear. Finite-n ballooning modes are studied in plasmas with the toroidal flow shear effect included. The stabilization of the ballooning modes by toroidal rotation is found to agree well with earlier predictions [Webster et al., Phys. Plasmas 11, 2135 (2004)]. The effect of high flow shear is analyzed for a sawtoothing discharge typical in the Mega Ampere Spherical Tokamak (MAST) [Sykes et al., Nucl. Fusion 41, 1423 (2001)]. It is found that the ideal n=1 internal kink mode can be stabilized by toroidal rotation at values observed experimentally.
Neoclassical transport coefficients for finite-aspect-ratio and bean-shaped tokamak plasmas
NASA Astrophysics Data System (ADS)
Crume, E. C., Jr.; Beasley, C. O., Jr.; Hirshman, S. P.; van Rij, W. I.
1987-04-01
Numerically calculated tokamak equilibria are used to compute banana-plateau transport coefficients for finite-aspect-ratio, finite-beta plasmas. Calculations are presented for the Spherical Torus Experiment (STX) (NTIS Document No. DE 86004663) and the Princeton Beta Experiment (PBX) (NTIS Document No. DE 86011173). In STX, the poloidal variation of B≡‖B‖ over a magnetic surface tends to be reduced in regions of large major radius R. The reduction of radial transport caused by this quasiomnigeneous condition is offset by increased drifts and trapping probabilities for smaller R. Thus the modulation Δ=(Bmax-Bmin)/(Bmax+Bmin) on a magnetic surface becomes the critical parameter determining neoclassical transport. In PBX, the bean-shaped topology of the magnetic surfaces leads to the presence of multiple magnetic wells. Numerical calculations confirm that analytic calculations of neoclassical transport based on the total fraction of circulating particles are valid even when geometrically distinct classes of trapped particles are present.
Aspect Ratio Scaling of Ideal No-wall Stability Limits in High Bootstrap Fraction Tokamak Plasmas
J.E. Menard; M.G. Bell; R.E. Bell; D.A. Gates; S.M. Kaye; B.P. LeBlanc; R. Maingi; S.A. Sabbagh; V. Soukhanovskii; D. Stutman; the NSTX National Research Team
2003-11-25
Recent experiments in the low aspect ratio National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40 (2000) 557] have achieved normalized beta values twice the conventional tokamak limit at low internal inductance and with significant bootstrap current. These experimental results have motivated a computational re-examination of the plasma aspect ratio dependence of ideal no-wall magnetohydrodynamic stability limits. These calculations find that the profile-optimized no-wall stability limit in high bootstrap fraction regimes is well described by a nearly aspect ratio invariant normalized beta parameter utilizing the total magnetic field energy density inside the plasma. However, the scaling of normalized beta with internal inductance is found to be strongly aspect ratio dependent at sufficiently low aspect ratio. These calculations and detailed stability analyses of experimental equilibria indicate that the nonrotating plasma no-wall stability limit has been exceeded by as much as 30% in NSTX in a high bootstrap fraction regime.
Shielding of External Magnetic Perturbations By Torque In Rotating Tokamak Plasmas
Park, Jong-Kyu; Boozer, Allen H.; Menard, Jonathan E.; Gerhardt, Stefan P.; Sabbagh, Steve A.
2009-08-24
The imposition of a nonaxisymmetric magnetic perturbation on a rotating tokamak plasma requires energy and toroidal torque. Fundamental electrodynamics implies that the torque is essentially limited and must be consistent with the external response of a plasma equilibrium ƒ = j x B. Here magnetic measurements on National Spherical Torus eXperiment (NSTX) device are used to derive the energy and the torque, and these empirical evaluations are compared with theoretical calculations based on perturbed scalar pressure equilibria ƒ = ∇p coupled with the theory of nonambipolar transport. The measurement and the theory are consistent within acceptable uncertainties, but can be largely inconsistent when the torque is comparable to the energy. This is expected since the currents associated with the torque are ignored in scalar pressure equilibria, but these currents tend to shield the perturbation.
Optical boundary reconstruction of tokamak plasmas for feedback control of plasma position and shape
NASA Astrophysics Data System (ADS)
Hommen, G.; de Baar, M.; Nuij, P.; McArdle, G.; Akers, R.; Steinbuch, M.
2010-11-01
A new diagnostic is developed to reconstruct the plasma boundary using visible wavelength images. Exploiting the plasma's edge localized and toroidally symmetric emission profile, a new coordinate transform is presented to reconstruct the plasma boundary from a poloidal view image. The plasma boundary reconstruction is implemented in MATLAB and applied to camera images of Mega-Ampere Spherical Tokamak discharges. The optically reconstructed plasma boundaries are compared to magnetic reconstructions from the offline reconstruction code EFIT, showing very good qualitative and quantitative agreement. Average errors are within 2 cm and correlation is high. In the current software implementation, plasma boundary reconstruction from a single image takes 3 ms. The applicability and system requirements of the new optical boundary reconstruction, called OFIT, for use in both feedback control of plasma position and shape and in offline reconstruction tools are discussed.
Hommen, G; de Baar, M; Nuij, P; McArdle, G; Akers, R; Steinbuch, M
2010-11-01
A new diagnostic is developed to reconstruct the plasma boundary using visible wavelength images. Exploiting the plasma's edge localized and toroidally symmetric emission profile, a new coordinate transform is presented to reconstruct the plasma boundary from a poloidal view image. The plasma boundary reconstruction is implemented in MATLAB and applied to camera images of Mega-Ampere Spherical Tokamak discharges. The optically reconstructed plasma boundaries are compared to magnetic reconstructions from the offline reconstruction code EFIT, showing very good qualitative and quantitative agreement. Average errors are within 2 cm and correlation is high. In the current software implementation, plasma boundary reconstruction from a single image takes 3 ms. The applicability and system requirements of the new optical boundary reconstruction, called OFIT, for use in both feedback control of plasma position and shape and in offline reconstruction tools are discussed.
Optical boundary reconstruction of tokamak plasmas for feedback control of plasma position and shape
Hommen, G.; Baar, M. de; Nuij, P.; Steinbuch, M.; McArdle, G.; Akers, R.
2010-11-15
A new diagnostic is developed to reconstruct the plasma boundary using visible wavelength images. Exploiting the plasma's edge localized and toroidally symmetric emission profile, a new coordinate transform is presented to reconstruct the plasma boundary from a poloidal view image. The plasma boundary reconstruction is implemented in MATLAB and applied to camera images of Mega-Ampere Spherical Tokamak discharges. The optically reconstructed plasma boundaries are compared to magnetic reconstructions from the offline reconstruction code EFIT, showing very good qualitative and quantitative agreement. Average errors are within 2 cm and correlation is high. In the current software implementation, plasma boundary reconstruction from a single image takes 3 ms. The applicability and system requirements of the new optical boundary reconstruction, called OFIT, for use in both feedback control of plasma position and shape and in offline reconstruction tools are discussed.
Euclidean, Spherical, and Hyperbolic Shadows
ERIC Educational Resources Information Center
Hoban, Ryan
2013-01-01
Many classical problems in elementary calculus use Euclidean geometry. This article takes such a problem and solves it in hyperbolic and in spherical geometry instead. The solution requires only the ability to compute distances and intersections of points in these geometries. The dramatically different results we obtain illustrate the effect…
Euclidean, Spherical, and Hyperbolic Shadows
ERIC Educational Resources Information Center
Hoban, Ryan
2013-01-01
Many classical problems in elementary calculus use Euclidean geometry. This article takes such a problem and solves it in hyperbolic and in spherical geometry instead. The solution requires only the ability to compute distances and intersections of points in these geometries. The dramatically different results we obtain illustrate the effect…
A Module in Spherical Trigonometry.
ERIC Educational Resources Information Center
Congleton, C. A.; Broome, L. E.
1980-01-01
This module, designed for use at the high school level as a four- to eight-hour topic, includes: the geometry of a sphere, the coordinate system used to describe points on the earth's surface, parallel and meridian sailing, and the solution of right spherical triangles. (Author/MK)
Spherical-Bearing Analysis Program
NASA Technical Reports Server (NTRS)
Kleckner, R. J.
1984-01-01
Computer program SPHERBEAN, developed to predict thermomechanical performance characteristics of double-row spherical roller bearings over wide range of operating conditions. Analysis allows six degrees of freedom for each roller and three for each half of an optionally split cage. Program capabilities provide sufficient generality to allow detailed simulation of both high-speed and conventional bearing operation.
Dispersion in Spherical Water Drops.
ERIC Educational Resources Information Center
Eliason, John C., Jr.
1989-01-01
Discusses a laboratory exercise simulating the paths of light rays through spherical water drops by applying principles of ray optics and geometry. Describes four parts: determining the output angles, computer simulation, explorations, model testing, and solutions. Provides a computer program and some diagrams. (YP)
Dispersion in Spherical Water Drops.
ERIC Educational Resources Information Center
Eliason, John C., Jr.
1989-01-01
Discusses a laboratory exercise simulating the paths of light rays through spherical water drops by applying principles of ray optics and geometry. Describes four parts: determining the output angles, computer simulation, explorations, model testing, and solutions. Provides a computer program and some diagrams. (YP)
A Module in Spherical Trigonometry.
ERIC Educational Resources Information Center
Congleton, C. A.; Broome, L. E.
1980-01-01
This module, designed for use at the high school level as a four- to eight-hour topic, includes: the geometry of a sphere, the coordinate system used to describe points on the earth's surface, parallel and meridian sailing, and the solution of right spherical triangles. (Author/MK)
A simulation study of a controlled tokamak plasma
NASA Astrophysics Data System (ADS)
Fujii, N.; Niwa, Y.
1980-03-01
A tokamak circuit theory, including results of numerical simulation studies, is applied to a control system synthesized for a Joule heated tokamak plasma. The treatment is similar to that of Ogata and Ninomiya (1979) except that in this case a quadrupole field coil current is considered coexisting with image induced on a vacuum chamber.
Advanced tokamak operating modes in TPX and ITER
Nevins, W.M.
1994-12-31
A program is described to develop the advanced tokamak physics required for an economic steady-state fusion reactor on existing (short-pulse) tokamak experiments; to extend these operating modes to long-pulse on TPX; and finally to demonstrate them in a long-pulse D-T plasma on ITER.
Numerical investigations of plasma parameters in the COMPASS tokamak
Havlickova, E.; Zagorski, R.; Panek, R.
2008-09-15
A numerical investigation of plasma parameters in a diverter configuration of COMPASS tokamak is presented. The plasma parameters in the device are analyzed in the frame of the self-consistent description of the central plasma and edge region. The possibility of achieving high recycling and detached regimes in the boundary layer of the COMPASS tokamak is discussed.
Fokker-Planck/Transport model for neutral beam driven tokamaks
Killeen, J.; Mirin, A.A.; McCoy, M.G.
1980-01-01
The application of nonlinear Fokker-Planck models to the study of beam-driven plasmas is briefly reviewed. This evolution of models has led to a Fokker-Planck/Transport (FPT) model for neutral-beam-driven Tokamaks, which is described in detail. The FPT code has been applied to the PLT, PDX, and TFTR Tokamaks, and some representative results are presented.
Physics design requirements for the Tokamak Physics Experiment (TPX)
Neilson, G.H.; Goldston, R.J.; Jardin, S.C.; Reiersen, W.T.; Nevins, W.M.; Porkolab, M.; Ulrickson, M.
1993-11-01
The design of TPX is driven by physics requirements that follow from its mission. The tokamak and heating systems provide the performance and profile controls needed to study advanced steady state tokamak operating modes. The magnetic control systems provide substantial flexibility for the study of regimes with high beta and bootstrap current. The divertor is designed for high steady state power and particle exhaust.
Advanced tokamak operating modes in TPX and ITER
NASA Astrophysics Data System (ADS)
Nevins, W. M.
1994-09-01
A program is described to develop the advanced tokamak physics required for an economic steady-state fusion reactor on existing (short-pulse) tokamak experiments; to extend these operating modes to long-pulse on TPX; and finally to demonstrate them in a long-pulse D-T plasma on ITER.
Progress and prospects in understanding the physics of tokamak experiments
Hutchinson, I.
1992-12-01
A whistle-stop tour of the diverse physics of tokamak plasma confinement. This talk will illustrate the way in which fusion research on tokamaks has led to important and interesting physics results, and discuss some of the scientific challenges still ahead before fusion`s potential can be established.
Recent progress on the Compact Ignition Tokamak (CIT)
Ignat, D.W.
1987-01-01
This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule.
Tokamak Physics Experiment (TPX) power supply design and development
Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.
1995-04-01
The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes.
Hybrid Fusion: The Only Viable Development Path for Tokamaks?
NASA Astrophysics Data System (ADS)
Manheimer, Wallace
2009-03-01
The world needs a great deal of carbon free energy, and soon, for civilization to continue. Fusion's goal is to develop such a carbon free energy source. For the last 4 decades, tokamaks have been the best magnetic fusion has to offer. But what if its development stops short of commercial fusion? This paper introduces `conservative design principles' for tokamaks. These are very simple, are reasonably based in theory, and have always constrained tokamak operation. Assuming they continue to do so, it is unlikely that tokamaks will ever make it as commercial reactors. This is independent of their confinement properties. However because of the large additional gain in hybrid fusion, tokamaks reactors look like they can make it as hybrid fuel producers, and provide large scale power by mid century or shortly thereafter.
US fusion effort hit by tokamak losses
NASA Astrophysics Data System (ADS)
Gwynne, Peter
2016-11-01
Stewart Prager, director of the Princeton Plasma Physics Laboratory (PPPL) in the US, resigned in late September just weeks after a major setback at the lab's National Spherical Torus Experiment Upgrade (NSTX-U).
Mathematical modeling plasma transport in tokamaks
Quiang, Ji
1997-01-01
In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10^{20}/m^{3} with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.
Halo Current Simulations for Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Paccagnella, R.; Strauss, H. R.; Torasso, R.; Park, W.; Jardin, S.; Breslau, J.; Pletzer, A.; Fu, G. Y.; Sugiyama, L.
2003-10-01
A 3D MHD multi-level code, M3D [1], has been used in this work to simulate a tokamak Vertical Displacement Event (VDE) scenario. These simulations are interesting both from the point of view of gaining physical insight on the plasma dynamics during a VDE or a disruption event and are also relevant in order to estimate the amount of non-axisymmetric torques on the vacuum chamber -- a particular important issue for next generation tokamak (like ITER) design. In this work we numerically simulate a single-null ITER-like plasma evolution using a nonlinear single-fluid model interfaced through thin shell resistive wall boundary conditions to the external vacuum solution. A complete reconstruction of the magnetic field time evolution is therefore possible in the plasma and vacuum regions. Work is in progress to benchmark the code with 2D simulations done with the TSC [2] code. We have previously verified the expected linear scaling of the vertical instability growth rate with the wall resistivity. We are studying the effect of different initial tokamak equilibria (i.e. with different q profiles) on the instability growth rate. We are also studying the nonlinear evolution of the VDE for different fault-scenarios involving the relative timing of the disruptive thermal quench and the loss of vertical control, and for different assumptions regarding the conductivity and width of the plasma halo region. An output of this work is a prediction of the Toroidal Peaking Factor (TPF) [3] of the halo currents and associated vessel forces characterizing the non-axisymmetric events. This work was supported in part by the USDOE. [1] PARK, W., et al., Phys.Plasmas 6, 1796 (1999). [2] Sayer, R.O., Peng, Y-K. M., Jardin, S. C., Kellman, A. G., Wesley, J. C., Nuclear Fusion 33, 969 (1993). [3] Pomphrey, N., Bialek, J., Park, W., Nuclear Fusion 38, 449 (1998).
Industry roles in the Tokamak Physics Experiment
Thomassen, K.; Lang, D.; Schmidt, J.; Burger, A.
1995-06-01
There are several distinguishing features of the Tokamak Physics Experiment (TPX) to be found in the TPX program and in the organizations for constructing and operating the machine. Programmatically, TPX addresses several issues critical to the viability of magnetic fusion power plants. Organizationally, it is a multi-institutional partnership to construct and operate the machine and carry out its program mission. An important part of the construction partnership is the integrated industrial responsibility for design, R&D, and construction. The TPX physics design takes advantage of recent research on advanced tokamak operating modes achieved for time scales of the order of seconds that are consistent with continuous operation. This synergism of high performance (higher power density) modes with plasma current driven mostly by internal pressure (boot-strap effect) points toward tokamak power plants that will be cost-competitive and operate continuously. A large fraction of the project is subcontracted to industry. By policy, these contracts are at a high level in the project breakdown of work, giving contractors much of the overall responsibility for a given major system. That responsibility often includes design and R&D in addition to the fabrication of the system in question. Each contract is managed through one of three national laboratories: PPPL, LLNL, and ORNL. Separate contracts for system integration and construction management round out the industry involvement in the project. This integrated, major responsibility attracts high-level corporate attention within each company, which are major corporations with long-standing interest in fusion. Through the contracts already established on the TPX project, a new standard for industry involvement in fusion has been set, and these industries will be well prepared for future fusion projects.
Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment
Lucia, Matthew James
2015-09-01
The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experiment (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (d ~ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H2 within minutes. For impurity sequestration, LTX plasma performance
Electrostatic analysis of the tokamak edge plasma
Motley, R.W.
1981-07-01
The intrusion of an equipotential poloidal limiter into the edge plasma of a circular tokamak discharge distorts the axisymmetry in two ways: (1) it (partially) shorts out the top-to-bottom Pfirsch-Schlueter driving potentials, and (2) it creates zones of back current flow into the limiter. The resulting boundary mismatch between the outer layers and the inner axisymmetric Pfirsch-Schlueter layer provides free energy to drive the edge plasma unstable. Special limiters are proposed to symmetrize the edge plasma and thereby reduce the electrical and MHD activity in the boundary layer.
Tokamak physics experiment: Diagnostic windows study
Merrigan, M.; Wurden, G.A.
1995-11-01
We detail the study of diagnostic windows and window thermal stress remediation in the long-pulse, high-power Tokamak Physics Experiment (TPX) operation. The operating environment of the TPX diagnostic windows is reviewed, thermal loads on the windows estimated, and cooling requirements for the windows considered. Applicable window-cooling technology from other fields is reviewed and its application to the TPX windows considered. Methods for TPX window thermal conditioning are recommended, with some discussion of potential implementation problems provided. Recommendations for further research and development work to ensure performance of windows in the TPX system are presented.
Self-Organized Stationary States of Tokamaks
Jardin, S. C.; Ferraro, N.; Krebs, I.
2015-11-17
We demonstrate that in a 3D resistive magnetohydrodynamic (MHD) simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to non-linearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary non-sawtoothing “hybrid” discharges, often referred to as “flux-pumping”.
An efficient transport solver for tokamak plasmas
Park, Jin Myung; Murakami, Masanori; St. John, H. E.; ...
2017-01-03
A simple approach to efficiently solve a coupled set of 1-D diffusion-type transport equations with a stiff transport model for tokamak plasmas is presented based on the 4th order accurate Interpolated Differential Operator scheme along with a nonlinear iteration method derived from a root-finding algorithm. Here, numerical tests using the Trapped Gyro-Landau-Fluid model show that the presented high order method provides an accurate transport solution using a small number of grid points with robust nonlinear convergence.
Overview of the Compact Ignition tokamak
Flanagan, C. A.; Peng, Yueng Kay Martin
1986-01-01
The Compact Ignition Tokamak (CIT) mission is to achieve ignition and provide the capability to experimentally study burning plasma behavior. A national team has developed a baseline concept including definition of the necessary research and development. The baseline concept satisfies the physics performance objectives established for the project and complies with defined design specifications. To ensure that the mission is achieved, the design requires large magnetic fields on axis (10 T) and use of large plasma currents (10 MA). The design is capable of accommodating significant auxiliary heating to enter the ignited regime. The CIT is designed to operate in plasma parameter regimes that a are directly relevant to future fusion power reactors.
Neoclassical Transport Properties of Tokamak Plasmas
Weyssow, B.
2004-03-15
The classical transport theory is strictly valid for a plasma in a homogeneous and stationary magnetic field. In the '60, experiments have shown that this theory does not apply as a local theory of transport in Tokamaks. It was shown that global geometric characteristics of the confining elements have a strong influence on the transport. Three regimes of collisionality are characteristic of the neoclassical transport theory: the banana regime (the electronic diffusion coefficient increases starting from zero), the plateau regime (the diffusion coefficient is almost independent of the collisionality) and the Pfirsch-Schlueter regime (the electronic diffusion coefficient again increases with the collisionality)
3D MHD Simulations of Tokamak Disruptions
NASA Astrophysics Data System (ADS)
Woodruff, Simon; Stuber, James
2014-10-01
Two disruption scenarios are modeled numerically by use of the CORSICA 2D equilibrium and NIMROD 3D MHD codes. The work follows the simulations of pressure-driven modes in DIII-D and VDEs in ITER. The aim of the work is to provide starting points for simulation of tokamak disruption mitigation techniques currently in the CDR phase for ITER. Pressure-driven instability growth rates previously observed in simulations of DIIID are verified; Halo and Hiro currents produced during vertical displacements are observed in simulations of ITER with implementation of resistive walls in NIMROD. We discuss plans to exercise new code capabilities and validation.
Diamagnetic flux measurement in Aditya tokamak
Kumar, Sameer; Jha, Ratneshwar; Lal, Praveen; Hansaliya, Chandresh; Gopalkrishna, M. V.; Kulkarni, Sanjay; Mishra, Kishore
2010-12-15
Measurements of diamagnetic flux in Aditya tokamak for different discharge conditions are reported for the first time. The measured diamagnetic flux in a typical discharge is less than 0.6 mWb and therefore it has required careful compensation for various kinds of pick-ups. The hardware and software compensations employed in this measurement are described. We introduce compensation of a pick-up due to plasma current of less than 20 kA in short duration discharges, in which plasma pressure gradient is supposed to be negligible. The flux measurement during radio frequency heating is also presented in order to validate compensation.
Diamagnetic flux measurement in Aditya tokamak.
Kumar, Sameer; Jha, Ratneshwar; Lal, Praveen; Hansaliya, Chandresh; Gopalkrishna, M V; Kulkarni, Sanjay; Mishra, Kishore
2010-12-01
Measurements of diamagnetic flux in Aditya tokamak for different discharge conditions are reported for the first time. The measured diamagnetic flux in a typical discharge is less than 0.6 mWb and therefore it has required careful compensation for various kinds of pick-ups. The hardware and software compensations employed in this measurement are described. We introduce compensation of a pick-up due to plasma current of less than 20 kA in short duration discharges, in which plasma pressure gradient is supposed to be negligible. The flux measurement during radio frequency heating is also presented in order to validate compensation.
Tokamak equilibria with reversed current density.
Martynov, A A; Medvedev, S Yu; Villard, L
2003-08-22
Observations of nearly zero toroidal current in the central region of tokamaks (the "current hole") raises the question of the existence of toroidal equilibria with very low or reversed current in the core. The solutions of the Grad-Shafranov equilibrium equation with hollow toroidal current density profile including negative current density in the plasma center are investigated. Solutions of the corresponding eigenvalue problem provide simple examples of such equilibrium configurations. More realistic equilibria with toroidal current density reversal are computed using a new equilibrium problem formulation and computational algorithm which do not assume nested magnetic surfaces.
Nonlinear gyrokinetic equations for tokamak microturbulence
Hahm, T.S.
1988-05-01
A nonlinear electrostatic gyrokinetic Vlasov equation, as well as Poisson equation, has been derived in a form suitable for particle simulation studies of tokamak microturbulence and associated anomalous transport. This work differs from the existing nonlinear gyrokinetic theories in toroidal geometry, since the present equations conserve energy while retaining the crucial linear and nonlinear polarization physics. In the derivation, the action-variational Lie perturbation method is utilized in order to preserve the Hamiltonian structure of the original Vlasov-Poisson system. Emphasis is placed on the dominant physics of the collective fluctuations in toroidal geometry, rather than on details of particle orbits. 13 refs.
Viscosity in the edge of tokamak plasmas
Stacey, W.M.
1993-05-01
A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the ``short-radial-gradient-scale-length`` (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates.
Viscosity in the edge of tokamak plasmas
NASA Astrophysics Data System (ADS)
Stacey, W. M.
1993-05-01
A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the 'short radial gradient scale length' (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates.
Self-Organized Stationary States of Tokamaks
Jardin, S. C.; Ferraro, N.; Krebs, I.
2015-11-17
We demonstrate that in a 3D resistive magnetohydrodynamic (MHD) simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to non-linearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary non-sawtoothing “hybrid” discharges, often referred to as “flux-pumping”.
Dust divertor for a tokamak fusion reactor
Tang, X Z; Delzanno, G L
2009-01-01
Micron-size tungsten particulates find equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic-ion-flow drag parallel to the divertor surface. The natural circulation of dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.
Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment
NASA Astrophysics Data System (ADS)
Boyle, Dennis Patrick
This thesis presents new measurements of core impurity concentrations and transport in plasmas with lithium coatings on all-metal plasma facing components (PFCs) in the Lithium Tokamak Experiment (LTX). LTX is a modest-sized spherical tokamak uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma (as opposed to just the divertor or limiter region in other devices). Lithium (Li) wall-coatings have improved plasma performance and confinement in several tokamaks with carbon (C) PFCs, including the National Spherical Torus Experiment (NSTX). In NSTX, contamination of the core plasma with Li impurities was very low (<0.1%) despite extensive divertor coatings. Low Li levels in NSTX were found to be largely due to neoclassical forces from the high level of C impurities. Studying impurity levels and transport with Li coatings on stainless steel surfaces in LTX is relevant to future devices (including future enhancements to NSTX-Upgrade) with all-metal PFCs. The new measurements in this thesis were enabled by a refurbished Thomson scattering system and improved impurity spectroscopy, primarily using a novel visible spectrometer monitoring several Li, C, and oxygen (O) emission lines. A simple model was used to account for impurities in unmeasured charge states, assuming constant density in the plasma core and constant concentration in the edge. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with 2-4% Li, 0.6-2% C, 0.4-0.7% O, and Z eff<1.2. Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, unlike in NSTX, where collisions with C dominated. Furthermore, neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two, in contrast to NSTX where they differed by an order of
Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment
Boyle, Dennis Patrick
2016-09-01
This thesis presents new measurements of core impurity concentrations and transport in plasmas with lithium coatings on all-metal plasma facing components (PFCs) in the Lithium Tokamak Experiment (LTX). LTX is a modest-sized spherical tokamak uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma (as opposed to just the divertor or limiter region in other devices). Lithium (Li) wall-coatings have improved plasma performance and confinement in several tokamaks with carbon (C) PFCs, including the National Spherical Torus Experiment (NSTX). In NSTX, contamination of the core plasma with Li impurities was very low (<0.1%) despite extensive divertor coatings. Low Li levels in NSTX were found to be largely due to neoclassical forces from the high level of C impurities. Studying impurity levels and transport with Li coatings on stainless steel surfaces in LTX is relevant to future devices (including future enhancements to NSTX-Upgrade) with all-metal PFCs. The new measurements in this thesis were enabled by a refurbished Thomson scattering system and improved impurity spectroscopy, primarily using a novel visible spectrometer monitoring several Li, C, and oxygen (O) emission lines. A simple model was used to account for impurities in unmeasured charge states, assuming constant density in the plasma core and constant concentration in the edge. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with~2-4% Li, ~0.6-2% C, ~0.4-0.7% O, and Z_eff<1.2. Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, unlike in NSTX, where collisions with C dominated. Furthermore, neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two, in contrast to NSTX where they differed by an order of
Buckling of spherical shells revisited
NASA Astrophysics Data System (ADS)
Hutchinson, John W.
2016-11-01
A study is presented of the post-buckling behaviour and imperfection sensitivity of complete spherical shells subject to uniform external pressure. The study builds on and extends the major contribution to spherical shell buckling by Koiter in the 1960s. Numerical results are presented for the axisymmetric large deflection behaviour of perfect spheres followed by an extensive analysis of the role axisymmetric imperfections play in reducing the buckling pressure. Several types of middle surface imperfections are considered including dimple-shaped undulations and sinusoidal-shaped equatorial undulations. Buckling occurs either as the attainment of a maximum pressure in the axisymmetric state or as a non-axisymmetric bifurcation from the axisymmetric state. Several new findings emerge: the abrupt mode localization that occurs immediately after the onset of buckling, the existence of an apparent lower limit to the buckling pressure for realistically large imperfections, and comparable reductions of the buckling pressure for dimple and sinusoidal equatorial imperfections.
Orthogonality of spherical harmonic coefficients
NASA Technical Reports Server (NTRS)
Mcleod, M. G.
1980-01-01
Orthogonality relations are obtained for the spherical harmonic coefficients of functions defined on the surface of a sphere. Following a brief discussion of the orthogonality of Fourier series coefficients, consideration is given to the values averaged over all orientations of the coordinate system of the spherical harmonic coefficients of a function defined on the surface of a sphere that can be expressed in terms of Legendre polynomials for the special case where the function is the sum of two delta functions located at two different points on the sphere, and for the case of an essentially arbitrary function. It is noted that the orthogonality relations derived have found applications in statistical studies of the geomagnetic field.
Orthogonality of spherical harmonic coefficients
NASA Technical Reports Server (NTRS)
Mcleod, M. G.
1980-01-01
Orthogonality relations are obtained for the spherical harmonic coefficients of functions defined on the surface of a sphere. Following a brief discussion of the orthogonality of Fourier series coefficients, consideration is given to the values averaged over all orientations of the coordinate system of the spherical harmonic coefficients of a function defined on the surface of a sphere that can be expressed in terms of Legendre polynomials for the special case where the function is the sum of two delta functions located at two different points on the sphere, and for the case of an essentially arbitrary function. It is noted that the orthogonality relations derived have found applications in statistical studies of the geomagnetic field.
Fresnel diffraction by spherical obstacles
NASA Technical Reports Server (NTRS)
Hovenac, Edward A.
1989-01-01
Lommel functions were used to solve the Fresnel-Kirchhoff diffraction integral for the case of a spherical obstacle. Comparisons were made between Fresnel diffraction theory and Mie scattering theory. Fresnel theory is then compared to experimental data. Experiment and theory typically deviated from one another by less than 10 percent. A unique experimental setup using mercury spheres suspended in a viscous fluid significantly reduced optical noise. The major source of error was due to the Gaussian-shaped laser beam.
Contractions of affine spherical varieties
Arzhantsev, I V
1999-08-31
The language of filtrations and contractions is used to describe the class of G-varieties obtainable as the total spaces of the construction of contraction applied to affine spherical varieties, which is well-known in invariant theory. These varieties are local models for arbitrary affine G-varieties of complexity 1 with a one-dimensional categorical quotient. As examples, reductive algebraic semigroups and three-dimensional SL{sub 2}-varieties are considered.
Hill, K. W.; Bitter, M. L.; Scott, S. D.; Ince-Cushman, A.; Reinke, M.; Rice, J. E.; Beiersdorfer, P.; Gu, M.-F.; Lee, S. G.; Broennimann, Ch.; Eikenberry, E. F.
2008-10-15
A new spatially resolving x-ray crystal spectrometer capable of measuring continuous spatial profiles of high resolution spectra ({lambda}/d{lambda}>6000) of He-like and H-like Ar K{alpha} lines with good spatial ({approx}1 cm) and temporal ({approx}10 ms) resolutions has been installed on the Alcator C-Mod tokamak. Two spherically bent crystals image the spectra onto four two-dimensional Pilatus II pixel detectors. Tomographic inversion enables inference of local line emissivity, ion temperature (T{sub i}), and toroidal plasma rotation velocity (v{sub {phi}}) from the line Doppler widths and shifts. The data analysis techniques, T{sub i} and v{sub {phi}} profiles, analysis of fusion-neutron background, and predictions of performance on other tokamaks, including ITER, will be presented.
Sphericity Tests and Repeated Measures Data.
ERIC Educational Resources Information Center
Robey, Randall R.; Barcikowski, Robert S.
The mixed model analysis of variance assumes a mathematical property known as sphericity. Several preliminary tests have been proposed to detect departures from the sphericity assumption. The logic of the preliminary testing procedure is to conduct the mixed model analysis of variance if the preliminary test suggests that the sphericity assumption…
Lagrangian Description of Nonadiabatic Particle Motion in Spherical Tori
R.B. White; Yu.V. Yakovenko; Ya.I. Kolesnichenko
2002-06-21
The ability of a device to provide adiabatic motion of charged particles is crucial for magnetic confinement. As the magnetic field in the present-day spherical tori, e.g., MAST and NSTX, is much lower than in the conventional tokamaks, effects of the finite Larmor radius (FLR) on the motion of fast ions are of importance in these devices, affecting the stochasticity threshold for the interaction of the ions with electromagnetic perturbations. In addition, FLR by itself may result in non-conservation (jumps) of the magnetic moment of particles [4]. In this work we propose a Lagrangian approach to description of the resonant collisionless motion of charged particles under a perturbation, allowing for FLR. The work generalizes results of Ref. [1], where only time-independent perturbations were considered. The approach is used to find the stochasticity thresholds for the Goldston-White-Boozer (GWB) diffusion [2] and the cyclotron-resonance-induced (CRI) diffusion (for the case of the firs t cyclotron resonance, the latter was discovered in Ref. [3]). In addition, a new expression for the magnetic moment variation caused by FLR is found.
Plasma Shape and Current Density Profile Control in Advanced Tokamak Operating Scenarios
NASA Astrophysics Data System (ADS)
Shi, Wenyu
The need for new sources of energy is expected to become a critical problem within the next few decades. Nuclear fusion has sufficient energy density to potentially supply the world population with its increasing energy demands. The tokamak is a magnetic confinement device used to achieve controlled fusion reactions. Experimental fusion technology has now reached a level where tokamaks are able to produce about as much energy as is expended in heating the fusion fuel. The next step towards the realization of a nuclear fusion tokamak power plant is ITER, which will be capable of exploring advanced tokamak (AT) modes, characterized by a high fusion gain and plasma stability. The extreme requirements of the advanced modes motivates researchers to improve the modeling of the plasma response as well as the design of feedback controllers. This dissertation focuses on several magnetic and kinetic control problems, including the plasma current, position and shape control, and data-driven and first-principles-driven modeling and control of plasma current density profile and the normalized plasma pressure ratio betaN. The plasma is confined within the vacuum vessel by an external electromagnetic field, produced primarily by toroidal and poloidal field coils. The outermost closed plasma surface or plasma boundary is referred to as the shape of the plasma. A central characteristic of AT plasma regimes is an extreme elongated shape. The equilibrium among the electromagnetic forces acting on an elongated plasma is unstable. Moreover, the tokamak performance is improved if the plasma is located in close proximity to the torus wall, which guarantees an efficient use of available volume. As a consequence, feedback control of the plasma position and shape is necessary. In this dissertation, an Hinfinity-based, multi-input-multi-output (MIMO) controller for the National Spherical Torus Experiment (NSTX) is developed, which is used to control the plasma position, shape, and X
Electrostatic Dust Detection and Removal in Tokamaks
NASA Astrophysics Data System (ADS)
Hensley, R.; Skinner, C. H.; Roquemore, A. L.
2006-10-01
The inventory of in-vessel dust particles in next-step tokamaks will increase with the rise in stored energy and pulse duration. Dust levels will need to be measured and controlled for safety reasons and to avoid plasma contamination. A novel electrostatic dust detector has been developed with a sensitivity appropriate for the carbon dust levels expected in next-step devices.^23 Higher sensitivity is desired for real-time measurements in contemporary tokamaks that have less dust. We report on results from a larger area, more sensitive detector. A 2 x 2 circuit board has two interlocking combs of copper traces spaced by 25 microns and biased at 30-50 V. The carbon test dust is delivered to the circuit board by a mesh tray vibrated at 60 Hz. The impinging dust creates a short circuit and the resulting current pulse is recorded. We will present results on the dust detection sensitivity and dust removal efficiency of these new detectors in three environments: air, vacuum, and inert gas. ^2 C. Voinier et al., J. Nucl. Mater. 346 (2005) 266-271. ^3 C. Parker et al., PPPL Report, PPPL-4169.
Constrained ripple optimization of Tokamak bundle divertors
Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.
1983-02-01
Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ..xi.. B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple (<0.2%) so that, now, most banana-trapped fast ions are confined. Only those ions with banana tips near the outside region (absolute value theta < or equal to 45/sup 0/) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded.
TIBER: tokamak ignition/burn experimental research
Henning, C.D.; Logan, B.G.; Barr, W.L.; Bulmer, R.H.; Doggett, J.N.; Johnston, B.M.; Hoard, R.W.; Lee, J.D.; Miller, J.R.; Slack, D.S.; Schultz, J.H.
1985-11-01
As part of a continuing effort by the Office of Fusion Energy to define an ignition experiment, a superconducting tokamak has been designed with thin neutron shielding and aggressive magnet and plasma parameters. By so minimizing the inner radial dimensions of the tokamak center post, coil, and shielding region, the plasma major radius is reduced, with a corresponding reduction in device costs. The peak nuclear-heating rate in the superconducting TF coils is 22 mW/cmT, which results in a steady heat load of 50 kW to the cryogenic system. Fast-wave, lower-hybrid heating would be used to induce a 10-MA current in a moderate density plasma. Then pellet fueling would raise the density to achieve ignition as the current decays in a few hundred seconds. Steady-state current drive in subignited conditions permits a 0.8 MW/mS average wall loading to study plasma and nuclear engineering effects. 10 refs., 6 figs., 3 tabs.
The ARIES-I Tokamak Reactor Study
Najmabadi, F; Peng, Yueng Kay Martin
1991-01-01
The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design si a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-{sup 3}He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.
RF Wave Propagation and Scattering in Tokamaks
NASA Astrophysics Data System (ADS)
Horton, Wendell; Goniche, Marc; Arefiev, Alex; Peysson, Yves; Ekedahl, Annika; InstituteFusion Studies Collaboration; IRFM CEA Collaboration
2016-10-01
The propagation, scattering and absorption of the lower hybrid and electron cyclotron RF waves used to control fusion plasmas is reviewed. Drift wave turbulence driven by the steep ion and electron temperature gradients in H-mode divertor tokamaks produces strong scattering of the RF waves used for heating and plasma currents drive Both the 3-5GHz lower-hybrid (LH) and the 170GHZ electron cyclotron (EC) waves experience scattering and diffraction as propagating through the statistically complex density of the plasma. Ray equations are used to calculate the spread of the rays and the associated change in the parallel phase, polarization and group velocity of the RF waves in the propagation through the fusion plasma. A Fokker Planck equation for the phase space of the RF plasmons is one method to describe the spread of the RF wave power in the complex geometry of a divertor tokamak using the ray tracing codes. The evolution of the electron distribution function from the resonant electron-wave interactions is summarized for several scenarios. The resulting X-ray spectrum is broaden giving better agreement with the measured X-ray spectrum than that calculated in the absence of the turbulent scattering of the RF waves. M. Goniche et al., and Tore Supra Team, Phys. Plasmas 21, 2014.
The external kink mode in diverted tokamaks
Turnbull, Alan D.; Hanson, Jeremy M.; Turco, Francesca; ...
2016-06-16
Here, an explanation is provided for the disruptive instability in diverted tokamaks when the safety factor at the 95% poloidal flux surface, q95, is driven below 2.0. The instability is a resistive kink counterpart to the current-driven ideal mode that traditionally explained the corresponding disruption in limited cross-sections when qedge, the safety factor at the outermost closed flux surface, lies just below a rational value. Experimentally, external kink modes are observed in limiter configurations as the current in a tokamak is ramped up and qedge decreases through successive rational surfaces. For qedge < 2, the instability is always encountered andmore » is highly disruptive. However, diverted plasmas, in which qedge is formally infinite in the magnetohydrodynamic (MHD) model, have presented a longstanding difficulty since the theory would predict stability, yet, the disruptive limit occurs in practice when q95, reaches 2. It is shown from numerical calculations that a resistive kink mode is linearly destabilized by the rapidly increasing resistivity at the plasma edge when q95 < 2, but qedge >> 2. The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the q = 2 surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.« less
Gyrokinetic simulation of microturbulence in EAST tokamak
NASA Astrophysics Data System (ADS)
Xiao, Yong; Zhang, Taige; Zhao, Chen
2014-10-01
A complete understanding of anomalous transport is critical for designing future magnetic fusion reactors. It is generally accepted that the micro-scale turbulence leads to anomalous transport. For low beta toroidal plasmas, the electrostatic modes may dominate and ion temperature gradient (ITG) mode and trapped electron mode (TEM) are two very important candidates accounting for ion and electron turbulent transport respectively. Recently the massively parallel gyrokinetic simulation has emerged as a major tool to investigate the nonlinear physics of the turbulent transport. The newly-developed capabilities enable the gyrokinetic code GTC to simulate the turbulent transport for real tokamak plasma shape and profiles. These capabilities include a new gyrokinetic Poisson solver and zonal flow solver suitable for general plasma shape and profiles, improvements on the conventional four-point gyroaverage and newly-developed nonuniform initial marker loading. The GTC code is now able to import experimental plasma profiles and equilibrium magnetic field that come from the EFIT or TRANSP equilibrium reconstruction. Linear and nonlinear gyrokinetic simulations are carried out with the new capabilities in GTC for the electron coherent mode (ECM) recently observed in the EAST tokamak (EAST shot # 38300). We found that in the pedestal region with strong electron temperature gradient, the unstable waves propagate in the electron diamagnetic direction, showing a trapped electron mode (TEM) feature. It is also found in the collisionless limit, the linear mode frequency is higher than that from the experiment.
Predicting temperature and density profiles in tokamaks
Bateman, G.; Kritz, A.H.; Kinsey, J.E.; Redd, A.J.; Weiland, J.
1998-05-01
A fixed combination of theory-based transport models, called the Multi-Mode Model, is used in the BALDUR [C. E. Singer {ital et al.}, Comput. Phys. Commun. {bold 49}, 275 (1988)] transport simulation code to predict the temperature and density profiles in tokamaks. The choice of the Multi-Mode Model has been guided by the philosophy of using the best transport theories available for the various modes of turbulence that dominate in different parts of the plasma. The Multi-Mode model has been found to provide a better match to temperature and density profiles than any of the other theory-based models currently available. A description and partial derivation of the Multi-Mode Model is presented, together with three new examples of simulations of the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire {ital et al.}, Phys. Plasmas {bold 2}, 2176 (1995)]. The first simulation shows the strong effect of recycling on the ion temperature profile in TFTR supershot simulations. The second simulation explores the effect of a plasma current ramp{emdash}where the plasma energy content changes slowly on the energy confinement time scale. The third simulation shows that the Multi-Mode Model reproduces the experimentally measured profiles when tritium is used as the hydrogenic isotope in L-mode (low confinement mode) plasmas. {copyright} {ital 1998 American Institute of Physics.}
Midplane Faraday rotation: A tokamak densitometer
NASA Astrophysics Data System (ADS)
Jobes, F. C.
1995-01-01
The density in a tokamak can be determined by measuring the Faraday rotation of a laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then ne(R) can be readily obtained with a simple Abel inversion about the center line of the tokamak. For a large machine, such as ITER, TPX, or JT-60, a 10.6 μm laser would be appropriate. If the machine operated at a full field of 10-50 T m and a peak density of 2.5×1020/m3, the rotation angle would be quite large—about 15°-75° per pass. An elegant measurement system can be made up from a single laser beam diffracted off a moving grating to form a fan of ˜10 probe beams. With the addition of a few optical components to the system, the return beams can be recombined and sent to a single detector. In the detector there is a separate frequency component for both the right and left hand component of each ray. These can be separated electronically to provide a reference and probe signal for each ray; the difference in phase between the two signals is twice the Faraday rotation angle.
System studies for quasi-steady-state advanced physics tokamak
Reid, R.L.; Peng, Y.K.M.
1983-11-01
Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated.
/sup 3/He functions in tokamak-pumped laser systems
Jassby, D.L.
1986-10-01
/sup 3/He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the /sup 3/He(n,p)T reaction, and thereby excite gaseous lasants mixed with the /sup 3/He while simultaneously breeding tritium. The total /sup 3/He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak.
Effect of ICRF Heating on Single Particle Confinement in Tokamaks,
1980-03-01
80 K W WHANG. 6 .J MORALES N0001IG75C4476 W4CLASSIFlED PP"-73 ML IND 4, LAa4 5 057 Effect of ICRP Heating on Single Particle Confinement in Tokamaks ...bas =akapmwd a b 11 1Af b tmhmdin4 -2- ABSTRACT The simultaneous effect of ion heating and spatial diffusion due to ICRF heating in tokamak geometry... Tokamak confinement devices. Previous theoretical studies I1 2 of this topic have dealt with important questions related to the efficiency of the heating
Nonlinear stabilization of tokamak microturbulence by fast ions.
Citrin, J; Jenko, F; Mantica, P; Told, D; Bourdelle, C; Garcia, J; Haverkort, J W; Hogeweij, G M D; Johnson, T; Pueschel, M J
2013-10-11
Nonlinear electromagnetic stabilization by suprathermal pressure gradients found in specific regimes is shown to be a key factor in reducing tokamak microturbulence, augmenting significantly the thermal pressure electromagnetic stabilization. Based on nonlinear gyrokinetic simulations investigating a set of ion heat transport experiments on the JET tokamak, described by Mantica et al. [Phys. Rev. Lett. 107, 135004 (2011)], this result explains the experimentally observed ion heat flux and stiffness reduction. These findings are expected to improve the extrapolation of advanced tokamak scenarios to reactor relevant regimes.
Lu, B.; Wang, F.; Fu, J.; Li, Y.; Wan, B.; Shi, Y.; Bitter, M.; Hill, K. W.; Lee, S. G.
2012-10-15
Two imaging x-ray crystal spectrometers, the so-called 'poloidal' and 'tangential' spectrometers, were recently implemented on experimental advanced superconducting tokamak (EAST) to provide spatially and temporally resolved impurity ion temperature (T{sub i}), electron temperature (T{sub e}) and rotation velocity profiles. They are derived from Doppler width of W line for Ti, the intensity ratio of Li-like satellites to W line for Te, and Doppler shift of W line for rotation. Each spectrometer originally consisted of a spherically curved crystal and a two-dimensional multi-wire proportional counter (MWPC) detector. Both spectrometers have now been upgraded. The layout of the tangential spectrometer was modified, since it had to be moved to a different port, and the spectrometer was equipped with two high count rate Pilatus detectors (Model 100 K) to overcome the count rate limitation of the MWPC and to improve its time resolution. The poloidal spectrometer was equipped with two spherically bent crystals to record the spectra of He-like and H-like argon simultaneously and side by side on the original MWPC. These upgrades are described, and new results from the latest EAST experimental campaign are presented.
Lu, B; Wang, F; Shi, Y; Bitter, M; Hill, K W; Lee, S G; Fu, J; Li, Y; Wan, B
2012-10-01
Two imaging x-ray crystal spectrometers, the so-called "poloidal" and "tangential" spectrometers, were recently implemented on experimental advanced superconducting tokamak (EAST) to provide spatially and temporally resolved impurity ion temperature (T(i)), electron temperature (T(e)) and rotation velocity profiles. They are derived from Doppler width of W line for Ti, the intensity ratio of Li-like satellites to W line for Te, and Doppler shift of W line for rotation. Each spectrometer originally consisted of a spherically curved crystal and a two-dimensional multi-wire proportional counter (MWPC) detector. Both spectrometers have now been upgraded. The layout of the tangential spectrometer was modified, since it had to be moved to a different port, and the spectrometer was equipped with two high count rate Pilatus detectors (Model 100 K) to overcome the count rate limitation of the MWPC and to improve its time resolution. The poloidal spectrometer was equipped with two spherically bent crystals to record the spectra of He-like and H-like argon simultaneously and side by side on the original MWPC. These upgrades are described, and new results from the latest EAST experimental campaign are presented.
ECH by FEL and gyrotron sources on the Microwave Tokamak Experiment (MTX) tokamak
Stallard, B.W.; Turner, W.C.; Allen, S.L.; Byers, J.A.; Felker, B.; Fenstermacher, M.E.; Ferguson, S.W.; Hooper, E.G.; Thomassen, K.I.; Throop, A.L. ); Makowski, M.A. )
1990-08-09
The Microwave Tokamak Experiment (MTX) at LLNL is studying the physics of intense pulse ECH is a high-density tokamak plasma using a microwave FEL. Related technology development includes the FEL, a windowless quasi-optical transmission system, and other microwave components. Initial plasma experiments have been carried out at 140 GHz with single rf pulses generated using the ETA-II accelerator and the ELF wiggler. Peak power levels up to 0.2 GW and pulse durations up to 10 ns were achieved for injection into the plasma using as untapered wiggler. FEL pulses were transmitted over 33 m from the FEL to MTX using six mirrors mounted in a 50-cm-diam evacuated pipe. Measurements of the microwave beam and transmission through the plasma were carried out. For future rapid pulse experiments at high average power (4 GW peak power, 5kHz pulse rate, and {bar P} > 0.5 MW) using the IMP wiggler with tapered magnetic field, a gyrotron (140 GHz, 400 kW cw or up to 1 MW short pulse) is being installed to drive the FEL input or to directly heat the tokamak plasma at full gyrotron power. Quasi-optic techniques will be used to couple the gyrotron power. For direct plasma heating, the gyrotron will couple into the existing mirror transport system. Using both sources of rf generation, experiments are planned to investigate intense pulse absorption and tokamak physics, such as the ECH of a pellet-fueled plasma and plasma control using localized heating. 12 refs., 9 figs.
ECH by FEL and gyrotron sources on the Microwave Tokamak Experiment (MTX) tokamak
NASA Astrophysics Data System (ADS)
Stallard, B. W.; Turner, W. C.; Allen, S. L.; Byers, J. A.; Felker, B.; Fenstermacher, M. E.; Ferguson, S. W.; Hooper, E. G.; Thomassen, K. I.; Throop, A. L.
1990-08-01
The Microwave Tokamak Experiment (MTX) at LLNL is studying the physics of intense pulse ECH is a high-density tokamak plasma using a microwave FEL. Related technology development includes the FEL, a windowless quasi-optical transmission system, and other microwave components. Initial plasma experiments have been carried out at 140 GHz with single RF pulses generated using the ETA-2 accelerator and the ELF wiggler. Peak power levels up to 0.2 GW and pulse durations up to 10 ns were achieved for injection into the plasma using as untapered wiggler. FEL pulses were transmitted over 33 m from the FEL to MTX using six mirrors mounted in a 50 cm diam evacuated pipe. Measurements of the microwave beam and transmission through the plasma were carried out. For future rapid pulse experiments at high average power (4 GW peak power, 5 kHz pulse rate, and bar P is greater than 0.5 MW) using the IMP wiggler with tapered magnetic field, a gyrotron (140 GHz, 400 kW CW or up to 1 MW short pulse) is being installed to drive the FEL input or to directly heat the tokamak plasma at full gyrotron power. Quasi-optic techniques will be used to couple the gyrotron power. For direct plasma heating, the gyrotron will couple into the existing mirror transport system. Using both sources of RF generation, experiments are planned to investigate intense pulse absorption and tokamak physics, such as the ECH of a pellet-fueled plasma and plasma control using localized heating.
A Spherical Aerial Terrestrial Robot
NASA Astrophysics Data System (ADS)
Dudley, Christopher J.
This thesis focuses on the design of a novel, ultra-lightweight spherical aerial terrestrial robot (ATR). The ATR has the ability to fly through the air or roll on the ground, for applications that include search and rescue, mapping, surveillance, environmental sensing, and entertainment. The design centers around a micro-quadcopter encased in a lightweight spherical exoskeleton that can rotate about the quadcopter. The spherical exoskeleton offers agile ground locomotion while maintaining characteristics of a basic aerial robot in flying mode. A model of the system dynamics for both modes of locomotion is presented and utilized in simulations to generate potential trajectories for aerial and terrestrial locomotion. Details of the quadcopter and exoskeleton design and fabrication are discussed, including the robot's turning characteristic over ground and the spring-steel exoskeleton with carbon fiber axle. The capabilities of the ATR are experimentally tested and are in good agreement with model-simulated performance. An energy analysis is presented to validate the overall efficiency of the robot in both modes of locomotion. Experimentally-supported estimates show that the ATR can roll along the ground for over 12 minutes and cover the distance of 1.7 km, or it can fly for 4.82 minutes and travel 469 m, on a single 350 mAh battery. Compared to a traditional flying-only robot, the ATR traveling over the same distance in rolling mode is 2.63-times more efficient, and in flying mode the system is only 39 percent less efficient. Experimental results also demonstrate the ATR's transition from rolling to flying mode.
Electronic switching spherical array antenna
NASA Technical Reports Server (NTRS)
Stockton, R.
1978-01-01
This work was conducted to demonstrate the performance levels attainable with an ESSA (Electronic Switching Spherical Array) antenna by designing and testing an engineering model. The antenna was designed to satisfy general spacecraft environmental requirements and built to provide electronically commandable beam pointing capability throughout a hemisphere. Constant gain and beam shape throughout large volumetric coverage regions are the principle characteristics. The model is intended to be a prototype of a standard communications and data handling antenna for user scientific spacecraft with the Tracking and Data Relay Satellite System (TDRSS). Some additional testing was conducted to determine the feasibility of an integrated TDRSS and GPS (Global Positioning System) antenna system.
Spherically symmetric canonical quantum gravity
NASA Astrophysics Data System (ADS)
Brahma, Suddhasattwa
2015-06-01
Canonical quantization of spherically symmetric space-times is carried out, using real-valued densitized triads and extrinsic curvature components, with specific factor-ordering choices ensuring in an anomaly free quantum constraint algebra. Comparison with previous work [Nucl. Phys. B399, 211 (1993)] reveals that the resulting physical Hilbert space has the same form, although the basic canonical variables are different in the two approaches. As an extension, holonomy modifications from loop quantum gravity are shown to deform the Dirac space-time algebra, while going beyond "effective" calculations.
Radiative transfer in spherical atmospheres
NASA Astrophysics Data System (ADS)
Kalkofen, W.; Wehrse, R.
A method for defining spherical model atmospheres in radiative/convective and hydrostatic equilibrium is presented. A finite difference form is found for the transfer equation and a matrix operator is developed as the discrete space analog (in curvilinear coordinates) of a formal integral in plane geometry. Pressure is treated as a function of temperature. Flux conservation is maintained within the energy equation, although the correct luminosity transport must be assigned for any given level of the atmosphere. A perturbed integral operator is used in a complete linearization of the transfer and constraint equations. Finally, techniques for generating stable solutions in economical computer time are discussed.
APPARATUS FOR GRINDING SPHERICAL BODIES
Burch, R.F. Jr.
1963-09-24
A relatively inexpensive device is described for grinding rough ceramic bodies into accurate spherical shapes using a conventional drill press and a belt sander. A horizontal disk with an abrasive-surfaced recess in its lower face is mounted eccentrically on a vertical shaft which is forced downward against a stop by a spring. Bodies to be ground are placed in the recess and are subjected to the abrasive action of the belt sander as the disk is rotated by the drill press. (AEC)
TFTR/JET INTOR workshop on plasma transport tokamaks
Singer, C.E.
1985-01-01
This report summarizes the proceedings of a Workshop on transport models for prediction and analysis of tokamak plasma confinement. Summaries of papers on theory, predictive modeling, and data analysis are included.
On Stochastic Control of Tokamak and Artificial Intelligence
NASA Astrophysics Data System (ADS)
Rastovic, Danilo
2007-12-01
Instead of the theory of invariant manifolds and entropy reduction, the theory of fractional Brownian motions and artificiall neural networks is used for description of advanced methods for control of tokamak plasma behaviour.
Tokamak reactor cost model based on STARFIRE/WILDCAT costing
Evans, K. Jr.
1983-03-01
A cost model is presented which is useful for survey and comparative studies of tokamak reactors. The model is heavily based on STARFIRE and WILDCAT costing guidelines, philosophies, and procedures and reproduces the costing for these devices quite accurately.
Improvement of tokamak performance by injection of electrons
Ono, Masayuki
1992-12-01
Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas.
An emerging understanding of H-mode discharges in tokamaks
Groebner, R.J.
1992-12-01
A remarkable degree of consistency of experimental results from tokamaks throughout the world has developed with regard to the phenomenology of the transition from L-mode to H-mode confinement in tokamaks. The transition is initiated in a narrow layer at the plasma periphery where density fluctuations are suppressed and steep gradients of temperature and density form in a region with large first and second radial derivatives in the [upsilon][sub E][sup [yields
Experimental observations and modelling of intrinsic rotation reversals in tokamaks
NASA Astrophysics Data System (ADS)
Camenen, Y.; Angioni, C.; Bortolon, A.; Duval, B. P.; Fable, E.; Hornsby, W. A.; McDermott, R. M.; Na, D. H.; Na, Y.-S.; Peeters, A. G.; Rice, J. E.
2017-03-01
The progress made in understanding spontaneous toroidal rotation reversals in tokamaks is reviewed and current ideas to solve this ten-year-old puzzle are explored. The paper includes a summarial synthesis of the experimental observations in AUG, C-Mod, KSTAR, MAST and TCV tokamaks, reasons why turbulent momentum transport is thought to be responsible for the reversals, a review of the theory of turbulent momentum transport and suggestions for future investigations.
Design of a microwave calorimeter for the microwave tokamak experiment
Marinak, M. )
1988-10-07
The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs.
Fast tomographic methods for the tokamak ISTTOK
NASA Astrophysics Data System (ADS)
Carvalho, P. J.; Thomsen, H.; Gori, S.; Toussaint, U. v.; Weller, A.; Coelho, R.; Neto, A.; Pereira, T.; Silva, C.; Fernandes, H.
2008-04-01
The achievement of long duration, alternating current discharges on the tokamak IST-TOK requires a real-time plasma position control system. The plasma position determination based on magnetic probes system has been found to be inadequate during the current inversion due to the reduced plasma current. A tomography diagnostic has been therefore installed to supply the required feedback to the control system. Several tomographic methods are available for soft X-ray or bolo-metric tomography, among which the Cormack and Neural networks methods stand out due to their inherent speed of up to 1000 reconstructions per second, with currently available technology. This paper discusses the application of these algorithms on fusion devices while comparing performance and reliability of the results. It has been found that although the Cormack based inversion proved to be faster, the neural networks reconstruction has fewer artifacts and is more accurate.
Tearing mode analysis in tokamaks, revisited
Nishimura, Y.; Callen, J.D.; Hegna, C.C.
1998-12-01
A new {Delta}{sup {prime}} shooting code has been developed to investigate tokamak plasma tearing mode stability in a cylinder and large aspect ratio ({epsilon}{le}0.25) toroidal geometries, neglecting toroidal mode coupling. A different computational algorithm is used (shooting out from the singular surface instead of into it) to resolve the strong singularities at the mode rational surface, particularly in the presence of the finite pressure term. Numerical results compare favorably with Furth {ital et al.} [H. P. Furth {ital et al.}, Phys. Fluids {bold 16}, 1054 (1973)] results. The effects of finite pressure, which are shown to decrease {Delta}{sup {prime}}, are discussed. It is shown that the distortion of the flux surfaces by the Shafranov shift, which modifies the geometry metric elements, stabilizes the tearing mode significantly, even in a low-{beta} regime before the toroidal magnetic curvature effects come into play. {copyright} {ital 1998 American Institute of Physics.}
Decommissioning of the Tokamak Fusion Test Reactor
E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola
2003-10-28
The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.
Cooldown of the Compact Ignition Tokamak
Keeton, D.C.
1987-08-01
Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs.
Diagnostics modules for tokamak disruption experiments
Nahm, M.L.; Buchanan, C.D.; Bourham, M.A.; Gilligan, J.G.
1994-11-01
Diagnostic modules equipped with various sensors can provide useful information on key parameters for disruption events, e.g. energy deposition, vapor shielding effect, plasma pressure and force distribution. The modules are, basically, DIMES samples (Divertor Materials Evaluation System) equipped with sensors, coupled to digitizing units and interfaced to a data acquisition system. The DIMES samples are part of the lower diverter diagnostics on the DIII-D tokamak. Three top-cap prototype diagnostics modules have been designed and fabricated. The initial testing and calibration have been performed using the SIRENS plasma gun at an energy deposition of 1 to 12 MJ/m{sup 2} over 0.1 to 1.0 ms, with a plasma pressure >100 MPa.
Passive runaway electron suppression in tokamak disruptions
Smith, H. M.; Helander, P.
2013-07-15
Runaway electrons created in disruptions pose a serious problem for tokamaks with large current. It would be desirable to have a runaway electron suppression method which is passive, i.e., a method that does not rely on an uncertain disruption prediction system. One option is to let the large electric field inherent in the disruption drive helical currents in the wall. This would create ergodic regions in the plasma and increase the runaway losses. Whether these regions appear at a suitable time and place to affect the formation of the runaway beam depends on disruption parameters, such as electron temperature and density. We find that it is difficult to ergodize the central plasma before a beam of runaway current has formed. However, the ergodic outer region will make the Ohmic current profile contract, which can lead to instabilities that yield large runaway electron losses.
Argonne Plasma Engineering Experiment (APEX) Tokamak
Norem, J.H.; Balka, L.J.; Kulovitz, E.E.; Magill, S.R.; McGhee, D.G.; Moretti, A.; Praeg, W.F.
1981-03-01
The Argonne Plasma Engineering Experiment (APEX) Tokamak was designed to provide hot plasmas for reactor-relevant experiments with rf heating (current drive) and plasma wall experiments, principally in-situ low-Z wall coating and maintenance. The device, sized to produce energetic plasmas at minimum cost, is small (R = 51 cm, r = 15 cm) but capable of high currents (100 kA) and long pulse durations (100 ms). A design using an iron central core with no return legs, pure tension tapewound toroidal field coils, digital radial position control, and UHV vacuum technology was used. Diagnostics include monochrometers, x-ray detectors, and a microwave interferometer and radiometer for density and temperature measurements. Stable 100 ms shots were produced with electron temperatures in the range 500 to 1000 eV. Initial results included studies of thermal desorption and recoating of wall materials.
Toroidal microinstability studies of high temperature tokamaks
Rewoldt, G.; Tang, W.M.
1989-07-01
Results from comprehensive kinetic microinstability calculations are presented showing the effects of toroidicity on the ion temperature gradient mode and its relationship to the trapped-electron mode in high-temperature tokamak plasmas. The corresponding particle and energy fluxes have also been computed. It is found that, although drift-type microinstabilities persist over a wide range of values of the ion temperature gradient parameter /eta//sub i/ /equivalent to/ (dlnT/sub i//dr)/(dlnn/sub i//dr), the characteristic features of the dominant mode are those of the /eta//sub i/-type instability when /eta//sub i/ > /eta//sub ic/ /approximately/1.2 to 1.4 and of the trapped-electron mode when /eta//sub i/ < /eta//sub ic/. 16 refs., 7 figs.
Control of Asymmetric Magnetic Perturbations in Tokamaks
Park, Jong-kyu; Schaffer, Michael J.; Menard, Jonathan E.; Boozer, Allen H.
2007-10-03
The sensitivity of tokamak plasmas to very small deviations from the axisymmetry of the magnetic field |δ→(over)Β/→(over)Β|≈ 10–4 is well known. What was not understood until very recently is the importance of the perturbation to the plasma equilibrium in assessing the effects of externally produced asymmetries in the magnetic field, even far from a stability limit. DIII-D and NSTX experiments find that when the deleterious effects of asymmetries are mitigated, the external asymmetric field was often made stronger and with an increased interaction with the magnetic field of the unperturbed equilibrium fields. This paper explains these counter intuitive results. The explanation using ideal perturbed equilibria has important implications for the control of field errors in all toroidal plasmas.
Nusselt number scaling in tokamak plasma turbulence
Takeda, K.; Benkadda, S.; Hamaguchi, S.; Wakatani, M.
2005-05-15
Anomalous heat transport caused by ion temperature gradient (ITG) driven turbulence in tokamak plasmas is evaluated from numerical simulations of the two-dimensional (2D) partial-differential equations of the ITG model and of a reduced 1D version derived from a quasilinear approximation. In the strongly turbulent state, intermittent bursts of thermal transport are observed in both cases. In the strongly turbulent regime, the reduced model as well as the direct numerical simulation show that the Nusselt number Nu (normalized heat flux) scales with the normalized ion pressure gradient K{sub i} as Nu{proportional_to}K{sub i}{sup 1/3}. Since the Rayleigh number for ITG turbulence is proportional to K{sub i}, the Nusselt number scaling for ITG turbulence is thus similar to the classical thermal transport scaling for Rayleigh-Benard convections in neutral fluids.
Control of asymmetric magnetic perturbations in tokamaks.
Park, Jong-Kyu; Schaffer, Michael J; Menard, Jonathan E; Boozer, Allen H
2007-11-09
The sensitivity of tokamak plasmas to very small deviations from the axisymmetry of the magnetic field |deltaB/B| approximately 10{-4} is well known. What was not understood until very recently is the importance of the perturbation to the plasma equilibrium in assessing the effects of externally produced asymmetries in the magnetic field, even far from a stability limit. DIII-D and NSTX experiments find that when the deleterious effects of asymmetries are mitigated, the external asymmetric field was often made stronger and had an increased interaction with the magnetic field of the unperturbed equilibrium. This Letter explains these counterintuitive results. The explanation using ideal perturbed equilibria has important implications for the control of field errors in all toroidal plasmas.
Transport bifurcation in a rotating tokamak plasma.
Highcock, E G; Barnes, M; Schekochihin, A A; Parra, F I; Roach, C M; Cowley, S C
2010-11-19
The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.
Fast tomographic methods for the tokamak ISTTOK
Carvalho, P. J.; Coelho, R.; Neto, A.; Pereira, T.; Silva, C.; Fernandes, H.; Gori, S.; Toussaint, U. v.
2008-04-07
The achievement of long duration, alternating current discharges on the tokamak IST-TOK requires a real-time plasma position control system. The plasma position determination based on magnetic probes system has been found to be inadequate during the current inversion due to the reduced plasma current. A tomography diagnostic has been therefore installed to supply the required feedback to the control system. Several tomographic methods are available for soft X-ray or bolo-metric tomography, among which the Cormack and Neural networks methods stand out due to their inherent speed of up to 1000 reconstructions per second, with currently available technology. This paper discusses the application of these algorithms on fusion devices while comparing performance and reliability of the results. It has been found that although the Cormack based inversion proved to be faster, the neural networks reconstruction has fewer artifacts and is more accurate.