Activation analysis for JT-60U experiments with deuterium gases

International Nuclear Information System (INIS)

Miya, Naoyuki

1993-11-01

Identification of radionuclides and evaluation of dose rate level have been made on the structural materials of the JT-60U tokamak device. A one-dimensional neutron and gamma-ray transport code ANISN and an induced activation calculation code CINAC are used in this work. Radionuclides of 56 Mn (High-Mn steel toroidal field coil case), 58 Co (Inconel-625 vessel) and 60 Co (SS-316 first wall supporting material) appeared on the structures, which contribute to the dose rate around a vacuum vessel. Cobalt-58 and 60 Co with long half-life time intensely make residual activation in the time of 3 days to 3 months corresponding to the maintenance time after shutdown. The calculated dose rate on the vessel agreed well with the measured data in the first 2 years D-D operations. The one-dimensional code provided a sufficient prediction for the dose rate, although an error due to the toroidal field coil modeling in the calculation is estimated within ∼30%. (author)

The H-mode pedestal, ELMs and TF ripple effects in JT-60U/JET dimensionless identity experiments

International Nuclear Information System (INIS)

Saibene, G.; Oyama, N.; Loennroth, J.; Andrew, Y.; Luna, E. de la; Giroud, C.; Huysmans, G.T.A.; Kamada, Y.; Kempenaars, M.A.H.; Loarte, A.; Donald, D. Mc; Nave, M.M.F.; Meiggs, A.; Parail, V.; Sartori, R.; Sharapov, S.; Stober, J.; Suzuki, T.; Takechi, M.; Toi, K.; Urano, H.

2007-01-01

This paper summarizes results of dimensionless identity experiments in JT-60U and JET, aimed at the comparison of the H-mode pedestal and ELM behaviour in the two devices. Given their similar size, dimensionless matched plasmas are also similar in their dimensional parameters (in particular, the plasma minor radius a is the same in JET and JT-60U). Power and density scans were carried out at two values of I p , providing a q scan (q 95 = 3.1 and 5.1) with fixed (and matched) toroidal field. Contrary to initial expectations, a dimensionless match between the two devices was quite difficult to achieve. In general, p ped in JT-60U is lower than in JET and, at low q, the pedestal pressure of JT-60U with a Type I ELMy edge is matched in JET only in the Type III ELM regime. At q 95 = 5.1, a dimensionless match in ρ*, ν* and β p,ped is obtained with Type I ELMs, but only with low power JET H-modes. These results motivated a closer investigation of experimental conditions in the two devices, to identify possible 'hidden' physics that prevents obtaining a good match of pedestal values over a large range of plasmas parameters. Ripple-induced ion losses of the medium bore plasma used in JT-60U for the similarity experiments are identified as the main difference with JET. The magnitude of the JT-60U ripple losses is sufficient to induce counter-toroidal rotation in co-injected plasma. The influence of ripple losses was demonstrated at q 95 = 5.1: reducing ripple losses by ∼2 (from 4.3 to 1.9 MW) by replacing positive with negative neutral beam injection at approximately constant P in resulted in an increased p ped in JT-60U, providing a good match to full power JET H-modes. At the same time, the counter-toroidal rotation decreased. Physics mechanisms relating ripple losses to pedestal performance are not yet identified, and the possible role of velocity shear in the pedestal stability, as well as the possible influence of ripple on thermal ion transport are briefly

Design study of an AC power supply system in JT-60SA

International Nuclear Information System (INIS)

Shimada, Katsuhiro; Baulaigue, Olivier; Cara, Philippe; Coletti, Alberto; Coletti, Roberto; Matsukawa, Makoto; Terakado, Tsunehisa; Yamauchi, Kunihito

2011-01-01

In the initial research phase of JT-60SA, which is the International Thermonuclear Experimental Reactor (ITER) satellite Tokamak with superconducting toroidal and poloidal magnetic field coils, the plasma heating operation of 30 MW-60 s or 20 MW-100 s is planned for 5.5 MA single null divertor plasmas. To achieve this operation, AC power source of the medium voltage of 18 kV and ∼7 GJ has to be provided in total to the poloidal field coil power supplies and additional heating devices such as neutral beam injection (NBI) and electron cyclotron radio frequency (ECRF). In this paper, the proposed AC power supply system in JT-60SA was estimated from the view point of available power, and harmonic currents based on the standard plasma operation scenario during the initial research phase. This AC power supply system consists of the reused JT-60 power supply facilities including motor generators with flywheel, AC breakers, harmonic filters, etc., to make it cost effective. In addition, the conceptual design of the upgraded AC power supply system for the ultimate heating power of 41 MW-100 s in the extended research phase is also described.

Mechanism of vertical displacement events in JT-60U disruptive discharges

International Nuclear Information System (INIS)

Nakamura, Y.; Yoshino, R.; Neyatani, Y.; Tsunematsu, T.; Azumi, M.; Pomphrey, N.; Jardin, S.C.

1996-01-01

Enhanced vertical displacement events (VDEs), which are frequently observed in JT-60U disruptive discharges, are investigated using the Tokamak Simulation Code (TSC). The rapid plasma current quench can accelerate the vertical displacement, owing to both the up/down asymmetry of the eddy current distribution arising from the asymmetric geometry of the JT-60U vacuum vessel and the degradation of magnetic field decay index n, leading to high growth rates of positional instability. For a slightly elongated configuration (n = -0.9), the asymmetry of attractive forces on the toroidal plasma plays a dominant role in the VDE mechanism. For a more elongated configuration (n = -1.7), the degradation of field decay index n plays an important role on VDEs, in addition to the effect of asymmetric attractive forces. It is shown that the VDE characteristics of a highly elongated configuration with a rapid plasma current quench can be dominated by the field decay index degradation. It is also pointed out that both the softening of current quenches as was experimentally developed in the JT-60U tokamak, and the optimization of the allowable elongation of the plasma cross-section are critical issues in the development of a general control strategy of discharge termination. (author). 21 refs, 10 figs

Analysis of ELM stability with extended MHD models in JET, JT-60U and future JT-60SA tokamak plasmas

Science.gov (United States)

Aiba, N.; Pamela, S.; Honda, M.; Urano, H.; Giroud, C.; Delabie, E.; Frassinetti, L.; Lupelli, I.; Hayashi, N.; Huijsmans, G.; JET Contributors, the; Research Unit, JT-60SA

2018-01-01

The stability with respect to a peeling-ballooning mode (PBM) was investigated numerically with extended MHD simulation codes in JET, JT-60U and future JT-60SA plasmas. The MINERVA-DI code was used to analyze the linear stability, including the effects of rotation and ion diamagnetic drift ({ω }* {{i}}), in JET-ILW and JT-60SA plasmas, and the JOREK code was used to simulate nonlinear dynamics with rotation, viscosity and resistivity in JT-60U plasmas. It was validated quantitatively that the ELM trigger condition in JET-ILW plasmas can be reasonably explained by taking into account both the rotation and {ω }* {{i}} effects in the numerical analysis. When deuterium poloidal rotation is evaluated based on neoclassical theory, an increase in the effective charge of plasma destabilizes the PBM because of an acceleration of rotation and a decrease in {ω }* {{i}}. The difference in the amount of ELM energy loss in JT-60U plasmas rotating in opposite directions was reproduced qualitatively with JOREK. By comparing the ELM affected areas with linear eigenfunctions, it was confirmed that the difference in the linear stability property, due not to the rotation direction but to the plasma density profile, is thought to be responsible for changing the ELM energy loss just after the ELM crash. A predictive study to determine the pedestal profiles in JT-60SA was performed by updating the EPED1 model to include the rotation and {ω }* {{i}} effects in the PBM stability analysis. It was shown that the plasma rotation predicted with the neoclassical toroidal viscosity degrades the pedestal performance by about 10% by destabilizing the PBM, but the pressure pedestal height will be high enough to achieve the target parameters required for the ITER-like shape inductive scenario in JT-60SA.

Status of the cold test facility for the JT-60SA tokamak toroidal field coils

Energy Technology Data Exchange (ETDEWEB)

Abdel Maksoud, Walid, E-mail: walid.abdelmaksoud@cea.fr; Bargueden, Patrick; Bouty, André; Dispau, Gilles; Donati, André; Eppelle, Dominique; Genini, Laurent; Guiho, Patrice; Guihard, Quentin; Joubert, Jean-Michel; Kuster, Olivier; Médioni, Damien; Molinié, Frédéric; Sinanna, Armand; Solenne, Nicolas; Somson, Sébastien; Vieillard, Laurence

2015-10-15

Highlights: • The 5 K cryogenic loop includes a 500 W refrigerator and a She cold pump. • The coils are energized thanks to a 25.7 kA power supply and HTS current leads. • Temperature margin tests between 5 K and 7.5 K will be made on each coil. • A magnet safety system protects each double pancake of the coil in case of quench. • Instrumentation is monitored on a 1 Hz to 10 kHz fast acquisition system. - Abstract: JT-60SA is a fusion experiment which is jointly constructed by Japan and Europe and which shall contribute to the early realization of fusion energy, by providing support to the operation of ITER, and by addressing key physics issues for ITER and DEMO. In order to achieve these goals, the existing JT-60U experiment will be upgraded to JT-60SA by using superconducting coils. The 18 TF coils of the JT-60SA device will be provided by European industry and tested in a Cold Test Facility (CTF) at CEA Saclay. The coils will be tested at the nominal current of 25.7 kA and will be cooled with supercritical helium between 5 K and 7.5 K to check the temperature margin against a quench. The main objective of these tests is to check the TF coils performance and hence mitigate the fabrication risks. The most important components of the facility are: a 11.5 m × 6.5 m large cryostat in which the TF coils will be thermally insulated by vacuum; a 500 W helium refrigerator and a valve box to cool the coils down to 5 K and circulate 24 g/s of supercritical helium through the winding pack and through the casing; a power supply and HTS current leads to energize the coil; the control and instrumentation equipment (sensors, PLC's, supervision system, fast data acquisition system, etc.) and the Magnet Safety System (MSS) that protects the coils in case of quench. The paper will give an overview of the design of this large facility and the status of its realization.

Tritium distribution on plasma facing graphite tiles of JT-60U

International Nuclear Information System (INIS)

Tanabe, T.; Sugiyama, K.; Masaki, K.; Gotoh, Y.; Tobita, K.; Miya, N.

2003-01-01

Tritium distributions on the graphite divertor tiles, the dome units and the baffle plates of JT-60U were successfully measured. Poloidally, the highest tritium level was found at the dome top tiles and the outer baffle plates, where the plasma did not hit directly. On the other hand, although the toroidal tritium profiles on each tiles appeared uniform, detailed profiles in full toroidal direction clearly showed a periodic variation corresponding to the position of the magnetic field coils, indicating the ripple loss of high energy tritons as suggested by the OFMC code. Finally, the temperature increase owing to the plasma heat load was found to release the once retained tritium. (author)

Disruption Studies in JT-60U

International Nuclear Information System (INIS)

Kawano, Y.; Yoshino, R.; Neyatani, Y.; Nakamura, Y.; Tokuda, S.; Tamai, H.

2002-01-01

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

Technical aspects and manufacturing methods for JT-60SA toroidal field coil casings

International Nuclear Information System (INIS)

Rossi, Paolo; Cucchiaro, A.; Brolatti, G.; Cocilovo, V.; Ginoulhiac, G.; Polli, G.; Gabriele, M.; Di Muzio, F.; Philips, G.; Tomarchio, V.

2014-01-01

Highlights: • A contract between ENEA and Walter Tosto started on July 2012 for the construction of 18 TF coil casings for JT-60SA. • Design and manufacturing of mock-ups representative of straight and curved legs of the casings have been completed. • Final design of the casings has been completed and manufacturing activities have already started and are ongoing. • The completion of the first three casings will be completed within the end of 2013 and the production of all the 18 casings is foreseen by the end of 2015. - Abstract: JT-60SA is a superconducting tokamak machine to be assembled in Naka site, Japan, designed to contribute to the early realization of fusion energy by supporting the exploitation of ITER and research toward DEMO. In the frame of the Broader Approach Agreement a contract between ENEA and Walter Tosto (Chieti, Italy) started on July 2012 for the construction of 18 TF coil casings for JT-60SA. Two different sets of 9 casings each will be progressively delivered, from 2013 to the end of 2015, to ASG Superconductors (Genoa, Italy) and to Alstom (Belfort, France), where the integration of the winding pack into the casing will be carried out. Each TF coil casing (height 7.5 m and width 4.5 m) consists of four main components: one “Straight Leg Outboard” and one “Curved Leg Outboard” both with their own covers, “Straight Leg Inboard” and “Curved Leg Inboard”. The casing components are segmented in forgings and plates made of FM316LNL. The straight leg outboard is composed of two wings welded to a central core and two elbows welded at the ends with a cooling channel installed inside. Elbows of straight leg outboard are segmented in two half-elbows machined from 1 rough forging and welded to the central core made by plate. Welding of wings to the central core is performed in EBW (electron beam welding) and the straight part is welded to the elbows by NGTIG (TIG narrow gap) process. The curved leg outboard is composed of two

Technical aspects and manufacturing methods for JT-60SA toroidal field coil casings

Energy Technology Data Exchange (ETDEWEB)

Rossi, Paolo, E-mail: paolo.rossi@enea.it [ENEA, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Cucchiaro, A.; Brolatti, G.; Cocilovo, V.; Ginoulhiac, G.; Polli, G. [ENEA, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Gabriele, M.; Di Muzio, F. [Walter Tosto, Via Erasmo Piaggio, 66100 Chieti (Italy); Philips, G.; Tomarchio, V. [JT-60SA European Home Team, Boltzmannstrasse 2, D-85748 Garching (Germany)

2014-10-15

Highlights: • A contract between ENEA and Walter Tosto started on July 2012 for the construction of 18 TF coil casings for JT-60SA. • Design and manufacturing of mock-ups representative of straight and curved legs of the casings have been completed. • Final design of the casings has been completed and manufacturing activities have already started and are ongoing. • The completion of the first three casings will be completed within the end of 2013 and the production of all the 18 casings is foreseen by the end of 2015. - Abstract: JT-60SA is a superconducting tokamak machine to be assembled in Naka site, Japan, designed to contribute to the early realization of fusion energy by supporting the exploitation of ITER and research toward DEMO. In the frame of the Broader Approach Agreement a contract between ENEA and Walter Tosto (Chieti, Italy) started on July 2012 for the construction of 18 TF coil casings for JT-60SA. Two different sets of 9 casings each will be progressively delivered, from 2013 to the end of 2015, to ASG Superconductors (Genoa, Italy) and to Alstom (Belfort, France), where the integration of the winding pack into the casing will be carried out. Each TF coil casing (height 7.5 m and width 4.5 m) consists of four main components: one “Straight Leg Outboard” and one “Curved Leg Outboard” both with their own covers, “Straight Leg Inboard” and “Curved Leg Inboard”. The casing components are segmented in forgings and plates made of FM316LNL. The straight leg outboard is composed of two wings welded to a central core and two elbows welded at the ends with a cooling channel installed inside. Elbows of straight leg outboard are segmented in two half-elbows machined from 1 rough forging and welded to the central core made by plate. Welding of wings to the central core is performed in EBW (electron beam welding) and the straight part is welded to the elbows by NGTIG (TIG narrow gap) process. The curved leg outboard is composed of two

Development of magnetic sensors for JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Takechi, M., E-mail: takechi.manabu@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Matsunaga, G.; Sakurai, S.; Sasajima, T.; Yagyu, J.; Hoshi, R.; Kawamata, Y.; Kurihara, K. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Nishikawa, T.; Ryo, T.; Kagamihara, S. [Okazaki Manufacturing Company, Kobe, Hyogo 651-0087 (Japan); Nakamura, K. [RIAM, Kyushu Univ., Kasuga, Fukuoka 816-8580,Japan (Japan)

2015-10-15

JT-60SA has been designed and is being constructed to demonstrate and develop steady-state high-beta operation. Resistive wall mode (RWM) control, error field correction, and edge-localized mode (ELM) control will be performed using in-vessel coils. For these controls, we have developed a biaxial magnetic sensor to determine 3D magnetic configuration of the plasma. Moreover, for obtaining basic information about JT-60SA plasma, magnetic sensors, in particular, one-turn loops, Rogowski coils, diamagnetic loops, and saddle coils have been developed. Because the length of the vacuum vessel in the poloidal direction of JT-60SA is 16 m and almost twice as long as that of JT-60U, the length of the Rogowski coil and the diamagnetic loop of JT-60SA are also twice as long as those on JT-60U. We have devised new types of sensors and a connector for the mineral-insulated cable because construction and installation of these sensors are much more difficult in JT-60SA. We will report the design and specification of the magnetic sensors for JT-60SA from the physics and engineering aspects.

Development of a linear motion antenna for the JT-60SA ECRF system

International Nuclear Information System (INIS)

Moriyama, Shinichi; Kobayashi, Takayuki; Isayama, Akihiko; Hoshino, Katsumichi; Suzuki, Sadaaki; Hiranai, Shinichi; Yokokura, Kenji; Sawahata, Masayuki; Terakado, Masayuki; Hinata, Jun; Wada, Kenji; Sato, Yoshikatsu

2013-01-01

Highlights: ► Development of an antenna featuring linear motion (LM) concept for long pulse electron cyclotron range of frequency (ECRF) heating and current drive in JT-60SA is in progress. ► A mock-up using a metallic sliding bearing with solid lubricant was fabricated. ► A vacuum pumping test with mass analyzer showed evidence of some hydrocarbons during shaft motion. ► Injection beam profile in toroidal beam scan was checked by low power measurement with mock-up. ► Current drive characteristics with the LM antenna for typical experimental scenarios of JT-60SA have been investigated by calculation. -- Abstract: Development of an antenna that features the linear motion (LM) concept for long-pulse electron-cyclotron range of frequency heating and current drive for the JT-60SA is in progress. Combining a linearly movable first mirror and a fixed curved second mirror allows the injection-beam angle to be controlled. Cooling water is fed through the drive shaft for the first mirror and through the fixed support for the second mirror. The shaft support structure uses a metallic sliding bearing with a solid lubricant. The sliding bearing supports combined linear and rotational motion, whereas a conventional ball bearing supports either linear or rotational motion. Therefore, the sliding bearing offers the advantage of reducing the support-structure volume, which is important in the design of the relatively narrow port duct of the JT-60SA. Recently, the sliding bearing has been installed into the mockup. Results of a vacuum test with a mass analyzer indicate the presence of hydrocarbons during shaft motion. The injection-beam profile obtained from a toroidal beam scan is checked against low-power measurements taken on the mockup. Finally, for typical JT-60SA experimental scenarios, heating- and current-drive characteristics of the LM antenna are investigated theoretically

ELM frequency dependence on toroidal rotation in the grassy ELM regime in JT-60U

International Nuclear Information System (INIS)

Oyama, N; Kamada, Y; Isayama, A; Urano, H; Koide, Y; Sakamoto, Y; Takechi, M; Asakura, N

2007-01-01

A systematic study of the effect of the level of toroidal plasma rotation at the top of the ion temperature pedestal ( T i ped ) on the edge localised mode (ELM) characteristics in JT-60U has been performed. The level of toroidal plasma rotation was varied by using different combinations of tangential and perpendicular neutral beam injection (NBI). In the grassy ELM regime at high triangularity (δ) and high safety factor (q), the ELM frequency clearly increased up to 1400 Hz, when counter (ctr) plasma rotation was increased. The response of the ELM frequency was independent of poloidal beta (β p ) in the range 0.84 p 0.53. Even in non-rotating plasma with balanced-NBIs, a high ELM frequency of ∼400 Hz was observed without a large energy loss. When the frequency of the plasma rotation in the co-direction of the plasma current became higher than ∼1 kHz, type I ELMs with a frequency of ∼20 Hz was observed. The achieved pedestal pressure and plasma confinement were similar both in plasmas with type I ELMs and in plasmas with grassy ELMs. The energy loss due to grassy ELMs was evaluated from the reduction in the electron temperature, and the ratio of the energy loss to the pedestal stored energy was less than 1%

Design of JT-60SA magnets and associated experimental validations

International Nuclear Information System (INIS)

Zani, L.; Barabaschi, P.; Peyrot, M.; Meunier, L.; Tomarchio, V.; Duglue, D.; Decool, P.; Torre, A.; Marechal, J.L.; Della Corte, A.; Di Zenobio, A.; Muzzi, L.; Cucchiaro, A.; Turtu, S.; Ishida, S.; Yoshida, K.; Tsuchiya, K.; Kizu, K.; Murakami, H.

2011-01-01

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak toward a fully superconducting configuration, the detailed design phase led to adopt for the three main magnet systems a brand new design. Europe (EU) is expected to provide to Japan (JA) the totality of the toroidal field (TF) magnet system, while JA will provide both Equilibrium field (EF) and Central Solenoid (CS) systems. All magnet designs were optimized trough the past years and entered in parallel into extensive experimentally-based phases of concept validation, which came to maturation in the years 2009 and 2010. For this, all magnet systems were investigated by mean of dedicated samples, e.g. conductor and joint samples designed, manufactured and tested at full scale in ad hoc facilities either in EU or in JA. The present paper, after an overall description of magnet systems layouts, presents in a general approach the different experimental campaigns dedicated to qualification design and manufacture processes of either coils, conductors and electrical joints. The main results with the associated analyses are shown and the main conclusions presented, especially regarding their contribution to consolidate the triggering of magnet mass production. The status of respective manufacturing stages in EU and in JA are also evoked. (authors)

Study on usage of fluorocarbon for toroidal field coil cooling

International Nuclear Information System (INIS)

Miyata, Hiroshi; Arai, Takashi

1998-09-01

In JT-60 machine, usage of fluorocarbon as an alternate coolant to a cooling channel of toroidal field coil (TF coil) in which a crack was detected is investigated. Fluorinert (a registered trademark of 3M) liquid which is one of fluorocarbon was reviewed, and liquid 'FC-43' was found as an appropriate one for TF coils cooling because of its physical properties about boiling point and thermal capacity. Fortunately, Fluorinert does not have impact on the greenhouse effect for the earth under the temperature of its boiling point. And thermal analysis shows that the cooling effectiveness obtained with liquid 'FC-43' for TF coils is rather well. Moreover, corrosion tests were carried out between liquid 'FC-43' and materials used in JT-60 by considering deterioration of TF coils. The test results demonstrate that there is no problem in applying liquid 'FC-43' as a coolant to cooling channel of TF coils. Results obtained above conclude that usage of fluorocarbon is one of the effective means to perform further experiments in JT-60. (author)

Qualification and preparatory activities for the manufacturing of 9 TF coils of the JT-60SA magnet

International Nuclear Information System (INIS)

Cucchiaro, Antonio; Polli, Gian Mario; Cocilovo, Valter; Drago, Giovanni; Cuneo, Stefano; Terzi, Franco; Peyrot, Marc; Phillips, Guy; Tomarchio, Valerio

2013-01-01

In the framework of the Broader Approach Agreement for the construction of the JT-60SA tokamak, ENEA is in charge to provide 9 of the 18 Toroidal Field (TF) coils. The 9 coils are being manufactured by ASG superconductors in Genoa under the supervision of ENEA in collaboration with the JT-60SA European home team. Prior the manufacturing, a preparatory activity has been carried out aimed at designing, constructing and setting-up the relevant components to be realized. In order to get confidence of some special manufacturing process, several qualification activities have been performed. In this paper an overview of the principal milestones reached during the preparatory phase and a description of the qualification activities with relevant test results are presented

Advanced tokamak research with integrated modeling in JT-60 Upgrade

International Nuclear Information System (INIS)

Hayashi, N.

2010-01-01

Researches on advanced tokamak (AT) have progressed with integrated modeling in JT-60 Upgrade [N. Oyama et al., Nucl. Fusion 49, 104007 (2009)]. Based on JT-60U experimental analyses and first principle simulations, new models were developed and integrated into core, rotation, edge/pedestal, and scrape-off-layer (SOL)/divertor codes. The integrated models clarified complex and autonomous features in AT. An integrated core model was implemented to take account of an anomalous radial transport of alpha particles caused by Alfven eigenmodes. It showed the reduction in the fusion gain by the anomalous radial transport and further escape of alpha particles. Integrated rotation model showed mechanisms of rotation driven by the magnetic-field-ripple loss of fast ions and the charge separation due to fast-ion drift. An inward pinch model of high-Z impurity due to the atomic process was developed and indicated that the pinch velocity increases with the toroidal rotation. Integrated edge/pedestal model clarified causes of collisionality dependence of energy loss due to the edge localized mode and the enhancement of energy loss by steepening a core pressure gradient just inside the pedestal top. An ideal magnetohydrodynamics stability code was developed to take account of toroidal rotation and clarified a destabilizing effect of rotation on the pedestal. Integrated SOL/divertor model clarified a mechanism of X-point multifaceted asymmetric radiation from edge. A model of the SOL flow driven by core particle orbits which partially enter the SOL was developed by introducing the ion-orbit-induced flow to fluid equations.

Magnetic field measurements of JT-60SA CS model coil

Energy Technology Data Exchange (ETDEWEB)

Obana, Tetsuhiro, E-mail: obana.tetsuhiro@LHD.nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Takahata, Kazuya; Hamaguchi, Shinji; Chikaraishi, Hirotaka; Mito, Toshiyuki; Imagawa, Shinsaku [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Kizu, Kaname; Murakami, Haruyuki; Natsume, Kyohei; Yoshida, Kiyoshi [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2015-01-15

Highlights: • Magnetic fields of the JT-60SA CS model coil were measured. • While the coil current was held constant at 20 kA, magnetic fields varied slightly with several different long time constants. • We investigated coils consisting of CIC conductors and having long time constants. - Abstract: In a cold test of the JT-60SA CS model coil, which has a quad-pancake configuration consisting of a Nb{sub 3}Sn cable-in-conduit (CIC) conductor, magnetic fields were measured using Hall sensors. For a holding coil current of 20 kA, measured magnetic fields varied slightly with long time constants in the range 17–571 s, which was much longer than the time constant derived from a measurement using a short straight sample. To validate the measurements, the magnetic fields of the model coil were calculated using a computational model representing the positions of Nb{sub 3}Sn strands inside the CIC conductor. The calculated results were in good agreement with the measurements. Consequently, the validity of the magnetic field measurements was confirmed. Next, we investigated other coils consisting of CIC conductors and having long time constants. The only commonality among the coils was the use of CIC conductors. At present, there is no obvious way to prevent generation of such magnetic-field variations with long time constants.

Conceptual design of divertor cassette handling by remote handling system for JT-60SA

International Nuclear Information System (INIS)

Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

2007-01-01

The JT-60SA aims to contribute and supplement ITER toward DEMO reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is prohibited. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor modules. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor module, which is 10 degrees wide in toroidal direction and weighs 500kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor module to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the module by a pallet installed from outside the VV. (author)

Conceptual design of divertor cassette handling by remote handling system of JT-60SA

International Nuclear Information System (INIS)

Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

2008-01-01

The JT-60SA aims to contribute and supplement ITER toward demonstration fusion reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is restricted. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor cassettes. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor cassette, which is 10 degrees wide in toroidal direction and weighs 500 kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor cassette to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the cassette by a pallet installed from outside the VV. (author)

Development of JT-60 diagnostics system

International Nuclear Information System (INIS)

Suzuki, Yasuo

1988-01-01

The various kinds of plasma diagnostics have been developed and utilized in the JT-60 experiments. The features of JT-60 diagnostics system and the historical proceeding of the development are described in this paper. Taking account of the design consideration, JT-60 diagnostics system is sorted out into eight groups, which include six diagnostics systems, the data processing system and diagnostics supporting system. The all devices in the JT-60 diagnostics system were instrumented on schedule in the end of the fiscal year of 1985 and have contributed to JT-60 experiments. (author)

Control of divertor configuration in JT-60

International Nuclear Information System (INIS)

Yoshino, R.; Kukuchi, M.; Ninomiya, H.; Yoshida, H.; Tsuji, S.; Hosogane, N.; Seki, S.

1985-01-01

The control algorithm of JT-60 divertor configuration is presented. JT-60 has five types of poloidal magnetic field coil with each power supply in order to regulate the control objectives mentioned above. However, if one controls each objective by each coil current independently, there must inevitably occur large interaction between control objectives. Because the relation between control objectives and coil currents is complicated. This situation may be the same with a fusion reactor device. For making it possible to control each objective independently without causing large interaction, the authors adopt the noninteracting control algorithm. Hence, this report demonstrates the availability of this method to the control of JT-60 divertor configuration

Validation of special processes for the integration activities of the JT-60SA TF coils manufactured in Italy

Energy Technology Data Exchange (ETDEWEB)

Polli, Gian Mario, E-mail: gianmario.polli@enea.it [ENEA, UT-FUS, Via E. Fermi 45, Frascati (Italy); Cucchiaro, Antonio; Cocilovo, Valter [ENEA, UT-FUS, Via E. Fermi 45, Frascati (Italy); Drago, Giovanni; Pesenti, Paolo; Cuneo, Stefano; Terzi, Franco [ASG Superconductors, Corso Perrone 73 r, Genova (Italy); Phillips, Guy; Tomarchio, Valerio [JT-60SA European Home Team, 85748 Garching bei Munchen (Germany)

2015-10-15

Highlights: • Insertion. • Casing welding. • Casing embedding. - Abstract: In the framework of the Broader Approach Agreement for the construction of the JT-60SA tokamak, ENEA provides 9 of the 18 toroidal field (TF) coils of the JT-60SA magnet system. The 9 coils are being manufactured by ASG superconductors in Genoa under the supervision of ENEA in collaboration with the JT-60SA European home team. The manufacturing is composed of two main steps: one concerning winding pack assembly and impregnation, and the other devoted to the integration into the casing structure and associated final coil preparation. This paper describes the results of the validation activities set-up for the integration phase. Specifically, welding of casing components has been retained particularly critical for at least three reasons: (i) during welding the WP may be damaged by the intense heating; (ii) distortion caused by heating may determine incorrect coil geometry and then field errors; and (iii) flaws may reduce structural strength and then the overall lifetime of the machine. Similarly, final embedding has been demonstrated on a 1 m long mock-up of the coil. Main results and lessons learned are here described.

Progress in JT-60 innovative technologies

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-09-01

This review report provides the synthetic archives of innovative technologies in 20-year facility developments for the large tokamak experimental device JT-60, first founded as the major magnetic fusion device in the Second Basic Program for Fusion Research and Development of Japan. Manufacture of JT-60 was started in 1978, and the first plasma was achieved on April 8, 1985. In 1989-1991, the vacuum vessel and poloidal field coils were entirely reconfigured to improve the plasma performance. The major original mission of the JT-60 project, a breakeven condition for a D-T equivalent plasma, was finally attained in 1996. After this, JT-60, as a leading device for magnetic fusion research in the world, continues to challenge many experimental issues, which has been achieved by collaboration of innovative facility developments and experimental improvements. In addition, at this time to start the ITER construction phase in 2005, this review is expected to contribute the construction and operation activities for the next generation tokamak by providing the basic ideas in facility developments. We classified a tremendous number of development items into the selected sections for this review. Since the authors have been in charge of each development activity of their own, the contents are full of essential stories, points, and keywords in spite of its compact handbook size. We believe this review could provide highly sophisticated, informative ideas matured in JT-60 technological developments. (author)

Fabrication of the vacuum vessel for JT-60 machine upgrade

International Nuclear Information System (INIS)

Uchikawa, T.; Takanabe, K.; Tsujimura, S.; Ue, K.; Oka, K.; Kuri, S.; Ioki, K.; Namiki, K.; Suzuki, Y.; Horliike, H.; Ninomiya, H.; Yamamoto, M.; Neyatani, Y.; Ando, T.; Matsukawa, M.

1992-01-01

The JT-60 tokamak was upgraded to double the plasma current to 6 MA. In the JT-60 machine upgrade (JT-60U), the vacuum vessel and poloidal field (PF) coils were renewed. The new vacuum vessel features a three-dimensionally curved, thin double-skin torus with multi-arc D-shaped cross section. The double-skin structure is strengthened with square pipes placed in between the outer and inner skins. Fabrication and site installation of the vessel was smoothly completed in February, 1991. This paper describes an overview of the JT-60U vacuum vessel construction

Quench propagation and quench detection in the TF system of JT-60SA

International Nuclear Information System (INIS)

Lacroix, Benoit; Duchateau, Jean-Luc; Meuris, Chantal; Ciazynski, Daniel; Nicollet, Sylvie; Zani, Louis; Polli, Gian-Mario

2013-01-01

Highlights: • The JT-60SA primary quench detection system will be based on voltage measurements. • The early quench propagation was studied in the JT-60SA TF conductor. • The impact of the conductor jacket on the hot spot criterion was quantified. • The detection parameters were investigated for different quench initiations. -- Abstract: In the framework of the JT-60SA project, France and Italy will provide to JAEA 18 Toroidal Field (TF) coils including NbTi cable-in-conduit conductors. During the tokamak operation, these coils could experience a quench, an incidental event corresponding to the irreversible transition from superconducting state to normal resistive state. Starting from a localized disturbance, the normal zone propagates along the conductor and dissipates a large energy due to Joule heating, which can cause irreversible damages. The detection has to be fast enough (a few seconds) to trigger the current discharge, so as to dump the stored magnetic energy into an external resistor. The JT-60SA primary quench detection system will be based on voltage measurements, which are the most rapid technology. The features of the detection system must be adjusted so as to detect the most probable quenches, while avoiding inopportune fast safety discharges. This requires a reliable simulation of the early quench propagation, performed in this study with the Gandalf code. The conductor temperature reached during the current discharge must be kept under a maximal value, according to the hot spot criterion. In the present study, a hot spot criterion temperature of 150 K was taken into account and the role of each conductor component (strands, helium and conduit) was analyzed. The detection parameters were then investigated for different hypotheses regarding the quench initiation

The design study of the JT-60SU device. No. 4. The vacuum vessel and cryostat of JT-60SU

International Nuclear Information System (INIS)

Neyatani, Yuzuru; Ushigusa, Kenkichi; Tobita, Kenji

1997-03-01

The vacuum vessel and the cryostat for the JT-60 Super Upgrade (JT-60SU) have been designed. Two types of the complex materials for the vacuum vessel were chosen on the basis of the avoidance of tritium occlusion and the low irradiation, i.e. (1) SUS316 covered by tungsten plate (30mm thickness) as a γ-ray shielding, (2) Ti-6Al-4V alloy covered by SUS430 plate (1mm thickness) as a tritium protector. Selecting the double skin type of vacuum vessel with toroidally continued structure gave the basic design of the vacuum vessel satisfying the design criteria of the vessel strength for the electromagnetic force, heat load and the property of radiation shielding. The characteristics of the SUS316 covered by tungsten plate type is that as the tungsten can shield the γ-ray, the dose rate inside the vacuum vessel during the maintenance can reduce effectively. The advantage of the Ti-6Al-4V alloy covered by SUS430 plate type vacuum vessel is the quick reduction of the radioactive isotope because of no production of the isotopes with long half-life periods. Channel type and vertical type of the divertor were designed. The sector type of toroidally separated structure was selected for the remote handling. The material of the armor plate was not determined because no material endure the high heat load on the divertor. The cryostat composing the dome and the tank was designed. The electromagnetic force by the eddy current, generated at the plasma start up phase and at the quench of CS super-conducting coil, were small compared to the force produced by the stress limit. (author)

Long-pulse hybrid scenario development in JT-60U

International Nuclear Information System (INIS)

Oyama, N.; Isayama, A.; Matsunaga, G.; Suzuki, T.; Takenaga, H.; Sakamoto, Y.; Nakano, T.; Kamada, Y.; Ide, S.

2009-01-01

The performance and sustained duration of long-pulse discharges for the 'ITER hybrid scenario' have been improved in JT-60U. The modification of power supply systems for three perpendicular neutral beam (NB) injections provides a long period of central NB heating up to 30 s, which is important for keeping the internal transport barrier (ITB). The peaked density profile in the core plasma can be maintained even when the density at the pedestal increased in the latter phase of the discharge due to the increase in the divertor recycling. Then, the peaked pressure profile attributed to the ITB can be kept constant through the discharge with the peaked power deposition profile. In these long-pulse discharges, MHD activity with toroidal mode number n = 1 is observed even when neoclassical tearing modes (NTMs) are suppressed. When the amplitude of the mode in the peripheral region becomes large, the pedestal pressure is degraded. The mode amplitude is sensitive to the toroidal magnetic field (or edge safety factor) and heating power. After the adjustment of the toroidal magnetic field so as to reduce the mode amplitude, a high normalized beta (β N ) of 2.6 and a high thermal confinement enhancement factor (H H98(y,2) > 1) are sustained for 25 s (∼14τ R , where τ R is the current diffusion time) under the ITER relevant small toroidal rotation condition. The peaked pressure profile in low safety factor plasma (safety factor at 95% flux surface q 95 ∼ 3.2) is stable against NTMs up to β N ∼ 3. A high β N H H98(y,2) of 2.6 gives a high G-factor ( β N H H98(y,2) /q 95 2 ) of 0.25 and a peaked pressure profile gives a large bootstrap current fraction (f BS > 0.43).

Ion transport analysis of a high beta-poloidal JT-60U discharge

International Nuclear Information System (INIS)

Horton, W.; Tajima, T.; Dong, J.-Q.; Kim, J.-Y.; Kishimoto, Y.

1997-01-01

The high beta-poloidal discharge number 17110 in JT-60U (JT-60 Team, IAEA, Vienna, 1993) that developes an internal transport barrier is analysed for the transport of ion energy and momentum. First, the classical ion temperature gradient stability properties are calculated in the absence of sheared plasma flows to establish the L-mode transport level prior to the emergence of the transport barrier. Then the evolving toroidal and poloidal velocity profiles reported by Koide et al (1994 Phys. Rev. Lett. 72 3662) are used to show how the sheared mass flows control the stability and transport. Coupled energy-momentum transport equations predict the creation of a transport barrier. The balance of the steep ion temperature gradient against the magnetic shear and sheared mass flow is calculated for the profiles in the 17110 discharge. (Author)

Construction and testing of JT-60

International Nuclear Information System (INIS)

Kishimoto, H.; Aikawa, H.; Oikawa, A.; Miya, N.; Suzuki, K.; Ozeki, T.; Tokutake, T.; Kunieda, S.; Hiruta, K.; Hosoda, R.

1987-01-01

The JT-60 project is reviewed in terms of design, R and D, construction, commissioning and project management. Design features of JT-60 have been refined and renewed through periodic assessments. Engineering targets have been achieved by R and D efforts. Construction and commissioning have progressed on schedule with intensive project management and control. JT-60 obtained high performance and has entered into the experimental phase after completion of machine construction. (orig.)

Characteristics of internal transport barrier in JT-60U reversed shear plasmas

International Nuclear Information System (INIS)

Sakamoto, Y.; Kamada, Y.; Ide, S.; Fujita, T.; Shirai, H.; Takizuka, T.; Koide, Y.; Fukuda, T.; Oikawa, T.; Suzuki, T.; Shinohara, K.; Yoshino, R.

2001-01-01

Characteristics of internal transport barrier (ITB) structure are studied and the active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of the ITB with steep T i gradient is limited to the minimum safety factor location (ρ qmin ). However the ITB with reduced T i gradient can move to the outside of ρ qmin . Lower boundary of ITB width is proportional to the ion poloidal gyroradius at the ITB center. Furthermore the demonstration of the active control of the ITB strength based on the modification of the radial electric field shear profile is successfully performed by the toroidal momentum injection in different directions or the increase of heating power by neutral beams. (author)

JT-60SA vacuum vessel manufacturing and assembly

Energy Technology Data Exchange (ETDEWEB)

Masaki, Kei, E-mail: masaki.kei@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Shibama, Yusuke K.; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

2012-08-15

Highlights: Black-Right-Pointing-Pointer The design of the JT-60SA vacuum vessel body was completed with the demonstration of manufacturing procedure by the mock-up fabrication of the 20 Degree-Sign upper half of VV. Black-Right-Pointing-Pointer The actual VV manufacturing has started since November 2009. Black-Right-Pointing-Pointer The first product of the VV 40 Degree-Sign sector was completed in May 2011. Black-Right-Pointing-Pointer A basic VV assembly scenario and procedure were studied to complete the 360 Degree-Sign VV including positioning method and joint welding. - Abstract: The JT-60SA vacuum vessel (VV) has a D-shaped poloidal cross section and a toroidal configuration with 10 Degree-Sign segmented facets. A double wall structure is adopted to ensure high rigidity at operational load and high toroidal one-turn resistance. The material is 316L stainless steel with low cobalt content (<0.05%). The design temperatures of the VV at plasma operation and baking are 50 Degree-Sign C and 200 Degree-Sign C, respectively. In the double wall, boric-acid water is circulated at plasma operation to reduce the nuclear heating of the superconducting magnets. For baking, nitrogen gas is circulated in the double wall after draining of the boric-acid water. The manufacturing of the VV started in November 2009 after a fundamental welding R and D and a trial manufacturing of 20 Degree-Sign upper half mock-up. The manufacturing of the first VV 40 Degree-Sign sector was completed in May 2011. A basic concept and required jigs of the VV assembly were studied. This paper describes the design and manufacturing of the vacuum vessel. A plan of VV assembly in torus hall is also presented.

Progress in JT-60 joint research

International Nuclear Information System (INIS)

Kimura, Haruyuki; Kikuchi, Mitsuru; Inutake, Masaaki

2007-01-01

It consists of five chapters; 1) introduction, 2) management system of joint plan and researches, 3) progress of joint researches, 4) results of researches and 5) summary. The second chapter stated the structure of management system of JT-60 joint researches, progress of management of the JT-60 experimental theme system, invitation the public to joint researches and selection of the subjects. The progress of joint researches contained the number of subjects, research members and organizations, change of joint research fields, remote control system of experiments, analysis code group, and number of reports. The main results of researches such as development of operation without center solenoid, Magneto-Hydro-Dynamics (MHD) control by electron cyclotron wave, plasma-wall interaction, application of laser technologies to plasma measurement, and comparison between tokamak and helical are reported. (S.Y.)

Surface studies of tungsten erosion and deposition in JT-60U

International Nuclear Information System (INIS)

Ueda, Y.; Fukumoto, M.; Nishikawa, M.; Tanabe, T.; Miya, N.; Arai, T.; Masaki, K.; Ishimoto, Y.; Tsuzuki, K.; Asakura, N.

2007-01-01

In order to study tungsten erosion and migration in JT-60U, 13 W tiles have been installed in the outer divertor region and tungsten deposition on graphite tiles was measured. Dense local tungsten deposition was observed on a CFC tile toroidally adjacent to the W tiles, which resulted from prompt ionization and short range migration of tungsten along field lines. Tungsten deposition with relatively high surface density was found on an inner divertor tile around standard inner strike positions and on an outer wing tile of a dome. On the outer wing tile, tungsten deposition was relatively high compared with carbon deposition. In addition, roughly uniform tungsten depth distribution near the upper edge of the inner divertor tile was observed. This could be due to lift-up of strike point positions in selected 25 shots and tungsten flow in the SOL plasma

Data Processing and Analysis Systems for JT-60U

International Nuclear Information System (INIS)

Matsuda, T.; Totsuka, T.; Tsugita, T.; Oshima, T.; Sakata, S.; Sato, M.; Iwasaki, K.

2002-01-01

The JT-60U data processing system is a large computer complex gradually modernized by utilizing progressive computer and network technology. A main computer using state-of-the-art CMOS technology can handle ∼550 MB of data per discharge. A gigabit ethernet switch with FDDI ports has been introduced to cope with the increase of handling data. Workstation systems with VMEbus serial highway drivers for CAMAC have been developed and used to replace many minicomputer systems. VMEbus-based fast data acquisition systems have also been developed to enlarge and replace a minicomputer system for mass data.The JT-60U data analysis system is composed of a JT-60U database server and a JT-60U analysis server, which are distributed UNIX servers. The experimental database is stored in the 1TB RAID disk of the JT-60U database server and is composed of ZENKEI and diagnostic databases. Various data analysis tools are available on the JT-60U analysis server. For the remote collaboration, technical features of the data analysis system have been applied to the computer system to access JT-60U data via the Internet. Remote participation in JT-60U experiments has been successfully conducted since 1996

Performance verification tests of JT-60SA CS model coil

Energy Technology Data Exchange (ETDEWEB)

Obana, Tetsuhiro, E-mail: obana.tetsuhiro@LHD.nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Murakami, Haruyuki [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Takahata, Kazuya; Hamaguchi, Shinji; Chikaraishi, Hirotaka; Mito, Toshiyuki; Imagawa, Shinsaku [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Kizu, Kaname; Natsume, Kyohei; Yoshida, Kiyoshi [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2015-11-15

Highlights: • The performance of the JT-60SA CS model coil was verified. • The CS model coil comprised a quad-pancake wound with a Nb{sub 3}Sn CIC conductor. • The CS model coil met the design requirements. - Abstract: As a final check of the coil manufacturing method of the JT-60 Super Advanced (JT-60SA) central solenoid (CS), we verified the performance of a CS model coil. The model coil comprised a quad-pancake wound with a Nb{sub 3}Sn cable-in-conduit conductor. Measurements of the critical current, joint resistance, pressure drop, and magnetic field were conducted in the verification tests. In the critical-current measurement, the critical current of the model coil coincided with the estimation derived from a strain of −0.62% for the Nb{sub 3}Sn strands. As a result, critical-current degradation caused by the coil manufacturing process was not observed. The results of the performance verification tests indicate that the model coil met the design requirements. Consequently, the manufacturing process of the JT-60SA CS was established.

Operation and Development on the Positive-Ion Based Neutral Beam Injection System for JT-60 and JT-60U

International Nuclear Information System (INIS)

Kuriyama, M.; Akino, N.; Ebisawa, N.; Honda, A.; Itoh, T.; Kawai, M.; Mogaki, K.; Ohga, T.; Oohara, H.; Umeda, N.; Usui, K.; Yamamoto, M.; Yamamoto, T.; Matsuoka, M.

2002-01-01

The positive-ion based neutral beam injection (NBI) system for JT-60, which consists of 14 beamline units and has a beam energy of 70 to 100 keV, started operation in 1986 with hydrogen beams and injected a neutral beam power of 27 MW at 75 keV into the JT-60 plasma. In 1991, the NBI system was modified to be able to handle deuterium beams as part of the JT-60 upgrade modification. After executing some research and developments, deuterium beams of 40 MW at 95 keV were injected in 1996. As a result, NBI has contributed to the achievement of the highest performance plasmas, a DT-equivalent fusion power gain of 1.25 and a fusion triple product of 1.55 x 10 21 keVs/m 3 , in the world on JT-60U

Assembly work and transport of JT-60SA cryostat base

International Nuclear Information System (INIS)

Okano, Fuminori; Masaki, Kei; Yagyu, Jun-ichi; Shibama, Yusuke; Sakasai, Akira; Miyo, Yasuhiko; Kaminaga, Atsushi; Nishiyama, Tomokazu; Suzuki, Sadaaki; Nakamura, Shigetoshi; Shibanuma, Kiyoshi

2013-11-01

Japan Atomic Energy Agency started to construct a fully superconducting tokamak experiment device, JT-60SA, to support the ITER since January, 2013 at the Fusion Research and Development Directorate in Naka, Japan. The JT-60SA will be constructed with enhancing the previous JT-60 infrastructures, in the JT-60 torus hall, where the ex-JT-60 machine was disassembled. The JT-60SA Cryostat Base, for base of the entire tokamak structure, were assembly as the first step of this construction. The Cryostat Base (CB, 250tons) is consists of 7 main components made of stainless steel, in 12 m diameter and 3 m height. The CB was built in the Spain and transported to the Naka site, via Hitachi port. After pre-assembly work including preliminary measurements and sole plate adjustments of its height/flatness, the JT-60SA CB was carefully set on the sole plate. JT-60SA CB was assembled with high accuracy by using a laser tracker. The CB was adjusted in the height and flatness against the assembly reference position and determined by the absolute coordinates. This report introduces the concrete result of assembly work and transport of JT-60SA CB. (author)

User's manual of JT-60 experimental data analysis system

International Nuclear Information System (INIS)

Hirayama, Takashi; Morishima, Soichi; Yoshioka, Yuji

2010-02-01

In the Japan Atomic Energy Agency Naka Fusion Institute, a lot of experiments have been conducted by using the large tokamak device JT-60 aiming to realize fusion power plant. In order to optimize the JT-60 experiment and to investigate complex characteristics of plasma, JT-60 experimental data analysis system was developed and used for collecting, referring and analyzing the JT-60 experimental data. Main components of the system are a data analysis server and a database server for the analyses and accumulation of the experimental data respectively. Other peripheral devices of the system are magnetic disk units, NAS (Network Attached Storage) device, and a backup tape drive. This is a user's manual of the JT-60 experimental data analysis system. (author)

ERATO-code analysis of vacuum magnetic field oscillations in JT-60 divertor configuration

International Nuclear Information System (INIS)

Ozeki, Takahisa; Tokuda, Shinji; Tsunematsu, Toshihide; Ishida, Shinichi; Neyatani, Yuzuru; Itami, Kiyoshi; Azumi, Masafumi

1989-07-01

Magnetic field oscillations caused by external kink instabilities are numerically studied by using the ideal MHD stability code ERATO-J. Dependence of a spatial distribution of their amplitude and phase on aspect-ratio, beta-poloidal, shaping of conducting shell and divertor/limiter configurations is examined in detail. In the low aspect ratio plasma, the amplitude of magnetic oscillations in the inner side of the torus is larger than that in the outer. On the contrary, as the poloidal beta increases, the amplitude in the outer side of the torus becomes larger than that in the inner. In the divertor configuration, the amplitude of oscillations reduces near the X-point and the phase is locally modulated. The coherent magnetic oscillations observed in JT-60 agree well with the theoretical results, where the vacuum vessel is assumed to be an ideal conducting shell. The non-uniformity of the poloidal distribution observed in JT-60 can be explained by the combined effects of the finite beta, the X-point and the shape of shell. (author)

Baking technique of JT-60

Energy Technology Data Exchange (ETDEWEB)

Arai, Takashi; Masaki, Kei; Miyachi, Kengo [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1998-10-01

It needs to make low ultimate pressure to decrease impurities in a plasma in a nuclear fusion device. Baking technique is very effective method to decrease outgassing rate from first walls and inner surface of the vacuum vessel. However, in such large vacuum vessel, e.g. JT-60, the non-uniform heating and the deformation due to thermal expansion might be very important problems. The baking technique of JT-60 is described. Two baking systems were applied to JT-60 to solve the problem of non-uniform heating. One is a circulation system of the hot nitrogen gas which is located between the inner and outer skins of the double-skin structured vacuum vessel. The other is an electric heater which apply the thick of the vessel. In order to prevent the deformation, the difference of temperature was as small as possible. By both the baking at 300degC and the conditioning such as discharge cleaning, the ultimate pressure was achieved 7.9x10{sup -7} Pa. (author)

Baking technique of JT-60

International Nuclear Information System (INIS)

Arai, Takashi; Masaki, Kei; Miyachi, Kengo

1998-01-01

It needs to make low ultimate pressure to decrease impurities in a plasma in a nuclear fusion device. Baking technique is very effective method to decrease outgassing rate from first walls and inner surface of the vacuum vessel. However, in such large vacuum vessel, e.g. JT-60, the non-uniform heating and the deformation due to thermal expansion might be very important problems. The baking technique of JT-60 is described. Two baking systems were applied to JT-60 to solve the problem of non-uniform heating. One is a circulation system of the hot nitrogen gas which is located between the inner and outer skins of the double-skin structured vacuum vessel. The other is an electric heater which apply the thick of the vessel. In order to prevent the deformation, the difference of temperature was as small as possible. By both the baking at 300degC and the conditioning such as discharge cleaning, the ultimate pressure was achieved 7.9x10 -7 Pa. (author)

Stable existence of central current hole in the JT-60U tokamak

International Nuclear Information System (INIS)

Miura, Y.; Fujita, T.; Oikawa, T.

2003-01-01

In an extreme state of a reversed magnetic shear configuration, it was found in JT-60U that there is almost no plasma current in the central region (called Current Hole). The Current Hole region extends to 40% of the plasma minor radius and it exists stably for several seconds. The Current Hole is formed by the growth of the bootstrap current and it is impossible to drive current in either positive or negative direction by ECH or N-NB inside the Current Hole. In that region, there is almost no gradient of density, temperature and toroidal rotation velocity. It means that there is almost no confinement in the Current Hole and the large energy in that region is sustained only by an internal transport barrier (ITB). The effects of the Current Hole on particle orbits and the effects on an error field on the Current Hole are also discussed. (author)

Low energy, high power injection in JT-60 NBI

International Nuclear Information System (INIS)

Mizuno, Makoto; Dairaku, Masayuki; Horiike, Hiroshi

1988-05-01

JT-60 neutral beam injector (JT-60 NBI) is designed to inject 20 MW neutral hydrogen beam at energies of 70 ∼ 100 keV and the injection power decreases significantly at low energies (∼40 keV). For the extention of operation region aiming at the low density plasma heating and achieving H-mode by plasma periphery heating, increment of the injection power at low beam energies was required. The single-stage acceleration system was investigated in advance at the Prototype Injector Unit. From this result, the total injection power of 17 MW at 40 keV, 48 A per source was expected at the JT-60 NBI. This system was adopted in the JT-60 NBI from June, 1987 to July, 1987 and 17.6 MW neutral beam injection power was achieved. In the NB heating experiment, the H-mode transition phenomena was observed in JT-60 plasma. (author)

Study of grounding system of large tokamak device JT-60

International Nuclear Information System (INIS)

Arakawa, Kiyotsugu; Shimada, Ryuichi; Kishimoto, Hiroshi; Yabuno, Kohei; Ishigaki, Yukio.

1982-01-01

In the critical plasma testing facility JT-60 constructed by the Japan Atomic Energy Research Institute, high voltage, large current is required in an instant. Accordingly, for the protection of human bodies and the equipment, and for realizing the stable operation of the complex, precise control and measurement system, a large scale facility of grounding system is required. In case of the JT-60 experimental facility, the equipments with different functions in separate buildings are connected, therefore, it is an important point to avoid high potential difference between buildings. In the grounding system for the JT-60, a reticulate grounding electrode is laid for each building, and these electrodes are connected with a low impedance metallic duct called grounding trunk line. The power supply cables for various magnetic field coils, control lines and measurement lines are laid in the duct. It is a large problem to grasp quantitatively the effect of a grounding trunk line by analysis. The authors analyzed the phenomenon that large current flows into a grounding system by lightning strike or grounding. The fundamental construction of the grounding system for the JT-60, the condition for the analysis and the result of simulation are reported. (Kako, I.)

A four-pellet pneumatic injection system in the JT-60

International Nuclear Information System (INIS)

Hiratsuka, Hajime; Kawasaki, Kouzo; Miyo, Yasuhiko; Yoshioka, Yuji; Ohta, Kazuya; Shimizu, Masatsugu; Kondo, Ikuo; Onozuka, Masanori; Shimomura, Tomoyoshi; Iwamoto, Syuichi; Hashiri, Noboru

1991-01-01

A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter x 3.0 mm long for two pellets and 4.0 mm in diameter x 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector. (orig.)

A four-pellet pneumatic injection system in the JT-60

Energy Technology Data Exchange (ETDEWEB)

Hiratsuka, Hajime; Kawasaki, Kouzo; Miyo, Yasuhiko; Yoshioka, Yuji; Ohta, Kazuya; Shimizu, Masatsugu; Kondo, Ikuo (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)); Onozuka, Masanori; Shimomura, Tomoyoshi; Iwamoto, Syuichi; Hashiri, Noboru (Mitsubishi Heavy Industries Ltd., Kobe (Japan))

1991-05-01

A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter x 3.0 mm long for two pellets and 4.0 mm in diameter x 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector. (orig.).

Physical design of JT-60 Super Upgrade

International Nuclear Information System (INIS)

Nagashima, K.; Kikuchi, M.; Kurita, G.; Ozeki, T.; Aoyagi, T.; Ushigusa, K.; Neyatani, Y.; Kubo, T.; Mori, K.; Nakagawa, S.; Kuriyama, M.; Nagami, M.

1997-01-01

The JT-60 Super Upgrade (JT-60SU) is an upgraded tokamak device of JT-60U for developing the steady-state reactor and advanced tokamak operation in the International Thermonuclear Experimental Reactor. The device is planned to utilize the JT-60 facilities fully and to minimize the needed modification. The major radius is 4.8 m and the maximum plasma current is 10 MA. Neutral beam injection with 750 keV beam energy is the primary heating method. The machine is capable of steady-state operation with high density up to 8.8 x 10 19 m -3 at 5 MA plasma current. The high operating density, over the Greenwald et al. limit, is critically important in order to achieve high bootstrap current fraction. Ballooning mode and low n ideal magnetohydrodynamic (MHD) mode including the bootstrap current were analyzed for steady-state operation. The current profile must be optimized to obtain a normalized beta up to 3. The plasma configuration with high triangularity was adopted in order to get good MHD stability and high energy confinement. A compact divertor was designed in order to get the large plasma space. (orig.)

Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

Energy Technology Data Exchange (ETDEWEB)

Matsukawa, M. E-mail: matsukaw@naka.jaeri.go.jp; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T

2003-09-01

The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control.

Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

International Nuclear Information System (INIS)

Matsukawa, M.; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T.

2003-01-01

The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control

Data management facility for JT-60

International Nuclear Information System (INIS)

Ohasa, K.; Kurimoto, K.; Mochizuki, O.

1983-01-01

This study considers the Data Management Facility which is provided for unified management of various diagnostics data with JT-60 experiments. This facility is designed for the purpose of data access. There are about 30 kinds of diagnostic devices that are classified by diagnostic objects equipped for JT-60 facility. It gathers the diagnostic date about 10 Mega Byte per each discharge. Those diagnostic data are varied qualitatively and quantitatively by experimental purpose. Other fundamental information like discharge condition, adjustive value for diagnostic devices is required to process those gathered data

Development of multilayer piezoelectric actuator valve for JT-60

International Nuclear Information System (INIS)

Miyo, Yasuhiko; Hiratsuka, Hajime; Masui, Hiroshi; Hosogane, Nobuyuki; Miya, Naoyuki

2001-11-01

In order to improve the gas injection valve used for the operation of JT-60, a new type of valve (multilayer piezoelectric actuator valve) was developed. The conventional valve (bimorph piezoelectric valve) has been used for 15 years since the beginning of experimental operation in April, 1985. However, it came to be hard to keep the performance of the valve because of the deterioration of the driving source, i.e. piezoelectric element. Developments of the new valve were carried out based on experiences through experimental operations in JT-60. Requirements for the design are: (1) to be hard structure for making a sheet leak, (2) to allow a repair work at atmosphere side without an air vent of the vacuum vessel, (3) to be more smaller and lighter compared with the conventional one, and (4) to have a high maintenance efficiency by utilizing of the commercial piezoelectric elements and power supplies. The newly developed valve was examined with various tests such as gas flow characteristic test, high magnetic field proof test, high temperature proof test and gas flow rate test for aged deterioration. Results, confirm that the performance of the valve is applicable for JT-60 operations. (author)

Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

International Nuclear Information System (INIS)

Kawashima, H.; Shimizu, K.; Takizuka, T.

2007-01-01

To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

The archives of operational achievements in JT-60

International Nuclear Information System (INIS)

Seimiya, Munetaka

2007-08-01

Since the first plasma in JT-60 was achieved in April 1985, various experimental challenges have been successfully conducted, and currently producing many new findings. These achievements have been realized by large modifications for lower X-point divertor in 1987, for large plasma current upgrade in 1989-1991, for W-shaped divertor in 1997, and for long pulse discharge in 2002. Such developments contribute to have established JT-60 as the leading tokamak in the world. As a consequence of the 22-year operation, we have accumulated many operational and experimental data. This reports the operational records including troubles and availability, the outline of planning management, the safety control and the promotion procedure of operation in JT-60. (author)

Operation and management manual of JT-60 experimental data analysis system

International Nuclear Information System (INIS)

Hirayama, Takashi; Morishima, Soichi

2014-03-01

In the Japan Atomic Energy Agency Naka Fusion Institute, a lot of experiments have been conducted by using the large tokamak device JT-60 aiming to realize fusion power plant. In order to optimize the JT-60 experiment and to investigate complex characteristics of plasma, JT-60 experimental data analysis system was developed and used for collecting, referring and analyzing the JT-60 experimental data. Main components of the system are a data analysis server and a database server for the analyses and accumulation of the experimental data respectively. Other peripheral devices of the system are magnetic disk units, NAS (Network Attached Storage) device, and a backup tape drive. This is an operation and management manual the JT-60 experimental data analysis system. (author)

Simple multijunction launcher with oversized waveguides for lower hybrid current drive on JT-60U

International Nuclear Information System (INIS)

Ikeda, Y.; Naito, O.; Seki, M.; Kondoh, T.; Ide, S.; Anno, K.; Fukuda, H.; Ikeda, Y.; Kitai, T.; Kiyono, K.; Sawahata, M.; Shinozaki, S.; Suganuma, K.; Suzuki, N.; Ushigusa, K.

1994-01-01

A multijunction technique with oversized waveguides has been developed for the lower hybrid current drive launcher on JT-60U. The launcher consists of 4 (toroidal)x4 (poloidal) multijunction modules. RF power in the module is divided toroidally into 12 sub-waveguides at a junction point through an oversized waveguide. This method simplifies the structure of the multijunction launcher with a large number of subwaveguides. A maximum power density up to 25 MW m -2 has been achieved with a low reflection coefficient of less than 2%. The coupling and current drive efficiency are well explained by the designed wave spectra without taking account of higher modes in the oversize waveguides. Thus, the simple multijunction launcher has been demonstrated to excite expected wave spectra with high power handling capability. ((orig.))

Heating experiments of JT-60

International Nuclear Information System (INIS)

1987-01-01

In JT-60, after the finish of the first stage Joule experiment, the heating facilities were installed, and the heating experiment was started in August, 1986. As to neutral beam injection, the beam injection experiment at the maximum rating 20 MW carried out, and also as to RF, the injection experiment up to 1.4 MW was carried out in both ion cyclotron and low band hybrid waves. The results worthy of special mention in the heating experiment were the success in the current drive up to 1.7 MA at maximum using low band hybrid waves and the improvement of plasma confinement characteristics obtained by the compound heating of NBI and RF. In this paper, the main results of these heating experiments and their significance are explained. The JT-60 is the testing facilities for attaining the critical plasma condition by additionally heating the plasma which is generated by Joule electric discharge with NBI and RF heatings. The experimental operation cycle of the JT-60 consists of the unit cycle of two weeks, and the number of days in operation is nine days. The temperature of heated plasma rose to 70 million deg C in the 20 MW NBI heating. Hereafter, the improvement of confinement time by increasing the stored energy of plasma is attempted. (Kako, I.)

Discharge optimization in JT-60

International Nuclear Information System (INIS)

Ninomiya, H.; Hosogane, N.; Kikuchi, M.; Yoshino, R.; Seki, S.; Kurihara, K.; Kimura, T.; Shimada, R.; Matsukawa, M.

1986-01-01

For the optimization of the feedback control gains of the plasma control system in JT-60, the plasma modelling by the regression analysis, the matrix transfer function analysis and the simulation study are performed. The experimental results of plasma control are well consistent with these estimations and the usefulness of a modelling by the regression analysis, the matrix transfer function analysis and the simulation study is experimentally confirmed. It is also shown that the regression analysis is useful for development of the sensor algorithm of plasma shape and location of separatrix line in a feedback control system. Some topics are also presented about plasma engineering obtained in JT-60: possibility to suppress the uncontrollability of plasma density, αI/sub p/ control for plasma position and volt-sec consumption

Development of the pellet injector for JT-60

International Nuclear Information System (INIS)

Kawasaki, Kouzo; Hiratsuka, Hajimo; Takatsu, Hideyuki; Shimizu, Masatsugu; Onozuka, Masanori; Uchikawa, Takashi; Iwamoto, Syuichi; Hashiri, Nobuo

1989-01-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proved that the device provides high speed hydrogen pellets just as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 1.6 km/sec at 50 bar propellant gas. The device is now in use for JT-60 contributing to plasma study. In this paper the outline of features and performance of the device is presented. (author). 4 refs.; 8 figs

Conceptual design of a new supervisory control system for JT-60SA

International Nuclear Information System (INIS)

Totsuka, Toshiyuki; Sakata, Shinya

2009-05-01

The functions of JT-60 discharge control computer system and the data processing computer system will be integrated into a new JT-60SA supervisory control system to improve the operational efficiency of the JT-60 control computer system. In this report, we first show the necessary requirements for the new JT-60SA supervisory control system that should have high cost performance and maintainability. Next, overall system image of the new JT-60SA supervisory control system is presented and the necessary functions and the issues to be solved in the development are shown. Finally, the necessary manpower for this development and performance of the computer hardware, and the expected reduction of maintenance cost of the computer system are described. (author)

Resistive instabilities in reversed shear discharges and wall stabilization on JT-60U

International Nuclear Information System (INIS)

Takeji, S.; Tokuda, S.; Fujita, T.; Suzuki, T.; Isayama, A.; Ide, S.; Ishii, Y.; Kamada, Y.; Koide, Y.; Matsumoto, T.; Oikawa, T.; Ozeki, T.; Sakamoto, Y.

2001-01-01

Resistive instabilities and wall stabilization of ideal low toroidal mode number, n, kink modes are investigated in JT-60U reversed shear discharges. Resistive interchange modes with n=1 are found to appear in reversed shear discharges with large pressure gradient at the normalized beta, β N , of about unity or even lower. The resistive interchange modes appear as intermittent burst-like magnetohydrodynamic (MHD) activities and higher n≤3 modes are observed occasionally in higher β N regime. No clear degradation of the plasma stored energy is observed by the resistive interchange modes themselves. It is also found that resistive interchange modes can lead to major collapse owing to a coupling with tearing modes at the outer mode rational surface over the minimum safety factor. Stability analysis revealed that stability parameter of tearing modes, Δ' , at the outer mode rational surface is affected by the free-boundary condition. The result is consistent with the experimental evidence that major collapse tends to occur when plasma edge safety factor, q*, is near integer values. Stabilization of ideal low n kink modes by the JT-60U wall is demonstrated. Magnetohydrodynamic perturbations that are attributed to resistive wall modes are observed followed by major collapse in wall-stabilized discharges. (author)

Development of fast opening magnetic valve for JT-60 pellet injector

International Nuclear Information System (INIS)

Hiratsuka, Hajime; Kawasaki, Kouzo; Takatsu, Hideyuki; Miyo, Yasuhiko; Yoshioka, Yuji; Ohta, Kazuya; Shimizu, Masatsugu; Onozuka, Masanori; Uchikawa, Takashi; Iwamoto, Syuichi; Hashiri, Noboru

1989-01-01

A pneumatic four-pellet injector (JT-60 pellet injector) has been constructed for JT-60 in May, 1988. A fast opening magnetically driven propellant gas injection valve has been developed for JT-60 pellet injector. This valve can accelerate four cylindrical pellets, two 3.8 mm diameter by 3.8 mm and two 2.7 mm diameter by 2.7 mm, to greater than 1.6 km/s with propellent gas of up to 50 bar. It is now successfully in use in JT-60, contributing to plasma studies. In this paper the outline of a newly developed fast opening magnetic valve and the results of performance tests are presented. (author). 6 figs.; 1 tab

Vacuum leak test technique of JT-60

International Nuclear Information System (INIS)

Kaminaga, Atsushi; Arai, Takashi; Kodama, Kozo; Sasaki, Noboru; Saidoh, Masahiro

1998-01-01

Since a vacuum vessel of JT-60 is very large (167 m 3 ) and is combined with many components, such as magnetic coils, neutral beam injection systems and RF heating systems, etc., the position of leak testing exceeds 700. The two kind of techniques for vacuum leak test used in JT-60 has been described. Firstly the probe helium gas can be fed remotely in the three-dimensionally sectioned 54 regions of the JT-60 torus. The leak test was very rapidly performed by using this method. Secondly the helium detector system has been modified by the additional installation of the cryopump, which reduced the background level of the deuterium gas. The sensitivity of vacuum leak test with the cryopump was two orders of magnitude larger than that of without it. The examples of the performed vacuum leak test are stated. The vacuum leaks during experiments were 9 times. They were caused by thermal strain and plasma discharge. The vacuum leaks just after maintenance are 36 times which mainly caused by mis-installation. (author)

Design study of the vertical field power supply for JT-60

International Nuclear Information System (INIS)

Yabuno, Kohei; Tani, Keiji; Shimada, Ryuichi; Kishimoto, Hiroshi; Yoshida, Hidetoshi

1977-09-01

The results of a basic design study of the vertical field power supply for JT-60 (JAERI large tokamak) are described. The objective of the study is to evaluate several types of power supply circuits for fast excitation and control of the vertical field. A design requirement is to produce a rapidly increasing vertical field within accuracy of +-5% around the proper field strength required to center the plasma in the vacuum vessel. The plasma current is assumed to increase at the rate of about 100 MA/sec. To meet the requirement, a maximum voltage of 15 kV is necessary in the current build-up time, while generally relatively low voltage is necessary after the current flattop is reached. A hybrid power supply which consists of a dc power source (a thyristor converter) and an inductive energy storage system is proposed. The maximum voltage of the dc power source is determined as 4 kV from the voltage required in the current flattop time. This is sufficient also in the current build-up time if the dc power source is used together with the inductive energy storage system. (auth.)

Characteristics of halo current in JT-60U

International Nuclear Information System (INIS)

Neyatani, Y.; Nakamura, Y.; Yoshino, R.; Hatae, T.

1999-01-01

Halo currents and their toroidal peaking factor (TPF) have been measured in JT-60U by Rogowski coil type halo current sensors. The electron temperature in the halo region was around 10 eV at 1 ms before the timing of the maximum halo current. The maximum TPF*I h /I p0 was 0.52 in the operational range of I p = 0.7 ∼ 1.8 MA, B T = 2.2 ∼ 3.5 T, including ITER design parameters of κ > 1.6 and q 95 = 3, which was lower than that of the maximum value of ITER data base (0.75). The magnitude of halo currents tended to decrease with the increase in stored energy just before the energy quench and with the line integrated electron density at the time of the maximum halo current. A termination technique in which the current channel remains stationary was useful to avoid halo current generation. Intense neon gas puffing during the VDE was effective for reducing the halo currents. (author)

Characteristics of halo current in JT-60U

International Nuclear Information System (INIS)

Neyatani, Y.; Nakamura, Y.; Yoshino, R.; Hatae, T.

2001-01-01

Halo currents and their toroidal peaking factor (TPF) have been measured in JT-60U by Rogowski coil type halo current sensors. The electron temperature in the halo region was around 10 eV at 1 ms before the timing of the maximum halo current. The maximum TPF *I h /I p0 was 0.52 in the operational range of I p =0.7∼1.8MA, B T =2.2∼3.5T, including ITER design parameters of κ>1.6 and q 95 =3, which was lower than that of the maximum value of ITER data base (0.75). The magnitude of halo currents tended to decrease with the increase in stored energy just before the energy quench and with the line integrated electron density at the time of the maximum halo current. A termination technique in which the current channel remains stationary was useful to avoid halo current generation. Intense neon gas puffing during the VDE was effective for reducing the halo currents. (author)

Review of JT-60U experimental results from January to October, 1992

International Nuclear Information System (INIS)

1993-03-01

Heating experiments have been carried out in JT-60U with plasma current up to 4 MA, toroidal field up to 4.2 T and neutral beam heating power of 30 MW. H-mode was observed at high toroidal field of 4.2 T. Confinement improvement factor of up to 1.6 over the L-mode scaling has been obtained during continuous ELMy phase at 4.2 T, 2.7 MA and an injection power of 25 MW. Confinement enhancement of 2.2 has been obtained. At plasma current of 3.5 MA, diamagnetic stored energy of 7.7 MJ and energy confinement time of 0.36 s have been obtained. In the discharges with high poloidal beta exhibit, high ion temperatures of 38 keV, electron temperatures of 12 keV, maximum diamagnetic stored energy of 6.1 MJ and neutron yield of 2.8 x 10 16 s -1 were obtained. Schemes to avoid locked modes and disruptions have been checked experimentally. Divertor measurement indicates that high q, high density operation is favorable for divertor heat handling. Quantitative analysis of divertor spectroscopy data indicates that the carbon production rate at the divertor plate is explained by deuterium, oxygen and carbon sputtering. Toroidal ripple loss power of NB-injected fast ions to the first wall was investigated by using an infrared TV camera. Absolute values of the ripple loss power and dependence on plasma parameters were consistent with values calculated by an orbit following Monte-Carlo code. In the LHCD area, the linear dispersion relation and accessibility condition for LH wave were experimentally validated. Furthermore, the power directly lost via high energy electron was shown to scale as the slowing down time. Current drive of 400 kA was realized with tangential NB. ICRF experiments using two new antennas started. Sawtooth stabilization at a high n-bar e /P tot value was achieved. (J.P.N.)

JT-60SA TF magnet industrial manufacturing preparation and qualifications

International Nuclear Information System (INIS)

Decool, P.; Cloez, H.; Gros, G.; Marechal, J.L.; Torre, A.; Verger, J.M.; Nusbaum, M.; Billotte, G.; Crepel, B.; Bourquard, A.; Davis, S.; Phillips, G.

2014-01-01

The general design of the JT-60SA toroidal field system was defined in agreement with all the participants in the project (CEA, ENEA, F4E), the detailed design was issued by the Voluntary Contributors. For the French part including the procurement of 9 of the 18 TF winding packs and their integration in the casings, an industrial contract was signed mid-2011 with Alstom (France). After agreement on manufacturing drawings and QA documentation, the manufacturing process was defined giving the guidelines for the workshop organization and the definition of the required tooling. The critical manufacturing points were identified in the process and, regarding technical requirements, have led to the definition of a set of qualification mockups. They are related to helium inlets, conductor winding and insulation, local conductor bending, electrical joint and terminal areas for the winding pack (WP), as well as winding embedding, case welding, and impregnations for WP integration in the casing. The fabrication processes have been improved and shall be qualified thanks to the manufacture and testing of 12 corresponding mockups. The successful achievement of several key mock-ups gives confidence in the feasibility of the manufacture, and their completion will give the green light to the industrial coils manufacture. (authors)

Accomplishment of JT-60U disassembly work dealing with radioactive components

International Nuclear Information System (INIS)

Ikeda, Yoshitaka

2015-01-01

The upgrade of the JT-60U to the superconducting tokamak 'JT-60SA' has been carried out to contribute the early realization of fusion energy by addressing key physics issues relevant for ITER and DEMO. Disassembly of the JT-60U tokamak was required so as to newly install the JT-60SA torus at the same position in the torus hall. The JT-60U tokamak was featured by the complicated and welded structure against the strong electromagnetic force, and by the radioactivation due to deuterium-deuterium (D-D) reactions of 1.5x10"2"0 (n) in total. Since this work was the first experience of disassembling a large radioactivated fusion device in Japan, careful preparations of disassembly activities, including treatment of the radioactivated materials and safety work, have been made. About 13,000 components with a total weight of more than 5,400 tonnes were removed from the torus hall and stored safely in storage facilities. All disassembly components were stored with recording the data such as dose rate, weight and kind of material, so as to apply the clearance level regulation in future. It was confirmed that the main radioactive material of the disassembly components was the stainless steel and that its dose rate was almost background level (∼0.1 μSv/h) at ∼10 m far from the vacuum vessel. It seems that the disassembly components with background dose level are in the clearance level. The assembly of JT-60SA tokamak has started in January 2013 after this disassembly of the JT-60U tokamak. (author)

Recent results of JT-60U ICRF antenna operation

International Nuclear Information System (INIS)

Fujii, T.; Saigusa, M.; Kimura, H.

1994-01-01

Ion cyclotron range of frequencies (ICRF) heating is one of attractive plasma heating methods for reactor grade tokamaks, because it is quite effective in the wide ranges of plasma density and temperature. An antenna which should inject high power into plasma has been developed intensively because the heating efficiency and the coupling properties depend on its design. The antenna was operated at a small antenna-plasma gap in the JT-60 in out of phase mode, which showed the high heating efficiency to obtain high loading resistance, and similarly to other tokamaks. However, in order to reduce heat load to the antenna from plasma, a wide gap is required in reactor grade tokamaks such as ITER, in which the gap is designed to be 0.15 m in CDA. Two new antennas were fabricated for the JT-60U, which were designed to obtain high loading resistance at a wide gap for (π,0) phasing. The JT-60U ICRF heating system is explained. Also the JT-60U antenna is described. Antenna conditioning has been conducted well in the initial operation period. The phasing mode was set at (π,0) phasing, in which high heating efficiency is expected. The procedure is explained. The coupling and radiation loss properties during ICRF heating are reported. The JT-60U ICRF antennas were conditioned quickly with about 70 shots. The maximum coupled power was 6.4 MW for (π,0) phasing, and the power density was 6.1 MW/m 2 . (K.I.)

Maximum β limited by ideal MHD ballooning instabilites in JT-60

International Nuclear Information System (INIS)

Seki, Shogo; Azumi, Masashi

1986-03-01

Maximum β limited by ideal MHD ballooning instabilities is investigated on divertor configurations in JT-60. Maximum β against ballooning modes in JT-60 has strong dependecy on the distribution of the safety factor over the magnetic surfaces. Maximum β is ∼ 2 % for q 0 = 1.0, while more than 3 % for q 0 = 1.5. These results suggest that the profile control of the safety factor, especially on the magnetic axis, is attractive to the higher β operation in JT-60. (author)

Pneumatic pellet injector for JT-60

International Nuclear Information System (INIS)

Onozuka, Masanori; Hiratsuka, Hajime; Kawasaki, Kouzo.

1990-01-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas. (author)

Pneumatic pellet injector for JT-60

Energy Technology Data Exchange (ETDEWEB)

Onozuka, Masanori (Mitsubishi Heavy Industries Ltd., Tokyo (Japan)); Hiratsuka, Hajime; Kawasaki, Kouzo

1990-11-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas. (author).

Core and edge toroidal rotation study in JT-60U

International Nuclear Information System (INIS)

Yoshida, M.; Sakamoto, Y.; Honda, M.; Kamada, Y.; Takenaga, H.; Oyama, N.; Urano, H.

2012-01-01

The relation between toroidal rotation velocities (V t ) in the core and edge regions is investigated in H-mode plasmas with a small external torque input from the viewpoint of momentum transport. The toroidal rotation velocity in the core region (core-V t ) gradually varies on a timescale of ∼20 ms after a rapid change in the toroidal rotation velocity in the edge region (edge-V t ) at the L–H transition. This timescale of ∼20 ms is consistent with a transport timescale using the momentum diffusivity (χ φ ) and convection velocity (V conv ). In steady state, a linear correlation between the core- and edge-V t is observed in H-mode plasmas when the ion pressure gradient (∇P i ) is small. This relation between core- and edge-V t is also explained by momentum transport. The V t profiles with a large ∇P i are reproduced in the core region of r/a ∼ 0.2–0.7 by adopting a residual stress term 'Π res = α k χ φ ∇P i ' proposed in this paper. Here r/a is the normalized plasma radius and α k1 is a radial constant. Using this formula, V t profiles are reproduced over a wide range of plasma conditions. Parameter dependences of the edge-V t are investigated at a constant ripple loss power, ripple amplitude and plasma current. A reduction in the CTR-rotation is observed with decreasing ion temperature gradient (∇T i ). Here CTR refers to the counter-I P direction.

Localized MHD activity near transport barriers in JT-60U and TFTR

International Nuclear Information System (INIS)

Manickam, J.

2001-01-01

Localized MHD activity observed in JT-60U and TFTR near transport barriers with their associated large pressure gradients is investigated. Stability analysis of equilibria modeling the experiments supports an identification of this MHD as being due to an ideal MHD n=1 instability. The appearance of the instability depends on the local pressure gradient, local shear in the q profile and the proximity of rational surfaces where q∼m/n and m and n are the poloidal and toroidal mode numbers respectively. The mode width is shown to depend on the local value of q, and is larger when q is smaller. In addition the role of the edge current density in coupling the internal mode to the plasma edge and of the energetic particles which can drive fishbone like modes is investigated. (author)

Optical design for divertor Thomson scattering system for JT-60SA

International Nuclear Information System (INIS)

Kajita, Shin; Enokuchi, Akito; Hatae, Takaki; Itami, Kiyoshi; Hamano, Takashi; Kado, Shinichiro; Ohno, Noriyasu; Takeyama, Norihide

2014-01-01

Highlights: •A detailed designing for collection optical system of divertor Thomson scattering system in JT-60SA is conducted. •The assessment of the density and temperature errors of the measurement system is conducted. •It is shown that the measurement could be done with the temperature error of 50% when the density was 10 20 m −3 . •The availability of the laser transmission mirrors for the measurement system is discussed. •Several guidelines to improve the measurement system are discussed. -- Abstract: Optical design for divertor Thomson scattering system in JT-60SA has been conducted. The measurement system will use a Nd:YAG laser at 1064 nm, and scattered photons are collected by a collection optical system. The collection optics consists of primary mirror, secondary mirror, relay optics, and fiber collection optics. The laser transmission mirror and collection optics were designed to be installed in a slender lower port of JT-60SA. The assessment of the measurement errors in temperature was conducted for the designed collection optical system. Because of spatial limitation, the solid angle from the measurement points would be small especially for the measurement points in high field side, and consequently, the temperature errors in the high field side would be considerably large. The effects of several improvements on the error are discussed. Moreover, an assessment for the in-vessel laser transmission metallic mirrors is conducted for the present design

FIR-laser scattering for JT-60

International Nuclear Information System (INIS)

Itagaki, Tokiyoshi; Matoba, Tohru; Funahashi, Akimasa; Suzuki, Yasuo

1977-09-01

An ion Thomson scattering method with far infrared (FIR) laser has been studied for measuring the ion temperature in large tokamak JT-60 to be completed in 1981. Ion Thomson scattering has the advantage of measuring spatial variation of the ion temperature. The ion Thomson scattering in medium tokamak (PLT) and future large tokamak (JET) requires a FIR laser of several megawatts. Research and development of FIR high power pulse lasers with power up to 0.6 MW have proceeded in ion Thomson scattering for future high-temperature tokamaks. The FIR laser power will reach to the desired several megawatts in a few years, so JAERI plans to measure the ion temperature in JT-60 by ion Thomson scattering. A noise source of the ion Thomson scattering with 496 μm-CH 3 F laser is synchrotron radiation of which the power is similar to NEP of the Schottky-barrier diode. However, the synchrotron radiation power is one order smaller than that when a FIR laser is 385 μm-D 2 O laser. The FIR laser power corresponding to a signal to noise ratio of 1 is about 4 MW for CH 3 F laser, and 0.4 MW for D 2 O laser if NEP of the heterodyne mixer is one order less. A FIR laser scattering system for JT-60 should be realized with improvement of FIR laser power, NEP of heterodyne mixer and reduction of synchrotron radiation. (auth.)

Development of upgraded pellet injector for JT-60

International Nuclear Information System (INIS)

Onozuka, M.; Shimomura, T.; Tanaka, N.; Iwamoto, S.; Hashiri, N.; Oda, Y.; Minami, M.; Hiratsuka, H.; Kawasaki, K.; Takatsu, H.; Shimizu, M.

1989-01-01

The pneumatic 4-shot pellet injector had been in use for JT-60 (JAERI Tokamak-60) contributing to plasma studies in 1988. It could propel the pellets up to 1.6 km/sec at 50 bar propellant gas. In 1989, the new gun assembly has been reinstalled in the upgraded system to provide higher performance and reliability. The supply pressure of the propellant gas is to be raised to 100 bar to obtain higher pellet velocity up to 2.3 km/sec. The device is now in use for JT-60, and is expected to contribute to further plasma studies. In this paper the outline of features and performance of the device is presented. 5 refs., 9 figs

Prototype ion source for JT-60 neutral beam injectors

International Nuclear Information System (INIS)

Akiba, M.

1981-01-01

A prototype ion source for JT-60 neutral beam injectors has been fabricated and tested. Here, we review the construction of the prototype ion source and report the experimental results about the source characteristics that has been obtained at this time. The prototype ion source is now installed at the prototype unit of JT-60 neutral beam injection units and the demonstration of the performances of the ion source and the prototype unit has just started

Overview of engineering design, manufacturing and assembly of JT-60SA machine

Energy Technology Data Exchange (ETDEWEB)

Di Pietro, Enrico, E-mail: enrico.dipietro@jt60sa.org [JT-60SA EU Home Team, Fusion for Energy, Boltzmannstrasse 2, Garching 85748 (Germany); Barabaschi, Pietro [JT-60SA EU Home Team, Fusion for Energy, Boltzmannstrasse 2, Garching 85748 (Germany); Kamada, Yutaka [JT-60SA JA Home Team, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Ishida, Shinichi [JT-60SA JA Project Team, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2014-10-15

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The JT-60SA device is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I{sub p} = 5.5 MA. This makes JT-60SA capable to support and complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. After a complex start-up phase due to the necessity to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, in 2009 detailed design could start. With the majority of time-critical industrial contracts in place, in 2012, it was possible to establish a credible time plan, and now, the project is progressing on schedule towards the first plasma in March 2019. After careful and focused R and D and qualification tests, the procurement of the major components and plant is now well advanced in manufacturing design and/or fabrication. In the meantime the disassembly of the JT-60U machine has been completed and the engineering of the JT-60SA assembly process has been developed. The actual assembly of JT-60SA started in January 2013 with the installation of the cryostat base. The paper gives an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.

JT-60 plasma control system

International Nuclear Information System (INIS)

Kurihara, K.

1988-01-01

JT-60 plasma control can be performed by the supervisory controller, the measurement system and actuators such as the poloidal field coil power supplies, gas injectors, neutral beam injection (NBI) heating system and radio frequency (RF) heating system. One of the most important characteristics of this system is a perfect digital control one composed of mini-computers, fast array processors and CAMAC modules, and it has large flexibility and few troubles to adjust the system. This system started to be operated in April 1985, after the six-year-long design, construction and testing, and have been operated and improved many times for two years. In this paper, the final system specification and its performance are presented aiming at the technological aspect of hardware and software. In addition, and experienced troubles are also presented. (author)

Plasma equilibrium response modelling and validation on JT-60U

International Nuclear Information System (INIS)

Lister, J.B.; Sharma, A.; Limebeer, D.J.N.; Wainwright, J.P.; Nakamura, Y.; Yoshino, R.

2002-01-01

A systematic procedure to identify the plasma equilibrium response to the poloidal field coil voltages has been applied to the JT-60U tokamak. The required response was predicted with a high accuracy by a state-space model derived from first principles. The ab initio derivation of linearized plasma equilibrium response models is re-examined using an approach standard in analytical mechanics. A symmetric formulation is naturally obtained, removing a previous weakness in such models. RZIP, a rigid current distribution model, is re-derived using this approach and is compared with the new experimental plasma equilibrium response data obtained from Ohmic and neutral beam injection discharges in the JT-60U tokamak. In order to remove any bias from the comparison between modelled and measured plasma responses, the electromagnetic response model without plasma was first carefully tuned against experimental data, using a parametric approach, for which different cost functions for quantifying model agreement were explored. This approach additionally provides new indications of the accuracy to which various plasma parameters are known, and to the ordering of physical effects. Having taken these precautions when tuning the plasmaless model, an empirical estimate of the plasma self-inductance, the plasma resistance and its radial derivative could be established and compared with initial assumptions. Off-line tuning of the JT-60U controller is presented as an example of the improvements which might be obtained by using such a model of the plasma equilibrium response. (author)

Recent developments in the JT-60 data processing system

International Nuclear Information System (INIS)

Matsuda, T.; Saitoh, N.; Tsugita, T.; Oshima, T.; Sakata, S.; Sato, M.; Koiwa, M.; Watanabe, K.

1999-01-01

The JT-60 data processing system was originally a large computer complex system including a lot of micro-computers, several mini-computers, and a mainframe computer. Recently, several improvements have been made to the original system to modernize the system. Many sub-systems composed of aged mini-computers have been replaced with workstations by utilizing recent progress in computer and network technologies. The system can handle ∝300 MB of raw data per discharge, which is three times larger than the amount in the original system. These improvements have been applied to develop element technologies necessary to the remote participation in JT-60 experiments. A remote diagnostics control and monitoring system and a computer system for access to JT-60 data from the Internet are used together with video conferencing systems for a real-time communication. In 1996, the remote participation based on them was successfully demonstrated in collaboration with Japan Atomic Energy Research Institute, Los Alamos National Laboratory, and Princeton Plasma Physics Laboratory. (orig.)

Conceptual design of JT-60SA cryostat

International Nuclear Information System (INIS)

Shibama, Y.K.; Sakurai, S.; Masaki, K.; Sukekawa, A.M.; Kaminaga, A.; Yoshida, K.; Matsukawa, M.

2007-01-01

JT-60U modification program to fully superconducting device has been proceeded, namely ''JT-60SA'', toward early realization of fusion energy based on tokamak concept. The design of JT-60SA cryostat is expected to achieve a vacuum thermal insulation for super conducting coils, a bio-shielding boundary and structural gravity support. The cryostat is required to cover JT-60SA tokamak device, which is 15 m of total height and 7 m of radius, but there is geometrical limit due to surrounding devices reutilized. Although the cryostat consists of vessel body and gravity support, and the structural material is low cobalt 304 stainless steel (Co: 2 , and the design of the leaf spring is considered to reduce thermal stress, and to withstand the mechanical loads of plasma disruption and seismic loads. The coolant is 80 K gas helium and both sides of panel are covered with multi-layers super insulation (SI) to reduce heat load (radiation) up to 1/100. Fraction of non-covered region is assumed to be 2% due to many port-joints and supports for the vacuum vessel. Total heat load for inner surface of cryostat (600 m 2 ) is 9kW and the heat load for the port-joints (-300 m 2 ) is assumed up to 9 kW. The operational pressure of the cryostat is required to keep less than 10 -2 Pa and about 100,000 m 2 of structural surfaces is considered for exhaust system specification. Another role of the cryostat is the radiation protection. Biological shielding up to 10 micro-Sv/h (for maintenance acceptance) is required of the cryostat surface after the 10 years operation. Thus the cryostat consists of boron (2 wt%) doped concrete of 220 mm thickness and structural SS304 of total 40 mm thickness. The concrete reduces the air activation (41Ar) in the torus hall by 90% rather than the normal one by the thermal neutron absorption of boron. (orig.)

Gyrokinetic analyses of core heat transport in JT-60U plasmas with different toroidal rotation direction

International Nuclear Information System (INIS)

Narita, Emi; Fukuda, Takeshi; Honda, Mitsuru; Hayashi, Nobuhiko; Urano, Hajime; Ide, Shunsuke

2015-01-01

Tokamak plasmas with an internal transport barrier (ITB) are capable of maintaining improved confinement performance. The ITBs formed in plasmas with the weak magnetic shear and the weak radial electric field shear are often observed to be modest. In these ITB plasmas, it has been found that the electron temperature ITB is steeper when toroidal rotation is in a co-direction with respect to the plasma current than when toroidal rotation is in a counter-direction. To clarify the relationship between the direction of toroidal rotation and heat transport in the ITB region, we examine dominant instabilities using the flux-tube gyrokinetic code GS2. The linear calculations show a difference in the real frequencies; the counter-rotation case has a more trapped electron mode than the co-rotation case. In addition, the nonlinear calculations show that with this difference, the ratio of the electron heat diffusivity χ_e to the ion's χ_i is higher for the counter-rotation case than for the co-rotation case. The difference in χ_e /χ_i agrees with the experiment. We also find that the effect of the difference in the flow shear between the two cases due to the toroidal rotation direction on the linear growth rate is not significant. (author)

Studies on first wall and plasma wall interaction in JT-60

International Nuclear Information System (INIS)

Nakamura, Hiroo

1988-12-01

This paper describes studies on first wall and plasma wall interaction in JT-60. Main results are as follows; (1) To select JT-60 first wall material, various RandD were done in FY1975 ∼ 1976. Mo was selected as first wall materials of limiters and divertor plates because of its reliability under a high heat flux condition. (2) Development of low-Z material has been done to reduce impurity problem of Mo first wall. As a result, titanium carbide (TiC) was selected as a coating material on the Mo. High heat load testing has been done for TiC coated Mo limiter same as JT-60. This material can survive under the condition of 1 kW/cm 2 x 10 s, expected in JT-60 limiter design. (3) To reduce high heat load on the divertor plate, separatrix swing is proposed. Optimum frequency of the sweeping is evaluated to be 2 Hz in JT-60. For a discharge with heating power of 30 MW and duration time of 10 s, in addition to the separatrix swing, remote radiative cooling in the divertor region is necessary. Moreover, calculations of erosion thickness have been done for stainless steel, Mo, graphite, TiC and silicon caibide under high heat flux during plasma disruption. (4) In divertor experiments in JT-60, divertor functions on particle, heat load and impurity controls have been demonstrated. In elctron density of 6 x 10 19 m -3 , particle fueling rate of 20 MW NB heating (3 Pa m 3 /s) can be exhausted by divertor pumping system. Effectiveness of remote radiative cooling is demonstrated under the condition of 20 MW NB heating power. Also, separatrix swing is demonstrated to reduce heat load on the divertor plate. Total radiation in main plasma is 5 ∼ 10% of total absorbed power. (author) 120 refs

Alfven eigenmodes driven by Alfvenic beam ions in JT-60U

International Nuclear Information System (INIS)

Shinohara, K.; Kusama, Y.; Takechi, M.

2001-01-01

Instabilities with frequency chirping in the frequency range of Alfven eigenmodes have been found in the domain 0.1% h > bparallel /υ A ∼ 1 with high energy neutral beam injection in JT-60U. One instability with a frequency inside the Alfven continuum spectrum appears and its frequency increases slowly to the toroidicity induced Alfven eigenmode (TAE) gap on the timescale of an equilibrium change (∼ 200 ms). Other instabilities appear with a frequency inside the TAE gap and their frequencies change very quickly by 10-20 kHz in 1-5 ms. During the period when these fast frequency sweeping (fast FS) modes occur, abrupt large amplitude events (ALEs) often appear with a drop of neutron emission rate and an increase in fast neutral particle fluxes. The loss of energetic ions increases with a peak fluctuation amplitude of B-tilde θ /B θ . An energy dependence of the loss ions is observed and suggests a resonant interaction between energetic ions and the mode. (author)

Design of the movable limiters for JT-60

International Nuclear Information System (INIS)

Takashima, Tetsuo; Yamamoto, Masahiro; Nakamura, Hiroo; Ohkubo, Minoru; Ohta, Mitsuru

1976-07-01

Two fast-acting movable rail limiters will be used in JT-60 to suppress skin effect of the plasma current with a large radius. They travel safely through a stroke of about 1 m for 0.1 sec in the build-up phase of plasma current. The movable limiter system consists of a driving mechanism, a vacuum seal, a bearing used at high temperatures in a vacuum, a molybdenum rail limiter weighing 200 kg and its auxiliary members. Many problems are involved in construction of the system because the design specifications exceed the present technology. Described are design of the movable limiter system for JT-60 and problems in the mechanical, electrical and vacuum aspects. (auth.)

Waveguide circuit for LHRF heating in 'JT-60'

International Nuclear Information System (INIS)

Uehara, Kazuya; Saegusa, Mikio; Mizuno, Takenori; Sano, Keigo; Hara, Mitsuru; Oishi, Isamu; Kanai, Takao.

1985-01-01

As the heating method for attaining the critical condition in the critical plasma experiment apparatus 'JT-60' in the Japan Atomic Energy Research Institute, in addition to Joule heating, as the second heating method, neutral beam injection heating and high frequency heating have been adopted. For this high frequency heating, several tens to 200 MHz band of ICRF heating, several hundreds MHz to several GHz band of LHRF heating and several tens to 200 GHz band of ECR heating were considered, and in the JT-60, 100 MHz band (ICRF) and 2 GHz band (LHRF) have been adopted. Furukawa Electric Co., Ltd. has engaged in the development and manufacture of the waveguides of transmission system used for this high frequency heating through NEC Corp. This high frequency heating is to heat plasma by injecting high frequency radio waves into plasma proper, and reaches 10 MW for the whole high frequency heating. The system efficiently transmitting the radio waves of large power from a Klystron as a high frequency source to the JT-60 is the transmission system. The outline of the waveguides of the 2 GHz band transmission system and the individual performance of respective waveguides are reported. (Kako, I.)

JT-60 database system, 2

International Nuclear Information System (INIS)

Itoh, Yasuhiro; Kurihara, Kenichi; Kimura, Toyoaki.

1987-07-01

The JT-60 central control system, ''ZENKEI'' collects the control and instrumentation data relevant to discharge and device status data for plant monitoring. The former of the engineering data amounts to about 3 Mbytes per shot of discharge. The ''ZENKEI'' control system which consists of seven minicomputers for on-line real-time control has little performance of handling such a large amount of data for physical and engineering analysis. In order to solve this problem, it was planned to establish the experimental database on the Front-end Processor (FEP) of general purpose large computer in JAERI Computer Center. The database management system (DBMS), therefore, has been developed for creating the database during the shot interval. The engineering data are shipped up from ''ZENKEI'' to FEP through the dedicated communication line after the shot. The hierarchical data model has been adopted in this database, which consists of the data files with tree structure of three keys of system, discharge type and shot number. The JT-60 DBMS provides the data handling packages of subroutines for interfacing the database with user's application programs. The subroutine packages for supporting graphic processing and the function of access control for security of the database are also prepared in this DBMS. (author)

Heat structural problems in JT-60

International Nuclear Information System (INIS)

Takatsu, Hideyuki; Shimizu, Masaomi; Yamamoto, Masahiro; Nakamura, Hiroo; Miyauchi, Yasuyuki.

1980-01-01

The construction of JT-60 is in progress to study the behavior of hydrogen plasma. The D-T reaction does not occur in this device, therefore the considerations for neutron damage, tritium leakage and so on are not necessary. The long-pulse operation will be done, and the suppression of the production and mixing of impurity is considered in the design of the JT-60. The high temperature baking is possible, and the magnetic limiter is set. The vacuum container has the complex structure consists of 8 sector type thick rings and 8 U-shaped bellows, and has egg-shaped cross section. The main radius of the torus is about 3 m. The material of the vacuum container is INCONEL 625. The analyses of various stresses due to such as atmospheric pressure, eddy current and thermal expansion were made. It is also necessary to consider the thermal stress due to the leakage of neutral beam. The thermal input of about 20 MW per one discharge to the first wall is taken into consideration. The material of the first wall is molybdenum. (Kato, T.)

Vacuum pumping system for the JT-60 radio-frequency heating system

International Nuclear Information System (INIS)

Yokokura, Kenji; Ikeda, Yoshitaka; Imai, Tuyoshi; Suganuma, Kazuaki; Nagashima, Takashi

1988-01-01

The basic design requirements set up for the JT-60 radio-frequency heating system included: (1) rapid pumping of gas released upon application of a radio-frequency power to maintain the pressure in the launchers at 10 -2 - 10 -3 Pa or less, (2) incorporation of a gas analysis system that can operate under a strong field and high pressure (>10 -2 Pa) to permit remote controlled data collection and processing, and (3) low cost, multiple functions and high reliability. The vacuum pumping system, consisting of three units for low hybrid radio-frequency (LHRF) and one unit for ion cyclotron radio-frequency (ICRF), is connected to each launcher provided at the four ports of JT-60. The LHRF unit is composed of a main pump, an alumina joint for electrical insulation from the launcher, a metallic gate valve for isolation from the JT-60 vacuum region, and various vacuum gauges. Only a turbo-molecular pump is used for the ICRF system because a large-scale differential pumping is not required. A gas measuring system is incorporated which consists of a mass filter, personal computer, turbo-molecular pump, and variable flow valve equipped with an APG control. This system is designed to identify and make use of gas impurities released during the launcher aging process. The control system employed consists of a personal computer, interlock control board, data logger and other devices such as vacuum gages. (Nogami, K.)

Present status of the JT-60 control system

International Nuclear Information System (INIS)

Kimura, T.

1992-01-01

The present status of the control system for a large fusion device of the JT-60 upgrade tokamak is reported including its original design concept, the progress of the system in the past five-year operation and modification for the upgrade. The control system has the features of hierarchical structure, computer control, adoption of CAMAC interfaces and protective interlock by both software and hard-wired systems. Plant monitoring and control are performed by an efficient data communication via CAMAC highways. Sequential discharge control of is executed by a combination of computers and a timing system. A plasma feedback control system with fast 32-bit microprocessors and a man/machine interface with modern workstations have been newly developed for the operation of the JT-60 upgrade. (author)

Review of JT-60U experimental results in 1997

International Nuclear Information System (INIS)

Adachi, H.; Akasaka, H.; Akino, N.

1998-08-01

The JT-60U experiments in 1997 focused mainly on the steady-state tokamak research with the newly installed W-shaped pumped divertor and the negative ion based neutral beam (NNB) in addition to the existing profile and shape control techniques developed in JT-60U. In particular, the research on divertor physics was accelerated under the new divertor system with many of fine diagnostics: Detachment characteristics, pumping control, impurity control, recycling characteristics, etc. in the W-shaped divertor were investigated in detail. The main purpose of confinement and stability studies in 1997 was to improve steadiness of high confinement plasmas with the new divertor. Researches progressed also for the formation conditions of the internal and the surface transport barriers in the high-β p mode, the reversed shear mode and the H-mode. Toward the advanced feedback controls of multiple parameters, the JT-60U started new feedback controls of central line density and divertor neutral gas pressure in addition to the existing controls of off-axis line density, radiation power and neutron production rate. The JT-60U team also carefully studied characteristics of halo current during disruptions. Optimization of NNB operation progressed steadily and injection power increased up to 4.2MW. The NNB-driven current was identified directly from the internal magnetic measurement and driven current profile was confirmed to be consistent with the ACCOME calculation. The current profile control with LHCD successfully sustained the internal transport barrier in reversed shear plasmas. Continuous TAE modes were observed with NNB for the first time as beam-driven TAE modes. (J.P.N.)

Review of JT-60U experimental results in 1997

Energy Technology Data Exchange (ETDEWEB)

Adachi, H.; Akasaka, H.; Akino, N. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1998-08-01

The JT-60U experiments in 1997 focused mainly on the steady-state tokamak research with the newly installed W-shaped pumped divertor and the negative ion based neutral beam (NNB) in addition to the existing profile and shape control techniques developed in JT-60U. In particular, the research on divertor physics was accelerated under the new divertor system with many of fine diagnostics: Detachment characteristics, pumping control, impurity control, recycling characteristics, etc. in the W-shaped divertor were investigated in detail. The main purpose of confinement and stability studies in 1997 was to improve steadiness of high confinement plasmas with the new divertor. Researches progressed also for the formation conditions of the internal and the surface transport barriers in the high-{beta}{sub p} mode, the reversed shear mode and the H-mode. Toward the advanced feedback controls of multiple parameters, the JT-60U started new feedback controls of central line density and divertor neutral gas pressure in addition to the existing controls of off-axis line density, radiation power and neutron production rate. The JT-60U team also carefully studied characteristics of halo current during disruptions. Optimization of NNB operation progressed steadily and injection power increased up to 4.2MW. The NNB-driven current was identified directly from the internal magnetic measurement and driven current profile was confirmed to be consistent with the ACCOME calculation. The current profile control with LHCD successfully sustained the internal transport barrier in reversed shear plasmas. Continuous TAE modes were observed with NNB for the first time as beam-driven TAE modes. (J.P.N.)

Overview of JT-60U progress towards steady-state advanced tokamak

International Nuclear Information System (INIS)

Ide, S.

2005-01-01

Recent experimental results on steady state advanced tokamak (AT) research on JT-60U are presented with emphasis on longer time scale in comparison with characteristics time scales in plasmas. Towards this, modification on control in operation, heating and diagnostics systems have been done. As the results, ∼ 60 s I p flat top and an ∼ 30 s H-mode are obtained. The long pulse modification has opened a door into a new domain for JT-60U. The high normalized beta (β N ) of 2.3 is maintained for 22.3 s and 2.5 for 16.5 s in a high β p H-mode plasma. A standard ELMy H-mode plasma is also extended and change in wall recycling in such a longer time scale has been unveiled. Development and investigation of plasmas relevant to AT operation has been continued in former 15 s discharges as well in which higherNB power (≤ 10 s) is available. Higher β N ∼ 3 is maintained for 6.2 s in high β p H-mode plasmas. High bootstrap current fraction (f BS ) of ∼ 75% is sustained for 7.4 s in an RS plasma. On NTM suppression by localized ECCD, ECRF injection preceding the mode saturation is found to be more effective to suppress the mode with less power compared to the injection after the mode saturated. The domain of the NTM suppression experiments is extended to the high β N regime, and effectiveness of m/n=3/2 mode suppression by ECCD is demonstrated at β N ∼ 2.5-3. Genuine center-solenoid less tokamak plasma start up is demonstrated. In a current hole region, it is shown that no scheme drives a current in any direction. Detailed measurement in both spatial and energy spaces of energetic ions showed dynamic change in the energetic ion profile at collective instabilities. Impact of toroidal plasma rotation on ELM behaviors is clarified in grassy ELM and QH domains. (author)

2. Interferometry and polarimetry. 2.3. Polarimetry on JT-60U

International Nuclear Information System (INIS)

Kawano, Yasunori

2000-01-01

In order to establish an electron density measurement method with high reliability and stability for magnetic-confinement fusion devices, studies on infrared polarimetry have been carried out in JT-60U. Electron density measurement based on tangential Faraday rotation has been verified using a CO 2 laser polarimeter developed for JT-60U. In this article, basic ideas of studies, results from polarimetry experiments, and suggestions for future devices are presented. (author)

Conceptual design of JT-60SA cryostat

International Nuclear Information System (INIS)

Shibama, Y.K.; Sakurai, S.; Masaki, K.; Sukekawa, A.M.; Kaminaga, A.; Sakasai, A.; Matsukawa, M.

2008-01-01

This paper describes the conceptual design of cryostat for the JT-60SA, which is a research device for the commercial production of electricity from the controlled fusion reaction in the future. JT-60SA is designed to be a fully superconducting device and cryostat is one of the main components to allow the normal operation. Cryostat covers up the tokamak device, which is 15 m of total height and 7 m of radius, and supports the total weight of 25 MN. Cryostat components consist of vessel body, gravity support and auxiliary systems, such as 80 K thermal shield and vacuum exhaust. The functions required of cryostat are these three, thermal insulation for superconducting magnets, gravity support for the tokamak device, and bio-shielding. The design conditions for each cryostat component are outlined and the features of auxiliary systems such as capacity of vacuum exhaust related to 80 K thermal shield design are summarized

Divertor pumping system with NBI cryopump for JT-60

International Nuclear Information System (INIS)

Akino, Noboru; Kuriyama, Masaaki; Ohga, Tokumichi; Seki, Hiroshi; Tanai, Yutaka

1998-08-01

The pumping system for JT-60 W-shape divertor with the NBI cryopump have been developed. The pumping speed achieved in the divertor region was 13-15 m 3 /s for deuterium gas with three units of the NBI cryopumps. In a simulation experiment of helium ash exhaust through the divertor, pumping of a mixed gas of helium and deuterium has been demonstrated using the NBI cryosorption pumps covered with an argon condensed layer. Control of neutral particle pressure in the divertor region became possible by having remodeled an aperture of the existing fast shutter, which is installed between the JT-60 vacuum vessel and NBI beam-line, to be regulated. (author)

Plasma regimes and research goals of JT-60SA towards ITER and DEMO

International Nuclear Information System (INIS)

Kamada, Y.; Ide, S.; Fujita, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; Barabaschi, P.; Lackner, K.; Ishida, S.; Bolzonella, T.

2011-01-01

The JT-60SA device has been designed as a highly shaped large superconducting tokamak with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) in order to satisfy the central research needs for ITER and DEMO. In the ITER- and DEMO-relevant plasma parameter regimes and with DEMO-equivalent plasma shapes, JT-60SA quantifies the operation limits, plasma responses and operational margins in terms of MHD stability, plasma transport and confinement, high-energy particle behaviour, pedestal structures, scrape-off layer and divertor characteristics. By integrating advanced studies in these research fields, the project proceeds 'simultaneous and steady-state sustainment of the key performances required for DEMO' with integrated control scenario development applicable to the highly self-regulating burning high-β high bootstrap current fraction plasmas.

Injection control development of the JT-60U electron cyclotron heating system

Energy Technology Data Exchange (ETDEWEB)

Hiranai, Shinichi; Shinozaki, Shin-ichi; Yokokura, Kenji; Moriyama, Shinichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sato, Fumiaki [Nippon Advanced Technology Co., Ltd., Tokai, Ibaraki (Japan); Suzuki, Yasuo [Atomic Energy General Service Co., Ltd., Tokai, Ibaraki (Japan); Ikeda, Yoshitaka [Japan Atomic Energy Research Inst., Kashiwa, Chiba (Japan)

2003-03-01

The JT-60U electron cyclotron heating (ECH) System injects a millimeteric wave at 110 GHz into the JT-60 Plasma, and heats the plasma or drives a current locally to enhance the confinement performance of the JT-60 plasma. The system consists of four sets of high power gyrotrons, high voltage power supplies and transmission lines, and two antennas that launch electron cyclotron (EC) beams toward the plasma. The key features of the injection control system are streering of the direction of the EC beam by driving the movable mirror in the antenna, and capability to set any combination of polarization angle and ellipticity by rotating the two grooved mirrors in the polarizers. This report represents the design, fabrication and improvements of the injection control system. (author)

Review of JT-60U experimental results from March to October, 1991

International Nuclear Information System (INIS)

1992-06-01

Experimental results achieved in the initial operation of JT-60U are described in this paper. Experiments of JT-60U were initiated in March 1991, and deuterium experiments started in the middle of July. Multivariable non-interacting control, H-mode experiments, the high-q and high-β p regime with hot ion enhanced confinement, the divertor heat flux, etc. are reported. Achieved values of the first experiment of the JT-60U LHCD in 1991 were P LH = 1.5 MW, driven current I RF = 2MA, current drive efficiency η CD (=n-bar e R p I RF /P LH ) = 0.25 x 10 20 m -2 A/W and current driven product CDP(=n-bar e R p I RF ) = 3 x 10 20 m -2 MA. (J.P.N.)

Review of JT-60 experiment (March 1986)

International Nuclear Information System (INIS)

1986-11-01

Results of JT-60 experiment with ohmic heating in March 1986 are summarized. A maximum plasma current of 2 MA, an average plasma density of 5.7 x 10 19 m -3 and energy confinement time of 0.4 - 0.5 sec were obtained. Detailed characteristics of ohmic plasmas are discussed. (author)

Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Higashijima, S. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)], E-mail: higashijima.satoru@jaea.go.jp; Sakurai, S.; Suzuki, S.; Yokoyama, K.; Kashiwa, Y.; Masaki, K.; Shibama, Y.K.; Takechi, M.; Shibanuma, K.; Sakasai, A.; Matsukawa, M.; Kikuchi, M. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2009-06-15

The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program 'Broader Approach'. The maximum heat flux to JT-60SA divertor is estimated to {approx}15 MW/m{sup 2} for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R and D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R and D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m{sup 2} as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

JT-60 configuration parameters for feedback control determined by regression analysis

Energy Technology Data Exchange (ETDEWEB)

Matsukawa, Makoto; Hosogane, Nobuyuki; Ninomiya, Hiromasa (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment)

1991-12-01

The stepwise regression procedure was applied to obtain measurement formulas for equilibrium parameters used in the feedback control of JT-60. This procedure automatically selects variables necessary for the measurements, and selects a set of variables which are not likely to be picked up by physical considerations. Regression equations with stable and small multicollinearity were obtained and it was experimentally confirmed that the measurement formulas obtained through this procedure were accurate enough to be applicable to the feedback control of plasma configurations in JT-60. (author).

JT-60 configuration parameters for feedback control determined by regression analysis

International Nuclear Information System (INIS)

Matsukawa, Makoto; Hosogane, Nobuyuki; Ninomiya, Hiromasa

1991-12-01

The stepwise regression procedure was applied to obtain measurement formulas for equilibrium parameters used in the feedback control of JT-60. This procedure automatically selects variables necessary for the measurements, and selects a set of variables which are not likely to be picked up by physical considerations. Regression equations with stable and small multicollinearity were obtained and it was experimentally confirmed that the measurement formulas obtained through this procedure were accurate enough to be applicable to the feedback control of plasma configurations in JT-60. (author)

Design of Plasma Facing Components for Superconducting Modification of JT-60

International Nuclear Information System (INIS)

Shinji Sakurai; Kei Masaki; Yusuke-Kudo Shibama; Hiroshi Tamai; Makoto Matsukawa; Cordier, J.J.

2006-01-01

JT-60 is planning to modify the machine as a fully superconducting coil tokamak (JT-60 Super Advanced, the former JT-60SC and NCT) to establish scientific and technological bases for an economically and environmentally attractive DEMO reactor. It will be also a satellite tokamak in a part of broader approach for ITER. It is designed for high beta (betaN = 3.5-5.5) and steady-state research in a break-even class DD plasma for 100 s or longer. Nominal plasma parameters are I p =5.5 MA, B t =2.7 T, R=3.01 m, a=1.14 m with double-null configuration. An ITER-like single-null configuration with I p =3.5 MA, B t =2.6 T can be also operated. In order to study the ITER-relevant high confinement plasma with high density, designed plasma heating power was enhanced from 25 MW to 41 MW for 100 s through the design review with EU and Japan. The heat flux onto outer divertor target exceeds 10 MW/m 2 with moderate radiative fraction of 50-60% in single-null configuration. Therefore, the ITER-like mono-block CFC target will be adopted to aim at power handling of 15 MW/m 2 . A cooling water system should be reinforced 3 times from original design for double null divertor with high coolant flow velocity of ∼10 m/s. The peak heat flux onto the neutral beam armor for perpendicular injected positive NB is evaluated to be 2 MW/m 2 , which needs to be actively water-cooled. A bolt-fixed CFC tile was tested at the heat flux of 1-3 MW/m 2 and will be applied to the NB armor. In order to improve plasma beta value by enhancing wall stabilization effect, passive-stabilizing plates, which are electrically and mechanically connected in poloidal and toroidal direction, will be installed near the plasma surface (r wall /a=1.1-1.3) at the outboard side. Stabilizing plate has double-wall ribbed structure and can be operated at 573 K with heating nitrogen gas instead of cooling water between double walls. It has crank-type support legs to allow thermal expansion at high temperature operation. The

Identity physics experiment on internal transport barriers in JT-60U and JET

Energy Technology Data Exchange (ETDEWEB)

De Vries, P C; Beurskens, M N A; Brix, M; Giroud, C; Hawkes, N C; Parail, V [EURATOM/UKAEA Association, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Sakamoto, Y; Fujita, T; Hayashi, N; Matsunaga, G; Oyama, N; Shinohara, K; Suzuki, T; Takechi, M [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Litaudon, X; Joffrin, E [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Crombe, K [Department of Applied Physics, Ghent University, Rozier 44, 9000 Gent (Belgium); Mantica, P [Istituto di Fisica del Plasma, EURATOM/ENEA-CNR Association, Milano (Italy); Salmi, A [Association Euratom-Tekes, Helsinki University of Technology, PO Box 4100 (Finland); Strintzi, D, E-mail: Peter.de.Vries@jet.u [National Technical University of Athens, EURATOM Association, GR-15773, Athens (Greece)

2009-12-15

A series of experiments have been carried out in 2008 at JT-60U and JET to find common characteristics and explain differences between internal transport barriers (ITBs). The identity experiments succeeded in matching the profiles of most dimensionless parameters at the time ITBs were triggered. Thereafter the q-profile development deviated due to differences in non-inductive current density profile, affecting the ITB. Furthermore, the ITBs in JET were more strongly influenced by the H-mode pedestal or edge localized modes. It was found to be difficult to match the plasma rotation characteristics in both devices. However, the wide range of Mach numbers obtained in these experiments shows that the rotation has little effect on the triggering of ITBs in plasmas with reversed magnetic shear. On the other hand the toroidal rotation and more specifically the rotational shear had an impact on the subsequent growth and allowed the formation of strong ITBs.

Requirements for tokamak remote operation: Application to JT-60SA

International Nuclear Information System (INIS)

Innocente, Paolo; Barbato, Paolo; Farthing, Jonathan; Giruzzi, Gerardo; Ide, Shunsuke; Imbeaux, Frédéric; Joffrin, Emmanuel; Kamada, Yutaka; Kühner, Georg; Naito, Osamu; Urano, Hajime; Yoshida, Maiko

2015-01-01

Highlights: • We analyzed the data management system (DMS) appropriate for international collaboration. • We define the principal requirements for all components of the DMS. • We evaluated application of DMS requirements to the JT-60SA experiment. • We evaluated the role network bandwidth and time delay between EU and Japan. - Abstract: Remote operation and data analysis are becoming key requirements of any fusion devices. In this framework a well-conceived data management system integrated with a suite of analysis and support tools are essential components for an efficient remote exploitation of any fusion device. The following components must be considered: data archiving data model architecture; remote data and computers access; pulse schedule, data analysis software and support tools; remote control room specifications and security issues. The definition of a device-generic data model plays also important role in improving the ability to share solution and reducing learning time. As for the remote control room, the implementation of an Operation Request Gateway has been identified as an answer to security issues meanwhile remotely proving all the required features to effectively operate a device. Previous requirements have been analyzed for the new JT-60SA tokamak device. Remote exploitation is paramount in the JT-60SA case which is expected to be jointly operated between Japan and Europe. Due to the geographical distance of the two parties an optimal remote operation and remote data-analysis is considered as a key requirement in the development of JT-60SA. It this case the effects of network speed and delay have been also evaluated and tests have confirmed that the performance can vary significantly depending on the technology used.

Requirements for tokamak remote operation: Application to JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Innocente, Paolo, E-mail: paolo.innocente@igi.cnr.it [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Barbato, Paolo [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Farthing, Jonathan [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Giruzzi, Gerardo [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Ide, Shunsuke [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Imbeaux, Frédéric; Joffrin, Emmanuel [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Kamada, Yutaka [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Kühner, Georg [Max-Planck-Institute for Plasma Physics, EURATOM Association, Wendelsteinstr. 1, 17491 Greifswald (Germany); Naito, Osamu; Urano, Hajime; Yoshida, Maiko [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

2015-10-15

Highlights: • We analyzed the data management system (DMS) appropriate for international collaboration. • We define the principal requirements for all components of the DMS. • We evaluated application of DMS requirements to the JT-60SA experiment. • We evaluated the role network bandwidth and time delay between EU and Japan. - Abstract: Remote operation and data analysis are becoming key requirements of any fusion devices. In this framework a well-conceived data management system integrated with a suite of analysis and support tools are essential components for an efficient remote exploitation of any fusion device. The following components must be considered: data archiving data model architecture; remote data and computers access; pulse schedule, data analysis software and support tools; remote control room specifications and security issues. The definition of a device-generic data model plays also important role in improving the ability to share solution and reducing learning time. As for the remote control room, the implementation of an Operation Request Gateway has been identified as an answer to security issues meanwhile remotely proving all the required features to effectively operate a device. Previous requirements have been analyzed for the new JT-60SA tokamak device. Remote exploitation is paramount in the JT-60SA case which is expected to be jointly operated between Japan and Europe. Due to the geographical distance of the two parties an optimal remote operation and remote data-analysis is considered as a key requirement in the development of JT-60SA. It this case the effects of network speed and delay have been also evaluated and tests have confirmed that the performance can vary significantly depending on the technology used.

Enhanced performance and control issues in JT-60U long pulse discharges

International Nuclear Information System (INIS)

Sakamoto, Y

2005-01-01

Recent experimental results are reported on control issues involved in long timescales and enhanced performance in JT-60U. The control issues in neoclassical tearing mode (NTM) suppression in the weak shear plasma regime include background optimization through decreasing β p (L q /L p ) at the rational surface and active stabilization of NTMs using ECCD. By optimizing β p (L q /L p ), a condition of β N ∼ 2.5 was sustained for 10 times the current profile relaxation time and one of β N ∼ 2.4 with q min ∼ 1.5 was sustained for 2.8 times the current profile relaxation time, with nearly full non-inductive current drive. In addition, a condition of β N ∼ 3 was sustained for 5.5 s through stabilization of NTMs using ECCD, and an EC driven current nearly equal to the bootstrap current was required for complete stabilization. In the reversed shear plasma regime, the issue is the existence of the steady state solution with a large f BS value. By controlling the pressure gradient at the internal transport barrier through toroidal rotation to avoid the disruption, a large f BS value of approximately 75% was sustained for 2.7 times the current profile relaxation time, with nearly full non-inductive current drive, and a steady-state solution with a large f BS value is confirmed. The control issues for the edge pedestal and edge localized modes (ELMs) are control of the pedestal pressure and the energy loss through ELMs. The pedestal pressure increases by >40% through the change in toroidal rotation. The type of ELM can be controlled by toroidal rotation from type-I to grassy

Impurity and particle recycling reduction by boronization in JT-60U

International Nuclear Information System (INIS)

Higashijima, S.; Sugie, T.; Kubo, H.; Tsuji, S.; Shimada, M.; Asakura, N.; Hosogane, N.; Kawano, Y.; Nakamura, H.; Itami, K.; Sakasai, A.; Shimizu, K.; Ando, T.; Saidoh, M.

1995-01-01

In JT-60U boronization using decaborane was carried out. Boronization reduced the oxygen impurity in OH discharges and shortened the wall conditioning after the vacuum vessel vent and consequently enabled JT-60U to produce clean plasmas easily except for NB heated plasmas. After boronization, particle recycling was reduced drastically in OH and NB discharges. High confinement plasmas were obtained including high β p mode and H-mode discharges. In the latest boronization part of divertor plates were replaced with B 4 C coated tiles with a B 4 C thickness similar 300 μm. After introducing B 4 C divertor tiles, an explosive generation of boron particles from the tiles was observed. By the combined effects of boronization with decaborane and boron generation from B 4 C tiles, oxygen impurity was so low that oxygen line signals were reduced to noise levels after the latest boronization. On the other hand, boron burst from the B 4 C tiles restricted the operation of JT-60U. ((orig.))

Comparison of transient electron heat transport in LHD helical and JT-60U tokamak plasmas

International Nuclear Information System (INIS)

Inagaki, S.; Ida, K.; Tamura, N.; Shimozuma, T.; Kubo, S.; Nagayama, Y.; Kawahata, K.; Sudo, S.; Ohkubo, K.; Takenaga, H.; Isayama, A.; Takizuka, T.; Kamada, Y.; Miura, Y.

2005-01-01

Transient transport experiments are performed in plasmas with and without Internal Transport Barrier (ITB) on LHD and JT-60U. The dependence of χ e on electron temperature, T e , and electron temperature gradient, ∇T e , is analyzed by an empirical non-linear heat transport model. In plasmas without ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation. The χ e depends on T e and ∇T e in JT-60U, while the ∇T e dependence is weak in LHD. Inside the ITB region, there is no or weak ∇T e dependence both in LHD and JT-60U. A cold pulse growing driven by the negative T e dependence of χ e is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U). (author)

Analysis of JT-60SA operational scenarios

Science.gov (United States)

Garzotti, L.; Barbato, E.; Garcia, J.; Hayashi, N.; Voitsekhovitch, I.; Giruzzi, G.; Maget, P.; Romanelli, M.; Saarelma, S.; Stankiewitz, R.; Yoshida, M.; Zagórski, R.

2018-02-01

Reference scenarios for the JT-60SA tokamak have been simulated with one-dimensional transport codes to assess the stationary state of the flat-top phase and provide a profile database for further physics studies (e.g. MHD stability, gyrokinetic analysis) and diagnostics design. The types of scenario considered vary from pulsed standard H-mode to advanced non-inductive steady-state plasmas. In this paper we present the results obtained with the ASTRA, CRONOS, JINTRAC and TOPICS codes equipped with the Bohm/gyro-Bohm, CDBM and GLF23 transport models. The scenarios analysed here are: a standard ELMy H-mode, a hybrid scenario and a non-inductive steady state plasma, with operational parameters from the JT-60SA research plan. Several simulations of the scenarios under consideration have been performed with the above mentioned codes and transport models. The results from the different codes are in broad agreement and the main plasma parameters generally agree well with the zero dimensional estimates reported previously. The sensitivity of the results to different transport models and, in some cases, to the ELM/pedestal model has been investigated.

Triton burnup study using scintillating fiber detector on JT-60U

Energy Technology Data Exchange (ETDEWEB)

Harano, Hideki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1997-09-01

The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV {alpha} particles behave. The {alpha} particles` behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton`s behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector`s directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R{sub p}-scan and n{sub e}-scan experiments have been performed. The R{sub p}-scan experiment indicates that the triton`s transport was increased as the ripple amplitude over the triton became larger. In the n{sub e}-scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs.

Triton burnup study using scintillating fiber detector on JT-60U

International Nuclear Information System (INIS)

Harano, Hideki

1997-09-01

The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV α particles behave. The α particles' behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton's behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector's directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R p -scan and n e -scan experiments have been performed. The R p -scan experiment indicates that the triton's transport was increased as the ripple amplitude over the triton became larger. In the n e -scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs

Construction of negative-ion based NBI for JT-60U

International Nuclear Information System (INIS)

Kawai, Mikito; Akino, Noboru; Ebisawa, Noboru

2001-11-01

The world's first negative-ion based neutral beam injector (N-NBI) system has been developed for studies of non-inductive current drive and plasma core heating with high energy neutral beam injection in higher density plasma. Construction of the N-NBI system for JT-60U was completed in March 1996. The system is composed of a beamline with two ion sources, a set of ion source power supplies, control system and auxiliary sub-system such as cooling water, refrigeration and vacuum system. In July 2001, deuterium neutral beam injection of 400keV and 5.8MW into JT-60U plasma was achieved. In order to increase both beam power and energy we have to go on more improvement of the N-NBI. (author)

Diagnostic system for passive charge-exchange particle measurements on JT-60

International Nuclear Information System (INIS)

Nemoto, Masahiro; Tobita, Kenji; Kusama, Yoshinori; Takeuchi, Hiroshi

1988-01-01

In order to measure energy distributions of the charge-exchange neutral particles in the JT-60 experiments, a compact size electrostatic energy analyzer which the measurable energy range is from 1 keV to 100 keV is developed successfully. Compactness of an analyzer is accomplished by setting an accelerator between a gas stripping cell and a deflector of 45deg injection type. The calibration of the analyzer was carried out owing to confirm the capability of energy analysis and stripping efficiency. This analyzer was applied to measure the energy distribution in additionally heated plasmas in JT-60. The usefullness of the analyzer was confirmed. (author)

Neural-net disruption predictor in JT-60U

International Nuclear Information System (INIS)

Yoshino, R.

2003-01-01

The prediction of major disruptions caused by the density limit, the plasma current ramp-down with high internal inductance l i , the low density locked mode and the β-limit has been investigated in JT-60U. The concept of 'stability level', newly proposed in this paper to predict the occurrence of a major disruption, is calculated from nine input parameters every 2 ms by the neural network and the start of a major disruption is predicted when the stability level decreases to a certain level, the 'alarm level'. The neural network is trained in two steps. It is first trained with 12 disruptive and six non-disruptive shots (total of 8011 data points). Second, the target output data for 12 disruptive shots are modified and the network is trained again with additional data points generated by the operator. The 'neural-net disruption predictor' obtained has been tested for 300 disruptive shots (128 945 data points) and 1008 non-disruptive shots (982 800 data points) selected from nine years of operation (1991-1999) of JT-60U. Major disruptions except for those caused by the -limit have been predicted with a prediction success rate of 97-98% at 10 ms prior to the disruption and higher than 90% at 30 ms prior to the disruption while the false alarm rate is 2.1% for non-disruptive shots. This prediction performance has been confirmed for 120 disruptive shots (56 163 data points), caused by the density limit, as well as 1032 non-disruptive shots (1004 611 data points) in the last four years of operation (1999-2002) of JT-60U. A careful selection of the input parameters supplied to the network and the newly developed two-step training of the network have reduced the false alarm rate resulting in a considerable improvement of the prediction success rate. (author)

Density limit in JT-60

International Nuclear Information System (INIS)

Kamada, Yutaka; Hosogane, Nobuyuki; Hirayama, Toshio; Tsunematsu, Toshihide

1990-05-01

This report studies mainly the density limit for a series of gas- and pellet-fuelled limiter discharges in JT-60. With the pellet injection into high-current/low-q (q(a)=2.3∼2.5) discharges, the Murakami factor reaches up to 10∼13 x 10 19 m -2 T -1 . The values are about factors of 1.5∼2.0 higher than those for usual gas-fuelled discharges. The pellet injected discharges have high central density, whereas the electron density in the outer region (a/2 abs and n e 2 (r=50 cm) x Z eff (r=50 cm). (author)

JT-60 database system, 1

International Nuclear Information System (INIS)

Kurihara, Kenichi; Kimura, Toyoaki; Itoh, Yasuhiro.

1987-07-01

Naturally, sufficient software circumstance makes it possible to analyse the discharge result data effectively. JT-60 discharge result data, collected by the supervisor, are transferred to the general purpose computer through the new linkage channel, and are converted to ''database''. Datafile in the database was designed to be surrounded by various interfaces. This structure is able to preserve the datafile reliability and does not expect the user's information about the datafile structure. In addition, the support system for graphic processing was developed so that the user may easily obtain the figures with some calculations. This paper reports on the basic concept and system design. (author)

Ion beam dump for JT-60 NBI

International Nuclear Information System (INIS)

Kuriyama, Masaaki; Horiike, Hiroshi; Matsuda, Shinzaburo; Morita, Hiroaki; Shibanuma, Kiyoshi

1981-10-01

The design of the active cooling type ion beam dump for JT-60 NBI which receives the total beam power of 5.6 MW for 10 sec continuously is described. It is composed of array of many finned tubes which is made of oxygen free copper with 0.2% silver content. The safety margin against thermal and mechanical troubles is estimated by the heat transfer and the thermal stress calculation. (author)

Quench detection of fast plasma events for the JT-60SA central solenoid

International Nuclear Information System (INIS)

Murakami, Haruyuki; Kizu, Kaname; Tsuchiya, Katsuhiko; Kamiya, Koji; Takahashi, Yoshikazu; Yoshida, Kiyoshi

2012-01-01

Highlights: ► Pick-up coil method is used for the quench detection of JT-60SA magnet system. ► Disk-shaped pick-up coils are inserted in CS module to compensate inductive voltage. ► Applicability of pick-up coil is evaluated by two dimensional analysis. ► Pick-up coil is applicable whenever disruption, mini collapse and other plasma event. - Abstract: The JT-60 is planned to be modified to a full-superconducting tokamak referred to as the JT-60 Super Advanced (JT-60SA). The maximum temperature of the magnet during its quench might reach the temperature of higher than several hundreds Kelvin that will damage the superconducting magnet itself. The high precision quench detection system, therefore, is one of the key technologies in the superconducting magnet protection system. The pick-up coil method, which is using voltage taps to detect the normal voltage, is used for the quench detection of the JT-60SA superconducting magnet system. The disk-shaped pick-up coils are inserted in the central solenoid (CS) module to compensate the inductive voltage. In the previous study, the quench detection system requires a large number of pick-up coils. The reliability of quench detection system would be higher by simplifying the detection system such as reducing the number of pick-up coils. Simplifying the quench detection system is also important to reduce the total cost of the protection system. Hence the design method is improved by increasing optimizing parameters. The improved design method can reduce the number of pick-up coils without reducing the sensitivity of detection; consequently the protection system can be designed with higher reliability and lower cost. The applicability of the disk-shaped pick-up coil for quench detection system is evaluated by the two dimensional analysis. In the previous study, however, the analysis model only took into account the CS, EF (equilibrium field) coils and plasma. Therefore, applicability of the disk-shaped pick-up coil for

Structural analysis of the JT-60SA cryostat vessel body

Energy Technology Data Exchange (ETDEWEB)

Botija, José, E-mail: jose.botija@ciemat.es [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Alonso, Javier; Fernández, Pilar; Medrano, Mercedes; Ramos, Francisco; Rincon, Esther; Soleto, Alfonso [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Davis, Sam; Di Pietro, Enrico; Tomarchio, Valerio [Fusion for Energy, JT-60SA European Home Team, 85748 Garching bei Munchen (Germany); Masaki, Kei; Sakasai, Akira; Shibama, Yusuke [JAEA – Japan Atomic Energy Agency, Naka Fusion Institute, Ibaraki 311-0193 (Japan)

2013-10-15

Highlights: ► Structural analysis to validate the JT-60SA cryostat vessel body design. ► Design code ASME 2007 “Boiler and Pressure Vessel Code. Section VIII”. ► First buckling mode: load multiplier of 10.644, higher than the minimum factor 4.7. ► Elastic and elastic–plastic stress analysis meets ASME against plastic collapse. ► Bolted fasteners have been analyzed showing small gaps closed by strong welding. -- Abstract: The JT-60SA cryostat is a stainless steel vacuum vessel (14 m diameter, 16 m height) which encloses the Tokamak providing the vacuum environment (10{sup −3} Pa) necessary to limit the transmission of thermal loads to the components at cryogenic temperature. It must withstand both external atmospheric pressure during normal operation and internal overpressure in case of an accident. The paper summarizes the structural analyses performed in order to validate the JT-60SA cryostat vessel body design. It comprises several analyses: a buckling analysis to demonstrate stability under the external pressure; an elastic and an elastic–plastic stress analysis according to ASME VIII rules, to evaluate resistance to plastic collapse including localized stress concentrations; and, finally, a detailed analysis with bolted fasteners in order to evaluate the behavior of the flanges, assuring the integrity of the vacuum sealing welds of the cryostat vessel body.

Experimental study of radiation losses on the JT-60 tokamak

International Nuclear Information System (INIS)

Nishitani, Takeo

1990-06-01

Bolometric measurement system and associated diagnostics, soft x-ray pulse-height-analyzer, soft x-ray intensity and Balmer α line measurement systems, were developed to investigate the radiation losses of the JT-60 plasmas. The bolometric measurement is the most important diagnostics in the radiation loss study. The soft x-ray pulse-height-analyzer is useful to estimate the metallic impurity concentration, and the soft x-ray intensity and Balmer α line measurements are monitors of radiation in x-ray region and particle recycling in the plasma edge, respectively. In JT-60, the marfe has been observed frequently in high-Ip and high density limited discharges with NB heating after the replacement of the first wall from TiC coated molybdenum tiles to graphite ones. The threshold electron density of the marfe onset increased with the NB power. The empirical scaling of the marfe onset taking account of the NB power was obtained. This scaling was useful to predict the marfe onset condition in NB heated discharges on JT-60. The marfe was modelled based on the radiative thermal instability. The simple model can explain the marfe onset condition. The radiated power from the plasma with marfe was about 90 % of the absorbed power. Both stored energy and central electron temperatures did not change by the marfe onset in spite of the such intense radiation loss. Finally, this study revealed that the most clean plasma was obtained in the metallic first wall with the divertor on JT-60. This fact is suggesting the capability of the metallic material for the first wall of next devices. Enhance radiation localized in the peripheral plasma such as marfe and IDC does not degrade the core plasma confinement or somewhat improves it, so that marfe and IDC are suitable operational regime in the high density region for future devices because they have strong remote-radiative-cooling-effect. (J.P.N.)

Study of neutral particle behavior and particle confinement in JT-60U

International Nuclear Information System (INIS)

Takenaga, Hidenobu; Shimizu, Katsuhiro; Asakura, Nobuyuki; Shimada, Michiya; Kikuchi, Mitsuru; Tsuji-Iio, Shunji; Uchino, Kiichiro; Muraoka, Katsunori.

1995-07-01

In order to understand the particle confinement properties in JT-60U, the particle confinement time was estimated through analyses of the neutral particle behavior. First, the neutral particle transport simulation code DEGAS using a Monte-Carlo technique was combined with the simple divertor code for calculating the edge plasma parameters, and was developed to calculate under the experimental conditions in JT-60U. Then, the charged particle source in the main plasma due to the ionization of the neutral particles was evaluated from the analyses of the neutral particle penetration to the main plasma based on results of the simulation code and measurements of D α emission intensities. Finally, the particle confinement time was estimated from the analysis of particle balance. The analyses were performed systematically for the L-mode plasma and H-mode plasma of JT-60U, and a data base of the particle confinement time was obtained. The dependence of the particle confinement time on the plasma parameters and the relationship between the properties of the particle confinement and the energy confinement were examined. (author)

Parametric thermo-hydraulic analysis of the TF system of JT-60SA during fast discharge

International Nuclear Information System (INIS)

Polli, Gian Mario; Lacroix, Benoit; Zani, Louis; Besi Vetrella, Ugo; Cucchiaro, Antonio

2013-01-01

Highlights: • We modeled the central clock-wise pancake of JT-60SA TF magnet at the EOB. • We simulated a quench followed by a fast discharge. • We evaluated the temperature and pressure rises in the nominal configuration. • We evaluated the effect of several parameter changes on the thermal-hydraulic response of the system. -- Abstract: The evolution of the conductor temperature and of the helium pressure of the central pancake of the TF superconducting magnet of the JT-60SA tokamak in a quench scenario are here discussed. The quench is triggered by a heat disturbance applied at the end of burning and followed by a fast safety discharge. A parametric study aimed at assessing the robustness of the calculation is also addressed with special regard to the voltage threshold, used to define the occurrence of the quench, and to the time delay, that cover all the possible delays in the fast discharge after quench detection. Finally, due to sensitivity analyses the influences of different parameters were assessed: the material properties of the strands (RRR, copper fraction), the magnitude and the spatial length of the triggering disturbance and the magnetic field distribution. The numerical evaluations were performed in the framework of the Broader Approach Agreement in collaboration with CEA, ENEA and the JT-60SA European Home Team using the 1D code Gandalf [1

Improvement of voltage holding capability in the 500 keV negative ion source for JT-60SA.

Science.gov (United States)

Tanaka, Y; Hanada, M; Kojima, A; Akino, N; Shimizu, T; Ohshima, K; Inoue, T; Watanabe, K; Taniguchi, M; Kashiwagi, M; Umeda, N; Tobari, H; Grisham, L R

2010-02-01

Voltage holding capability of JT-60 negative ion source that has a large electrostatic negative ion accelerator with 45 cm x 1.1 m acceleration grids was experimentally examined and improved to realize 500 keV, 22 A, and 100 s D- ion beams for JT-60 Super Advanced. The gap lengths in the acceleration stages were extended to reduce electric fields in a gap between the large grids and at the corner of the support flanges from the original 4-5 to 3-4 kV/mm. As a result, the voltage holding capability without beam acceleration has been successfully improved from 400 to 500 kV. The pulse duration to hold 500 kV reached 40 s of the power supply limitation.

Development of virtual private network for JT-60SA CAD integration

International Nuclear Information System (INIS)

Oshima, Takayuki; Fujita, Takaaki; Seki, Masami; Kawashima, Hisato; Hoshino, Katsumichi; Shibanuma, Kiyoshi; Verrecchia, M.; Teuchner, B.

2010-01-01

The CAD models will be exchanged and integrated at Naka for JT-60SA, a common computer network efficiently connected between Naka site and the Garching site is needed to be established. Virtual Private Network (VPN) was introduced with LAN on computer network physically-separated from JAEA intranet area and firewall. In July 2009, a new VPN connection between the Naka and Garching sites has been successfully demonstrated using IPSec-VPN technology with a commercial and cost-effective firewall/router for security. It was found that the introduction of the Wide Area File Service (WAFS) could solve the issue of the data transmission time and enhance the usability of the VPN for design integration in JT-60SA. (author)

Retention characteristics of hydrogen isotopes in JT-60U

International Nuclear Information System (INIS)

Masaki, K.; Sugiyama, K.; Hayashi, T.; Ochiai, K.; Gotoh, Y.; Shibahara, T.; Hirohata, Y.; Oya, Y.; Miya, N.; Tanabe, T.

2005-01-01

Erosion/deposition distribution and hydrogen isotope behavior in the JT-60U plasma-facing wall were investigated. Distribution of the tritium, which was produced by D-D nuclear reaction, was not correlated with erosion/deposition distribution. The tritium distribution can be explained by the distribution of high-energy tritium ion-implantation due to ripple loss. Deuterium distribution in the divertor region was different from the tritium distribution and not well correlated with the deposition. The highest D/C was ∼0.05 at the bottom of the outer dome wing, which is much less than that observed in other tokamaks. For the deuterium retention, at least two retention processes (ion-implantation and co-deposition) were found on the dome region. The systematic dust collection gave the small amount of dust (∼7 g: 0.2 mg/s production) in the whole vessel of JT-60U. Deposition was observed at the remote area of the outer divertor region

Characteristics of large scale ionic source for JT-60

Energy Technology Data Exchange (ETDEWEB)

Fujiwara, Yukio; Honda, Atsushi; Inoue, Takashi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1997-02-01

The Neutral Beam Injection (NBI) apparatus is expected for important role sharing apparatus to realize the plasma electric current drive and the plasma control in not only temperature upgrading of the plasma but also Tokamak nuclear fusion reactor for the next generation such as JT-60, ITER and so forth. Japan Atomic Energy Research Institute has developed the ionic source with high energy and large electric current for about 10 years. Some arrangement tests of the large negative ion source for JT-60 No. 1 were executed from June to October, 1995. As a series of arrangement tests, 400 KeV and 13.5 A of deuterium negative ion beam was successfully accelerated for 0.12 sec. under 0.22 Pa of low gas pressure. And, it was elucidated that electron electric current could be controlled efficiently even in deuterium negative ion beam. Here is described on the testing results in details. (G.K.)

Quality control of the software in the JT-60 computer control system

International Nuclear Information System (INIS)

Isaji, Nobuaki; Kurihara, Kenichi; Kimura, Toyoaki

1990-07-01

The JT-60 Control System should be improved corresponding to the experimental requirements. In order to keep the integrity of the system even in the modification the concept of quality control (QC) was introduced in the software development. What we have done for QC activity are (1) to establish standard procedures of the software development, (2) to develop support tools for grasping the present status of the program structure, and (3) to develop a document system, and a source program management system. This paper reports these QC activities and their problems for the JT-60 control system. (author)

Study of carbon impurity generation by chemical sputtering in JT-60U

International Nuclear Information System (INIS)

Higashijima, S.; Kubo, H.; Sugie, T.; Shimizu, K.; Asakura, N.; Itami, K.; Hosogane, N.; Sakasai, A.; Konoshima, S.; Sakurai, S.; Takenaga, H.

1997-01-01

CD/CH-band intensities emitted from hydrocarbon molecules have been measured in the divertor region of JT-60U and the chemical sputtering yield of methane was estimated as a function of the surface temperature and the deuterium ion flux. The chemical sputtering yield increases with the surface temperature and decreases with increasing ion flux density in the L-mode plasmas. The B 4 C converted CFC tiles are installed in JT-60U and it is found that the chemical sputtering of B 4 C converted CFC tiles is suppressed in comparison to normal CFC tiles. (orig.)

Development of a compact W-shaped pumped divertor in JT-60U

Energy Technology Data Exchange (ETDEWEB)

Sakurai, S.; Hosogane, N.; Masaki, K.; Kodama, K.; Sasajima, T.; Kishiya, K.; Takahashi, S.; Shimizu, K.; Akino, N.; Miyo, Y.; Hiratsuka, H.; Saidoh, M. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Inoue, M.; Umakoshi, T.; Onozuka, M.; Morimoto, M. [Mitsubishi Heavy Industries, Wadasaki-cho, Hyogo-ku, Kobe-shi, 642 (Japan)

1998-09-01

In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current. (orig.) 11 refs.

Very fast feedback control of coil-current in JT-60 tokamak

International Nuclear Information System (INIS)

Aoyagi, T.; Terakado, T.; Takahashi, M.; Nobusaka, H.; Yagyu, J.; Matsuzaki, Y.

1992-01-01

A direct digital control (DDC) system is adopted for controlling thyristor converters of power supplies in the JT-60 tokamak built in 1984. Microcomputers of the DDC were 5 MHz i8086 microprocessor and programs were written by assembler language and the processing time was under 1ms. They were, however, too old in hardware and too complicated in software. New DDC system has been made in the JT-60 Upgrade (JT-60U) to control the power supplies more quickly under 0.25 and 0.5 ms of the processing time and also to write the programs used by high-level language. The new system consists of a host computer and five microcomputers with microprocessor on VME bus system. The host computer AS3260 performs on-line processing such as setting the DDC under the discharge conditions and so on. Functions of the microcomputers with a 32-bit, 20 MHz microprocessor MC68030, whose OS are VxWorks and programs are written by C language, are real-time processing such as taking in instructions from a ZENKEI computer and in feedback control of currents and voltages of coils every 0.25 and 0.5 ms. The system is now operating very smoothly. (author)

Simulation of collisional effects on divertor pumping in JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Gleason-González, C., E-mail: cristian.gleason@kit.edu [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, Baden-Württemberg 76344 (Germany); Varoutis, S.; Luo, X. [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, Baden-Württemberg 76344 (Germany); Shimizu, K.; Nakano, T.; Hoshino, K.; Kawashima, H.; Asakura, N. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka, Ibaraki 311-0193 (Japan); Day, Chr. [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, Baden-Württemberg 76344 (Germany); Sakurai, S. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2016-11-01

Highlights: • The exhausted (sub-divertor) gas flows calculations in tokamaks by means of three approaches: ProVac3D, DIVGAS and NEUT2D. • Exhausted neutral gas flow is modeled for two scenarios of a simplified JT-60SA sub-divertor geometry. • A modelled scenario with a simplified setup is done by using two intrinsic collisionless approaches: ProVac3D and NEUT2D and a third approach: DIVGAS, which has been used without its collision module for a direct comparison with the other two. The solvers are cross-checked in terms of the reproduction of the transmission probability. • A second case study is based on the Scenario # 2 of JT-60SA, where the assessment of collisionality in the sub-divertor was done. The gas flow is simulated by using DIVGAS with and without consideration of collisions. • The results include the transmission probability in JT-60SA sub-divertor, the Knudsen number, which characterizes the collisionality of the flow, velocity profiles, pressure and temperature distributions. - Abstract: In this work, the exhausted neutral gas flow is modeled for two cases of a simplified JT-60SA sub-divertor geometry and compared via three different approaches, namely (i) a collisionless approach based on the ProVac3D code, (ii) the DSMC approach based on the DIVGAS code that can be run with and without consideration of particle collisions, and (iii) the NEUT2D approach which has been extensively used in the past for the JT-60 design. In a first case study, the transmission probability was calculated by the 3 approaches and very good agreement is found between NEUT2D-ProVac3D whereas discrepancies between DIVGAS and NEUT2D are found and further analyzed. In the second case, the assessment of collisions is done by means of DIVGAS. Simulations showed that the flow is found in the transitional regime with Kn numbers between 0.1 and 0.4. The DIVGAS collisionless case yielded lower values of temperature than the collisional one by factors of 0.5–0.8 in

Development of the piezoelectric gas injection valve for JT-60

International Nuclear Information System (INIS)

Kawasaki, Kazuo; Hiratuka, Hajime

1986-01-01

Piezoelectric gas injection valve (PEV) for JT-60 have been developed which was a piezo-electric element. The raliability of the PEV under the actual condition of high magnetic fields and high temperatures are veryfied, and it became clear that the PEV had enough throughput range and sufficient repetability for long life throughput characteristics. Remarkables of the developed PEV are summarized as follows, (1) The maximum throughput rate, responce time and helium leakage rate satisfy the desiged specifications. (2) Throughput equation for PEV is clarified by comparison with experiment. (3) Reliabilities of PEV under the actual condition during coil power test become clear. (author)

Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas

International Nuclear Information System (INIS)

Tamai, H.; Fujita, T.; Kikuchi, M.

2006-01-01

JT-60SA, the former JT-60SC and NCT, a superconducting tokamak positioned as the satellite machine of ITER, collaborating with Japan and EU fusion community, aims at contribution to ITER and DEMO through the demonstration of advanced plasma operation scenario and the plasma applicability test with advanced materials. After the discussions in JA-EU Satellite Tokamak Working Group in 2005, the increased heating power, higher heat removal capacity for the plasma facing components, improvement of the radiation shielding, the remote handling system for the maintenance of in-vessel components, and related equipment are planed to be additionally installed. With such full equipment towards the increased heating power of 41 MW (34 MW-NBI and 7 MW-ECH) with 100 s, the prospective plasma performances, analysed by the equilibrium and transport analysis codes, are rather improved in the view point of the contribution to ITER and DEMO relevant research. Accessibility for higher heating power in a higher density region enables the lower normalized Larmor radius and normalized collision frequency close to the reactor relevant plasma with the ITER-similar configuration of single null divertor plasma with the aspect ratio of A = 3.1, elongation of k95 = 1.7, triangularity of d95 (q95) in the plasma current of I p = 3.5 MA, toroidal magnetic field of B T = 2.59 T and the major radius of Rp=3.16 m. The increases in the electron temperature, beam driven and bootstrap current fraction by the increase of the power of Negative ion based NBI (10 MW) reduce the loop voltage and contribute to increase the maximum plasma current of ITER similar shape. Full non-inductive current drive controllability is also extended into the high density and high plasma current operation towards high beta plasma. Flexibility in aspect ratio and shape parameter is kept the same as NCT, i.e. a double null divertor plasma with A = 2.6, k95 = 1.83, d95 = 0.57, I p = 5.5 MA, B T = 2.72 T, and R p = 3.01 m which

Performance of RF power and phase control on JT-60 LHRF heating system

International Nuclear Information System (INIS)

Fujii, T.; Ikeda, Y.; Imai, T.; Honda, M.; Kiyono, K.; Maebara, S.; Saigusa, M.; Sakamoto, K.; Sawahata, M.; Seki, M.

1987-01-01

The performance of RF power and phase control on the JT-60 LHRFD heating system are presented. The JT-60 LHRF heating system has three units of huge RF source with a total output of 24 MW, each unit consisting of eight amplifier chains. A high power klystron generating 1 MW for 10 s at 2 GHz is used in each chain. Automatic gain control is employed to regulate the output power not only against gain fluctuations in the chain but also against the unstable plasma load without any output circulator for the klystron

Diagnostic planning in JT-60 project

International Nuclear Information System (INIS)

Matoba, Tohru; Suzuki, Yasuo; Funahashi, Akimasa; Itagaki, Tokiyoshi

1977-08-01

The diagnostic plans of JT-60 were made along with design of the main machine. Basic requirements of the diagnostic program are (1) multiple measurement of respective plasma parameters, (2) efficient usage of the discharge, (3) capable data acquisition system, (4) high reliability of the diagnostic equipments, and (5) systematic development of new diagnostic techniques. Dimensions of the diagnostic ports were determined in detailed design of the vacuum vessel, anticipating the possible diagnostic methods. The proposed diagnostic systems and the plans are shown in table and figures respectively. Problems in the diagnostics are also described. (auth.)

Review of JT-60U experimental results from February to October, 1995

International Nuclear Information System (INIS)

1996-03-01

Renewed theme group organization started from October 1994 for the upcoming experiments in JT-60U. This regime has three theme groups each of which is composed of two sub-theme groups as; (1) Plasma Operation Theme (Leader Y. Neyatani) with Operation Sub-Theme and Disruption Sub-Theme, (2) High Performance (Leader S. Ishida) with Confinement and MHD Sub-Theme and High Energy Particle Sub-Theme and (3) Steady State Theme (Leader A. Sakasai) with Current Drive Sub-Theme and Divertor Sub-Theme. The main results from the JT-60U experiments in 1995 are summarized in the overviews of the three theme group activities. (J.P.N.)

Review of JT-60U experimental results from February to October, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-03-01

Renewed theme group organization started from October 1994 for the upcoming experiments in JT-60U. This regime has three theme groups each of which is composed of two sub-theme groups as; (1) Plasma Operation Theme (Leader Y. Neyatani) with Operation Sub-Theme and Disruption Sub-Theme, (2) High Performance (Leader S. Ishida) with Confinement and MHD Sub-Theme and High Energy Particle Sub-Theme and (3) Steady State Theme (Leader A. Sakasai) with Current Drive Sub-Theme and Divertor Sub-Theme. The main results from the JT-60U experiments in 1995 are summarized in the overviews of the three theme group activities. (J.P.N.).

Advanced control scenario of high-performance steady-state operation for JT-60 superconducting tokamak

International Nuclear Information System (INIS)

Tamai, H.; Kurita, G.; Matsukawa, M.; Urata, K.; Sakurai, S.; Tsuchiya, K.; Morioka, A.; Miura, Y.M.; Kizu, K.; Kamada, Y.; Sakasai, A.; Ishida, S.

2004-01-01

Plasma control on high-β N steady-state operation for JT-60 superconducting modification is discussed. Accessibility to high-β N exceeding the free-boundary limit is investigated with the stabilising wall of reduced-activated ferritic steel and the active feedback control of the in-vessel non-axisymmetric field coils. Taking the merit of superconducting magnet, advanced plasma control for steady-state high performance operation could be expected. (authors)

Design and Structural Analysis for the Vacuum Vessel of Superconducting Tokamak JT-60SC

International Nuclear Information System (INIS)

Kudo, Y.; Sakurai, S.; Masaki, K.; Urata, K.; Sasajima, T.; Matsukawa, M.; Sakasai, A.; Ishida, S.

2003-01-01

A modification of the JT-60 is planned to be a superconducting tokamak (JT-60SC) in order to establish steady-state operation of high beta plasma for 100 s, and to ensure the applicability of ferritic steel as a reduced activation material for reactor relevant break-even class plasmas. This paper describes the detailed design of the vacuum vessel, which has a unique structure for cost effective manufacturing, as well as structural analysis results for a feasibility study

High performance experiments in JT-60U reversed shear discharges

International Nuclear Information System (INIS)

Fujita, T.; Kamada, Y.; Ishida, S.

2001-01-01

The operation of JT-60U reversed shear discharges has been extended to a high plasma current, low-q regime keeping a large radius of the internal transport barrier (ITB) and the record value of equivalent fusion multiplication factor in JT-60U, Q DT eq =1.25, has been achieved at 2.6 MA. Operational schemes to reach the low-q regime with good reproducibility have been developed. The reduction of Z eff was obtained in the newly installed W-shaped pumped divertor. The beta limit in the low-q min regime, which limited the performance of L-mode edge discharges, has been improved in H-mode edge discharges with a broader pressure profile, which was obtained by power flow control with ITB degradation. Sustainment of ITB and improved confinement for 5.5 seconds has been demonstrated in an ELMy H reversed shear discharge. (author)

Progress of data processing system in JT-60 utilizing the UNIX-based workstations

International Nuclear Information System (INIS)

Sakata, Shinya; Kiyono, Kimihiro; Oshima, Takayuki; Sato, Minoru; Ozeki, Takahisa

2007-07-01

JT-60 data processing system (DPS) possesses three-level hierarchy. At the top level of hierarchy is JT-60 inter-shot processor (MSP-ISP), which is a mainframe computer, provides communication with the JT-60 supervisory control system and supervises the internal communication inside the DPS. The middle level of hierarchy has minicomputers and the bottom level of hierarchy has individual diagnostic subsystems, which consist of the CAMAC and VME modules. To meet the demand for advanced diagnostics, the DPS has been progressed in stages from a three-level hierarchy system, which was dependent on the processing power of the MSP-ISP, to a two-level hierarchy system, which is decentralized data processing system (New-DPS) by utilizing the UNIX-based workstations and network technology. This replacement had been accomplished, and the New-DPS has been started to operate in October 2005. In this report, we describe the development and improvement of the New-DPS, whose functions were decentralized from the MSP-ISP to the UNIX-based workstations. (author)

Analyses of plasma parameter profiles in JT-60U

Energy Technology Data Exchange (ETDEWEB)

Shirai, Hiroshi; Shimizu, Katsuhiro; Hayashi, Nobuhiko [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Itakura, Hirofumi; Takase, Keizou [CSK Co. Ltd., Tokyo (Japan)

2001-01-01

The methods how diagnostics data are treated as the surface quantity of magnetic surface and processed to the profile data in the JT-60U plasmas are summarized. The MHD equilibrium obtained by solving Grad-Shafranov equation on the MHD equilibrium calculation and registration software FBEQU are saved shot by shot as a database. Various experimental plasma data measured at various geometrical positions on JT-60 are mapped onto the MHD equilibrium and treated as functions of the volume averaged minor radius {rho} on the experimental data time slice monitoring software SLICE. Experimental data are integrated and edited on SLICE. The experimental data measured as the line integral values are transformed by Able inversion. The mapped data are fitted to a functional form and saved to the profile database MAP-DB. SLICE can also read data from MAP-DB and redisplay and transform them. In addition, SLICE can generate the profile data TOKRD as run data for orbit following Monte-Carlo (OFMC) code, analyzer for current drive consistent with MHD equilibrium (ACCOME) code and tokamak predictive and interpretive code system (TOPICS). (author)

Analyses of plasma parameter profiles in JT-60U

International Nuclear Information System (INIS)

Shirai, Hiroshi; Shimizu, Katsuhiro; Hayashi, Nobuhiko

2001-01-01

The methods how diagnostics data are treated as the surface quantity of magnetic surface and processed to the profile data in the JT-60U plasmas are summarized. The MHD equilibrium obtained by solving Grad-Shafranov equation on the MHD equilibrium calculation and registration software FBEQU are saved shot by shot as a database. Various experimental plasma data measured at various geometrical positions on JT-60 are mapped onto the MHD equilibrium and treated as functions of the volume averaged minor radius ρ on the experimental data time slice monitoring software SLICE. Experimental data are integrated and edited on SLICE. The experimental data measured as the line integral values are transformed by Able inversion. The mapped data are fitted to a functional form and saved to the profile database MAP-DB. SLICE can also read data from MAP-DB and redisplay and transform them. In addition, SLICE can generate the profile data TOKRD as run data for orbit following Monte-Carlo (OFMC) code, analyzer for current drive consistent with MHD equilibrium (ACCOME) code and tokamak predictive and interpretive code system (TOPICS). (author)

Structure of the radial electric field and toroidal/poloidal flow in high temperature toroidal plasma

International Nuclear Information System (INIS)

Ida, Katsumi

2001-01-01

The structure of the radial electric field and toroidal/poloidal flow is discussed for the high temperature plasma in toroidal systems, tokamak and Heliotron type magnetic configurations. The spontaneous toroidal and poloidal flows are observed in the plasma with improved confinement. The radial electric field is mainly determined by the poloidal flow, because the contribution of toroidal flow to the radial electric field is small. The jump of radial electric field and poloidal flow are commonly observed near the plasma edge in the so-called high confinement mode (H-mode) plasmas in tokamaks and electron root plasma in stellarators including Heliotrons. In general the toroidal flow is driven by the momentum input from neutral beam injected toroidally. There is toroidal flow not driven by neutral beam in the plasma and it will be more significant in the plasma with large electric field. The direction of these spontaneous toroidal flows depends on the symmetry of magnetic field. The spontaneous toroidal flow driven by the ion temperature gradient is in the direction to increase the negative radial electric field in tokamak. The direction of spontaneous toroidal flow in Heliotron plasmas is opposite to that in tokamak plasma because of the helicity of symmetry of the magnetic field configuration. (author)

Effects of pressure profile and plasma shaping on the n=1 internal kink mode in JT-60/JT-60U pellet fuelled plasmas

International Nuclear Information System (INIS)

Ozeki, Takahisa; Azumi, Masafumi

1990-10-01

The stability of the n=1 internal kink mode in a tokamak is numerically analyzed for plasmas with a centrally peaked pressure profile. These studies are carried out with the strongly peaked pressure inside the q=1 surface, which is based on the experimentally observed plasmas by means of injections of hydrogen-ice pellets in JT-60 tokamak. The effects of peaked pressure and shaping, i.e., elongation and triangularity, are also studied for JT-60U tokamak. The plasma with the strongly peaked pressure profile has higher critical value of poloidal beta defined within the q=1 surface than that with a parabolic pressure profile. Though the beta limit reduces with the increase of the elongation, the plasma with the peaked pressure profile has larger improvement due to the triangularity than that with the parabolic pressure profile. To access the second stability of the n=1 internal kink mode, the plasma with a flat pressure profile and the large minor radius of the q=1 surface is effective. (author)

Development of TiC coated wall materials for JT-60

International Nuclear Information System (INIS)

Abe, T.; Murakami, Y.; Obara, K.; Hiroki, S.; Nakamura, K.; Inagawa, K.

1985-01-01

Development of titanium carbide (TiC, 20 μm thick) coated wall materials has been carried out for JT-60. Application of TiC coatings onto molybdenum and Inconel 625 substrates requires a deposition temperature below 950 0 C and 600 0 C respectively, because recrystallization of molybdenum and age hardening of Inconel 625 occur above these temperatures. Through this process of coating we develop a new type plasma CVD(TP-CVD method) for molybdenum and a new type PVD(HCD-ARE method) for Inconel 625 which could successfully reduce the deposition temperature to 900 0 C and 500 0 C, respectively. The TiC coated wall samples were characterized by AES, ESCA, X-ray diffractometer, EPMA, SEM, metalography, tensile tests, thermal shock tests, and other techniques. As a result of the above measurements, it was demonstrated that the characteristics of those TiC coated walls satisfy the requirements arising from JT-60 operation conditions. (orig.)

Review of ICRF antenna development and heating experiments up to advanced experiment I, 1989 on the JT-60 tokamak

International Nuclear Information System (INIS)

Fujii, Tsuneyuki

1992-03-01

Two main subjects of ion cyclotron range of frequencies (ICRF) heating on JT-60 are described in this paper from development phase of the JT-60 ICRF heating system up to advanced experiment I, 1989. One is antenna design and development for the high power JT-60 ICRF heating system (6 MW for 10 s at a frequency range of 108 - 132 MHz). The other is the experimental investigation of characteristics of second harmonic ICRF heating in a large tokamak. (J.P.N.)

JT-60 negative ion beam NBI apparatus. Present state of its construction and initial experimental results

Energy Technology Data Exchange (ETDEWEB)

Kuriyama, Masaaki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1997-02-01

The NBI (Neutral Beam Injection) apparatus used for negative ion at first in the world, has an aim to actually prove heating and electric current drive with high density plasma at the JT-60 and to constitute physical and technical bases for selection and design of heating apparatus of ITER (International Thermal Nuclear Fusion Experimental Reactor). Construction of 500 KeV negative ion NBI apparatus for the JT-60 started to operate on 1993 was completed at March, 1996. On the way, at a preliminary test on forming and acceleration of the negative ion beam using a portion of this apparatus, 400 KeV and 13.5 A/D of the highest deuterium negative ion beam acceleration in the world was obtained successfully, which gave a bright forecasting of the plasma heating and electric current drive experiment using the negative ion NBI apparatus. After March, 1996, some plans to begin beam incident experiment at the JT-60 using the negative ion NBI apparatus and to execute the heating and electric current drive experiment at the JT-60 under intending increase of beam output are progressed. (G.K.)

Manufacturing of central control system of 'JT-60' a plasma feasibility experiment device

International Nuclear Information System (INIS)

Kondo, Ikuo; Kimura, Toyoaki; Murai, Katsuji; Iba, Daizo; Takemaru, Koichi.

1984-01-01

For constructing a critical-plasma-experiment apparatus JT-60, it was necessary to develop a new control system which enables to operate safely and smoothly a large scale nuclear fusion apparatus and to carry out efficient experiment. For the purpose, the total system control facility composed of such controllers as CAMAC system, timing system and protective interlock panel with multi-computer system as the core was developed. This system generalizes, keeps watch on and controls the total facilities as the key point of the control system of JT-60, and allows flexible operation control corresponding to the diversified experimental projects. At the same time, it carries out the fast real-time control of high temperature, high density plasma. In this paper, the system constitution, function and the main contents of development of the total system control facility are reported. JT-60 is constructed to attain the critical plasma condition as the premise of nuclear fusion reactors and to scientifically verify controlled nuclear fusion. Plasma expe riment will be started in April, 1985. The real-time control of plasma for carrying out high beta operation is planned, intending to develop future economical practical reactors. (Kako, I.)

Web-based Java application to advanced JT-60 Man-Machine Interfacing System for remote experiments

International Nuclear Information System (INIS)

Totsuka, Toshiyuki; Suzuki, Yoshio; Sakata, Shinya; Oshima, Takayuki; Iba, Katsuyuki

2008-01-01

Since remote participation in ITER experiments is planned, it is expected to demonstrate that the JT-60SA experiment is controlled from a Japanese remote experiment center located in Rokkasho-mura, Aomori-ken, Japan as a part of the ITER-BA project. Functions required for this experiment are monitoring of the discharge sequence status, handling of the discharge parameter, checking of experiment data, and monitoring of plant data, all of which are included in the existing JT-60 Man-Machine Interfacing System (MMIF). The MMIF is now only available to on-site users at the Naka site due to network safety. The motivation for remote MMIF is prompted by the issue of developing and achieving compatibility with network safety. The Java language has been chosen to implement this task. This paper deals with details of the JT-60 MMIF for the remote experiment that has evolved using the Java language

Progress of JT-60SA Project: EU-JA joint efforts for assembly and fabrication of superconducting tokamak facilities and its research planning

Energy Technology Data Exchange (ETDEWEB)

Shirai, Hiroshi, E-mail: shirai.hiroshi@jaea.go.jp [JT-60SA Project Team, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Barabaschi, Pietro [JT-60SA EU-Home Team, Fusion for Energy, Boltsmannstr 2, Garching 85748 (Germany); Kamada, Yutaka [JT-60SA JA-Home Team, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2016-11-01

Highlights: • JT-60SA Project is promoted under the BA Agreement and JA national programme. • JT-60SA is designed to operate in break-even equivalent condition for a long period. • JT-60SA Project supports and complements the ITER project, and promotes DEMO design. • Fabrication of JT-60SA components and assembly in Naka are steadily going on. • JT-60SA Research Plan has been developed jointly by EU and JA fusion communities. - Abstract: Aiming at supporting the early realization of fusion energy, the JT-60SA Project has shown steady progress for several years toward the first plasma in 2019 under the dual frameworks: the Satellite Tokamak Programme of the Broader Approach Agreement between EU and Japan, and the Japanese national programme. JT-60SA is a superconducting tokamak designed to operate in break-even equivalent conditions for a long pulse duration (typically 100 s) with a maximum plasma current of 5.5 MA. A variety of plasma control capabilities enable JT-60SA to contribute directly to the ITER project and also to DEMO by addressing key engineering and physics issues for advanced plasma operation. Design and fabrication of JT-60SA components, shared by the EU and Japan, started in 2007. Assembly in the torus hall started in January 2013, and welding work of the vacuum vessel sectors (seven 40° sectors and two 30° sectors) is currently ongoing on the cryostat base. Other components such as TF coils, PF coils, power supplies, cryogenic system, cryostat vessel, thermal shields and so on were or are being delivered to the Naka site for installation, assembly and commissioning. This paper gives technical progress on fabrication, installation and assembly of tokamak components and ancillary systems, as well as progress of the JT-60SA Research Plan being developed jointly by European and Japanese fusion communities.

Tritium processing in JT-60U

International Nuclear Information System (INIS)

Miya, Naoyuki; Masaki, Kei

1997-01-01

Tritium retention analysis and tritium concentration measurement have been made during the large Tokamak JT-60U deuterium operations. This work has been carried out to evaluate the tritium retention for graphite tiles inside the vacuum vessel and tritium release characteristics in the tritium cleanup operations. JT-60U has carried out D-D experiments since July 1991. In the deuterium operations during the first two years, about 1.7 x 10 19 D-D fusion neutrons were produced by D (d, p) T reactions in plasma, which are expected to produce ∼31 GBq of tritium. The tritium produced is evacuated by a pumping system. A part of tritium is, however, trapped in the graphite tiles. Several sample tiles were removed from the vessel and the retained tritium Distribution in the tiles was measured using a liquid scintillator. The results of poloidal distribution showed that the tritium concentration in the divertor tiles was higher than that in the first wall tiles and it peaked in the tiles between two strike points of divertor magnetic lines. Tritium concentration in the exhaust gas from the vessel have also been measured with an ion chamber during the tritium cleanup operations with hydrogen divertor discharges and He-GDC. Total of recovered tritium during the cleanup operations was ∼ 7% of that generated. The results of these measurements showed that the tritium of 16-23 GBq still remained in the graphite tiles, which corresponded to about 50-70% of the tritium generated in plasma. The vessel is ventilated during the in-vessel maintenance works, then the atmosphere is always kept lower than the legal concentration guide level of 0.7 Bq/cm 3 for radiation work permit requirements. (author)

Bow-shaped toroidal field coils

International Nuclear Information System (INIS)

Bonanos, P.

1981-05-01

Design features of Bow-Shaped Toroidal Field Coils are described and compared with circular and D shaped coils. The results indicate that bow coils can produce higher field strengths, store more energy and be made demountable. The design offers the potential for the production of ultrahigh toroidal fields. Included are representative coil shapes and their engineering properties, a suggested structural design and an analysis of a specific case

Review of JT-60 experiment (April-June, 1985)

International Nuclear Information System (INIS)

1986-11-01

Initial ohmic heating experiments in JT-60 were performed for a three month period of April-June 1985. A maximum plasma current of 1.6 MA was obtained for both divertor and limiter discharges. Low-q discharges of q eff = 2.5 and high density discharges of 4.8 x 10 19 m -3 were obtained in the divertor configuration. Typical divertor actions, i.e. particle exhaust, heat exhaust, impurity reduction and remote radiative cooling were demonstrated. (author)

The design study of the JT-60SU device. No.8. Nuclear shielding and safety design

Energy Technology Data Exchange (ETDEWEB)

Miya, Naoyuki; Kikuchi, Mitsuru; Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1998-03-01

Results of nuclear shielding design study and safety analysis for the steady-state tokamak device JT-60SU are described. D-T operation (option) for two years is adopted in addition to ten years operation using deuterium. Design work has been done in accordance with general laws for radioisotopes handling in Japan as a guideline of safety evaluation, which is applied to the operation of present JT-60U device. Optimization of the shielding design for the device structure including vacuum vessel has been presented to meet with allowable limits of biological shielding determined in advance. It is shown that JT-60SU can be operated safely in the present JT-60 experimental building. It is planed to use 100g/year of tritium in D-T operation phase. A concept of multiple -barrier system is applied to the facility design to prevent propagation of tritium, in which the torus hall and the tritium removal room provide the tertiary confinement. From the design of atmosphere detritiation system for accidental tritium release, it is shown that tritium concentration level can be reduced to the allowable level after two weeks with reasonable compact size components. Safety assessment related to activation of coolant/air, and atmospheric tritium effluents are discussed. (author)

Design, fabrication and test of double-wall vacuum vessel for JT-60U

International Nuclear Information System (INIS)

Uchikawa, Takashi; Ioki, Kimihiro; Ninomiya, Hiromasa.

1994-01-01

A double-wall vacuum vessel was designed and fabricated for JT-60U (an upgraded machine of JT-60), which has a plasma current up to 6 MA and a large plasma volume (100 m 3 ). A new concept of Inconel 625 all-welded structure was adopted to the vessel, that comprises an inner plate, square tubes and an outer plate. The vacuum vessel with a multi-arc D-shaped cross section was fabricated by using hot-sizing press. The electromagnetic and structural analysis has been performed for plasma disruption loads. Dynamic responses of the vessel were measured during plasma disruptions, and the observed displacement had a good agreement with the result of FEM analysis. (author)

Core transport properties in JT-60U and JET identity plasmas

NARCIS (Netherlands)

Litaudon, X.; Sakamoto, Y.; de Vries, P. C.; Salmi, A.; Tala, T.; Angioni, C.; Benkadda, S.; Beurskens, M. N. A.; Bourdelle, C.; Brix, M.; Crombe, K.; Fujita, T.; Futatani, S.; Garbet, X.; Giroud, C.; Hawkes, N. C.; Hayashi, N.; Hoang, G. T.; Hogeweij, G. M. D.; Matsunaga, G.; Nakano, T.; Oyama, N.; Parail, V.; Shinohara, K.; Suzuki, T.; Takechi, M.; Takenaga, H.; Takizuka, T.; Urano, H.; Voitsekhovitch, I.; Yoshida, M.

2011-01-01

The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma

Development of a new discharge control system utilizing UNIX workstations and VME-bus systems for JT-60

Energy Technology Data Exchange (ETDEWEB)

Akasaka, Hiromi; Sueoka, Michiharu; Takano, Shoji; Totsuka, Toshiyuki; Yonekawa, Izuru; Kurihara, Kenichi; Kimura, Toyoaki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

2002-01-01

The JT-60 discharge control system, which had used HIDIC-80E 16 bit mini-computers and CAMAC systems since the start of JT-60 experiment in 1985, was renewed in March, 2001. The new system consists of a UNIX workstation and a VME-bus system, and features a distributed control system. The workstation performs message communication with a VME-bus system and controllers of JT-60 sub-systems and processing for discharge control because of its flexibility to construction of a new network and modifications of software. The VME-bus system performs discharge sequence control because it is suitable for fast real time control and flexible to the hardware extension. The replacement has improved the control function and reliability of the discharge control system and also has provided sufficient performance necessary for future modifications of JT-60. The new system has been running successfully since April 2001. The data acquisition speed was confirmed to be twice faster than the previous one. This report describes major functions of the discharge control system, technical ideas for developing the system and results of the initial operation in detail. (author)

Tokamak with liquid metal toroidal field coil

International Nuclear Information System (INIS)

Ohkawa, T.; Schaffer, M.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

Communication systems in JT-60 control

International Nuclear Information System (INIS)

Kimura, T.; Hosogane, N.; Kondo, I.; Kumahara, T.; Kurihara, K.; Yonekawa, I.; Yoshino, R.

1983-01-01

A new concept in communication is applied to the JT-60 control system which handles a large amount of data for the plant support and monitoring and for the discharge control including plasma feedback control. The communication systems are characterized by 1) adoption of an efficient protocol in the central highways which are composed of dual serial CAMAC ones, 2) standardization of the protocol and data format between the central controller and each subsystem one, 3) adoption of a polling method for plant monitoring and of block transfer for discharge conditions and results, and 4) use of novel modules for the fast data transfer in the real-time systems. A compact tool has also been developed for testing the data communication

Joule loss on a Faraday shield of JT-60 ICRF test antenna

International Nuclear Information System (INIS)

Fujii, Tsuneyuki; Saigusa, Mikio; Ikeda, Yoshitaka; Kimura, Haruyuki; Hirashima, Teruhisa; Uehara, Munenori.

1988-01-01

Joule loss on a Faraday shield of JT-60 ICRF test antenna with a conductive casing is investigated at the frequency range of 120 MHz. The magnetic field radiated from the antenna is measured by three-dimensionally scanning an rf probe both inside and outside the antenna casing. The magnetic field perpendicular to the Faraday shield, B x , is found to be the largest component near the Faraday shield. It consequently gives the major part of the joule loss on the Faraday shield. The temperature distribution of the Faraday shield due to joule loss is measured directly with a thermocamera. It is confirmed that the area of the high temperature rise is consistent with the peak positions of the B x field. Faraday shield resistance which is estimated from power measurements agrees with the theoretical value. (author)

Improvement of JT-60U Negative Ion Source Performance

International Nuclear Information System (INIS)

Grisham, L.R.; Kuriyama, M.; Kawai, M.; Itoh, T.; Umeda, N.

2000-01-01

The negative ion neutral beam system now operating on JT-60U was the first application of negative ion technology to the production of beams of high current and power for conversion to neutral beams, and has successfully demonstrated the feasibility of negative ion beam heating systems for ITER and future tokamak reactors [1, 2]. It also demonstrated significant electron heating[3] and high current drive efficiency in JT-60U[4]. Because this was such a large advance in the state of the art with respect to all system parameters, many new physical processes appeared during the earlier phases of the beam injection experiments. We have explored the physical mechanisms responsible for these processes, and implemented solutions for some of them, in particular excessive beam stripping, the secular dependence of the arc and beam parameters, and nonuniformity of the plasma illuminating the beam extraction grid. This has reduced the percentage of beam heat loading on the downstream grids by roug hly a third, and permitted longer beam pulses at higher powers. Progress is being made in improving the negative ion current density, and in coping with the sensitivity of the cesium in the ion sources to oxidation by tiny air or water leaks, and the cathode operation is being altered

A new workstation based man/machine interface system for the JT-60 Upgrade

International Nuclear Information System (INIS)

Yonekawa, I.; Shimono, M.; Totsuka, T.; Yamagishi, K.

1992-01-01

Development of a new man/machine interface system was stimulated by the requirements of making the JT-60 operator interface more 'friendly' on the basis of the past five-year operational experience. Eleven Sun/3 workstations and their supervisory mini-computer HIDIC V90/45 are connected through the standard network; Ethernet. The network is also connected to the existing 'ZENKEI' mini-computer system through the shared memory on the HIDIC V90/45 mini-computer. Improved software, such as automatic setting of the discharge conditions, consistency check among the related parameters and easy operation for discharge result data display, offered the 'user-friendly' environments. This new man/machine interface system leads to the efficient operation of the JT-60. (author)

Development of Gyrotron and JT-60U EC heating system for fusion reactor

International Nuclear Information System (INIS)

Sakamoto, K.; Kasugai, A.; Ikeda, Yo.

2003-01-01

The progress of ECH technology, for ITER and JT-60U tokamak, are presented. In the development of gyrotron, 0.9MW/9.2sec, 0.5MW/30sec, 0.3MW/60sec, etc. have been demonstrated at 170GHz. At 110GHz, 1.3MW/1.2sec, 1.2MW/4.1sec, 1MW/5sec were obtained. It is found that the reduction of the stray radiation and the enhancement of cooling capability are keys for CW operation. Four 110GHz gyrotrons are under operation in the ECH system of JT-60U. The power up to approximately 3MW/2.7sec was injected into the plasma through the poloidally movable mirrors, and contributed to the electron heating up to 26keV(n e ∼0.5x10 13 cm -3 ), and the suppression of the neo-classical tearing mode. (author)

Development of gyrotron and JT-60U EC heating system for fusion reactor

International Nuclear Information System (INIS)

Sakamoto, K.; Kasugai, A.; Ikeda, Yo.

2003-01-01

The progress of ECH technology, for ITER and JT-60U tokamak, are presented. In the development of gyrotron, 0.9MW/9.2sec, 0.5MW/30sec, 0.3MW/60sec, etc. have been demonstrated at 170GHz. At 110GHz, 1.3MW/1.2sec, 1.2MW/4. 1sec. 1MW/5sec were obtained. It is found that the reduction of the stray radiation and the enhancement of cooling capability are keys for CW operation. Four 110GHz gyrotrons are under operation in the ECH system of JT-60U. The power up to approximately 3MW/2.7sec was injected into the plasma through the poloidally movable mirrors, and contributed to the electron heating up to 26keV(n e ∼0.5x10 13 cm -3 ), and the suppression of the neo-classical tearing mode. (author)

Development of a VME and CAMAC based data acquisition and transfer system for JT-60 control

International Nuclear Information System (INIS)

Totsuka, Toshiyuki

1993-08-01

Development of a VME and CAMAC based data acquisition and transfer system for JT-60 Control is reported. The present data acquisition and transfer system in JT-60 control is basically composed of CAMAC devices. Since the system equipped with 16-bit microcomputers was manufactured more than ten years ago, the performance and program development environment of the system are apparently worse than those of modern 32-bit microcomputers. To improve these disadvantages, a new data acquisition and transfer system using VME-based 32-bit microcomputers and CAMAC drivers is under design. Corresponding to this design, a CAMAC handler, which runs on the microcomputer, for the VME based CAMAC driver was newly developed. Moreover, the functions of the driver and data transfer performance of the VME and CAMAC complex system were tested. The test results shown that the VME based microcomputer and CAMAC serial driver can be applied for the fast and reliable acquisition and transfer system for JT-60 control. (author)

Development of a VME multi-processor system for plasma control at the JT-60 Upgrade

International Nuclear Information System (INIS)

Takahashi, M.; Kurihara, K.; Kawamata, Y.; Akasaka, H.; Kimura, T.

1992-01-01

Design and initial operation results are reported of a VME multi-processor system [1] for plasma control at a large fusion device named 'the JT-60 Upgrade' utilizing three 32-bit MC88100 based RISC computers and VME components. Development of the system was stimulated by faster and more accurate computation requirements for the plasma position and current control. The RISC computers operate at 25 MHz along with two cashe memories named MC88200. We newly developed VME bus modules of up/down counter, analog-to-digital converter and clock pulse generator for measuring magnetic field and coil current and for synchronizing the processing in the three RISCs and direct digital controllers (DDCs) of magnet power supplies. We also evaluated that the speed of the data transfer between the VME bus system and the DDCs through CAMAC highways satisfies the above requirements. In the initial operation of the JT-60 upgrade, it has been proved that the VME multi-processor system well controls the plasma position and current with a sampling period of 250 μsec and a delay of 500 μsec. (author)

VME multiprocessor system for plasma control at the JT-60 Upgrade

International Nuclear Information System (INIS)

Kimura, T.; Kurihara, K.; Takahashi, M.; Kawamata, Y.; Akasaka, H.; Matsukawa, M.

1989-01-01

In this paper design and preliminary tests are reported of a VME multiprocessor system for the JT-60 Upgrade plasma control utilizing three MC88100 based RISC computers and VME buses. The design of the VME system was stimulated by faster and more accurate computation requirements for the plasma position and shape control

Global energy confinement in JT-60 neutral beam heated L-mode discharges

International Nuclear Information System (INIS)

Naito, O.; Hosogane, N.; Tsuji, S.; Ushigusa, K.; Yoshida, H.

1990-01-01

The global energy confinement characteristics of neutral beam heated JT-60 discharges are presented. There is a difference in the dependence of the energy confinement time on the plasma current between limiter and divertor discharges. For limiter discharges, the energy confinement increases with plasma current up to 3.2 MA, whereas for divertor discharges this improvement saturates when the safety factor drops below 3, independent of the location of the X-point. The JT-60 L-mode results indicate that the deterioration in energy confinement for q < 3, which is also found in H-mode regimes of other devices, may be a universal characteristic of divertor discharges. Regarding the scaling with plasma size, it is shown that the global/incremental confinement time increases with plasma minor radius. For sufficiently large plasmas, however, the global/incremental confinement time is no longer a function of minor radius. (author). 13 refs, 14 figs

PDX toroidal field coils stress analysis

International Nuclear Information System (INIS)

Nikodem, Z.D.; Smith, R.A.

1975-01-01

A method used in the stress analysis of the PDX toroidal field coil is developed. A multilayer coil design of arbitrary dimensions in the shape of either a circle or an oval is considered. The analytical model of the coil and the supporting coil case with connections to the main support structure is analyzed using the finite element technique. The three dimensional magnetic fields and the non-uniform body forces which are a loading condition on a coil due to toroidal and poloidal fields are calculated. The method of analysis permits rapid and economic evaluations of design changes in coil geometry as well as in coil support structures. Some results pertinent to the design evolution and their comparison are discussed. The results of the detailed stress analysis of the final coil design due to toroidal field, poloidal field and temperature loads are presented

Physics and operation oriented activities in preparation of the JT-60SA tokamak exploitation

Science.gov (United States)

Giruzzi, G.; Yoshida, M.; Artaud, J. F.; Asztalos, Ö.; Barbato, E.; Bettini, P.; Bierwage, A.; Boboc, A.; Bolzonella, T.; Clement-Lorenzo, S.; Coda, S.; Cruz, N.; Day, Chr.; De Tommasi, G.; Dibon, M.; Douai, D.; Dunai, D.; Enoeda, M.; Farina, D.; Figini, L.; Fukumoto, M.; Galazka, K.; Galdon, J.; Garcia, J.; Garcia-Muñoz, M.; Garzotti, L.; Gil, C.; Gleason-Gonzalez, C.; Goodman, T.; Granucci, G.; Hayashi, N.; Hoshino, K.; Ide, S.; Imazawa, R.; Innocente, P.; Isayama, A.; Itami, K.; Joffrin, E.; Kamada, Y.; Kamiya, K.; Kawano, Y.; Kawashima, H.; Kobayashi, T.; Kojima, A.; Kubo, H.; Lang, P.; Lauber, Ph.; de la Luna, E.; Maget, P.; Marchiori, G.; Mastrostefano, S.; Matsunaga, G.; Mattei, M.; McDonald, D. C.; Mele, A.; Miyata, Y.; Moriyama, S.; Moro, A.; Nakano, T.; Neu, R.; Nowak, S.; Orsitto, F. P.; Pautasso, G.; Pégourié, B.; Pigatto, L.; Pironti, A.; Platania, P.; Pokol, G. I.; Ricci, D.; Romanelli, M.; Saarelma, S.; Sakurai, S.; Sartori, F.; Sasao, H.; Scannapiego, M.; Shimizu, K.; Shinohara, K.; Shiraishi, J.; Soare, S.; Sozzi, C.; Stępniewski, W.; Suzuki, T.; Suzuki, Y.; Szepesi, T.; Takechi, M.; Tanaka, K.; Terranova, D.; Toma, M.; Urano, H.; Vega, J.; Villone, F.; Vitale, V.; Wakatsuki, T.; Wischmeier, M.; Zagórski, R.

2017-08-01

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised.

ASDEX Upgrade-JT-60U comparison and ECRH power requirements for NTM stabilization in ITER

International Nuclear Information System (INIS)

Urso, L.; Zohm, H.; Maraschek, M.; Poli, E.; Isayama, A.

2010-01-01

Neoclassical tearing modes (NTMs) are experimentally controlled with local electron cyclotron current drive (ECCD) and the island width decay during NTM stabilization is modelled using the so-called modified Rutherford equation (MRE). In this paper, a modelling of the MRE is carried out and simulations of the island width decay are compared with the experimentally observed ones in order to fit the two free machine-independent parameters present in the equation. A systematic study on a database of NTM stabilization discharges from ASDEX Upgrade and JT-60U is done for extrapolating the ECCD power requirements for ITER. The extrapolation to ITER of the NTM stabilization results from ASDEX Upgrade and JT-60U shows that 10 MW of ECCD power are enough to stabilize large NTMs. The 10 MW power estimate for ITER is based on the assumption that the free parameters in the MRE are machine independent. Indeed, this assumption is verified in this paper for ASDEX Upgrade and JT-60U. An interesting consequence of the relatively modest power requirement for ITER is that the installed 20 MW will suffice for simultaneous 2/1 and 3/2 NTM stabilization.

Estimation of the lifetime of resin insulators against baking temperature for JT-60SA in-vessel coils

Energy Technology Data Exchange (ETDEWEB)

Sukegawa, Atsuhiko M., E-mail: morioka.atsuhiko@jaea.go.jp; Murakami, Haruyuki; Matsunaga, Go; Sakurai, Shinji; Takechi, Manabu; Yoshida, Kiyoshi; Ikeda, Yoshitaka

2015-10-15

Highlights: • The lifetime of resin insulators at about 200 °C was estimated. • We make use of the Arrhenius plot by the Weibull analysis for the estimation. • A suitable temperatures for the in-vessel coils were discussed. - Abstract: In the present study, the thermal endurance of epoxy-based, bismaleimides, and cyanate ester resins for the current design of the in-vessel coils was measured by performing acceleration tests to assess their insulation properties using the thermal endurance defined by the International Electrotechnical Commission (IEC-60216 Part1–Part 6) for a minimum of 5,000 h in the 180–240 °C temperature range. It was found that none of the resin insulators could tolerate the baking conditions of 40,000 h at ∼200 °C in the JT-60SA vacuum vessel. Therefore, the design of the in-vessel coils, including the error field correction coils (EFCC), was changed from the type without water cooling to with water cooling on JT-60SA.

JT-60 power tests from mechanical and thermal viewpoints of tokamak machine

International Nuclear Information System (INIS)

Takatsu, H.; Yamamoto, M.; Ohkubo, M.

1986-01-01

JT-60 power tests were carried out, to demonstrate, in advance of actual plasma operation, satisfactory performance of the tokamak machine, power suppliers and control system in combination. The tests began with low power ones of individual coil systems, progressed to full power ones and concluded successfully. The present paper describes the principal results of JT-60 power tests from mechanical and thermal viewpoints of tokamak machine. All of the coil systems were raised up to full power operation in combination and system performance was verified including thermal and mechanical integrity of tokamak machine. Measured strain and displacement showed good agreements with those predicted in the design, which was an evidence that electromagnetic loads were supported adequately as expected in the design. Vibration of the vacuum vessel was found to be large up to 48 m/s/sup 2/ and caused excessive vibration of the lateral port gate-valves. A few limitations to machine operation were also made clear quantatively

The healthiness of JT-60 ICRF antenna and development of its temperature measurement device

International Nuclear Information System (INIS)

Hiranai, Shinichi; Yokokura, Kenji; Moriyama, Shinichi; Sato, Tomio; Ishii, Kazuhiro; Fujii, Tsuneyuki

1998-03-01

Ion Cyclotron Range of Frequency (ICRF) heating system in JT-60 employs two antennas to couple RF power in the range of 100 MHz to the plasma. The antennas are installed in the vacuum vessel of JT-60, facing to the high temperature plasma. Due to the severe heat load from the plasma, parts of the antenna surface are suffering from melt. It is important to investigate the mechanism of the heat load and the melting. 'Temperature measurement for ICRF antenna surface' employing an infrared thermographic camera has been developed, in order to investigate the heat load to the antenna and to maintain the antenna available. We have succeeded in minimizing the melting damage of the antenna surface using the temperature measurement device. (author)

Design study of a new P-NBI control system for 100-s injection in JT-60SA

International Nuclear Information System (INIS)

Honda, Atsushi; Okano, Fuminori; Shinozaki, Shinichi; Ooshima, Katsumi; Ikeda, Yoshitaka; Numazawa, Susumu

2007-03-01

The modification of the JT-60U to a fully superconducting coil tokamak, JT-60SA (Super Advanced), has been programmed as the satellite devise for the ITER (International Thermonuclear Experimental Reactor) and as the national centralized tokamak. The present positive-ion-based NBI system (P-NBI), which has been operated for 20 years and will be the main heating system on JT-60SA, is required to manage the long pulse injection extended from 30 s to 100 s at the power of 24 MW with 12 units. To realize such a requirement, the original control system handling more than 4000 digital data is to be fully remodeled. Design study of the new control system has been conducted from viewpoint of market availability, system extensibility, cost-effectiveness and independent development in programming. It has been concluded that a distributed control system using PLC (Programmable Logic Controller) could be applied to the large-scale control system for 100-s operations with satisfaction of the evaluation viewpoints. (author)

Results of the H-mode experiments with JT-60 outer and lower divertors

International Nuclear Information System (INIS)

Nakamura, Hiroo; Tsuji, Shunji; Nagami, Masayuki

1989-08-01

In JT-60, hydrogen H-mode experiments with outer and lower divertors were performed. In the outer divertor, H-mode were obtained, similar to the ones observed in the other lower/upper divertors. Its threshold absorbed power and electron density were 16 MW and 1.8 x 10 19 m -3 . In the two combined heatings with NB+ICRF and NB+LHRF, H-mode discharges are also obtained. Moreover, in new configuration of lower divertor, H-mode phases without and with ELM are obtained. Typical results of the lower divertor are shown to compare the H-mode characteristics between the two configurations. Improvement of the energy confinement time in the two divertors was limited to 10 %. Analyses on ballooning/interchange instabilities were carried out with precise equlibria of JT-60. These results showed that the both modes were enough stable. (author)

Suppression of m = 0 in a RFP by toroidal field coils

International Nuclear Information System (INIS)

Alexander, D.; Robertson, S.

1993-01-01

The Reversatron RFP is normally operated with the toroidal field coils connected in series. The time-integrated voltage applied to the circuit determines the sum of the fluxes linking each turn but not the flux within each turn. Each winding may have a different flux determined by the external drive and by currents within the plasma. A parallel connection of the field coils results in the flux within each coil being determined by the volt-seconds applied to the windings; thus the toroidal flux is the same within each coil. This configuration suppresses any toroidal variation in the toroidal flux and effectively reduces the level of the m = 0 component of the radial field. The m = 0 fluctuations are expected to arise due to nonlinear coupling of the m = 1 modes. A parallel connection of field coils is impractical due to the low impedance required for driving the coils. The authors have tested the effect of parallel connected coils by adding an auxiliary set of 36 coils. These are connected in parallel but are not connected to any supply. The toroidal flux is generated by the series-connected coils which generate voltage but not current in the parallel-connected coils. With the auxiliary coils, the discharge duration is increased from 500 to 550 μsec, the plasma current is increased from 50 kA to 60 kA, F is more negative, Θ is larger, and there is less shot-to-shot variation in the discharges. The m = 0 fluctuations measured by 43 surface coils are, however, only slightly reduced

Spherical tokamak without external toroidal fields

International Nuclear Information System (INIS)

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

2001-01-01

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

Machine performance and its effects on experiments in JT-60U

International Nuclear Information System (INIS)

Kondo, I.

1995-01-01

The operational results of JT-60U were reviewed in light of the strategy made at the design stage. The operational plan for better confinement shifted from that of low q to high poloidal beta plasma configuration with higher q value according to the revealed machine properties. Some technical and operational skills helped bring about the recent results out of the machine. (orig.)

Effects of Resonant Helical Field on Toroidal Field Ripple in IR-T1 Tokamak

Science.gov (United States)

Mahdavipour, B.; Salar Elahi, A.; Ghoranneviss, M.

2018-02-01

The toroidal magnetic field which is created by toroidal coils has the ripple in torus space. This magnetic field ripple has an importance in plasma equilibrium and stability studies in tokamak. In this paper, we present the investigation of the interaction between the toroidal magnetic field ripple and resonant helical field (RHF). We have estimated the amplitude of toroidal field ripples without and with RHF (with different q = m/n) ( m = 2, m = 3, m = 4, m = 5, m = 2 & 3, n = 1) using “Comsol Multiphysics” software. The simulations show that RHF has effects on the toroidal ripples.

Steady state toroidal magnetic field at earth's core-mantle boundary

Science.gov (United States)

Levy, Eugene H.; Pearce, Steven J.

1991-01-01

Measurements of the dc electrical potential near the top of earth's mantle have been extrapolated into the deep mantle in order to estimate the strength of the toroidal magnetic field component at the core-mantle interface. Recent measurements have been interpreted as indicating that at the core-mantle interface, the magnetic toroidal and poloidal field components are approximately equal in magnitude. A motivation for such measurements is to obtain an estimate of the strength of the toroidal magnetic field in the core, a quantity important to our understanding of the geomagnetic field's dynamo generation. Through the use of several simple and idealized calculation, this paper discusses the theoretical relationship between the amplitude of the toroidal magnetic field at the core-mantle boundary and the actual amplitude within the core. Even with a very low inferred value of the toroidal field amplitude at the core-mantle boundary, (a few gauss), the toroidal field amplitude within the core could be consistent with a magnetohydrodynamic dynamo dominated by nonuniform rotation and having a strong toroidal magnetic field.

Power and particle control in JT-60SA to support and supplement ITER and DEMO

International Nuclear Information System (INIS)

Sakurai, Shinji

2007-01-01

JT-60 is planned to be modified as a fully superconducting coil tokamak (JT-60 Super Advanced, JT-60SA). Divertor targets are water-cooled to handle heat flux up to 15 MW/m 2 . JT-60SA allows exploitation of high beta regime with stabilizing shell covered with ferritic plates and internal resistive wall mode (RWM) stabilizing coils. A remote handling system is equipped to maintain in-vessel components even for high dose rate due to a substantial annual neutron production. Divertor cassettes are introduced to be maintained by a remote handling. In the present design, a monoblock type carbon fibre composite (CFC) divertor target will be used to withstand a heat load of ∼15 MW/m 2 . CFC divertor targets and other bolted armor tiles will be mounted on the divertor cassette. All of the plasma facing components including the first wall armor are water-cooled to handle heat load during 100s or more. Divertor heat load and pumping efficiency for an ITER-like configuration has been evaluated, using 2D plasma fluid (SOLDOR) and neutral Monte-Carlo (NEUT2D) code. The pumping speed of 50 m 3 /s is specified at an albedo for neutrals in front of the in-vessel cryopanel. In the simulation for the divertor with a V -shaped corner' like as that in ITER, the plasma detachment occurs near the outer-strike point within the 'V-shaped corner', as well as near the inner-strike point, which results in low peak heat flux density 5.8 MW/m 2 for the case with additional gas puff of 5x10 21 /s compared to 11.4 MW/m 2 for the case without 'V-shaped corner'. (author)

Study of particle pumping characteristics for different pumping geometries in JT-60U and DIII-D divertors

International Nuclear Information System (INIS)

Takenaga, H.; Sakasai, A.; Kubo, H.

2001-01-01

Particle pumping characteristics were compared between pumping from the inner side private flux region (IPP) and pumping from both sides of the private flux region (BPP) in the JT-60U W shaped divertor, and between JT-60U IPP and pumping in the DIII-D lower baffled divertor. The pumping flux for BPP is smaller than that for IPP by about a factor of 2 with weak in-out asymmetry of recycling neutral flux and by a factor of 3.5-6.5 with strong in-out asymmetry. The reduction of the pumping flux for BPP is consistent with Monte Carlo simulations, where backflow at the outer pumping slot is observed due to in-out recycling asymmetry. The pumping flux in DIII-D at I p =0.8 MA and B T =1.6 T is comparable to or smaller than that for JT-60U IPP at I p =1.0 MA, B T =3.8 T and I p =1.5 MA, B T =3.5 T in the same density regime. In the DIII-D divertor with pumping from the private flux region, the pumping flux decreases with increasing in-out asymmetry. The pumping flux normalized by the integrated D α emission over the whole plasma exhibits a similar dependence on the distance between the pumping slot and the strike point in JT-60U IPP and the DIII-D lower divertor with pumping through the outer divertor plasma region. (author)

High Mach flow associated with plasma detachment in JT-60U

International Nuclear Information System (INIS)

Hatayama, A.; Hoshino, K.; Miyamoto, K.

2003-01-01

Recent new results of the high Mach flows associated with plasma detachment are presented on the basis of numerical simulations by a 2-D edge simulation code (the B2-Eirene code) and their comparisons with experiments in JT-60U W-shaped divertor plasma. High Mach flows appear near the ionization front away from the target plate. The plasma static pressure rapidly drops, while the total pressure is kept almost constant near the ionization front, because the ionization front near the X-point is clearly separated from the momentum loss region near the target plate. Redistribution from static to dynamic pressure without a large momentum loss is confirmed to be a possible mechanism of the high Mach flows. It has been also shown that the radial structure of the high Mach flow near the X point away from the target plate has a strong correlation with the DOD (Degree of Detachment) at the target plate. Also, we have made systematic analyses on the high Mach flows for both the 'Open' geometry and the 'W-shaped' geometry of JT-60U in order to clarify the geometric effects on the flows. (author)

Relativistic stars with purely toroidal magnetic fields

International Nuclear Information System (INIS)

Kiuchi, Kenta; Yoshida, Shijun

2008-01-01

We investigate the effects of the purely toroidal magnetic field on the equilibrium structures of the relativistic stars. The basic equations for obtaining equilibrium solutions of relativistic rotating stars containing purely toroidal magnetic fields are derived for the first time. To solve these basic equations numerically, we extend the Cook-Shapiro-Teukolsky scheme for calculating relativistic rotating stars containing no magnetic field to incorporate the effects of the purely toroidal magnetic fields. By using the numerical scheme, we then calculate a large number of the equilibrium configurations for a particular distribution of the magnetic field in order to explore the equilibrium properties. We also construct the equilibrium sequences of the constant baryon mass and/or the constant magnetic flux, which model the evolution of an isolated neutron star as it loses angular momentum via the gravitational waves. Important properties of the equilibrium configurations of the magnetized stars obtained in this study are summarized as follows: (1) For the nonrotating stars, the matter distribution of the stars is prolately distorted due to the toroidal magnetic fields. (2) For the rapidly rotating stars, the shape of the stellar surface becomes oblate because of the centrifugal force. But, the matter distribution deep inside the star is sufficiently prolate for the mean matter distribution of the star to be prolate. (3) The stronger toroidal magnetic fields lead to the mass shedding of the stars at the lower angular velocity. (4) For some equilibrium sequences of the constant baryon mass and magnetic flux, the stars can spin up as they lose angular momentum.

Carbon transport and fuel retention in JT-60U with high temperature operation based on postmortem analysis

Energy Technology Data Exchange (ETDEWEB)

Yoshida, M., E-mail: yoshida.masafumi@jaea.go.jp [Japan Atomic Energy Agency, Mukoyama 801-1, Naka-shi, Ibaraki-ken 311-0193 (Japan); Tanabe, T.; Adachi, A. [Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Hayashi, T.; Nakano, T.; Fukumoto, M.; Yagyu, J.; Miyo, Y.; Masaki, K.; Itami, K. [Japan Atomic Energy Agency, Mukoyama 801-1, Naka-shi, Ibaraki-ken 311-0193 (Japan)

2013-07-15

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas, or tile gaps and remote areas, in JT-60U were measured. The total fuel retention rate in the plasma shadowed areas was more than two times higher than that in the carbon re-deposited layers on the plasma facing surfaces, or the divertor tiles. This is because of lower temperature in the plasma shadowed areas than in the plasma facing surfaces, which leads to high hydrogen saturation concentration, although the amount of the carbon re-deposited on the plasma shadowed areas was only 60% of that on the plasma facing surfaces. The total fuel retention rate in JT-60U, including previously determined for all the plasma facing areas, was evaluated to be 1.3 × 10{sup 20} H + D s{sup −1}, and this retention rate was lower than that in the other devices, due probably to high baking temperature operation in JT-60U. Distributions of the fuel retention and the carbon re-deposition in the whole in-vessel of a large tokamak were determined for the first time in the world.

Carbon transport and fuel retention in JT-60U with high temperature operation based on postmortem analysis

International Nuclear Information System (INIS)

Yoshida, M.; Tanabe, T.; Adachi, A.; Hayashi, T.; Nakano, T.; Fukumoto, M.; Yagyu, J.; Miyo, Y.; Masaki, K.; Itami, K.

2013-01-01

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas, or tile gaps and remote areas, in JT-60U were measured. The total fuel retention rate in the plasma shadowed areas was more than two times higher than that in the carbon re-deposited layers on the plasma facing surfaces, or the divertor tiles. This is because of lower temperature in the plasma shadowed areas than in the plasma facing surfaces, which leads to high hydrogen saturation concentration, although the amount of the carbon re-deposited on the plasma shadowed areas was only 60% of that on the plasma facing surfaces. The total fuel retention rate in JT-60U, including previously determined for all the plasma facing areas, was evaluated to be 1.3 × 10 20 H + D s −1 , and this retention rate was lower than that in the other devices, due probably to high baking temperature operation in JT-60U. Distributions of the fuel retention and the carbon re-deposition in the whole in-vessel of a large tokamak were determined for the first time in the world

High current density toroidal pinch discharges with weak toroidal fields

International Nuclear Information System (INIS)

Brunsell, P.; Brzozowski, J.; Drake, J.R.; Hellblom, G.; Kaellne, E.; Mazur, S.; Nordlund, P.

1990-01-01

Toroidal discharges in the ultralow q regime (ULQ) have been studied in the rebuilt Extrap TI device. ULQ discharges are sustained for pulse lengths exceeding 1 ms, which corresponds to more than 10 resistiv shell times. Values for the safety factor at the vacuum vessel wall are between rational values: 1/(n+1) -2 . The magnetic fluctuation level increases during the transition between rational values of q(a). For very low values of q(a), the loop voltage increases and the toroidal field development in the discharge exhibits the characteristic behaviour of the setting-up phase of a field reversed pinch. (author) 1 ref., 2 figs., 1 tab

Impact of arcing on carbon and tungsten. From the observations in JT-60U, LHD, and NAGDIS-II

International Nuclear Information System (INIS)

Kajita, Shin; Fukumoto, Masakatsu; Nakano, Tomohide; Tokitani, Masayuki; Masuzaki, Suguru; Ohno, Noriyasu; Takamura, Shuichi; Yoshida, Naoaki; Ueda, Yoshio

2012-11-01

This paper assesses the impact of arcing in fusion devices based on the observations in JT-60U, LHD, and the linear divertor simulator NAGDIS-II. In NAGDIS-II, the demonstration experiments of arcing/unipolar arcing have been conducted by simulating the transient heat load using a pulsed laser; it was found that the arcing can be easily initiated on the helium irradiated nanostructured tungsten. By measuring the field emission current property from the helium irradiated tungsten surface, the initiation conditions are discussed. From the detailed analysis of JT-60U tiles, it is found that arcing phenomena occurred on carbon baffle plates. From the observation of the arc trails recorded on the baffle plate, the amount of the eroded materials is discussed. The arcing seemed to occur frequently on inner baffles rather than the outer baffles. From LHD, it is shown that the arcing can be initiated on nanostructured tungsten even without transient events. The erosion of tungsten by arcing will become an important issue in a fusion reactor, where helium fluence is significantly increased. (author)

Design of tangential viewing phase contrast imaging for turbulence measurements in JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Tanaka, K., E-mail: ktanaka@nifs.ac.jp [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Department of Advanced Energy Engineering, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Coda, S. [EPFL–SPC, Lausanne (Switzerland); Yoshida, M.; Sasao, H.; Kawano, Y.; Imazawa, R.; Kubo, H.; Kamada, Y. [National Institutes for Quantum and Radiological Science and Technology, Naka, Ibaraki 311-0193 (Japan)

2016-11-15

A tangential viewing phase contrast imaging system is being designed for the JT-60SA tokamak to investigate microturbulence. In order to obtain localized information on the turbulence, a spatial-filtering technique is applied, based on magnetic shearing. The tangential viewing geometry enhances the radial localization. The probing laser beam is injected tangentially and traverses the entire plasma region including both low and high field sides. The spatial resolution for an Internal Transport Barrier discharge is estimated at 30%–70% of the minor radius at k = 5 cm{sup −1}, which is the typical expected wave number of ion scale turbulence such as ion temperature gradient/trapped electron mode.

Developmental prototype for replacement of JT-60 timing system

International Nuclear Information System (INIS)

Akasaka, H.; Kawamata, Y.; Yonekawa, I.

2004-01-01

The present CAMAC based timing system has been used for synchronizing sequential events of the discharge and the data collection of the interesting JT-60U experiment plasma phenomena. However, a more flexible and sophisticated state-of-the-art timing system now is required to realize advanced plasma control with minimal maintenance costs. In this context, the versa module Europe (VME-bus) system with a high-speed data communication network using reflective memory (RM) modules and user-friendly application software based on MATLAB TM tools has been selected to develop the new prototype timing system. In the ZENKEI, the supervisory control system of the JT-60, the supervisory timing system provides the 50-μs master clock pulses, the various timing signal preparation logic, which is built into the digital signal processing (DSP) module in conjunction with the discharge sequence event signals, and the 6.2 MB/s high-speed communication data link provided by the RM module. Except the clock pulse generator (CPG) module, no other special timing module is necessary for this new timing system. The timing signal is prepared by software logic in conjunction with sequential events and the preset timer, is transferred to the subsystems through the RM module, where it is synchronized to the 50-μs clock pulses. The timing system of the subsystems also consists of hardware similar in structure to the ZENKEI timing system. The fundamental timing system configuration, the necessary functions, and the preliminary test results of the prototype system are reported in this presentation

Physics of strong internal transport barriers in JT-60U reversed-magnetic-shear plasmas

Energy Technology Data Exchange (ETDEWEB)

Hayashi, N; Takizuka, T; Sakamoto, Y; Fujita, T; Kamada, Y; Ide, S; Koide, Y [Japan Atomic Energy Agency, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)

2006-05-15

The physics of strong internal transport barriers (ITBs) in JT-60U reversed-magnetic-shear (RS) plasmas has been studied through the modelling on the 1.5 dimensional transport simulation. The key physics to produce two scalings on the basis of the JT-60U box-type ITB database are identified. As for the scaling for the narrow ITB width proportional to the ion poloidal gyroradius, the following three physics are important: (1) the sharp reduction of the anomalous transport below the neoclassical level in the RS region, (2) the autonomous formation of pressure and current profiles through the neoclassical transport and the bootstrap current and (3) the large difference between the neoclassical transport and the anomalous transport in the normal-shear region. As for the scaling for the energy confinement inside ITB ({epsilon}{sub f}{beta}{sub p,core} {approx} 0.25, where {epsilon}{sub f} is the inverse aspect ratio at the ITB foot and {beta}{sub p,core} is the core poloidal beta value), the value of 0.25 is found to be a saturation value due to the MHD equilibrium. The value of {epsilon}{sub f}{beta}{sub p,core} reaches the saturation value, when the box-type ITB is formed in the strong RS plasma with a large asymmetry of the poloidal magnetic field, regardless of the details of the transport and the non-inductively driven current.

Kinetic global analysis of Alfven eigenmodes in toroidal plasmas

International Nuclear Information System (INIS)

Fukuyama, A.

2002-01-01

Systematic study on low to medium n (toroidal mode number) Alfven eigenmodes (AE) in tokamaks and helical systems is presented. Linear stability of AE in the presence of energetic ions was studied using the kinetic full-wave code TASK/WM.We have reproduced the destabilizing effect of toroidal co-rotation on TAE for JT-60U parameters. We have found the existence of reversed-shear-induced Alfven eigenmode (RSAE) which localizes near the q minimum in a reversed magnetic shear configuration. Two kinds of mode structures are identified for energetic particle mode (EPM) below the TAE frequency gap. The coupling to lower-frequency modes such as drift waves and MHD modes as well as the effect of trapped particles are also taken into account. For a helical plasma, the existence of GAE in the central region and TAE in the off-axis region was confirmed. (author)

Toroidal field ripple effects in large tokamaks

International Nuclear Information System (INIS)

Uckan, N.A.; Tsang, K.T.; Callen, J.D.

1975-01-01

In an experimental power reactor, the ripple produced by the finite number of toroidal field coils destroys the ideal axisymmetry of the configuration and is responsible for additional particle trapping, loss regions and plasma transport. The effects of toroidal field ripple on the plasma transport coefficient, the loss of alpha particles and energetic injection ions, and the relaxation of toroidal flows are investigated in a new and systematic way. The relevant results are applied to the ORNL-EPR reference design; the maximum ripple there of about 2.2 percent at the outer edge of the plasma column is found to be tolerable from plasma physics considerations

Improvement of the real-time processor in JT-60 data processing system

International Nuclear Information System (INIS)

Sakata, S.; Kiyono, K.; Sato, M.; Kominato, T.; Sueoka, M.; Hosoyama, H.; Kawamata, Y.

2009-01-01

Real-time processor, RTP is a basic subsystem in the JT-60 data processing system and plays an important role in JT-60 feedback control for plasma experiment. During the experiment, RTP acquires various diagnostic signals, processes them into a form of physical values, and transfers them as sensor signals to the particle supply and heating control supervisor for feedback control via reflective memory synchronization with 1 ms clock signals. After the start of RTP operation in 1997, to meet the demand for advanced plasma experiment, RTP had been improved continuously such as by addition of diagnostic signals with faster digitizers, reducing time for data transfer utilizing reflective memory instead of CAMAC. However, it is becoming increasingly difficult to maintain, manage, and improve the outdated RTP with limited system CPU capability. Currently, a prototype RTP system is being developed for the next real-time processing system, which is composed of clustered system utilizing VxWorks computer. The processes on the existing RTP system will be decentralized to the VxWorks computer to solve the issues of the existing RTP system. The prototype RTP system will start to operate in August 2008.

Characteristics of divertor plasma and scrape-off layer in JT-60U

International Nuclear Information System (INIS)

Itami, K.; Shimada, M.; Hosogane, N.

1992-01-01

Heat flux to the divertor is measured by thermography and the heat transport in the scrape-off layer is studied in beam heated discharges of JT-60U. The heat flux onto the divertor is ∝50% of total beam power at maximum. The in-out asymmetry of the heat flux P HEAT in /P HEAT out is as large as 20-40% when the ion grad-B drift is toward the divertor. Differences in P HEAT in /P HEA T out due to the direction of ion grad-B drift are as large as large as ∝40%. A scaling of the peaking factor Y of heat flux, defined by Y=2πRfq max /P HEAT , is obtained for beam heated discharges in JT-60U with a wide range of plasma parameters. The Y corresponds to the inverse of the thickness of the scrape-off layer. From a statistical analysis, it is found that the peaking factor Y of heat flux scales as P HEAT 0.49±0.18 anti n e -0.45±0.22 q eff -0.67±0.18 . (orig.)

Long Pulse Operation on NBI Systems for JT-60U

International Nuclear Information System (INIS)

Akino, N.; Ebisawa, N.; Honda, A.; Ikeda, Y.; Kawai, M.; Kazawa, M.; Mogaki, K.; Ohga, T.; Umeda, N.; Usui, K.; Yamamoto, T.; Grisham, L.

2005-01-01

In the neutral beam injection (NBI) system, an extension of the pulse duration up to 30 sec has been intended to study quasi-steady state plasma on JT-60U. The four positive-ion based (P-NBI) units, which tangentially inject neutral beam to plasma, were mainly modified on the electric power supplies and the beam limiters to extend the pulse duration up to 30 sec with 2 MW at 80 keV per each. The seven P-NBI units, each of which perpendicularly injects for 10 sec, were conducted to operate in series for the total pulse duration of 30 sec. The ion source of the negative-ion based (N-NBI) unit, whose target beam energy is 500 keV for 10 sec, was also modified to reduce the heat load of the grid for long pulse operation. The reduction of the re-ionization of the neutral beam in the beam drift duct was a key to achieve a long pulse injection. It was found that the pressure rise in the beam drift duct, which gives the re-ionization rate, depended on the temperature of the re-ionization plates during NBI injection. Up to now, it was attained successfully that the pulse duration of the tangential P-NBI unit was extended up to 30 sec. 310 MJ of the total integrated injection energy into JT-60U plasma was achieved, including the negative-ion based NBI operation for 17 sec at 366 keV

Formation conditions for electron internal transport barriers in JT-60U plasmas

Energy Technology Data Exchange (ETDEWEB)

Fujita, T [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Fukuda, T [Osaka University, Suita, Osaka 565-0871 (Japan); Sakamoto, Y [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Ide, S [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Suzuki, T [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Takenaga, H [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Ida, K [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Idei, H [Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Shimozuma, T [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Fujisawa, A [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Ohdachi, S [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Toi, K [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan)

2004-05-01

The formation of electron internal transport barriers (ITBs) was studied using electron cyclotron (EC) heating in JT-60U positive shear (PS) and reversed shear (RS) plasmas with scan of neutral beam (NB) power. With no or low values of NB power and with a small radial electric field (E{sub r}) gradient, a strong, box-type electron ITB was formed in RS plasmas while a peaked profile with no strong electron ITBs was observed in PS plasmas within the available EC power. When the NB power and the E{sub r} gradient were increased, the electron transport in strong electron ITBs with EC heating in RS plasmas was not affected, while electron thermal diffusivity was reduced in conjunction with the reduction of ion thermal diffusivity, and strong electron and ion ITBs were formed in PS plasmas.

The plasma movie database system for JT-60

International Nuclear Information System (INIS)

Sueoka, Michiharu; Kawamata, Yoichi; Kurihara, Kenichi; Seki, Akiyuki

2007-01-01

The real-time plasma movie with the computer graphics (CG) of plasma shape is one of the most effective methods to know what discharge have been made in the experiment. For an easy use of the movie in the data analysis, we have developed the plasma movie database system (PMDS), which automatically records plasma movie according to the JT-60 discharge sequence, and transfers the movie files on request from the web site. The file is compressed to about 8 MB/shot small enough to be transferred within a few seconds through local area network (LAN). In this report, we describe the developed system from the technical point of view, and discuss a future plan on the basis of advancing video technology

Toroidal field effects on the stability of Heliotron E

International Nuclear Information System (INIS)

Carreras, B.A.; Garcia, L.; Lynch, V.E.

1986-02-01

The addition of a small toroidal field to the Heliotron E configuration improves the stability of the n = 1 mode and increases the value of the stability beta critical. Total stabilization of this mode can be achieved with added toroidal fields between 5 and 15% of the total field. In this situation, the plasma can have direct access to the second stability regime. For the Heliotron E configuration, the self-stabilization effect is due to the shear, not to the magnetic well. The toroidal field threshold value for stability depends strongly on the pressure profile and the plasma radius. 21 refs., 15 figs

Dynamic simulations for preparing the acceptance test of JT-60SA cryogenic system

Science.gov (United States)

Cirillo, R.; Hoa, C.; Michel, F.; Poncet, J. M.; Rousset, B.

2016-12-01

Power generation in the future could be provided by thermo-nuclear fusion reactors like tokamaks. There inside, the fusion reaction takes place thanks to the generation of plasmas at hundreds of millions of degrees that must be confined magnetically with superconductive coils, cooled down to around 4.5 K. Within this frame, an experimental tokamak device, JT-60SA is currently under construction in Naka (Japan). The plasma works cyclically and the coil system is subject to pulsed heat loads. In order to size the refrigerator close to the average power and hence optimizing investment and operational costs, measures have to be taken to smooth the heat load. Here we present a dynamic model of the JT-60SA's Auxiliary Cold box (ACB) for preparing the acceptance tests of the refrigeration system planned in 2016 in Naka. The aim of this study is to simulate the pulsed load scenarios using different process controls. All the simulations have been performed with EcosimPro® and the associated cryogenic library: CRYOLIB.

Joint resistance measurements of pancake and terminal joints for JT-60SA EF coils

Energy Technology Data Exchange (ETDEWEB)

Obana, Tetsuhiro, E-mail: obana.tetsuhiro@LHD.nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Takahata, Kazuya; Hamaguchi, Shinji; Mito, Toshiyuki; Imagawa, Shinsaku [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Kizu, Kaname; Murakami, Haruyuki; Yoshida, Kiyoshi [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2013-11-15

Highlights: • To evaluate the joint fabrication technology for the JT-60SA EF coils, joint resistance measurements were conducted with a joint sample. • The joint sample was composed of pancake and terminal joints. • The measurements demonstrated that both joints fulfilled the design requirement. • Considering the measurements, the characteristics of both joints were investigated using an analytical model that represents the joints. -- Abstract: To evaluate the joint fabrication technology for the JT-60SA EF coils, joint resistance measurements were conducted using a sample consisting of pancake and terminal joints. Both joints are shake-hands lap joints composed of cable-in-conduit conductors and a pure copper saddle-shaped spacer. The measurements demonstrated that both joints fulfilled the design requirement. Considering these measurements, the characteristics of both joints were investigated using analytical models that represent the joints. The analyses indicated that the characteristics of the conductors used in the joints affect the characteristics of the joints.

Design and performance tests of gas circulation heating of JT-60U vacuum vessel

International Nuclear Information System (INIS)

Yotsuga, M.; Masuzaki, T.; Sago, H.; Nishikane, M.; Uchikawa, T.; Iritani, Y.; Murakami, T.; Horiike, H.; Neyatani, Y.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Miyachi, I.

1992-01-01

This paper reports that in the final stage of construction of the upgraded JT-60 device (JT-60U), baking tests of the vacuum vessel was performed. The vessel torus was heated-up to 300 degrees C by means of the nitrogen gas circulation system and electric heaters mounted on the outboard solid wall of the vessel. The design of the gas flow channels inside the double-wall structure of the vessel was done based on flow model tests, fluid analysis, and flow network analysis. The results of the baking tests were satisfactory. In maintaining 300 degrees C bake-out temperature, required heating power of the gas circulation system and outboard heaters was 520kW and 50kW, respectively. The temperature distribution over the vessel wall was within 300 ± 30 degrees C. It was also shown or suggested that heat-up and cool-down time is about 30 hours. The baking tests data have been reflected on operations for plasma experiments

Fabrication and performance tests of a prototype in-situ coating machine for JT-60

International Nuclear Information System (INIS)

Obara, Kenjiro; Abe, Tetsuya; Murakami, Yoshio

1987-09-01

Prior to the design and construction of the JT-60's in-situ coating device, a prototype machine was fabricated and tested to confirm the applicability of proposed driving methods and mechanical elements to the device which would be operated in very severe conditions including high ambient temperature and high vacuum. The machine basically consists of an in-vessel manipulator, a fiberscope and an ohmically heated titanium evaporator. From the test results, we recommended to use the combination of Inconel 625 and a self-lubricating alloy for the solid-lubricated bearings and MoS 2 -coated Inconel 625 for the solid-lubricated gears. It was also found that TiC coating showed a effect for the prevention of welding between bolts and nuts. In order to optimize the operating parameters of the machine, many wall inspection tests and titanium evaporation tests were carried out in a large vacuum vessel by simulating the JT-60 conditions. (author)

Fabrication and tests of EF conductors for JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Kizu, Kaname, E-mail: kizu.kaname@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Kashiwa, Yoshitoshi; Murakami, Haruyuki [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Obana, Tetsuhiro; Takahata, Kazuya [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Tsuchiya, Katsuhiko; Yoshida, Kiyoshi [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Hamaguchi, Shinji [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Matsui, Kunihiro [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Nakamura, Kazuya; Takao, Tomoaki [Sophia University, Tokyo 102-8554 (Japan); Yanagi, Nagato; Imagawa, Shinsaku; Mito, Toshiyuki [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan)

2011-10-15

The conductors for plasma equilibrium field (EF) coils of JT-60SA are NbTi cable-in-conduit (CIC) conductor with stainless steel 316L jacket. The production of superconductors for actual EF coils started from February 2010. Nine superconductors with 444 m in length were produced up to July 2010. More than 300 welding of jackets were performed. Six nonconformities were found by inspections as go gauge, visual inspection and X-ray test. In order to shorten the manufacturing time schedule, helium leak test was conducted at once after connecting the long length jacket not just after the welding. The maximum force to pull the cable into jacket was about 7.6 kN on average. The mass flow rates of 9 conductors showed almost same values indicating that there are no blockages in the conductors. The measured current sharing temperature agreed with the expectation values from strand performance indicating that no degradation was caused by production process. The coupling time constants of conductors ranged from 80 to 90 ms which are much smaller than the design value of 200 ms.

Extension of TFTR operations to higher toroidal field levels

International Nuclear Information System (INIS)

Woolley, R.D.

1995-01-01

For the past year, TFTR has sometimes operated at extended toroidal field (TF) levels. The extension to 5.6 Tesla (79 kA) was crucial for TFTR's November 1994 10.7 MW DT fusion power record. The extension to 6.0 Tesla (85 kA) was commissioned on 9 September 1995. There are several reasons that one could expect the TF coils to survive the higher stresses that develop at higher fields. They were designed to operate at 5.2 Tesla with a vertical field of 0.5 Tesla, whereas the actual vertical field needed for the plasma does not exceed 0.35 Tesla. Their design specification explicitly required they survive some pulses at 6.0 Tesla. TF coil mechanical analysis computer models available during coil design were crude, leading to conservative design. And design analyses also had to consider worst-case misoperations that TFTR's real time Coil Protection Calculators (CPCs) now positively prevent from occurring

Toroidal plasma reactor with low external magnetic field

International Nuclear Information System (INIS)

Beklemishev, A.D.; Khayrutdinov, R.R.; Petviashvili, V.I.; Tajima, T.; Gordin, V.A.; Tajima, T.

1991-01-01

A toroidal pinch configuration with safety factor q < 0.5 decreasing from the center to periphery without field reversal is proposed. This is capable of containing high pressure plasma with only small toroidal external magnetic field. Sufficient conditions for magnetohydrodynamic stability are fulfilled in this configuration. The stability is studied by constructing the Lyapunov functional and investigating its extrema both analytically and numerically. Comparison of the Lyapunov stability conditions with the conventional linear theory is carried out. Stable configurations are found with average β near 15%, with magnetic field associated mainly with plasma current. The β value calculated with the external magnetic field can be over 100%. Fast charged particles produced by fusion reactions are asymmetrically confined by the poloidal magnetic field (and due to the lack of strong toroidal field). They thus generate a current in the noncentral part of plasma to reinforce the poloidal field. This current drive can sustain the monotonic decrease of q with radius. 20 refs., 9 figs

Avoidance of VDEs during plasma current quench in JT-60U

International Nuclear Information System (INIS)

Yoshino, R.; Nakamura, Y.; Neyatani, Y.

1996-01-01

Vertical displacement events (VDEs) during plasma current quench (I p quench) are one of the serious problems encountered in designing tokamak fusion reactors, owing to the generation of enormously high electromagnetic forces on the vacuum vessel and in-vessel components, but they have been passively and actively avoided in JT-60U. In JT-60U 'slow I p quench' is ended with very fast plasma current termination (final I p termination), and the halo current is frequently measured at this final I p termination. VDEs make the final I p termination severe by increasing the halo current and the electromagnetic force. A strong dependence of VDE growth rate on the initial vertical position of the plasma current centre (Z J ) has been clarified experimentally, and a neutral point of Z J for VDE has been found at ∼ 15 cm above the midplane of the vacuum vessel. According to these measurements, VDE has been avoided by the selection of Z J at the start of I p quench (passive control) and by the control of Z J during I p quench (active control) eventually obtained owing to the small deviation of Z J in real time calculations from its actual value. Furthermore, passive avoidance of VDEs by the injection of a neon ice pellet has been demonstrated. (author). 29 refs, 14 figs

Structural analysis of TFTR toroidal field coil conceptual design

International Nuclear Information System (INIS)

Smith, R.A.

1975-10-01

The conceptual design evaluation of the V-shaped toroidal field coils on the Tokamak Fusion Test Reactor has been performed by detailed structural analysis with the finite element method. The innovation provided by this design and verified in this work is the capability to support toroidal field loads while simultaneously performing the function of twist restraint against the device axial torques resulting from the vertical field loads. The evaluations made for the conceptual design provide predictions for coil deflections and stresses. The results are available for the separate effects from toroidal fields, poloidal fields, and the thermal expansion of the coils as well as for the superposition of the primary loads and the primary plus thermal loads

LASL toroidal reversed-field pinch program

International Nuclear Information System (INIS)

Baker, D.A.; Buchenauer, C.J.; Burkhardt, L.C.

1978-01-01

The determination of the absolute energy loss due to radiation from impurities in the LASL toroidal reversed-field pinch experiment ZT-S is reported. The measurements show over half of the energy loss is accounted for by this mechanism. Thomson scattering electron density measurements indicate only a gradual increase in temperature as the filling pressure is reduced indicating an increased energy loss at lower pressures. Cylindrical and toroidal simulations of the experiment indicate either that a highly radiative pinch boundary or anomalous transport are needed to match the experimental results. New effects on the equilibrium due to plasma flows induced by the toroidal geometry are predicted by the toroidal simulations. The preliminary results on the low temperature discharge cleaning of the ZT-S torus are reported. A description of the upgrade of the ZT-S experiment and the objectives, construction and theoretical predictions for the new ZT-40 experiment are given

Stabilization of neoclassical tearing modes by electron cyclotron current drive in JT-60U

International Nuclear Information System (INIS)

Isayama, A.; Oyama, N.; Urano, H.; Suzuki, T.; Takechi, M.; Hayashi, N.; Nagasaki, K.; Kamada, Y.; Ide, S.; Ozeki, T.

2007-01-01

Results of active control of neoclassical tearing modes (NTMs) by electron cyclotron current drive (ECCD) in JT-60U are described. Growth of an NTM with poloidal mode number m = 3 and toroidal mode number n = 2 has been suppressed by ECCD inside the sawtooth inversion radius in the co-direction, showing the possibility of the coexistence of sawtooth oscillations and a small-amplitude m/n = 3/2 NTM without large confinement degradation. Stabilization of an m/n = 2/1 NTM by ECCD at the mode rational surface has been demonstrated with a small ratio of the current density driven by the electron cyclotron (EC) wave to the local bootstrap current density (∼ 0.5). In addition, dependence of the stabilization effect on ECCD location has been investigated in detail. It has been found that an m/n = 2/1 NTM can be completely stabilized with the misalignment of the ECCD location less than about half of the full island width, and that the m/n = 2/1 NTM is destabilized with the misalignment comparable to the full island width. Time-dependent, self-consistent simulation of magnetic island evolution using the TOPICS code has shown that the stabilization and destabilization of an m/n = 2/1 NTM are well reproduced with the same set of coefficients of the modified Rutherford equation. The TOPICS simulation has also clarified that EC wave power required for complete stabilization can be significantly reduced by narrowing the ECCD deposition width

Representation of magnetic fields with toroidal topology in terms of field-line invariants

International Nuclear Information System (INIS)

Lewis, H.R.

1990-01-01

Beginning with Boozer's representation of magnetic fields with toroidal topology [Phys. Fluids 26, 1288 (1983)], a general formalism is presented for the representation of any magnetic field with toroidal topology in terms of field-line invariants. The formalism is an application to the magnetic field case of results developed recently by Lewis et al. (submitted for publication to J. Phys. A) for arbitrary time-dependent Hamiltonian systems with one degree of freedom. Every magnetic field with toroidal topology can be associated with time-dependent Hamiltonian systems with one degree of freedom and every time-dependent Hamiltonian system with one degree of freedom can be associated with magnetic fields with toroidal topology. In the Hamiltonian context, given any particular function I(q,p,t), Lewis et al. derived those Hamiltonians for which I(q,p,t) is an invariant. In addition, for each of those Hamiltonians, they derived a function canonically conjugate to I(q,p,t) that is also an invariant. They applied this result to the case where I(q,p,t) is expressed as a function of two canonically conjugate functions. This general Hamiltonian formalism provides a basis for representing magnetic fields with toroidal topology in terms of field-line invariants. The magnetic fields usually contain plasma with flow and anisotropic pressure. A class of fields with or without rotational symmetry is identified for which there are magnetic surfaces. The formalism is developed for application to the case of vacuum magnetic fields

Heat analysis of the magnetic limiter plate for JT-60

International Nuclear Information System (INIS)

Nakamura, Hiroo; Ninomiya, Hiromasa; Shimizu, Masatsugu; Ohta, Mitsuru

1977-03-01

Heat analysis has been made of the magnetic limiter plate for JT-60. Test materials of the magnetic limiter plate are molybdenum, graphite, pyrolytic graphite and silicon carbide. It is assumed in calculation of the heat analysis that 10MW is deposited on the 2 cm wide surface of the magnetic limiter plate in about 10 sec. The magnetic limiter plate of pyrolytic graphite is a stack of pyrolytic graphite sheets, heat input is in the deposition plane to take advantage of the large heat conductivity along this plane. Pyrolytic graphite is the best in terms of temperature rise. The temperature of molybdenum and graphite rise up to 1800 0 C and 620 0 C, respectively, in an deposition of 10 MWx10sec. Silicon carbide is not suitable for the magnetic limiter plate. Because the plasma of the JT-60 discharges every 10 min, the average heat flux decreases to 17 w/cm 2 during the each interval. When the magnetic limiter plate has the above heat inflow, a maximum of above 1000 0 C occurs at the edge far from the joint to the thick ring of the vacuum vessel. To reduce heat load of the magnetic limiter plate, an alternating current (2 -- 5Hz) is superposed on the magnetic limiter coil current. The intersection of separatrix line and magnetic limiter plate then moves cyclically more than 10 cm. Concerning temperature distribution of the multi-groove magnetic limiter plate, its dimensions are determined by the limitation in vapor pressure to prevent the impurity inflow. (auth.)

Steady-state exhaust of helium ash in the W-shaped divertor of JT-60U

International Nuclear Information System (INIS)

Sakasai, A.; Takenaga, H.; Hosogane, N.

2001-01-01

By injecting a neutral beam of 60 keV helium (He) atoms as central fueling of helium into the ELMy H-mode plasmas, helium exhaust has been studied in the W-shaped pumped divertor on JT-60U. Efficient He exhaust was realized by He pumping using argon frosted cryopumps in the JT-60U new divertor. In steady state, good He exhaust capability (τ He */τ E =4 and high enrichment factor, where τ He * is a global particle confinement time of helium and τ E is the energy confinement time) was successfully demonstrated in attached ELMy H-mode plasmas. Good He exhaust capability was also obtained in detached ELMy H-mode plasmas, which was comparable to one in attached plasmas. This result of the helium exhaust is sufficient to support a detached divertor operation on ITER. After the divertor modification, helium exhaust in reversed shear plasmas has been investigated using He gas puff. Helium removal inside the internal transport barrier (ITB) is about two times as difficult as that outside the ITB in reversed shear discharges. (author)

Ion heating up to 1 MeV range with higher harmonic ICRF wave on JT-60U

International Nuclear Information System (INIS)

Nemoto, M.; Kusama, Y.; Hamamatsu, K.; Kimura, H.; Fujii, T.; Moriyama, S.; Saigusa, M.; Afanassiev, V.I.

1997-01-01

The properties of protons under accleration by an ion cyclotron range of frequency (ICRF) waves with the second to fourth hydrogen harmonics have been investigated in the JT-60U tokamak at the Japan Atomic Energy Research Institute (JAERI). Protons have been accelerated up to 1 MeV in the presence of an ICRF wave of fixed frequency, neutral beams (NB), and a fixed toroidal magnetic field which is scanned through several plasma discharges. The tail temperature of the protons, which is evaluated in the range 0.32-0.86 MeV, has been observed to increase in the second to third harmonics, however increase of the tail temperature in the third to fourth harmonics has not been observed clearly. Furthermore, the dependence of tail temperature on the harmonic number has been found to be in qualitative agreement with results from a simulation code analysis based upon the one-dimensional Fokker-Planck equation coupled with the kinetic wave equation. Experimental values for the stored energy of the accelerated ions have shown, however, that the response of stored energy to changes in absorbed ICRF power is much stronger than the response to changes in harmonic number. Also, the incremental energy confinement times for heating discharges matching the third and fourth harmonics (3 ω CH) and 4 ω CH) of hydrogen have been observed to be less than half that for those matching the second harmonic. It has been found that suppression of the absorbed ICRF power accompanied with the occurence of cavity resonance in the 3ω CH and 4ω CH heating discharges reduces the stored energy of the accelerated ions and the incremental energy confinement time. (Author)

Feedback control of plasma configuration in JT-60

International Nuclear Information System (INIS)

Ninomiya, Hiromasa; Kikuchi, Mitsuru; Yoshino, Ryuji; Hosogane, Nobuyuki; Kimura, Toyoaki; Kurihara, Kenichi; Takahashi, Minoru; Hayashi, Kazuo.

1986-08-01

Plasma current, plasma position (center of the outermost magnetic surface), decay index n index and width of the divertor throat are feedback controlled by using 5 kinds of poloidal field coils in JT-60. 5 control commands are calculated in a feedback control computer in each 1 msec. These feedback control functions are checked in ohmically heated plasma. The control characteristics of the plasma are well understood by the simplified control analysis and are consistent with the precise matrix transfer function analysis in the frequency domain and the simulation analysis which include the effects of eddy currents, delay time elements and mutual interactions between controllers. The usefulness of these analyses is experimentally confirmed. Each controlled variable is well feedback controlled to the command and the experimentally realized equilibrium configuration is checked by the well calibrated magnetic probes. Fast boundary identification code is used for the identification of the equilibrium and results are consistent with the precalculated plasma equilibria. By using this feedback control system of the plasma configuration and the equilibrium identification method, we have obtained the stable limiter and divertor configuration. The maximum parameters obtained during OH(I) experimental period are plasma current I p = 1.8 MA, the effective safety factor q eff e = 5.7 x 10 19 m -3 (Murakami parameter of 4.5) and the pulse length of 5 ∼ 10 sec. (author)

TFTR toroidal field coil design

International Nuclear Information System (INIS)

Smith, G.E.; Punchard, W.F.B.

1977-01-01

The design of the Tokamak Fusion Test Reactor (TFTR) Toroidal Field (TF) magnetic coils is described. The TF coil is a 44-turn, spiral-wound, two-pancake, water-cooled configuration which, at a coil current of 73.3 kiloamperes, produces a 5.2-Tesla field at a major radius of 2.48 meters. The magnetic coils are installed in titanium cases, which transmit the loads generated in the coils to the adjacent supporting structure. The TFTR utilizes 20 of these coils, positioned radially at 18 0 intervals, to provide the required toroidal field. Because it is very highly loaded and subject to tight volume constraints within the machine, the coil presents unique design problems. The TF coil requirements are summarized, the coil configuration is described, and the problems highlighted which have been encountered thus far in the coil design effort, together with the development tests which have been undertaken to verify the design

The life test of a DC circuit breaker of tokamak device JT-60 for a nuclear fusion research

International Nuclear Information System (INIS)

Shimada, Ryuichi; Tani, Keiji; Kishimoto, Hiroshi; Tamura, Sanae; Yanabu, Satoru.

1979-01-01

In the Tokamak devices for nuclear fusion research, the construction of the current transformer circuits having plasma as the secondary circuit and the change of the primary circuit current are necessary for generating current in the plasma. This is considered to be fairly difficult in practice if conventional methods using capacitor discharge and iron core coils are employed. Considering such circumstances, it was decided for JT-60 to use an air-core current transformer coil and to employ the method of storing energy in the form of current in the coil inductance instead of a capacitor. For this reason, a DC circuit breaker is required to interrupt coil current. The authors improved an AV vacuum breaker, which had been developed as the vacuum breaker of longitudinal magnetic field type applying a magnetic field in parallel with an arc, to get the one for DC circuit for the purpose of applying it to JT-60. In this paper, the operational characteristic of the DC breaker is described, the construction and function of the life test circuit is explained, and the test results are reported. Finally, interruptions of 10,000 times at 20 kA were carried out. It is successful that the restrike of arc occurring during tens of milli-seconds after interruptions was improved to 0.05% or less for 10,000 times operations. Further, it was found that the generation of arc restrike can be reduced practically to zero with two breakers in series. (Wakatsuki, Y.)

Energy confinement in JT-60 lower hybrid current driven plasmas

International Nuclear Information System (INIS)

Ushigusa, K.; Imai, T.; Naito, O.; Ikeda, Y.; Tsuji, S.; Uehara, K.

1990-01-01

The energy confinement in high power lower hybrid current driven (LHCD) plasmas has been studied in the JT-60 tokamak. At a plasma current of 1 MA, the diamagnetically estimated energy confinement time in LHCD plasmas has almost the same value as the confinement time in ohmically heated plasmas at n-bar e ∼ 1.0x10 19 m -3 . The confinement time of high power LHCD plasmas (P LH E varies as to P LH α n e β I p 0 with α + β ∼ -0.3. (author). Letter-to-the-editor. 12 refs, 5 figs

Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

International Nuclear Information System (INIS)

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

1985-01-01

Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line tracings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented

LASL toroidal reversed-field pinch programme

International Nuclear Information System (INIS)

Baker, D.A.; Buchenauer, C.J.; Burkhardt, L.C.

1979-01-01

The determination of the absolute energy loss due to radiation from impurities in the LASL toroidal reversed-field pinch experiment ZT-S is reported. The measurements show that over half the energy loss is accounted for by this mechanism. Thomson-scattering electron density measurements indicate only a gradual increase in temperature as the filling pressure is reduced, indicating an increased energy loss at lower pressures. Cylindrical and toroidal simulations of the experiment indicate either that a highly radiative pinch boundary or anomalous transport is needed to match the experimental results. New effects on the equilibrium due to plasma flows induced by the toroidal geometry are predicted by the toroidal simulations. The preliminary results on the low-temperature discharge cleaning of the ZT-S torus are reported. A description of the upgrade of the ZT-S experiment and the objectives, construction and theoretical predictions for the new ZT-40 experiment are given. (author)

Role of low order rational q values on the ITB-events in JT-60U plasmas

International Nuclear Information System (INIS)

Neudatchin, S.V.; Takizuka, T.; Hayashi, N.; Shirai, H.; Fujita, T.; Isayama, A.; Kamada, Y.; Koide, Y.; Suzuki, T.

2003-01-01

The formation of internal transport barriers (ITBs) near q=2,3 surfaces in normal (NrS) or optimized shear discharges of JT-60U and JET is well known. In reverse shear (RS) JT-60U plasmas, the role of q minimum (q min ) equal to 3, 5, 3, 2, 5, 2 is not obvious for ITB formation. ITB-events (non-local confinement bifurcations inside and around ITB in a ms timescale) are found in various JT-60U NrS and RS plasmas. Under sufficient power, ITB-events are seen at rational and not rational values of q min . The space-time evolution of T e and T i is similar even being strongly varied in space and time, suggesting same mechanism(s) of T e and T i transport. The temporal formation of strong ITB in H-mode under passing of q min =3 (after periodical improvements and degradations via ITB-events with 8ms period) in RS mode with P nbi =8MW is presented. Under smaller power, ITB-events are observed only at rational values of q min . In a weak RS shot with P nbi =4MW, abrupt rise of T e is seen at q min =3.5, while more cases of T i rise are observed. The difference of the T e and T i evolution seen regularly under the low power, suggests decoupling of T e and T i transport. (author)

Role of low order rational q values in the ITB-events in JT-60U plasmas

International Nuclear Information System (INIS)

Neudatchin, S.V.

2002-01-01

The formation of internal transport barriers (ITBs) near q=2,3 surfaces in normal (NrS) or optimized shear discharges of JT-60U and JET is well known. In reverse shear (RS) JT-60U plasmas, the role of q minimum (q min ) equal to 3.5, 3, 2.5, 2 is not obvious for ITB evolution. ITB-events (non-local confinement bifurcations inside and around ITB in a ms timescale) are found in various JT-60U NrS and RS plasmas. (a) Under sufficient power, ITB-events are seen at rational and not rational values of q min . The space-time evolution of Te and Ti is similar. The temporal creation of stronger ITB in H-mode (after periodical improvements and degradations via ITB-events with 8 ms period) under passing of q min = 3 is presented (P nbi = 8 MW, 1.5 MA / 3.7 T). (b) Under smaller power, the influence of some rational q min is seen clearly for ITB-events on T e . In 1.3MA/3.7T shot with very weak RS (Pnbi = 4 MW), abrupt rise of T e is seen at q min 3:5, while more cases of T i rise are observed. (c) The possible role of MHD-activity as ITB-events trigger (ms time scale correlation in some NrS and RS cases) is under investigation. (author)

Properties of internal transport barrier formation in JT-60U

International Nuclear Information System (INIS)

Sakamoto, Yoshiteru; Suzuki, T.; Ide, S.

2003-01-01

The dependence of the ion thermal diffusivity (χ i ) on the radial electric field (E r ) shear has been investigated in JT-60U plasmas. In positive magnetic shear (PS) plasmas, χ i in the core region generally increases with the heating power, similar to the L mode at low heating power. However, as a result of the intensive central heating, which is relevant to the enhancement of the E γ shear, a weak internal transport barrier (ITB) is formed, and χ i in the core region starts to decrease. Corresponding to a further increase of the heating power, a strong ITB is formed and χ i is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of χ i is observed in any of heating regimes. The electron thermal diffusivity (χ e ) is strongly correlated with χ i in PS and RS plasmas. There exists a threshold in the effective E γ shear to change the state from a weak to a strong ITB. It is found that the threshold of the effective E γ shear in the case of a PS plasma depends on the poloidal magnetic field at the ITB. There are multiple levels of reduced transport in the strong ITB for RS plasmas. (author)

Properties of internal transport barrier formation in JT-60U

International Nuclear Information System (INIS)

Sakamoto, Y.; Suzuki, T.; Ide, S.

2003-01-01

The dependence of the ion thermal diffusivity (χ i ) on the radial electric field (E r ) shear has been investigated in JT-60U plasmas. In positive magnetic shear (PS) plasmas, χ i in the core region generally increases with the heating power, similar to the L mode at low heating power. However, as a result of the intensive central heating, which is relevant to the enhancement of the E r shear, a weak internal transport barrier (ITB) is formed, and χ i in the core region starts to decrease. Corresponding to a further increase of the heating power, a strong ITB is formed and χ i is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of χ i is observed in any of the heating regimes. The electron thermal diffusivity (χ e ) is strongly correlated with χ i in PS and RS plasmas. There exists a threshold in the effective E r shear to change the state from a weak to a strong ITB. It is found that the threshold of the effective E r shear in the case of a PS plasma depends on the poloidal magnetic field at the ITB. There are multiple levels of reduced transport in the strong ITB for RS plasmas. (author)

Examples of data processing systems. Data processing system for JT-60

International Nuclear Information System (INIS)

Aoyagi, Tetsuo

1996-01-01

JT-60 data processing system is a large computer complex system including a lot of micro-computers, several mini-computers, and a main-frame computer. As general introduction of the original system configuration has been published previously, some improvements are described here. Transient mass data storage system, network database server, a data acquisition system using engineering workstations, and a graphic terminal emulator for X-Window are presented. These new features are realized by utilizing recent progress in computer and network technology and carefully designed user interface specification of the original system. (author)

Numerical analysis of gas puff modulation experiment on JT-60U

International Nuclear Information System (INIS)

Nagashima, Keisuke; Sakasai, Akira

1992-03-01

In tokamak transport physics, source modulation experiments are one of the most effective methods. For an analysis of these modulation experiments, a simple numerical method was developed to solve the general transport equations. This method was applied to gas puff modulation experiments on JT-60U. From the comparison between the measured and calculated density perturbations, it was found that the particle diffusion coefficient is about 0.8 m 2 /sec in the edge region and 0.1-0.2 m 2 /sec in the central region. (author)

Magnetic Fluctuations during plasma current rise of divertor discharge in JT-60

International Nuclear Information System (INIS)

Ushigusa, Kenkichi; Kikuchi, Mitsuru; Hosogane, Nobuyuki; Tsuji, Syunji; Hayashi, Kazuo.

1986-03-01

During a current rise phase in the JT-60 divertor discharge, a series of magnetic fluctuations which do not rotate poloidally (phase-locking) is observed. They cause a cooling of plasma periphery and an enhancement of H α emission in the divertor chamber. A significant increase in β P + 1 i /2 with minor disruptions during the phase-locked magnetic fluctuation suggests a relaxation of the current profile in the current rise phase of the divertor discharge. (author)

Design and characteristics of the drive mechanism for movable limiters of JT-60, (1)

International Nuclear Information System (INIS)

Takashima, Tetsuo; Morishita, Osamu; Yamamoto, Masahiro; Shimizu, Masatsugu; Ohta, Mitsuru

1976-10-01

Two fast-acting movable rail limiters will be installed in a large Tokamak JT-60 being designed in JAERI. The movable limiter consists of a drive mechanism, a vacuum seal, a bearing, and a molybdenum rail limiter. Design of the drive mechanism for the movable limiter and experimental results on the driving characteristics in full scale are described. (auth.)

Equilibrium poloidal-field distributions in reversed-field-pinch toroidal discharges

International Nuclear Information System (INIS)

Baker, D.A.; Mann, L.W.; Schoenberg, K.F.

1983-01-01

A comparison between the approximate analytic formulae of Shafranov for equilibrium in axisymmetric toroidal systems and fully toroidal numerical solutions of the Grad-Shafranov equation for reversed-field-pinch (RFP) configurations is presented as a function of poloidal beta, internal plasma inductance, and aspect ratio. The Shafranov formula for the equilibrium poloidal-field distribution at the conducting shell that surrounds the plasma is accurate to within 5% for aspect ratios greater than 2, poloidal betas less than 50%, and for plasma current channels that exceed one third of the minor toroidal radius. The analytic description for the centre shift of the innermost flux surface that encloses the plasma current (the Shafranov shift) is accurate to within 15% for aspect ratios greater than 2 and poloidal betas below 50%, provided the shift does not exceed one tenth of the minor conducting boundary radius. The Shafranov formulae provide a convenient method for describing the gross equilibrium behaviour of an axisymmetric RFP discharge, as well as an effective tool for designing the poloidal-field systems of RFP experiments. (author)

Development of the centrifugal pellet injector for JT-60U

Energy Technology Data Exchange (ETDEWEB)

Kizu, Kaname; Hiratsuka, Hajime; Ichige, Hisashi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

2001-03-01

For core fueling of JT-60U plasmas, a repetitive pellet injector which centrifugally accelerates D{sub 2} cubic pellets using a straight rod has been developed. This centrifugal pellet injector can eject trains of up to 40 cubic pellets at frequencies of 1-10 Hz and velocities of 0.3-1.0 km/s. The average pellet mass is 3.6x10{sup 20} atoms/pellet below 0.7 m/s. Key techniques for the development were a mesh structured acceleration component for removing gas sublimated from the pellet and a funnel with an appropriate angle connected just behind the acceleration chamber for introducing the pellet to plasma without destruction. Using the mesh structured components, the horizontal angular distribution of pellets ejected became narrow, because irregular pellet motion caused by sublimated gas was reduced. To investigate the performance of the injector, pellet injection experiments from the low magnetic field side (LFS) were conducted using ohmic heating plasmas. Central fueling and enhanced fueling rate have been observed. D{alpha} intensity around the divertor region was reduced in a pellet injection plasma compared to gas puffing, indicating low recycling rate was maintained with the pellet injection. (author)

Development of the centrifugal pellet injector for JT-60U

International Nuclear Information System (INIS)

Kizu, Kaname; Hiratsuka, Hajime; Ichige, Hisashi

2001-03-01

For core fueling of JT-60U plasmas, a repetitive pellet injector which centrifugally accelerates D 2 cubic pellets using a straight rod has been developed. This centrifugal pellet injector can eject trains of up to 40 cubic pellets at frequencies of 1-10 Hz and velocities of 0.3-1.0 km/s. The average pellet mass is 3.6x10 20 atoms/pellet below 0.7 m/s. Key techniques for the development were a mesh structured acceleration component for removing gas sublimated from the pellet and a funnel with an appropriate angle connected just behind the acceleration chamber for introducing the pellet to plasma without destruction. Using the mesh structured components, the horizontal angular distribution of pellets ejected became narrow, because irregular pellet motion caused by sublimated gas was reduced. To investigate the performance of the injector, pellet injection experiments from the low magnetic field side (LFS) were conducted using ohmic heating plasmas. Central fueling and enhanced fueling rate have been observed. Dα intensity around the divertor region was reduced in a pellet injection plasma compared to gas puffing, indicating low recycling rate was maintained with the pellet injection. (author)

Numerical simulation on current spike behaviour of JT-60U disruptive plasmas

International Nuclear Information System (INIS)

Takei, N; Nakamura, Y; Tsutsui, H; Yoshino, R; Kawano, Y; Ozeki, T; Tobita, K; Tsuji-Iio, S; Shimada, R; Jardin, S C

2004-01-01

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

Development of residual thermal stress-relieving structure of CFC monoblock target for JT-60SA divertor

Energy Technology Data Exchange (ETDEWEB)

Tsuru, Daigo, E-mail: tsuru.daigo@jaea.go.jp; Sakurai, Shinji; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Suzuki, Satoshi

2015-10-15

Highlights: • We carried out numerical simulations on residual thermal stress of targets for the JT-60SA divertor. • We developed three measures to reduce residual thermal stress. • We proposed two structures of CFC monoblock target for the JT-60SA divertor. • We confirmed the effectiveness of the structure by infrared thermography inspection and high heat flux test. - Abstract: Carbon fibre-reinforced carbon composite (CFC) monoblock target for JT-60SA divertor is under development towards the mass-production. CFC monoblocks, WCu interlayers and a CuCrZr cooling tube at the centre of the monoblocks were bonded by vacuum brazing in a high temperature, to a target. If residual thermal stress due to difference of thermal expansions between CFC and CuCrZr exceeds the maximum allowable stress of the CFC after the bonding, cracks are generated in the CFC monoblock and heat removal capacity of the target degrades. In this paper, new structures of the targets were proposed, to reduce residual thermal stress and to mitigate the degradation of heat removal capacity of the targets. Some measures, including slitting of the CFC monoblock aside of the cooling tube, replacement of the interlayer material and shifting the position of the cooling tube, were implemented. The effectiveness of the measures was evaluated by numerical simulations. Target mock-ups with the proposed structures were manufactured. Infrared thermography inspection and high heat flux test were carried out on the mock-ups in order to evaluate the heat removal capacity.

High coupling performance of JT-60U ICRF antennas

International Nuclear Information System (INIS)

Saigusa, M.; Moriyama, S.; Fujii, T.; Kimura, H.; Sato, M.; Hosogane, N.; Nemoto, M.; Yamamoto, T.

1994-01-01

Sufficient coupling of an ICRF antenna for high power experiments was obtained even for a wide gap between the separatrix and the antenna in JT-60U. The loading resistances for an out-of-phase mode are over 4 Ω for a gap of 13 cm between the separatrix and the Faraday shield over the wide range of electron density from 1 x 10 19 to 5.5 x 10 19 m -3 . In particular, the loading resistances for an in-phase mode are about 5 Ω for a gap of 33 cm between the separatrix and the Faraday shield for the same plasma parameters. However, the heating response for the out-of phase mode is much stronger than that for the in-phase mode. (author). Letter-to-the-editor. 11 refs, 6 figs

Air core poloidal magnetic field system for a toroidal plasma producing device

International Nuclear Information System (INIS)

Marcus, F.B.

1978-01-01

A poloidal magnetics system for a plasma producing device of toroidal configuration is provided that reduces both the total volt-seconds requirement and the magnitude of the field change at the toroidal field coils. The system utilizes an air core transformer wound between the toroidal field (TF) coils and the major axis outside the TF coils. Electric current in the primary windings of this transformer is distributed and the magnetic flux returned by air core windings wrapped outside the toroidal field coils. A shield winding that is closely coupled to the plasma carries a current equal and opposite to the plasma current. This winding provides the shielding function and in addition serves in a fashion similar to a driven conducting shell to provide the equilibrium vertical field for the plasma. The shield winding is in series with a power supply and a decoupling coil located outside the TF coil at the primary winding locations. The present invention requires much less energy than the usual air core transformer and is capable of substantially shielding the toroidal field coils from poloidal field flux

Plasma Heating and Losses in Toroidal Multipole Fields

International Nuclear Information System (INIS)

Armentrout, C. J.; Barter, J. D.; Breun, R. A.; Cavallo, A. J.; Drake, J. R.; Etzweiler,; Greenwood, J. R.

1974-01-01

The heating and loss of plasmas have been studied in three pulsed, toroidal multipole devices: a large levitated octupole, a small supported octupole and a very small supported quadrupole. Plasmas are produced by gun injection and heated by electron and ion cyclotron resonance heating and ohmic heating. Electron cyclotron heating rates have been measured over a wide range of parameters, and the results are in quantitative agreement with stochastic heating theory. Electron cyclotron resonance heating produces ions with energies larger than predicted by theory. With the addition of a toroidal field, ohmic heating gives densities as high as 10 13 cm -3 in the toroidal quadrupole and 10 12 cm -3 in the small octupole. Plasma losses for n=5 x 10 9 cm -3 plasmas are inferred from Langmuir probe and Fabry-Perot interferometer measurements, and measured with special striped collectors on the wall and rings. The loss to a levitated ring is measured using a modulated light beam telemeter. The confinement is better than Bohm but considerably worse than classical. Low frequency convective cells which are fixed in space are observed. These cells around the ring are diminished when a weak toroidal field is added, and loss collectors show a vastly reduced flux to the rings. Analysis of the spatial density profile shows features of B-independent diffusion. The confinement is sensitive to some kinds of dc field errors, but surprisingly insensitive to perturbations of the ac confining field

Neoclassical diffusion in toroidal three-cut magnetic field

International Nuclear Information System (INIS)

Nemov, V.V.; Shishkin, A.A.

1975-01-01

Quasi-classical diffusion is investigated in the regime of toroidal drift of 'bananas' in a three cut magnetic field. Unlike previous papers, it is supposed that the inhomogeneity of a helical magnetic field epsilonsub(k) is of the same order or less than that of the toroidal inhomogeneity epsilonsub(t). The case is considered when the efficient frequency of particle collisions exceeds that of the 'banana' precession around the magnetic axis. Expressions for diffusion flows and coefficients are obtained that transform into available ones at epsilonsub(h) > > epsilonsub(t) [ru

Equilibrium poloidal field distributions in reversed-field-pinch toroidal discharges

International Nuclear Information System (INIS)

Baker, D.A.; Mann, L.W.; Schoenberg, K.F.

1982-04-01

A comparison between the analytic formulae of Shafranov for equilibrium in axisymmetric toroidal reversed field pinch (RFP) systems and fully toroidal numerical solutions of the Grad-Shafranov equation is presented as a function of poloidal beta, internal plasma inductance, and aspect ratio. The Shafranov formula for the equilibrium poloidal field distribution is accurate to within 5% for aspect ratios greater than 2, poloidal betas less than 50%, and for plasma current channels that exceed one-third of the minor toroidal radius. The analytic description for the center shift of the innermost flux surface that encloses the plasma current (the Shafranov shift) is accurate to within 15% for aspect ratios greater than 2 and poloidal betas below 50%, provided the shift does not exceed one-tenth of the minor conducting boundary radius. The behavior of the magnetic axis shift as a function of plasma parameters is included. The Shafranov formulae provide a convenient method for describing the equilibrium behavior of an RFP discharge. Examples illustrating the application of the analytic formulae to the Los Alamos ZT-40M RFP experiment are given

High Beta Steady State Research and Future Directions on JT-60U and JFT-2M

Science.gov (United States)

Ishida, Shinichi

2003-10-01

JT-60U and JFT-2M research is focused on high beta steady state operation towards economically and environmentally attractive reactors. In JT-60U, a high-βp H-mode plasma was sustained with βN 2.7 for 7.4 s in which neoclassical tearing modes (NTMs) limited the attainable β_N. Real-time tracking NTM stabilization system using ECCD demonstrated complete suppression of NTM leading to recovery of βN before onset of NTM. Performance in a fully non-inductive H-mode plasma was improved up to n_i(0) τE T_i(0) = 3.1 x 10^20 keV s m-3 using N-NBCD with βN 2.4, HH_y,2=1.2 and bootstrap fraction f_BS 0.5. ECH experiments extended the confinement enhancement for dominantly electron heated reversed shear plasmas up to HH_y,2 2 at T_e/Ti 1.25. A world record ECCD efficiency, 4.2 x 10^18 A/W/m^2, was achieved at Te 23 keV with a highly localized central current density. Innovative initiation and current build-up without center solenoid currents were established by LHCD/ECH and bootstrap current up to f_BS 0.9. In JFT-2M, the inside of the vacuum vessel wall was fully covered with low-activation ferritic steel plates to investigate their use in plasmas near fusion conditions. High βN plasmas were produced up to βN = 3.3 with an internal transport barrier (ITB) and a steady H-mode edge. A new H-mode regime with steady high recycling (HRS) and an ITB was exploited leading to βN H_89P 6.2 at n_e/nG 0.7. In 2003, JT-60U will be able to operate for the duration up to 65 s at 1 MA/2.7 T and the heating/current-drive duration up to 30 s at 17 MW to prolong high-βN and/or high-f_BS discharges with feedback controls. JFT-2M is planning to implement wall stabilization experiments in 2004 to pursue plasmas above the ideal no-wall limit using a ferritic wall. The modification of JT-60 to a fully superconducting tokamak is under discussion to explore high-β steady state operation in collision-less plasmas well above no-wall limit with ferritic wall in a steady state.

Reference design of the power supply system for the resistive-wall-mode control in JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Ferro, Alberto, E-mail: alberto.ferro@igi.cnr.it [Consorzio RFX, C.so Stati Uniti 4, 35127 Padova (Italy); Gaio, Elena [Consorzio RFX, C.so Stati Uniti 4, 35127 Padova (Italy); Novello, Luca [Fusion for Energy, Broader Development of Fusion Department, Boltzmannstr 2, 85748 Garching (Germany); Matsukawa, Makoto; Shimada, Katsuhiro; Kawamata, Yoichi; Takechi, Manabu [Japan Atomic Energy Agency, Naka Fusion Institute, 801-1 Mukoyama, Naka, Ibaraki 311-019 (Japan)

2015-10-15

Highlights: • In JT-60SA, a power supply system (RWM-PS) will feed 18 coils to control the RWMs. • One power amplifier per coil will follow an arbitrary real-time reference. • Very fast dynamics is required (current bandwidth: 3 kHz; latency: 50 μs). • The requirements of the RWM-PS are updated and design solutions discussed. • The reference design of the RWM-PS is based on H-bridges operated at 20 – 30 kHz. - Abstract: The mission of JT-60SA, the satellite Tokamak under construction in Naka (Japan), includes the attainment of steady-state high-beta plasmas. For this purpose, an active control system based on 18 in-vessel sector coils (SC) is foreseen to suppress the resistive wall modes (RWM). Each coil will be independently fed by a dedicated converter, rated for 300 A and 240 V, which has to produce the required current/voltage following in real time the reference provided by the JT-60SA MHD Controller. To minimize the current rating, these converters shall be sufficiently fast to avoid an excessive growth of the RWM. This requires a very high dynamic performance, largely beyond that of standard industrial applications. This paper firstly reports the latest results of the studies on the requirements of the RWM active control system. Then, the reference design of the power supply system is presented, including the ac/dc conversion stage, the fast converters and the control section. The advantages of the proposed scheme are discussed and the main electrical parameters are presented.

Review of JT-60 experimental results from January to October, 1989

International Nuclear Information System (INIS)

1990-03-01

Emphases in recent JT-60 experiments are placed on 1) improvements in plasma confinement with profile control and 2) steady state operation study. Both limiter and lower X-point divertor configurations were employed. The operating gas was hydrogen and, in some cases, helium. Improvements in confinement were demonstrated with pellet injection, LH current drive, high-Ti mode operation or ICRF. Current profile controllability with energetic electrons has been improved by the new LH launcher. The H-mode was achieved in limiter discharges with LH current drive for the first time. Nearly steady-state ELM-free H-mode with durations up to 3.3. sec was established without significant impurity accumulation. High-Ti and high-βp discharges were obtained in high field and low plasma current. Major experimental issues for the steady-state operation research were non-inductive current drive and He-ash exhaust. The neoclassical bootstrap current was confirmed in the wide range of βp. 30 keV helium NB was injected into NB heated lower X-point discharges, which produced centrally peaked birth profile of α-particles. A simple extension of the present result is promising for the helium exhaust in future device. (J.P.N.)

Monte Carlo approach to define the refrigerator capacities for JT-60SA

International Nuclear Information System (INIS)

Wanner, Manfred; Barabaschi, Pietro; Lamaison, Valerie; Michel, Frederic; Reynaud, Pascal; Roussel, Pascal

2011-01-01

The JT-60SA cryogenic system shall provide refrigeration to keep the superconducting magnets and their structures at 4.4 K, cryo-pumps at 3.7 K, thermal shields at 80-100 K, and deliver a flow of 50 K helium to the current leads. A Monte Carlo method is proposed to determine the capacity contingencies for the refrigeration system. Attributing individual contingencies and distribution probability functions to the design variables allows the different load contributions to be statistically averaged. The total refrigeration contingency is derived for each temperature level from the 95% confidence level of the integrated distribution function.

Review of JT-60 experimental results from June to October, 1987

International Nuclear Information System (INIS)

1988-03-01

This is a prompt report on JT-60 experimental results from June to October, 1987. Experiments in hydrogen plasmas have been undertaken with up to 3.2 MA of plasma current in limiter discharges, 2.7 MA in divertor discharges, and 30 MW of total NB (H 0 → H + ) and RF injection power. In limiter discharges with ∼3 MA plasma current, the maximum central electron density of 1.3 x 10 20 m -3 and the energy confinement time of 0.15 - 0.18 sec were obtained with heating powers of 13 - 20 MW. The total plasma stored energy of 3.1 MJ, and n e (0)τ E of 1.4 - 1.8 x 10 19 m -3 sec were achieved. The best set of parameters achieved is n e (0)τ E = 1.8 x 10 19 m -3 sec and T i (0) = 3.7 KeV at plasma current of 3.2 MA. By applying deutrium discharge condition, n e (0)τ E and T i (0) enter the JT-60 target area determined by Atomic Energy Commission. Many short periods (50 - 100 ms) of H-mode phase are found in outside X-point divertor discharges with NB or NB + RF (LH or IC) heating power of more than 16 MW, although at present the energy confinement improvement is limited to within 10 %. In combined LH and NB heating of low-n-bar e discharge, the plasma stored energy increases with the same rate as NB heating only, with preferential absorption of LH wave to the high energy beam ions. In combined NB and on-axis ICRF heating of low-n-bar e discharge, a high incremental energy confinement time is obtained, with strong high energy beam ion acceleration and electron heating in the central region of the plasma. (author)

BPX toroidal field coil design

International Nuclear Information System (INIS)

Heitzenvoeder, D.J.

1992-01-01

This paper reports on the toroidal field (TF) coil system of the Burning Plasma Experiment (BPX) which consists of (18) beryllium copper magnets arrayed in a wedged configuration with a major radius of 2.6 meters and a field strength capability on axis of 9.0 Tesla. The toroidal array is constructed from six (3)-coil modules to facilitate remote recovery in the event of a magnet failure after nuclear activation precludes hands-on servicing. The magnets are of a modified Bitter plate design with partial cases of type 316-LN stainless steel welded with Inconel 182 weld wire. The coil turn plates are fabricated from CDA C17510 beryllium copper with optimized mechanical, thermal, and electrical characteristics. joints within the turns and between turns are made by welding with C17200 filler wire. Cryogenic cooling is employed to reduce power dissipation and to enhance performance. The magnets are cooled between experimental pulses by pressurized liquid nitrogen flowing through channels in the edges of the coil turns. This arrangement makes possible one full-power pulse per hour. Electrical insulation consists of polyimide-glass sheets bonded in place with vacuum-pressure impregnated epoxy/glass

Qualifying tests for TRIAM-1M superconducting toroidal magnetic field coil

Energy Technology Data Exchange (ETDEWEB)

Nakanura, Yukio; Hiraki, Naoji; Nakamura, Kazuo; Tanaka, Masayoshi; Nagao, Akihiro; Kawasaki, Shoji; Itoh, Satoshi

1984-09-01

In the strong toroidal magnetic field experimental facility ''TRIAM-1M'' currently under construction, construction of the superconducting toroidal magnetic field coil and the following qualifying tests conducted on the full-scale superconducting toroidal magnetic field coil actually fabricated are described: (1) coil excitation test, (2) superconducting stability test, (3) external magnetic field application test, and (4) high-speed excitation test. On the basis of these test results, stability was evaluated of the superconducting coil being operated in the tokamak device. In normal tokamak operation, there occurs no normal conduction transition. At the time of plasma disruption, though this transition takes place in part of the coil, the superconducting state is immediately restored. By its electromagnetic force analysis, the superconducting coil is also stable in structure.

Deuterated-decaborane using boronization on JT-60U

Energy Technology Data Exchange (ETDEWEB)

Yagyu, Jun-ichi; Arai, Takashi; Kaminaga, Atsushi; Miyata, Katsuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Arai, Masaru [Kaihatsu Denki Co., Ltd., Tokyo (Japan)

2001-03-01

In JT-60U, boronization using hydride-decaborane (B{sub 10}H{sub 14}) vaporization has been conducted for the first wall conditioning. Compared to other discharge cleaning (DC), boronization is claimed to be efficient in reduction of oxygen impurities and hydrogen recycling in plasma. However, there are some problems in reduction of hydrogen included in boron film and stabilization of DC glow discharge during the boronization. To solve these problems, a new boronization method using deuterated-decaborane (B{sub 10}D{sub 14}) was adopted instead of the conventional hydride-decaborane. As a result, hydrogen content in the boron film decreased clearly and discharge conditioning shots, for decreasing hydrogen content in plasmas, after the boronization were reduced to 1/10 in comparison to the conventional process. Furthermore, DC glow discharge became stable, with only helium carrier gas, and it was possible to save 30 hours in maximum of the time necessary to boronization. It is shown that the boronization using deuterated-decaborane is very efficient and effective method for the first wall conditioning. (author)

Review of JT-60U experimental results in 2003 and 2004

International Nuclear Information System (INIS)

2006-01-01

��2.8 τ R . In the reversed shear plasma, the large bootstrap current fraction (f BS - 75%) has been sustained for 7.4 s (∼2.7 τ R ) with HH 98y2 - 1.7 under nearly full non-inductive current drive condition. Compatibility of AT plasmas with high density and high radiation loss has been also investigated in both reversed shear and high-β p H-mode plasmas. In the reversed shear plasmas, the operation regime was extended to high density above the Greenwald density (n GW ) with high confinement (HH 98y2 > 1) and high radiation loss fraction (f rad > 0.9) by tailoring the internal transport barriers (ITBs). In the high-β p H-mode plasmas, high confinement ( HH 98y2 = 0.96) was maintained at high density (n-bar e /n GW =0.92) with high radiation loss fraction (f rad - 1) by utilizing high-field-side pellets and Ar injections. Electron ITB formation mechanism and transient electron heat transport were compared with Large Helical Device (LHD). Different power dependence of R/L Te (R is the major radius and L Te is the electron temperature gradient scale length) was observed due to the difference in the electron ITB formation mechanism. It was found that the electron thermal diffusivity (χ e ) depends not only on T e but also on ∇T e in JT-60U, while only on T e in LHD. Neo-classical tearing mode (NTM) suppression by localized ECCD was found to be more effective with ECRF injection preceding the mode saturation. The mode was suppressed with less power compared to that required after the mode saturates. The domain of the NTM suppression experiments was extended to the high-β N regime, and the effectiveness of m/n=3/2 mode suppression by ECCD was demonstrated at β N - 2.5 - 3. Genuine tokamak plasma start-up without a central solenoid was demonstrated. In a current hole region, it was shown that no scheme drives current in any direction. Detailed measurement of energetic ions in both space and energy showed dynamic change in the energetic ion profile due to

Poloidal and toroidal plasmons and fields of multilayer nanorings

International Nuclear Information System (INIS)

Garapati, K. V.; Salhi, M.; Kouchekian, S.; Siopsis, G.

2017-01-01

Composite and janus type metallodielectric nanoparticles are increasingly considered as a means to control the spatial and temporal behavior of electromagnetic fields in diverse applications such as coupling to quantum emitters, achieving invisibility cloaks, and obtaining quantum correlations between qubits. We investigate the surface modes of a toroidal nanostructure and obtain the canonical plasmon dispersion relations and resonance modes for arbitrarily layered nanorings. Unlike particle plasmon eigenmodes in other geometries, the amplitudes of the eigenmodes of tori exhibit a distinct forward and backward coupling. We present the plasmon dispersion relations for several relevant toroidal configurations in the quasistatic limit and obtain the dominant retarded dispersion relations of a single ring for comparison, discuss mode complementarity and hybridization, and introduce two new types of toroidal particles in the form of janus nanorings. The resonance frequencies for the first few dominant modes of a ring composed of plasmon supporting materials such as gold, silver, and aluminum are provided and compared to those for a silicon ring. A generalized Green's function is obtained for multilayer tori allowing for calculation of the scattering response to interacting fields. Employing the Green's function, the scalar electric potential distribution corresponding to individual poloidal and toroidal modes in response to an arbitrarily polarized external field and the field of electrons is obtained. The results are applied to obtain the local density of states and decay rate of a dipole near the center of the torus.

Poloidal and toroidal plasmons and fields of multilayer nanorings

Science.gov (United States)

Garapati, K. V.; Salhi, M.; Kouchekian, S.; Siopsis, G.; Passian, A.

2017-04-01

Composite and janus type metallodielectric nanoparticles are increasingly considered as a means to control the spatial and temporal behavior of electromagnetic fields in diverse applications such as coupling to quantum emitters, achieving invisibility cloaks, and obtaining quantum correlations between qubits. We investigate the surface modes of a toroidal nanostructure and obtain the canonical plasmon dispersion relations and resonance modes for arbitrarily layered nanorings. Unlike particle plasmon eigenmodes in other geometries, the amplitudes of the eigenmodes of tori exhibit a distinct forward and backward coupling. We present the plasmon dispersion relations for several relevant toroidal configurations in the quasistatic limit and obtain the dominant retarded dispersion relations of a single ring for comparison, discuss mode complementarity and hybridization, and introduce two new types of toroidal particles in the form of janus nanorings. The resonance frequencies for the first few dominant modes of a ring composed of plasmon supporting materials such as gold, silver, and aluminum are provided and compared to those for a silicon ring. A generalized Green's function is obtained for multilayer tori allowing for calculation of the scattering response to interacting fields. Employing the Green's function, the scalar electric potential distribution corresponding to individual poloidal and toroidal modes in response to an arbitrarily polarized external field and the field of electrons is obtained. The results are applied to obtain the local density of states and decay rate of a dipole near the center of the torus.

Review of JT-60 experimental results from January to March, 1987

International Nuclear Information System (INIS)

1987-08-01

This is a prompt report on JT-60 experimental results from January to March, 1987. Maximum LHRF injection power of 6.3 MW was achieved by using 3 launchers. The correlation between flattening of plasma current profile and energy confinement improvement was demonstrated by controlling N 11 spectrum of current drive and heating launchers. The maximum electron temperature of 6 kW was achieved with 2.4 MW LH injection in electron heating mode. Non-inductive driven current of 2 MA with 2.5 sec duration was demonstrated with 3 MW LHRF injection. Substantial acceleration of NB injected fast ions up to 160 kW was observed during ICRF + NB (60 keV) injection. In low density high power NB injected limiter discharges, high ion temperature of ∼ 11 keV and electron temperature of 5 keV were obtained with Z eff = 5. Various attempts were made to produce H-mode in divertor and limiter discharges. (author)

Particle transport analysis in lower hybrid current drive discharges of JT-60U

International Nuclear Information System (INIS)

Nagashima, K.; Ide, S.; Naito, O.

1996-01-01

Particle transport is modified in lower hybrid current drive discharges of JT-60U. The density profile becomes broad during the lower hybrid wave injection and the profile change depends on the injected wave spectrum. Particle transport coefficients (diffusion coefficient and profile peaking factor) were evaluated using gas-puff modulation experiments. The diffusion coefficient in the current drive discharges is about three times larger than in the ohmic discharges. The profile peaking factor decreases in the current drive discharges and the evaluated values are consistent with the measured density profiles. (author)

The Study of Spherical Cores with a Toroidal Magnetic Field Configuration

Energy Technology Data Exchange (ETDEWEB)

Gholipour, Mahmoud [Research Institute for Astronomy and Astrophysics of Maragha (RIAAM)—Maragha, P.O. Box 55134-441 (Iran, Islamic Republic of)

2017-04-01

Observational studies of the magnetic fields in molecular clouds have significantly improved the theoretical models developed for the structure and evolution of dense clouds and for the star formation process as well. The recent observational analyses on some cores indicate that there is a power-law relationship between magnetic field and density in the molecular clouds. In this study, we consider the stability of spherical cores with a toroidal magnetic field configuration in the molecular clouds. For this purpose, we model a spherical core that is in magnetostatic equilibrium. Herein, we propose an equation of density structure, which is a modified form of the isothermal Lane–Emden equation in the presence of the toroidal magnetic field. The proposed equation describes the effect of the toroidal magnetic field on the cloud structure and the mass cloud. Furthermore, we found an upper limit for this configuration of magnetic field in the molecular clouds. Then, the virial theorem is used to consider the cloud evolution leading to an equation in order to obtain the lower limit of the field strength in the molecular cloud. However, the results show that the field strength of the toroidal configuration has an important effect on the cloud structure, whose upper limit is related to the central density and field gradient. The obtained results address some regions of clouds where the cloud decomposition or star formation can be seen.

Destabilization of a peeling-ballooning mode by a toroidal rotation in tokamaks

International Nuclear Information System (INIS)

Aiba, N.; Hirota, M.; Tokuda, S.; Furukawa, M.

2009-01-01

Full text: From the viewpoint of the heat load on the divertor, Type-I edge localized mode (ELM) needs to be suppressed or the amplitude of this ELM needs to be reduced. In JT-60U, some experimental results showed that the ELM frequency depends on the toroidal rotation, and the rapid rotation in the counter direction of the plasma current changes from Type-I ELM to Grassy ELM, whose frequency is high and the amplitude is small. Recent experimental and theoretical/numerical studies in a static system have identified that both Type-I and Grassy ELMs are considered ideal magnetohydrodynamic (MHD) modes destabilizing near the plasma surface, called peeling-ballooning modes. To investigate the mechanism of the change of ELM frequency by a toroidal rotation, theoretical and numerical analyses are important for understanding the toroidal rotation effects on the peeling-ballooning mode. Previous works about the toroidal rotation effect on the edge MHD stability have illustrated that the toroidal rotation with shear can destabilize low/intermediate-n (<50) modes but can stabilize high-n modes, where n is the toroidal mode number. The stabilization of the high-n mode can be understood qualitatively in analogy with the infinite-n ballooning mode case. However, the destabilizing mechanism of the low/intermediate-n mode is not still clarified, and to understand the stability property related to ELM suppression/mitigation, it is important to clarify this destabilizing mechanism. In this paper, we investigate numerically the destabilizing effect of a toroidal rotation on the peeling-ballooning mode with a newly developed code MINERVA, which solves the Frieman-Rotenberg equation. Particularly, we pay attention to the effect of the centrifuged force on not only equilibrium but also change of equation of motion. (author)

Status of JT-60 data processing system

International Nuclear Information System (INIS)

Matsuda, T.; Tsugita, T.; Oshima, T.; Sakata, S.; Sato, M.; Koiwa, M.; Aoyagi, T.

2000-01-01

The JT-60 data processing system is a large computer complex and gradually modernized by utilizing progressing computer and network technology. There are two major changes in our system. A main computer of FACOM M-780 has been replaced with compatible GS8300 using state-of-art CMOS technology, which results in lower power and space usage with nearly the same performance. Now it can handle ∼500 MB of data per discharge. A gigabit ethernet switch with FDDI ports has been introduced to cope with the increase of handling data. The switch will connect a tera-byte (TB) data server at the bandwidth of a gigabit per second with the main computer and many data acquisition workstations. Other developments in our system are the realization of three workstation-based plans, the TB data server, the VME-based fast data acquisition system and a CICU. The TB data server is basically a UNIX workstation with ∼100 GB RAID disks and ∼900 GB MO auto-exchangers. The VME-based fast data acquisition system has been developed to enlarge the present TMDS. The CICU, which has a function of interfacing the main computer with the CAMAC system, has been replaced with the workstation-based system after the fine tuning

Operation experiences of the JT-60 first walls during high-power additional heating experiments

International Nuclear Information System (INIS)

Takatsu, H.; Ando, T.; Yamamoto, M.; Arai, T.; Kodama, K.; Suzuki, M.; Shimizu, M.

1989-01-01

JT-60 started its operation in May 1985 with TiC-coated molybdenum or Inconel 625 first walls. They provided very clean surfaces as well as superior plasma characteristics during Joule heating discharges. Though 20 μm-thick TiC coatings showed good adhesion characteristics, melting of the TiC coating and also the molybdenum or Inconel 625 substrate was observed at some specific spots, and an influx of heavy metals to the main plasma was inevitable during discharges. Initial results of the additional heating experiments showed degrading effects of locally melted TiC-coated molybdenum or Inconel 625 on plasma operation. Therefore, about a half of the TiC-coated first walls were removed and new graphite first walls were installed during the venting period from April to May 1987. The start-up of the discharge conditioning after installation of a significant number of graphite tiles was very rapid. Flexibility in plasma operation was increased, and JT-60 extended the operation region beyond its original specifications. The graphite first walls of the main chamber performed admirably and maintained their integrity under the conditions of plasma current and additional heating power up to 3.2 MA and 30 MW, respectively. On the other hand, the number of damaged divertor plates was much larger than that expected. The reason of unexpected failure is now under examination. (orig.)

Development of computer-aided software engineering tool for sequential control of JT-60U

International Nuclear Information System (INIS)

Shimono, M.; Akasaka, H.; Kurihara, K.; Kimura, T.

1995-01-01

Discharge sequential control (DSC) is an essential control function for the intermittent and pulse discharge operation of a tokamak device, so that many subsystems may work with each other in correct order and/or synchronously. In the development of the DSC program, block diagrams of logical operation for sequential control are illustrated in its design at first. Then, the logical operators and I/O's which are involved in the block diagrams are compiled and converted to a certain particular form. Since the block diagrams of the sequential control amounts to about 50 sheets in the case of the JT-60 upgrade tokamak (JT-60U) high power discharge and the above steps of the development have been performed manually so far, a great effort has been required for the program development. In order to remove inefficiency in such development processes, a computer-aided software engineering (CASE) tool has been developed on a UNIX workstation. This paper reports how the authors design it for the development of the sequential control programs. The tool is composed of the following three tools: (1) Automatic drawing tool, (2) Editing tool, and (3) Trace tool. This CASE tool, an object-oriented programming tool having graphical formalism, can powerfully accelerate the cycle for the development of the sequential control function commonly associated with pulse discharge in a tokamak fusion device

Overview of JT-60U results toward high integrated performance in reactor-relevant regime

International Nuclear Information System (INIS)

Fujita, T.

2002-01-01

Toward steady sustainment of high integrated performance, we have developed weak magnetic shear (high β p mode) and reversed magnetic shear plasmas. As a large-sized tokamak equipped with a variety of devices for heating, current drive and profile/shape control, JT-60U has high ability to approach the conditions required in reactors: low values of normalized Larmor radius and collisionality, high temperatures with T e > or approx. T i , etc. This paper reviews recent JT-60U results with the emphasis on the projection to the reactor-relevant regime. Full non-inductive current drive has been achieved in a 1.8 MA high β p H-mode plasma with β N 2:4, HH y2 =1.2 and high fusion triple product (3 x 10 20 m -3 keVs) owing to increased N-NB power. In a reversed shear plasma, HH y2 =1.4 at n e /n GW 0.8 under the full non-inductive current drive has been achieved with injection of LHRF and N-NB. In box-type ITBs with reversed shear, barriers for ions and electrons were sustained in a regime with T e > or approx. T i . The pedestal pressure was doubled with increased total poloidal beta in pellet-injected high triangularity plasmas with type I and II ELMs. Stable existence of current hole was demonstrated. (author)

The influence of the analog-to-digital conversion error on the JT-60 plasma position/shape feedback control system

International Nuclear Information System (INIS)

Yoshida, Michiharu; Kurihara, Kenichi

1995-12-01

In the plasma feedback control system (PFCS) and the direct digital controller (DDC) for the poloidal field coil power supply in the JT-60 tokamak, it is necessary to observe signals of all the poloidal field coil currents. Each of the signals, originally measured by a single sensor, is distributed to the PFCS and DDC through different cable routes and different analog-to-digital converters from each other. This produces the conversion error to the amount of several bits. Consequently, proper voltage from feedback calculation cannot be applied to the coil, and hence the control performance is possibly supposed to deteriorate to a certain extent. This paper describes how this error makes an influence on the plasma horizontal position control and how to improve the deteriorated control performance. (author)

Improvement of uniformity of the negative ion beams by tent-shaped magnetic field in the JT-60 negative ion source

International Nuclear Information System (INIS)

Yoshida, Masafumi; Hanada, Masaya; Kojima, Atsushi; Kashiwagi, Mieko; Akino, Noboru; Endo, Yasuei; Komata, Masao; Mogaki, Kazuhiko; Nemoto, Shuji; Ohzeki, Masahiro; Seki, Norikazu; Sasaki, Shunichi; Shimizu, Tatsuo; Terunuma, Yuto; Grisham, Larry R.

2014-01-01

Non-uniformity of the negative ion beams in the JT-60 negative ion source with the world-largest ion extraction area was improved by modifying the magnetic filter in the source from the plasma grid (PG) filter to a tent-shaped filter. The magnetic design via electron trajectory calculation showed that the tent-shaped filter was expected to suppress the localization of the primary electrons emitted from the filaments and created uniform plasma with positive ions and atoms of the parent particles for the negative ions. By modifying the magnetic filter to the tent-shaped filter, the uniformity defined as the deviation from the averaged beam intensity was reduced from 14% of the PG filter to ∼10% without a reduction of the negative ion production

Driving mechanism of SOL plasma flow and effects on the divertor performance in JT-60U

International Nuclear Information System (INIS)

Asakura, N.

2002-01-01

SOL plasma flow plays an important role in the plasma transport along the field lines, and influences control of the divertor plasma and impurity ions. Recently, mechanisms producing the SOL flow such as drifts produced by electric field and pressure gradient are pointed out. In JT-60U, three reciprocating Mach probes were installed at the high-field-side (HFS) baffle, low-field-side (LFS) midplane and just below the X-point. The measurements of the SOL flow and plasma profiles both at the HFS and LFS, for the first time, found out the SOL flow pattern and its driving mechanism. 'Flow reversal' was found near the separatrix of the HFS and LFS. Radial profiles of the SOL flow were similar to those calculated numerically using the UEDGE code with the plasma drifts included. SOL particle fluxes towards the HFS and LFS divertors were, for the first time, evaluated. Important physics issues for the divertor design and operation, such as in-out asymmetries of the heat and particle fluxes, and control of impurity ions with intense gas puff and divertor pump (puff and pump), were investigated. (author)

Development of a pellet cutting and loading device for the JT-60 repetitive pellet injector

International Nuclear Information System (INIS)

Hiratsuka, Hajime; Ichige, Hisashi; Kizu, Kaname; Iwahashi, Takaaki; Honda, Masao

2001-03-01

In JT-60, a pellet injector that repetitively injects deuterium pellets is under development to supply fuel to high temperature plasmas and sustain high-density plasmas. The pellet injector generates cubic pellets and accelerates them with a straight-arm rotor by centrifugal force. In this acceleration method, it is important to supply pellets reliably and stably, to prevent pellet orbits from disordering and to stabilize the launching direction. To achieve higher performance of the injector, a pellet cutting and loading device that cuts a deuterium ice rod into cubic pellets and loads them to the pellet injector successively and stably has been developed. The pellet cutting and loading device can cut a deuterium ice rod produced at low temperature of -8 Pam 3 /s, cutting time of <3 ms, cutting frequency of 1-20 Hz and cutter stroke of 2.5 mm were confirmed in the device test. In the operation test after assembling this device to the centrifugal pellet injector, the operational performance of pellet injection frequency of ∼10 Hz, pellet speed of ∼690 m/s and pellet injection duration time of ∼3.5 s was achieved. Thus, the development of the pellet cutting and loading device contributed to the upgrade of the JT-60 pellet injector. (author)

Improvement on control system of the JT-60 radio frequency heating system

Energy Technology Data Exchange (ETDEWEB)

Shinozaki, Shin-ichi; Moriyama, Shinichi; Hiranai, Shinichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sato, Fumiaki [Nippon Advanced Technology Co., Ltd., Tokai, Ibaraki (Japan)

2003-03-01

On the JT-60 radio frequency (RF) heating system, the decrease in the activity ratio was a problem because of the deterioration of the control system. To improve the reliability, we replaced CAMAC system for a power injection control system, which was a main cause of the trouble, with the microprocessor system. And, a function of computer supported programming function of RF power injection form was introduced, which contributed to reduce a load of operators. Furthermore, personal computers with network communication were introduced to improve a maintenance ability of the control system. As a result, the activity ratio of the RF heating system was improved significantly. (author)

Long-term erosion and re-deposition of carbon in the divertor region of JT-60U

International Nuclear Information System (INIS)

Gotoh, Y.; Tanabe, T.; Ishimoto, Y.; Masaki, K.; Arai, T.; Kubo, H.; Tsuzuki, K.; Miya, N.

2006-01-01

Erosion and redeposition profiles of carbon tiles used in the W-shaped divertor of JT-60U with all carbon plasma facing wall are studied. The inner divertor is mostly covered by carbon redeposited layers, while the outer divertor mostly eroded. In the dome region, the erosion dominates on the inner dome-wing, while redeposited on the outer dome-wing. The redeposited layers on the outer dome-wing show very clear columnar structures indicating local carbon transport form the outer divertor to the outer dome-wing. The weight gain by the redeposition extrapolated to the whole divertor area is 0.55 kg. Since the extrapolated total erosion is about 0.33 kg, the remaining 0.22 kg must be originated from the main chamber erosion. Significant amount of the redeposition is caused locally by multiple processes of erosion, ionization and prompt redeposition toward inboard direction owing to gyration along magnetic filed line. This inboard transport is one of the reasons for small redeposition on the plasma shadowed area of the W shaped divertor of JT-60U with pumping slots placed at the bottoms side

JT-60U high performance regimes

International Nuclear Information System (INIS)

Ishida, S.

1999-01-01

High performance regimes of JT-60U plasmas are presented with an emphasis upon the results from the use of a semi-closed pumped divertor with W-shaped geometry. Plasma performance in transient and quasi steady states has been significantly improved in reversed shear and high- βp regimes. The reversed shear regime elevated an equivalent Q DT eq transiently up to 1.25 (n D (0)τ E T i (0)=8.6x10 20 m-3·s·keV) in a reactor-relevant thermonuclear dominant regime. Long sustainment of enhanced confinement with internal transport barriers (ITBs) with a fully non-inductive current drive in a reversed shear discharge was successfully demonstrated with LH wave injection. Performance sustainment has been extended in the high- bp regime with a high triangularity achieving a long sustainment of plasma conditions equivalent to Q DT eq ∼0.16 (n D (0)τ E T i (0)∼1.4x10 20 m -3 ·s·keV) for ∼4.5 s with a large non-inductive current drive fraction of 60-70% of the plasma current. Thermal and particle transport analyses show significant reduction of thermal and particle diffusivities around ITB resulting in a strong Er shear in the ITB region. The W-shaped divertor is effective for He ash exhaust demonstrating steady exhaust capability of τ He */τ E ∼3-10 in support of ITER. Suppression of neutral back flow and chemical sputtering effect have been observed while MARFE onset density is rather decreased. Negative-ion based neutral beam injection (N-NBI) experiments have created a clear H-mode transition. Enhanced ionization cross- section due to multi-step ionization processes was confirmed as theoretically predicted. A current density profile driven by N-NBI is measured in a good agreement with theoretical prediction. N-NBI induced TAE modes characterized as persistent and bursting oscillations have been observed from a low hot beta of h >∼0.1-0.2% without a significant loss of fast ions. (author)

ATLAS Barrel Toroid magnet reached nominal field

CERN Multimedia

2006-01-01

 On 9 November the barrel toroid magnet reached its nominal field of 4 teslas, with an electrical current of 21 000 amperes (21 kA) passing through the eight superconducting coils as shown on this graph

Passing particle toroidal precession induced by electric field in a tokamak

International Nuclear Information System (INIS)

Andreev, V. V.; Ilgisonis, V. I.; Sorokina, E. A.

2013-01-01

Characteristics of a rotation of passing particles in a tokamak with radial electric field are calculated. The expression for time-averaged toroidal velocity of the passing particle induced by the electric field is derived. The electric-field-induced additive to the toroidal velocity of the passing particle appears to be much smaller than the velocity of the electric drift calculated for the poloidal magnetic field typical for the trapped particle. This quantity can even have the different sign depending on the azimuthal position of the particle starting point. The unified approach for the calculation of the bounce period and of the time-averaged toroidal velocity of both trapped and passing particles in the whole volume of plasma column is presented. The results are obtained analytically and are confirmed by 3D numerical calculations of the trajectories of charged particles

Passing particle toroidal precession induced by electric field in a tokamak

Energy Technology Data Exchange (ETDEWEB)

Andreev, V. V. [Peoples' Friendship University of Russia, Ordzhonikidze St. 3, Moscow 117198 (Russian Federation); Ilgisonis, V. I.; Sorokina, E. A. [Peoples' Friendship University of Russia, Ordzhonikidze St. 3, Moscow 117198 (Russian Federation); NRC “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation)

2013-12-15

Characteristics of a rotation of passing particles in a tokamak with radial electric field are calculated. The expression for time-averaged toroidal velocity of the passing particle induced by the electric field is derived. The electric-field-induced additive to the toroidal velocity of the passing particle appears to be much smaller than the velocity of the electric drift calculated for the poloidal magnetic field typical for the trapped particle. This quantity can even have the different sign depending on the azimuthal position of the particle starting point. The unified approach for the calculation of the bounce period and of the time-averaged toroidal velocity of both trapped and passing particles in the whole volume of plasma column is presented. The results are obtained analytically and are confirmed by 3D numerical calculations of the trajectories of charged particles.

Design study of a negative-ion based NBI system for JT-60U

International Nuclear Information System (INIS)

Akino, Noboru; Araki, Masanori; Ebisawa, Noboru

1994-03-01

A high energy negative-ion based NBI system for JT-60U has been designed. The objective of the NBI system is to demonstrate mega-ampere level NB current drive and plasma core heating in a reactor-grade high density plasma. This is the first negative-ion based NBI system in the world. The required specifications of the NBI system are; a beam energy of 500 keV, an injection power of 10 MW, a beam pulse duration of 10 sec with a duty cycle of 1/60 and a beam species of deuterium or hydrogen. The neutral beam power of 10 MW is injected tangentially using one beam-line with two large negative-ion sources. The construction of the NBI system has been started, and will be operational in 1996. (author)

Design study of a time-of-flight neutron spectrometer for JT-60U

International Nuclear Information System (INIS)

Elevant, T.; Hoek, M.; Nishitani, Takeo.

1993-06-01

A time-of-flight neutron spectrometer is proposed for measurements of neutron energy spectra from deuterium-deuterium reactions in JT-60U tokamak plasmas. The sensitivity of the instrument is 2 · 10 -2 cm 2 , energy resolution is 4.5 % (FWHM) and maximum useful count-rate is 6 kHz. Analysis of neutron energy spectra will provide information on central ion temperatures larger than ∼ 4 keV with an accuracy of ± 10 %, and neutron source fraction from reactions between thermal ions with an accuracy of ± 15 %. The minimum time required for data acquisition is 0.1 s. (author)

Experimental transport analysis code system in JT-60

International Nuclear Information System (INIS)

Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

1988-03-01

Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

Hamiltonian description of toroidal magnetic fields in vacuum

International Nuclear Information System (INIS)

Lewis, H.R.; Bates, J.W.

1996-01-01

An investigation of vacuum magnetic fields in toroidal geometry has been initiated. Previously, the general form of the magnetic scalar potential for fields regular at the poloidal axis was given. Here, these results have been expanded to obtain the magnetic scalar potential in a vacuum region that may surround a toroidal current distribution. Using this generalized magnetic scalar potential in conjunction with Boozer's canonical representation of a magnetic field, a field-line Hamiltonian for nonaxisymmetric vacuum fields has been derived. These fields axe being examined using a novel, open-quotes time-dependentclose quotes perturbation theory, which permits the iterative construction of invariants associated with magnetic field-line Hamiltonians that consist of an axisymmetric zeroth-order term, plus a nonaxisymmetric perturbation. By choosing appropriate independent variables, an explicit constructive procedure is developed which involves only a single canonical transformation. Such invariants are of interest because they provide a means of investigating the topology of magnetic field lines. Our objective is to elucidate the existence of magnetic surfaces for nonaxisymmetric vacuum configurations, as well as to provide an approach for studying the onset of stochastic behavior

Toroidal magnetic field system for a 2-MA reversed-field pinch experiment

International Nuclear Information System (INIS)

Melton, J.G.; Linton, T.W.

1983-01-01

The engineering design of the toroidal magnetic field (TF) system for a 2-MA Reversed-Field Pinch experiment (ZT-H) is described. ZT-H is designed with major radius 2.15 meters, minor radius 0.40 meters, and a peak toroidal magnetic field of 0.85 Tesla. The requirement for highly uniform fields, with spatial ripple <0.2% leads to a design with 72 equally spaced circular TF coils, located at minor radius 0.6 meters, carrying a maximum current of 9.0 MA. The coils are driven by a 12-MJ capacitor bank which is allowed to ring in order to aid the reversal of magnetic field. A stress analysis is presented, based upon calculated hoop tension, centering force, and overturning moment, treating these as a combination of static loads and considering that the periodic nature of the loading causes little amplification. The load transfer of forces and moments is considered as a stress distribution resisted by the coils, support structures, wedges, and the structural shell

On the stabilization of toroidal pinches by finite larmor radius effects and toroidal magnetic field

International Nuclear Information System (INIS)

Singh, R.; Weiland, J.

1989-01-01

The radial eigenvalue problem for internal modes in a large aspect ratio toriodal pinch has been solved. A particularly stable regime for a weak but nonzero toroidal magnetic field has been found. (31 refs.)

Detailed Analysis of the Transient Voltage in a JT-60SA PF Coil Circuit

International Nuclear Information System (INIS)

Yamauchi, K.; Shimada, K.; Terakado, T.; Matsukawa, M.; Coletti, R.; Lampasi, A.; Gaio, E.; Coletti, A.; Novello, L.

2013-01-01

A superconducting coil system is actually complicated by the distributed parameters, e.g. the distributed mutual inductance among turns and the distributed capacitance between adjacent conductors. In this paper, such a complicated system was modeled with a reasonably simplified circuit network with lumped parameters. Then, a detailed circuit analysis was conducted to evaluate the possible voltage transient in the coil circuit. As a result, an appropriate (minimum) snubber capacitance for the Switching Network Unit, which is a fast high voltage generation circuit in JT-60SA, was obtained. (fusion engineering)

Design optimization of JT-60SU for steady-state advanced operation

International Nuclear Information System (INIS)

Ushigusa, K.; Kurita, G.; Toyoshima, N.

2001-01-01

Design optimization of JT-60SU has been done for a steady-state advanced operation. A transport code simulation indicates that a fully non-inductive reversed shear plasmas with fractions of 70% of the bootstrap current and 30% of beam driven current can be sustained for more than 1,000s without any additional control. Investigations have been progressed on MHD stability, vertical positional stability and dynamics of the vertical displacement events. Significant progress has been achieved in the R and D of Nb 3 Al superconducting wires, low induced activation material (Fe-Cr-Mn steel). A design improvement has been made in TF coils to reduce a local stress on radial disk. Dynamic behaviors of the tokamak machine have been analyzed at emergency events such as an earthquake. (author)

Computer system for the beam line data processing at JT-60 prototype neutral beam injector

International Nuclear Information System (INIS)

Horiike, Hiroshi; Kawai, Mikito; Ohara, Yoshihiro

1987-08-01

The present report describes the hard and soft wares of the data acquisition computer system for the prototype neutral injector unit for JT-60. In order to operate the unit, more than hundreds of signals of the beam line components have to be measured. These are mainly differential thermometers for the coolant waters and thermocouples for the beam dump components but not include those for the cryo system. Since the unit operates in a series of pulses, the measurement should be conducted very quickly in order to ensure the simultaneity of large number of the measured data. The present system actualize fast data acquisition using a small computer of 128 kB and measuring instruments connected through the bus. The system is connected to the JAERI computer center since the data capacity is fairly large to completely process them by the small computer. Therefore the measured data can be transferred to the computer center to calculate there, and the results can be received. After the system was completed the computer quickly print out the power flow data, which needed much work to calculate with hands. This system was very useful. It enhanced the experiments at the unit and reduced the labor. It enables us to early demonstrate the rated operation of the unit and to accurately estimate such operation data of the JT-60 NBI as the injection power. (author)

Development of fast charge exchange recombination spectroscopy by using interference filter method in JT-60U

International Nuclear Information System (INIS)

Kobayashi, Shinji; Sakasai, Akira; Koide, Yoshihiko; Sakamoto, Yoshiteru; Kamada, Yutaka; Hatae, Takaki; Oyama, Naoyuki; Miura, Yukitoshi

2003-01-01

Recent developments and results of fast charge exchange recombination spectroscopy (CXRS) using interference filter method are reported. In order to measure the rapid change of the ion temperature and rotation velocity under collapse or transition phenomena with high-time resolution, two types of interference filter systems were applied to the CXRS diagnostics on the JT-60U Tokamak. One can determine the Doppler broadening and Doppler shift of the CXR emission using three interference filters having slightly different center wavelengths. A rapid estimation method of the temperature ad rotation velocity without non-linear least square fitting is presented. The modification of the three-filters system enables us to improve the minimum time resolution up to 0.8 ms, which is better than that of 16.7 ms for the conventional CXRS system using the CCD detector in JT-60U. The other system having seven wavelength channels is newly fabricated to crosscheck the results obtained by the three-filters assembly, that is, to verify that the CXR emission forms a Gaussian profile under collapse phenomena. In a H-mode discharge having giant edge localized modes, the results obtained by the two systems are compared. The applicability of the three-filters system to the measurement of rapid changes in temperature and rotation velocity is demonstrated. (author)

Radial electric field in JET advanced tokamak scenarios with toroidal field ripple

NARCIS (Netherlands)

Crombe, K.; Andrew, Y.; Biewer, T. M.; Blanco, E.; de Vries, P. C.; Giroud, C.; Hawkes, N. C.; Meigs, A.; Tala, T.; von Hellermann, M.; Zastrow, K. D.

2009-01-01

A dedicated campaign has been run on JET to study the effect of toroidal field (TF) ripple on plasma performance. Radial electric field measurements from experiments on a series of plasmas with internal transport barriers (ITBs) and different levels of ripple amplitude are presented. They have been

Recent results on steady state and confinement improvement research on JT-60U

International Nuclear Information System (INIS)

Ide, Shunsuke

2000-01-01

On the JT-60U tokamak, fusion plasma research for realization of a steady state tokamak reactor has been pursued. Towards that goal, confinement improved plasmas such as H-mode, high β p , reversed magnetic shear (RS) and latter two combined with H-mode edge pedestal have been developed and investigated intensively. A key issue to achieve non-inductive current drive relevant to a steady state fusion reactor is to increase the fraction of the bootstrap current and match the spatial profile to the optimum. In 1999, as the result of the optimization, the equivalent deuterium-tritium (D-T) fusion gain (Q DT eq ) of 0.5 was sustained for 0.8 s, which is roughly equal to the energy confinement time, in a RS plasma. In order to achieve a RS plasma in steady state two approach have been explored. One is to use external current driver such as lower hybrid current drive (LHCD), and by optimizing LHCD a quasi-steady RS discharge was obtained. The other approach is to utilize bootstrap current as much as possible, and with highly increased fraction of the bootstrap current, a confinement enhancement factor of 3.6 was maintained for 2.7 s in a RS plasma with H-mode edge. A heating and current drive system in the electron cyclotron range of frequency for localized heating and current drive has been installed on JT-60U, and in initial experiments a clear increase of the central electron temperature in a RS high density central region was confirmed only with injected power of 0.75 MW. (author)

Development of FEMAG. Calculation code of magnetic field generated by ferritic plates in the tokamak devices

Energy Technology Data Exchange (ETDEWEB)

Urata, Kazuhiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

2003-03-01

In design of the future fusion devises in which low activation ferritic steel is planned to use as the plasma facing material and/or the inserts for ripple reduction, the appreciation of the error field effect against the plasma as well as the optimization of ferritic plate arrangement to reduce the toroidal field ripple require calculation of magnetic field generated by ferritic steel. However iterative calculations concerning the non-linearity in B-H curve of ferritic steel disturbs high-speed calculation required as the design tool. In the strong toroidal magnetic field that is characteristic in the tokamak fusion devices, fully magnetic saturation of ferritic steel occurs. Hence a distribution of magnetic charges as magnetic field source is determined straightforward and any iteration calculation are unnecessary. Additionally objective ferritic steel geometry is limited to the thin plate and ferritic plates are installed along the toroidal magnetic field. Taking these special conditions into account, high-speed calculation code ''FEMAG'' has been developed. In this report, the formalization of 'FEMAG' code, how to use 'FEMAG', and the validity check of 'FEMAG' in comparison with a 3D FEM code, with the measurements of the magnetic field in JFT-2M are described. The presented examples are numerical results of design studies for JT-60 modification. (author)

Observation of plasma toroidal-momentum dissipation by neoclassical toroidal viscosity.

Science.gov (United States)

Zhu, W; Sabbagh, S A; Bell, R E; Bialek, J M; Bell, M G; LeBlanc, B P; Kaye, S M; Levinton, F M; Menard, J E; Shaing, K C; Sontag, A C; Yuh, H

2006-06-09

Dissipation of plasma toroidal angular momentum is observed in the National Spherical Torus Experiment due to applied nonaxisymmetric magnetic fields and their plasma-induced increase by resonant field amplification and resistive wall mode destabilization. The measured decrease of the plasma toroidal angular momentum profile is compared to calculations of nonresonant drag torque based on the theory of neoclassical toroidal viscosity. Quantitative agreement between experiment and theory is found when the effect of toroidally trapped particles is included.

Investigation and measures to noise on spectroscopic measurement system in JT-60U

International Nuclear Information System (INIS)

Nagaya, Susumu; Kubo, Hirotaka; Sugie, Tatsuo; Onizawa, Masami; Kawai, Isao; Nakata, Hisao.

1997-11-01

Breakdown of a negative-ion-based neutral beam injection (N-NBI) has caused noise trouble to several systems. The control circuit of a spectroscopic measurement system had not well worked because of the noise. The noise has been measured by an optical-fiber isolation system during operation of JT-60U. The amplitude and the frequency were 15-18 V and 15 MHz respectively. The transmission noise has been reduced by putting ferrite cores to all cables connecting with the control circuits. As a result, the trouble with the spectroscopic measurement system has completely been solved. Adding condensers and resistors to the circuit was not effective to reduce the noise. (author)

Mechanical properties of JT-60 tokamak machine in power tests

International Nuclear Information System (INIS)

Takatsu, Hideyuki; Ohkubo, Minoru; Yamamoto, Masahiro; Ohta, Mitsuru

1986-01-01

JT-60 power tests were carried out from Dec. 10, 1984 to Feb. 20, 1985 to demonstrate, in advance of actual plasma operation, satisfactory performance of tokamak machine, power suppliers and control system in combination. The tests began with low power test of individual coil systems and progressed to full power tests. The coil current was raised step by step, monitoring the mechanical, thermal, electrical and vacuum data. Power tests were concluded with successful results. All of the coil systems were raised up to full power operation in combination and system performance was verified including the structural integrity of tokamak machine. Measured strain and deflection showed good agreements with those predicted in the design, which was an evidence that electromagnetic forces were supported as expected in the design. A few limitations to machine operation was made clear quantitatively. And it was found that existing detectors were insufficient to monitor machine integrity and two kinds of detector were proposed to be installed. (author)

Radiation losses and global power balance of JT-60 plasmas

International Nuclear Information System (INIS)

Nishitani, T.; Itami, K.; Nagashima, K.; Tsuji, S.; Hosogane, N.; Yoshida, H.; Ando, T.; Kubo, H.; Takeuchi, H.

1990-01-01

The radiation losses and the global power balance for Ohmic and neutral beam heated plasmas have been investigated in different JT-60 configurations. Discharges with a TiC coated molybdenum wall and with a graphite wall, with limiter, outer and lower X-point configurations have been studied by bolometric measurements, thermocouples and an infrared TV camera. In neutral beam heated outer X-point discharges with a TiC coated molybdenum first wall, the radiation loss of the main plasma was very low (10% of the absorbed power). The radiation loss due to oxygen was dominant in this case. On the contrary, in discharges with TiC coated molybdenum limiters the radiation loss was very high (>60% of the absorbed power). In the discharges with a graphite wall the radiated power from the main plasma was 20-25% for both limiter and lower X-point configurations. In lower X-point discharges the main contributor to the radiation loss was oxygen, whereas in limiter discharges the loss due to carbon was equal to the loss due to oxygen. The radiation loss from the lower X-point divertor increased with increasing electron density of the main plasma. (author). 33 refs, 14 figs, 1 tab

Inward transport of a toroidally confined plasma subject to strong radial electric fields

Science.gov (United States)

Roth, J. R.; Krawczonek, W. M.; Powers, E. J.; Hong, J.; Kim, Y.

1977-01-01

The paper aims at showing that the density and confinement time of a toroidal plasma can be enhanced by radial electric fields far stronger than the ambipolar values, and that, if such electric fields point into the plasma, radially inward transport can result. The investigation deals with low-frequency fluctuation-induced transport using digitally implemented spectral analysis techniques and with the role of strong applied radial electric fields and weak vertical magnetic fields on plasma density and particle confinement times in a Bumpy Torus geometry. Results indicate that application of sufficiently strong radially inward electric fields results in radially inward fluctuation-induced transport into the toroidal electrostatic potential well; this inward transport gives rise to higher average electron densities and longer particle confinement times in the toroidal plasma.

ZONAL TOROIDAL HARMONIC EXPANSIONS OF EXTERNAL GRAVITATIONAL FIELDS FOR RING-LIKE OBJECTS

Energy Technology Data Exchange (ETDEWEB)

Fukushima, Toshio, E-mail: Toshio.Fukushima@nao.ac.jp [National Astronomical Observatory, Ohsawa, Mitaka, Tokyo 181-8588 (Japan)

2016-08-01

We present an expression of the external gravitational field of a general ring-like object with axial and plane symmetries such as oval toroids or annular disks with an arbitrary density distribution. The main term is the gravitational field of a uniform, infinitely thin ring representing the limit of zero radial width and zero vertical height of the object. The additional term is derived from a zonal toroidal harmonic expansion of a general solution of Laplace’s equation outside the Brillouin toroid of the object. The special functions required are the point value and the first-order derivative of the zonal toroidal harmonics of the first kind, namely, the Legendre function of the first kind of half integer degree and an argument that is not less than unity. We developed a recursive method to compute them from two pairs of seed values explicitly expressed by some complete elliptic integrals. Numerical experiments show that appropriately truncated expansions converge rapidly outside the Brillouin toroid. The truncated expansion can be evaluated so efficiently that, for an oval toroid with an exponentially damping density profile, it is 3000–10,000 times faster than the two-dimensional numerical quadrature. A group of the Fortran 90 programs required in the new method and their sample outputs are available electronically.

Neoclassical Drift of Circulating Orbits Due toToroidal Electric Field in Tokamaks

International Nuclear Information System (INIS)

Qin, Hong; Guan, Xiaoyin; Fisch, Nathaniel J.

2011-01-01

In tokamaks, Ware pinch is a well known neoclassical effect for trapped particles in response to a toroidal electric field. It is generally believed that there exists no similar neoclassical effect for circulating particles without collisions. However, this belief is erroneous, and misses an important effect. We show both analytically and numerically that under the influence of a toroidal electric field parallel to the current, the circulating orbits drift outward toward the outer wall with a characteristic velocity O ((var e psilon) -1 ) larger than the E x B velocity, where (var e psilon) is the inverse aspect-ratio of a tokamak. During a RF overdrive, the toroidal electric field is anti-parallel to the current. As a consequence, all charged particles, including backward runaway electrons, will drift inward towards the inner wall.

Work and safety managements for on-site installation, commissioning, tests by EU of quench protection circuits for JT-60SA

International Nuclear Information System (INIS)

Yamauchi, Kunihito; Okano, Jun; Shimada, Katsuhiro; Ohmori, Yoshikazu; Terakado, Tsunehisa; Matsukawa, Makoto; Koide, Yoshihiko; Kobayashi, Kazuhiro; Ikeda, Yoshitaka; Fukumoto, Masahiro; Kushita, Kouhei N.

2016-03-01

The superconducting Satellite Tokamak machine “JT-60SA” under construction in Naka Fusion Institute is an international collaborative project between Japan Atomic Energy Agency (JAEA) as the Implementing Agency (IA) of Japan (JA) and Fusion for Energy (F4E) as the IA of Europe (EU). The contributions for this project are based on the supply of components, and thus European manufacturer shall conduct the installation, commissioning and tests on Naka site under the general supervision by F4E via the designated institute in each EU nation. This means that JAEA had an issue to manage the works by European workers and their safety although there is no direct contract. This report describes the approaches for the work and safety managements, which were agreed with EU after the negotiation, and the completed on-site works for Quench Protection Circuits (QPC) as the first experience for EU in JT-60SA project. (author)

Review of JT-60U experimental results in 2000

International Nuclear Information System (INIS)

2002-11-01

The results of JT-60U experiments in 2000, from February to December, are reviewed. The performance under the full non-inductive current drive has been greatly advanced both in high β p H-mode plasmas and in reversed shear (RS) plasmas. In the high β p H-mode, with injection of the negative ion based neutral beam (NNB) of 360 keV and 4 MW into a high electron temperature plasma (T e (0) ∼ 13 keV), a high fusion triple product n D (0)T i (0)τ E = 2.0 x 10 20 keVm -3 s has been obtained at the plasma current of 1.5 MA, and the highest value of current drive efficiency of NNB (1.55 x 10 19 A/W/m 2 ) has been achieved. In RS, LHCD and NNB-CD were employed for current drive and high confinement (HH y2 ∼ 1.4) at high density (f GW ∼ 0.80) has been achieved. In the 110 GHz ECRF system, two more gyrotrons have been installed in addition to the one installed in 1999 and 1.5 MW was injected into the torus for 3 s. Complete stabilization of neoclassical tearing mode (NTM), realization of high confinement plasmas with T e ∼ T i , a high value (1 MA) of NNB-driven current in the high T e regime, and measurement of localized EC driven current were achieved with the upgraded EC system. The multiple pellet injection system has been newly installed. In high power NB heated plasmas, high-field-side pellet injection was more effective than low-field-side injection, and it extended the regime of high confinement high β p H-mode to a higher density. A new method, called CCS (Cauchy-condition surface method), for the control of the plasma position and shape in real-time became available and was found very useful especially for the control of plasma-wall clearance in LH and IC experiments. The active control of ITB strength by the switch of the injection direction of toroidal angular momentum was successfully demonstrated. In high triangularity H-mode plasmas, higher pressure and temperature at the edge pedestal were observed, which resulted in higher temperature and

Formation of magnetic islands due to field perturbations in toroidal stellarator configurations

International Nuclear Information System (INIS)

Lee, D.K.; Harris, J.H.; Lee, G.S.

1990-06-01

An explicit formulation is developed to determine the width of a magnetic island separatrix generated by magnetic field perturbations in a general toroidal stellarator geometry. A conventional method is employed to recast the analysis in a magnetic flux coordinate system without using any simplifying approximations. The island width is seen to be proportional to the square root of the Fourier harmonic of B ρ /B ζ that is in resonance with the rational value of the rotational transform, where B ρ and B ζ are contravariant normal and toroidal components of the perturbed magnetic field, respectively. The procedure, which is based on a representation of three-dimensional flux surfaces by double Fourier series, allows rapid and fairly accurate calculation of the island widths in real vacuum field configurations, without the need to follow field lines through numerical integration of the field line equations. Numerical results of the island width obtained in the flux coordinate representation for the Advanced Toroidal Facility agree closely with those determined from Poincare puncture points obtained by following field lines. 22 refs., 5 tabs

Design of a negative-ion based NBI system for JT-60U

International Nuclear Information System (INIS)

Kuriyama, M.; Araki, M.; Inoue, T.; Kunieda, S.; Matsuoka, M.; Mizuno, M.; Ohara, Y.; Okumura, Y.; Oohara, H.; Watanabe, K.

1992-01-01

This paper reports on a negative-ion based NBI system which is planned as a key device on the JT-60U in the experiments of current drive and plasma core heating with high density plasmas. The NBI system will inject neutral beams of 500keV, 10MW for 10sec from a beamline with two ion sources. The neutral beam will be injected tangentially in the codirection. Each ion source is a modified volume production-type negative-ion source with cesium vapor. The acceleration current is 22A with deuterium beam, and the current density is 13mA/cm 2 . An operational pressure in the negative-ion generator is less than 0.5 Pa. A three-stage electro static acceleration system is adopted as the accelerator. The beamline length between the ion source and the injection port is 24m. The beamline consists of an ion source tank, neutralizer cells of 10m in length, an ion dump tank and a drift duct. The ion source tank contains large cryopumps to maintain the exit of the ion source sufficiently low. The ion dump tank contains ion deflecting coils, ion dumps for positive and negative ions, a calorimeter, cryopumps and beam scrapers. Residual ions are deflected by the combined magnetic fields produced by the deflecting coils and the stray field form the tokamak. The two sources are connected to an acceleration power supply of 500kV/64A/10sec, while the negative-ion generator power, the extraction voltage, and electron-suppression voltage are fed individually

Liquid toroidal drop under uniform electric field

Science.gov (United States)

Zabarankin, Michael

2017-06-01

The problem of a stationary liquid toroidal drop freely suspended in another fluid and subjected to an electric field uniform at infinity is addressed analytically. Taylor's discriminating function implies that, when the phases have equal viscosities and are assumed to be slightly conducting (leaky dielectrics), a spherical drop is stationary when Q=(2R2+3R+2)/(7R2), where R and Q are ratios of the phases' electric conductivities and dielectric constants, respectively. This condition holds for any electric capillary number, CaE, that defines the ratio of electric stress to surface tension. Pairam and Fernández-Nieves showed experimentally that, in the absence of external forces (CaE=0), a toroidal drop shrinks towards its centre, and, consequently, the drop can be stationary only for some CaE>0. This work finds Q and CaE such that, under the presence of an electric field and with equal viscosities of the phases, a toroidal drop having major radius ρ and volume 4π/3 is qualitatively stationary-the normal velocity of the drop's interface is minute and the interface coincides visually with a streamline. The found Q and CaE depend on R and ρ, and for large ρ, e.g. ρ≥3, they have simple approximations: Q˜(R2+R+1)/(3R2) and CaE∼3 √{3 π ρ / 2 } (6 ln ⁡ρ +2 ln ⁡[96 π ]-9 )/ (12 ln ⁡ρ +4 ln ⁡[96 π ]-17 ) (R+1 ) 2/ (R-1 ) 2.

Design of a new P-NBI control system for 100-s injection in JT-60SA

International Nuclear Information System (INIS)

Okano, F.; Shinozaki, S.; Honda, A.; Ooshima, K.; Numazawa, S.; Ikeda, Y.

2008-01-01

Modification of JT-60U to a superconducting device (so-called JT-60SA) has been planned to contribute to ITER and DEMO. The positive-ion-based NBI system (P-NBI) is required to inject 24 MW for 100 s with 12 units. The P-NBI control system is to be fully remodeled with PLC (Programmable Logic Controller), which is featured by high market availability, system extensibility, cost-effectiveness, and independent development in programming. One of the critical issues to apply the PLC to the P-NBI control system is to control quickly the high voltage power supplies within 200 μs. For this purpose, the fastest PLC dealing with 4 refresh words at the processing time of 200 μs is to be employed. The second issue is to construct a data acquisition system for such a large number of data channels (∼2300 digital and ∼1300 analog data channels). The use of PLC linked with PC-based data measurement devices via Ethernet allows processing the large number of channels. The third issue is to make the man-machine interface simple. The marketed software giving an easy product of graphic menus is available for PLC programming. From these results, it is expected that commercial PLC could be applied to the large-scale control system of the P-NBI system for 100 s operations

Mechanical strength evaluation of the welded bellows for the ports of the JT-60 vacuum vessel

International Nuclear Information System (INIS)

Takatso, H.; Shimizu, M.; Yamamoto, M.

1983-01-01

Mechanical strength of the welded bellows for the ports of the JT-60 vacuum vessel was evaluated, laying the emphasis on the fatigue strength under the torsional electromagnetic force. The welded bellows were designed to be loaded with the forced deflection due to the relative displacement between the vacuum vessel and the external fixed point, the atmospheric pressure and the forced torsional angle due to the electromagnetic force. Stresses caused by the former two were estimated following the formulae proposed by the Kellogg Company. On the other hand, two formulae were established to estimate the stress caused by the last, after examining experimentally the behavior of the welded bellows under the torsional load; one is the shearing stress evaluation formula and the other is the axial bending stress evaluation formula. It was found that the welded bellows can easily buckle under the torsional load and the former formula corresponds to the case of non-buckling and the latter to the case of buckling. The present mechanical strength evaluation method was applied to the three kinds of the welded bellows to be used in the ports of the JT-60 vacuum vessel (neutral beam injection ports, vacuum pumping ports and the adjustable limiter ports) and it was confirmed that they have sufficient strength in the range of the design load conditions

Review of JT-60U experimental results in 2007 and 2008

International Nuclear Information System (INIS)

Isayama, Akihiko; Oyama, Naoyuki; Suzuki, Takahiro; Shinohara, Kouji; Sakamoto, Yoshiteru; Matsunaga, Go; Yoshida, Maiko; Asakura, Nobuyuki; Nakano, Tomohide; Kamiya, Kensaku; Itami, Kiyoshi

2010-02-01

Results in JT-60U experiments in 2007 and 2008 are reviewed. In this campaign, which is the final experimental period in JT-60U, development of advanced tokamak plasma was extensively performed toward establishment of physics basis of ITER and DEMO. High integrated performance plasma with high normalized beta (β N -2.6) and high confinement enhancement factor (H H98(y,2) -1.0-1.1), which are comparable to those in the ITER Hybrid Scenario, and at the same time with high bootstrap current fraction (f BS -40%) was sustained for 25 s. High density and high radiation loss fraction plasma was sustained for 12 s by adding argon and neon to a deuterium plasma. The duration of the high-performance plasmas is more than 10 times longer than the current diffusion time, τ R . In a high beta regime exceeding the ideal MHD limit without conducting wall (no-wall limit), a new instability was observed. By suppressing the instability a high beta plasma was sustained for 5 s, which corresponds to several times longer than τ R . Performance of reversed shear plasmas was significantly improved by utilizing the stabilizing effect of the conducting wall, and β N -2.7 and f BS -90% were obtained. These results significantly exceed those in the previous experimental campaign. In addition, real-time control system was improved, and ion temperature and current profile were independently or simultaneously controlled in real time. Development of new diagnostics was also continuously performed. For example, profiles of electron density and current were measured using the lithium beam probe diagnostic with high resolution. A number of important results from physics experiments were obtained in the area of transport, confinement, instability, plasma-wall interaction etc. Performance of heating and current drive systems was also extended significantly. In the electron cyclotron wave system, 2.9 MW for 5 s injection and 0.4 MW for 30 s injection to plasma were successfully demonstrated. Power

Calculation of modification to the toroidal magnetic field of the Tokamak Novillo. Part II

International Nuclear Information System (INIS)

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

1992-03-01

In a cylindrical magnetic topology. the confined plasma experiences 'classic' collisional transport phenomena. When bending the cylinder with the purpose of forming a toro, the magnetic field that before was uniform now it has a radial gradient which produces an unbalance in the magnetic pressure that is exercised on the plasma in the transverse section of the toro. This gives place to transport phenomena call 'neo-classicist'. In this work the structure of the toroidal magnetic field produced by toroidal coils of triangular form, to which are added even of coils of compensation with form of half moon is analyzed. With this type of coils it is looked for to minimize the radial gradient of the toroidal magnetic field. The values and characteristics of B (magnetic field) in perpendicular planes to the toro in different angular positions in the toroidal direction, looking for to cover all the cases of importance are exhibited. (Author)

Fast dynamics of Type I ELM and transport of ELM pulse in JT-60U

International Nuclear Information System (INIS)

Oyama, N.

2002-01-01

The mitigation of the large ELM heat load on the divertor target is one of the most important issues to be overcome on ITER. Since the ELM heat load strikes the divertor target not as a time-averaged load but as an instantaneous heat pulse, the evaluation of both ELM energy, and the time scale of the collapse and transport is very important. In JT-60U, the detailed dynamic behaviors of the collapse were measured using O-mode reflectometer. The duration of the collapse was within 0.35 ms and the lost pedestal density was recovered quickly within 0.5 ms. The collapse reached 10 cm inside the separatrix, which corresponds to twice the pedestal width of 5 cm. Dedicated edge density measurements on high- and low-field side revealed the poloidal asymmetry of the collapse of density pedestal for the first time. The measurement of SOL flow and heat load to the divertor target by using SOL Mach probe and IRTV showed that convective transport of the SOL plasma gave large contribution to the ELM heat deposition process. (author)

Baroclinic Instability in the Solar Tachocline for Continuous Vertical Profiles of Rotation, Effective Gravity, and Toroidal Field

Energy Technology Data Exchange (ETDEWEB)

Gilman, Peter A., E-mail: gilman@ucar.edu [High Altitude Observatory, National Center for Atmospheric Research, 3080 Center Green, Boulder, CO 80307-3000 (United States)

2017-06-20

We present results from an MHD model for baroclinic instability in the solar tachocline that includes rotation, effective gravity, and toroidal field that vary continuously with height. We solve the perturbation equations using a shooting method. Without toroidal fields but with an effective gravity declining linearly from a maximum at the bottom to much smaller values at the top, we find instability at all latitudes except at the poles, at the equator, and where the vertical rotation gradient vanishes (32.°3) for longitude wavenumbers m from 1 to >10. High latitudes are much more unstable than low latitudes, but both have e -folding times that are much shorter than a sunspot cycle. The higher the m and the steeper the decline in effective gravity, the closer the unstable mode peak to the top boundary, where the energy available to drive instability is greatest. The effect of the toroidal field is always stabilizing, shrinking the latitude ranges of instability as the toroidal field is increased. The larger the toroidal field, the smaller the longitudinal wavenumber of the most unstable disturbance. All latitudes become stable for a toroidal field exceeding about 4 kG. The results imply that baroclinic instability should occur in the tachocline at latitudes where the toroidal field is weak or is changing sign, but not where the field is strong.

Toroidal electric field in front of the lower hybrid grill of the castor tokamak

International Nuclear Information System (INIS)

Zacek, F.; Petrzilka, V.; Devynck, P.; Goniche, M.

2003-01-01

A small tokamak Castor (R/a = 0.4/0.85 m) with low plasma energy density and short pulses (20 ms) offers a unique possibility to carry out probe measurements in front of the grill antenna and as a consequence to provide direct information about the local electric fields in this region. For measurements of the toroidal electrical field, a small double probe with 2 tips separated by 3.5 mm in the toroidal direction has been used. The tips are oriented in the radial direction. The probe is radially movable in front of the central grill waveguide. Cross-correlations and FFT (fast Fourier transform) analysis of the measured V fl signals are given together with an attempt to investigate characteristics of toroidal electric field E tor (up to 500 kHz), derived from V fl measured by 2 toroidally separated tips

About the Toroidal Magnetic Field of a Tokamak Burning Plasma Experiment with Superconducting Coils

International Nuclear Information System (INIS)

Mazzucato, E.

2002-01-01

In tokamaks, the strong dependence on the toroidal magnetic field of both plasma pressure and energy confinement is what makes possible the construction of small and relatively inexpensive burning plasma experiments using high-field resistive coils. On the other hand, the toroidal magnetic field of tokamaks using superconducting coils is limited by the critical field of superconductivity. In this article, we examine the relative merit of raising the magnetic field of a tokamak plasma by increasing its aspect ratio at a constant value of the peak field in the toroidal magnet. Taking ITER-FEAT as an example, we find that it is possible to reach thermonuclear ignition using an aspect ratio of approximately 4.5 and a toroidal magnetic field of 7.3 T. Under these conditions, fusion power density and neutron wall loading are the same as in ITER [International Thermonuclear Experimental Reactor], but the normalized plasma beta is substantially smaller. Furthermore, such a tokamak would be able to reach an energy gain of approximately 15 even with the deterioration in plasma confinement that is known to occur near the density limit where ITER is forced to operate

Prospects for toroidal fusion reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.D.

1994-01-01

Work on the International Thermonuclear Experimental Reactor (ITER) tokamak has refined understanding of the realities of a deuterium-tritium (D-T) burning magnetic fusion reactor. An ITER-like tokamak reactor using ITER costs and performance would lead to a cost of electricity (COE) of about 130 mills/kWh. Advanced tokamak physics to be tested in the Toroidal Physics Experiment (TPX), coupled with moderate components in engineering, technology, and unit costs, should lead to a COE comparable with best existing fission systems around 60 mills/kWh. However, a larger unit size, ∼2000 MW(e), is favored for the fusion system. Alternative toroidal configurations to the conventional tokamak, such as the stellarator, reversed-field pinch, and field-reversed configuration, offer some potential advantage, but are less well developed, and have their own challenges

Development and upgrade of new real time processor in JT-60 data processing system

International Nuclear Information System (INIS)

Sakata, Shinya; Koiwa, Motonao; Matsuda, Toshiaki; Aoyagi, Tetsuo

2000-07-01

At the beginning of JT-60 experiments, the real time processor (RTP) in the data processing system was mainly constructed by PANAFACOM U-1500. As the computer became superannuated, however, it gradually became difficult to maintain both hardware and software. A performance of a recent UNIX workstation has been remarkably progressed. The UNIX workstation has a large flexibility for user application programs, an easiness for maintenance of the hardware and an ability of expansion to peripheral devices. Therefore, the RTP system is newly reconstructed by using the UNIX workstation. This report describes the overview, the basic design and the recent upgrade on the RTP in the data processing system. (author)

Development of the plasma movie database system in JT-60

International Nuclear Information System (INIS)

Sueoka, Michiharu; Kawamata, Yoichi; Kurihara, Kenichi; Seki, Akiyuki

2008-03-01

A plasma movie is generally expected as one of the most efficient methods to know what plasma discharge has been conducted in the experiment. The JT-60 plasma movie is composed of video camera picture looking at a plasma, computer graphics (CG) picture, and magnetic probe signal as a sound channel. In order to use this movie efficiently, we have developed a new system having the following functions: (a) To store a plasma movie in the movie database system automatically combined with the plasma shape CG and the sound according to a discharge sequence. (b) To make a plasma movie is available (downloadable) for experiment data analyses at the Web-site. Especially, this system aimed at minimizing the development cost, and it tried to develop the real-time plasma shape visualization system (RVS) without any operating system (OS) customized for real-time use. As a result, this system succeeded in working under Windows XP. This report deals with the technical details of the plasma movie database system and the real-time plasma shape visualization system. (author)

Effect of eddy currents in the toroidal field coils of a tokamak with an air-core transformer

International Nuclear Information System (INIS)

Tani, Keiji; Kobayashi, Tomofumi; Tamura, Sanae

1975-02-01

The effect of eddy currents in the copper parts of the toroidal field coils is evaluated for a tokamak with the air-core transformer windings located inside the bore of the toroidal field coils. By introducing appropriate weights to the solutions obtained for a simplified cylindrical model, calculation is made of the induction toroidal electric field on the plasma axis in the presence of the eddy currents. The result shows that, to reduce the influence of the eddy currents on the induction one-turn voltage to the permissible level, it is necessary to choose the optimal number of turns and shape of the single conductor of the toroidal field coil. (auth.)

Calculation of modification to the toroidal magnetic field of the Tokamak Novillo. Part II; Calculo de modificacion al campo magnetico toroidal del Tokamak nivillo. Parte II

Energy Technology Data Exchange (ETDEWEB)

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

1992-03-15

In a cylindrical magnetic topology. the confined plasma experiences 'classic' collisional transport phenomena. When bending the cylinder with the purpose of forming a toro, the magnetic field that before was uniform now it has a radial gradient which produces an unbalance in the magnetic pressure that is exercised on the plasma in the transverse section of the toro. This gives place to transport phenomena call 'neo-classicist'. In this work the structure of the toroidal magnetic field produced by toroidal coils of triangular form, to which are added even of coils of compensation with form of half moon is analyzed. With this type of coils it is looked for to minimize the radial gradient of the toroidal magnetic field. The values and characteristics of B (magnetic field) in perpendicular planes to the toro in different angular positions in the toroidal direction, looking for to cover all the cases of importance are exhibited. (Author)

Fast neutron-gamma discrimination on neutron emission profile measurement on JT-60U

International Nuclear Information System (INIS)

Ishii, K.; Okamoto, A.; Kitajima, S.; Sasao, M.; Shinohara, K.; Ishikawa, M.; Baba, M.; Isobe, M.

2010-01-01

A digital signal processing (DSP) system is applied to stilbene scintillation detectors of the multichannel neutron emission profile monitor in JT-60U. Automatic analysis of the neutron-γ pulse shape discrimination is a key issue to diminish the processing time in the DSP system, and it has been applied using the two-dimensional (2D) map. Linear discriminant function is used to determine the dividing line between neutron events and γ-ray events on a 2D map. In order to verify the validity of the dividing line determination, the pulse shape discrimination quality is evaluated. As a result, the γ-ray contamination in most of the beam heating phase was negligible compared with the statistical error with 10 ms time resolution.

Radiation loss and global energy balance of ohmically heated divertor discharge in JT-60 tokamak

International Nuclear Information System (INIS)

Koide, Yoshihiko; Yamada, Kimio; Yoshida, Hidetoshi; Nakamura, Hiroo; Niikura, Setsuo; Tsuji, Shunji

1986-03-01

Divertor experiment in JT-60 with a small divertor chamber has been successfully performed up to 1.6 MA discharge. Several divertor effects were experimentally confirmed as follows. Radiation loss in main plasma saturates with the increase of plasma current and its ratio to the input power is about 20 % at 1.5 MA. The rest of input power is exhausted into the divertor chamber and a half of it is dissipated as the radiation loss. Impurity accumulation is not observed during a few sec without internal MHD activity and gross impurity confinement time is several hundred msec. (author)

Structural design of vacuum bulkheads in piping penetration for the cryostat base of JT-60SA

International Nuclear Information System (INIS)

Nakamura, Shigetoshi; Shibama, Yusuke K.; Masaki, Kei

2016-11-01

This study examined the structure of the boundary box that is capable of installing the cryostat base of JT-60SA in a narrow space. Since other devices stand close in the neighborhood, it was designed to fit within a limited space to avoid interference. Spatial limitation and generated stress caused by each load were used as design conditions. From the calculation results of the generated stress with respect to each load, the maximum stress is generated by the displacement of the pipeline associated with the displacement of the vacuum container at the time of earthquake and 200degC baking, so bellows were designed to absorb the displacement of the piping. It was confirmed through 3-D finite element analysis that this generated stress is less than the allowable stress and there is no problem in structural integrity. This paper explained the composition of major equipment of JT-60SA and the structure of cryostat base. In the structural analysis of the boundary box, consideration was given to the pressure difference during vacuum closure or abnormal events, temperature distribution, pipe displacement associated with the deformation of vacuum vessel, and seismic load. As a result of finite element analysis, it was confirmed that the displacement amount and temperature distribution during plasma operation and baking were within the allowable range. In addition, the maximum stress during cryostat helium leak was also within the allowable range. (A.O.)

NCSX Toroidal Field Coil Design

International Nuclear Information System (INIS)

Kalish M; Rushinski J; Myatt L; Brooks A; Dahlgren F; Chrzanowski J; Reiersen W; Freudenberg K.

2005-01-01

The National Compact Stellarator Experiment (NCSX) is an experimental device whose design and construction is underway at the Department of Energy's Princeton Plasma Physics Laboratory (PPPL). The primary coil systems for the NCSX device consist of the twisted plasma-shaping Modular Coils, the Poloidal Field Coils, and the Toroidal Field (TF) Coils. The TF Coils are D-shaped coils wound from hollow copper conductor, and vacuum impregnated with a glass-epoxy resin system. There are 18 identical, equally spaced TF coils providing 1/R field at the plasma. They operate within a cryostat, and are cooled by LN2, nominally, to 80K. Wedge shaped castings are assembled to the inboard face of these coils, so that inward radial loads are reacted via the nesting of each of the coils against their adjacent partners. This paper outlines the TF Coil design methodology, reviews the analysis results, and summarizes how the design and analysis support the design requirements

Improvement of the protection devices for JT-60U LHRF antenna system

International Nuclear Information System (INIS)

Suzuki, Sadaaki; Seki, Masami; Shinozaki, Shinichi; Sato, Fumiaki; Hiranai, Shinichi; Hasegawa, Koichi; Moriyama, Shinichi; Ishii, Kazuhiro

2007-09-01

In the experiments featuring lower hybrid range of frequency (LHRF) system in JT-60U, carbon grills were attached to the plasma-facing part of the antenna in order to avoid the damage by the excessive heat load from the plasma. However some electric discharge traces were found there in the observation after the experiments. To avoid such discharges, improvements of the arc detector and the protection interlock by visible picture detection were tackled. In the arc detector, the amplification circuit was improved in order to obtain shorter response time and higher resolution of optical detection. Moreover, in visible picture detection, a new function of RF-on/off control utilizing PC image processing was added to distinguish the light of the arc from one of the plasma. This report summarizes improvement of the protection interlock device in a LHRF heating system. (author)

Numerical simulation of a high-brightness lithium ion gun for a Zeeman polarimetry on JT-60U

International Nuclear Information System (INIS)

Kojima, Atsushi; Kamiya, Kensaku; Fujita, Takaaki; Kamada, Yutaka; Iguchi, Harukazu

2007-01-01

A lithium ion gun is under construction for a lithium beam Zeeman polarimetry on JT-60U. The performance of the prototype ion gun has been estimated by the numerical simulation taking the space charge effects into account. The target values of the ion gun are the beam energy of 30 keV, the beam current of 10 mA and the beam divergence angle within 0.13 degrees. The low divergence of 0.13 degrees is required for the geometry of the Zeeman polarimetry on JT-60U where the observation area is 6.5 m away from the neutralizer. The numerical simulation needs to be carried out for the design study because the requirement of the divergence angle is severe for the development of the high-brightness ion gun. The simulation results show the beam loss of 50% caused by the clash to the electrode such as the cathode and the neutralizer. Moreover, the beam transport efficiency from the neutralizer to the observation area is low due to the broadening of the divergence angle. The total beam efficiency is about 5%. Extracted beam profile affects the beam focusing and the efficiency. The peaked profile achieves better efficiency than the hollow one. As a result, beam current of 1 mA is obtained at the observation area by the simulation for the prototype ion gun. (author)

Fast collimated neutron flux measurement using stilbene scintillator and flashy analog-to-digital converter in JT-60U

International Nuclear Information System (INIS)

Ishikawa, M.; Itoga, T.; Okuji, T.; Nakhostin, M.; Shinohara, K.; Hayashi, T.; Sukegawa, A.; Baba, M.; Nishitani, T.

2006-01-01

A line-integrated neutron emission profile is routinely measured using the radial neutron collimator system in JT-60U tokamak. Stilbene neuron detectors (SNDs), which combine a stilbene organic crystal scintillation detector (SD) with an analog neutron-gamma pulse shape discrimination (PSD) circuit, have been used to measure collimated neutron flux. Although the SND has many advantages as a neutron detector, the maximum count rate is limited up to ∼1x10 5 counts/s due to the analog PSD circuit. To overcome this issue, a digital signal processing system (DSPS) using a flash analog-to-digital converter (Acqiris DC252, 8 GHz, 10 bits) has been developed at Cyclotron and Radioisotope Center in Tohoku University. In this system anode signals from photomultiplier of the SD are directory stored and digitized. Then, the PSD between neutrons and gamma rays is performed using software. The DSPS has been installed in the vertical neutron collimator system in JT-60U and applied to deuterium experiments. It is confirmed that the PSD is sufficiently performed and collimated neutron flux is successfully measured with count rate up to ∼5x10 5 counts/s without the effect of pileup of detected pulses. The performance of the DSPS as a neutron detector, which supersedes the SND, is demonstrated

Regularity conditions of the field on a toroidal magnetic surface

International Nuclear Information System (INIS)

Bouligand, M.

1985-06-01

We show that a field B vector which is derived from an analytic canonical potential on an ordinary toroidal surface is regular on this surface when the potential satisfies an elliptic equation (owing to the conservative field) subject to certain conditions of regularity of its coefficients [fr

Eddy current calculations for the Tore Supra toroidal field magnet

International Nuclear Information System (INIS)

Blum, J.

1983-01-01

An outline is given of the calculation of the eddy currents in the magnetic structures of a Tokamak, which can be assimilated to thin conductors, so that the three-dimensional problem can be reduced mathematically to a two-dimensional one, the variables being two orthogonal coordinates of the considered surface. A finite element method has been used in order to treat the complicated geometry of the set of the 18 toroidal field coil casings and mechanical structures of Tore Supra. This eddy current code has been coupled with an axisymmetric equilibrium code in order to simulate typical phases of a Tokamak discharge (plasma current rise, additional heating, disruption, cleaning discharge) and the losses in the toroidal field magnet have thus been calculated. (author)

Balance of the stored energies sustained by the internal and edge transport barriers and effects of ELMs and L-H transitions in JT-60U

International Nuclear Information System (INIS)

Kamada, Y.; Yoshida, M.; Sakamoto, Y.; Koide, Y.; Oyama, N.; Urano, H.; Kamiya, K.; Suzuki, T.; Isayama, A.

2009-01-01

To understand key physics processes determining radial profiles of the kinetic plasma parameters in the advanced tokamak operation scenarios, correlations between the edge transport barrier (ETB) and the internal transport barrier (ITB) have been studied in the JT-60U tokamak device. It has been found that the edge pedestal poloidal beta, β p -ped, increases almost linearly with the total poloidal beta, β p -tot, over a wide range of the plasma current for type I ELMing H-mode plasmas, and this dependence becomes stronger with increasing triangularity. This dependence is not due to the profile stiffness, since the dependence is the same regardless of the existence of ITB. As the stored energy inside the ITB-foot radius (W ITB ) increases, the total thermal stored energy (W th ) increases and then the pedestal stored energy (W ped ) increases. On the other hand, as W ped increases, the ELM penetration expands more inwards and finally reaches the ITB-foot radius. At this situation, the ITB-foot radius cannot move outwards because of the erosion by ELMs. Then the fractions of W ITB /W th and W ped /W th become almost constant. It has also been found that the type I ELM expels/decreases the edge toroidal momentum larger than the edge ion thermal energy. The ELM penetration for the toroidal rotation tends to be deeper than that for the ion temperature and can exceed the ITB-foot radius. The ELM penetration is deeper for CO-rotating plasmas than CTR rotating plasmas. In both cases, the ELM penetration is deeper in the order of the toroidal rotation (V t ), the ion temperature (T i ) and then the electron temperature (T e ). The L-H transition also changes the V t profile more significantly than the T i profile. At the L-H transition, the pedestal V t shifts into the CTR-direction deeply and suddenly without a change in T i , and then the pedestal V t grows further together with a growth of the pedestal T i in a slower timescale. Such changes in V t by ELMs and L

Analyses of the impact of connections’ layout on the coil transient voltage at the Quench Protection Circuit intervention in JT-60SA

International Nuclear Information System (INIS)

Maistrello, Alberto; Gaio, Elena; Novello, Luca; Matsukawa, Makoto; Yamauchi, Kunihito

2015-01-01

The transient overvoltages associated to the interruption of high direct currents with high current derivative, at the base of the operation of a Quench Protection System for Superconducting Coils, have been studied, with particular reference to the JT-60SA project, which adopt edge technology solutions for current interruption: a Hybrid mechanical-static Circuit Breaker as main circuit breaker in series with a PyroBreaker as backup protection. The paper reports in particular on the analyses of the intervention of the backup circuit breaker in the final circuital conditions, considering the actual power connections that will be implemented on Site. The key elements which influence the peak value of the voltage and the relation existing among the different stray impedances of the circuit are identified, thus giving general guidelines for the design of the layout of the power connections. The specific case of JT-60SA is considered, but general criteria can be derived.

Toroidal rotation studies in KSTAR

Science.gov (United States)

Lee, S. G.; Lee, H. H.; Yoo, J. W.; Kim, Y. S.; Ko, W. H.; Terzolo, L.; Bitter, M.; Hill, K.; KSTAR Team

2014-10-01

Investigation of the toroidal rotation is one of the most important topics for the magnetically confined fusion plasma researches since it is essential for the stabilization of resistive wall modes and its shear plays an important role to improve plasma confinement by suppressing turbulent transport. The most advantage of KSTAR tokamak for toroidal rotation studies is that it equips two main diagnostics including the high-resolution X-ray imaging crystal spectrometer (XICS) and charge exchange spectroscopy (CES). Simultaneous core toroidal rotation and ion temperature measurements of different impurity species from the XICS and CES have shown in reasonable agreement with various plasma discharges in KSTAR. It has been observed that the toroidal rotation in KSTAR is faster than that of other tokamak devices with similar machine size and momentum input. This may due to an intrinsically low toroidal field ripple and error field of the KSTAR device. A strong braking of the toroidal rotation by the n = 1 non-resonant magnetic perturbations (NRMPs) also indicates these low toroidal field ripple and error field. Recently, it has been found that n = 2 NRMPs can also damp the toroidal rotation in KSTAR. The detail toroidal rotation studies will be presented. Work supported by the Korea Ministry of Science, ICT and Future Planning under the KSTAR project.

Atomic and molecular processes in JT-60U divertor plasmas

Energy Technology Data Exchange (ETDEWEB)

Takenaga, H.; Shimizu, K.; Itami, K. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1997-01-01

Atomic and molecular data are indispensable for the understanding of the divertor characteristics, because behavior of particles in the divertor plasma is closely related to the atomic and molecular processes. In the divertor configuration, heat and particles escaping from the main plasma flow onto the divertor plate along the magnetic field lines. In the divertor region, helium ash must be effectively exhausted, and radiation must be enhanced for the reduction of the heat load onto the divertor plate. In order to exhaust helium ash effectively, the difference between behavior of neutral hydrogen (including deuterium and tritium) and helium in the divertor plasma should be understood. Radiation from the divertor plasma generally caused by the impurities which produced by the erosion of the divertor plate and/or injected by gas-puffing. Therefore, it is important to understand impurity behavior in the divertor plasma. The ions hitting the divertor plate recycle through the processes of neutralization, reflection, absorption and desorption at the divertor plates and molecular dissociation, charge-exchange reaction and ionization in the divertor plasma. Behavior of hydrogen, helium and impurities in the divertor plasmas can not be understood without the atomic and molecular data. In this report, recent results of the divertor study related to the atomic and molecular processes in JT-60U were summarized. Behavior of neural deuterium and helium was discussed in section 2. In section 3, the comparisons between the modelling of the carbon impurity transport and the measurements of C II and C IV were discussed. In section 4, characteristics of the radiative divertor using Ne puffing were reported. The new diagnostic method for the electron density and temperature in the divertor plasmas using the intensity ratios of He I lines was described in section 5. (author)

Formation of toroidal pre-heat plasma without residual magnetic field for high-beta pinch experiments

International Nuclear Information System (INIS)

Ikeda, Nagayasu; Tamaru, Ken; Nagata, Akiyoshi.

1979-01-01

Formation of toroidal pre-heat plasma was studied. The pre-heat plasma without residual magnetic field was made by chopping the current for pre-heat, A small toroidal-pinch system was used for the experiment. The magnetic field was measured with a magnetic probe. One turn loop was used for the measurement of the toroidal one-turn electric field. A pair of Rogoski coil was used for the measurement of plasma current. The dependence of residual magnetic field on chopping time was measured. By fast chopping of the primary current in the pre-heating circuit, the poloidal magnetic field was reduced to several percent within 5 microsecond. After chopping, no instability was observed in the principal discharge plasma produced within several microsecond. As the conclusion, it can be said that the control of residual field can be made by current chopping. (Kato, T.)

Dynamic processes in field-reversed-configuration compact toroids

International Nuclear Information System (INIS)

Rej, D.J.

1987-01-01

In this lecture, the dynamic processes involved in field-reversed configuration (FRC) formation, translation, and compression will be reviewed. Though the FRC is related to the field-reversed mirror concept, the formation method used in most experiments is a variant of the field-reversed Θ-pinch. Formation of the FRC eqilibrium occurs rapidly, usually in less than 20 μs. The formation sequence consists of several coupled processes: preionization; radial implosion and compression; magnetic field line closure; axial contraction; equilibrium formation. Recent experiments and theory have led to a significantly improved understanding of these processes; however, the experimental method still relies on a somewhat empirical approach which involves the optimization of initial preionization plasma parameters and symmetry. New improvements in FRC formation methods include the use of lower voltages which extrapolate better to larger devices. The axial translation of compact toroid plasmas offers an attractive engineering convenience in a fusion reactor. FRC translation has been demonstrated in several experiments worldwide, and these plasmas are found to be robust, moving at speeds up to the Alfven velocity over distances of up to 16 m, with no degradation in the confinement. Compact toroids are ideal for magnetic compression. Translated FRCs have been compressed and heated by imploding liners. Upcoming experiments will rely on external flux compression to heat a translater FRC at 1-GW power levels. 39 refs

General Atomic's superconducting toroidal field coil concept

International Nuclear Information System (INIS)

Alcorn, J.; Purcell, J.

1978-01-01

General Atomic's concept for a superconducting toroidal field coil is presented. The concept is generic for large tokamak devices, while a specific design is indicated for a 3.8 meter (major radius) ignition/burn machine. The concept utilizes bath cooled NbTi conductor to generate a peak field of 10 tesla at 4.2 K. The design is simple and straightforward, requires a minimum of developmental effort, and draws extensively upon the perspective of past experience in the design and construction of large superconducting magnets for high energy physics. Thus, the primary emphasis is upon economy, reliability, and expeditious construction scheduling. (author)

Radial electric field in JET advanced tokamak scenarios with toroidal field ripple

Energy Technology Data Exchange (ETDEWEB)

Crombe, K [Postdoctoral Fellow of the Research Foundation - Flanders (FWO), Department of Applied Physics, Ghent University, Rozier 44, B-9000 Gent (Belgium); Andrew, Y; De Vries, P C; Giroud, C; Hawkes, N C; Meigs, A; Zastrow, K-D [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom); Biewer, T M [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169, TN (United States); Blanco, E [Laboratorio Nacional de Fusion, Asociacion EURATOM-CIEMAT, Madrid (Spain); Tala, T [VTT Technical Research Centre of Finland, Association EURATOM-Tekes, PO Box 1000, FIN-02044 VTT (Finland); Von Hellermann, M [FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands)], E-mail: Kristel.Crombe@jet.uk

2009-05-15

A dedicated campaign has been run on JET to study the effect of toroidal field (TF) ripple on plasma performance. Radial electric field measurements from experiments on a series of plasmas with internal transport barriers (ITBs) and different levels of ripple amplitude are presented. They have been calculated from charge exchange measurements of impurity ion temperature, density and rotation velocity profiles, using the force balance equation. The ion temperature and the toroidal and poloidal rotation velocities are compared in plasmas with both reversed and optimized magnetic shear profiles. Poloidal rotation velocity (v{sub {theta}}) in the ITB region is measured to be of the order of a few tens of km s{sup -1}, significantly larger than the neoclassical predictions. Increasing levels of the TF ripple are found to decrease the ion temperature gradient in the ITB region, a measure for the quality of the ITB, and the maximum value of v{sub {theta}} is reduced. The poloidal rotation term dominates in the calculations of the total radial electric field (E{sub r}), with the largest gradient in E{sub r} measured in the radial region coinciding with the ITB.

Validation of neutral point on JT-60U, Alcator C-Mod and ASDEX-Upgrade tokamaks

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Yoshino, Ryuji; Pautasso, Gabriella; Gruber, Otto; Jardin, Stephen

2002-01-01

Validation studies of a neutrally balanced vertical plasma position, so-called ''neutral point'', have been carried out by computational simulations and experiments under trilateral Japan-US-EU collaborations. It was clarified that the neutral point, where VDEs (Vertical Displacement Events) are hardly occurred, does exit in the Alcator C-Mod and ASDEX-Upgrade tokamaks as well as the JT-60U, consistent with the simulations. Meanwhile, precise details of the VDE behavior exhibit their own characters according to the individual of the tokamaks such as an up-down asymmetry of plasma shape. Sensitivity of the neutral point to the plasma shape and current profile was also addressed in detail. (author)

Edge safety factor at the onset of plasma disruption during VDEs in JT-60U

International Nuclear Information System (INIS)

Sugihara, Masayoshi; Lukash, Victor; Khayrutdinov, Rustam; Neyatani, Yuzuru

2004-01-01

Detailed examinations of the value of the edge safety factor (q a ) at the onset of thermal quench (TQ) during intentional vertical displacement event (VDE) experiments in JT-60U are carried out using two different reconstruction methods, FBI/FBEQU and DINA. The results from the two methods are very similar and show that the TQ occurs when the q a value is in the range between 1.5 and 2. This result suggests that the predictive simulations for VDEs should be performed within this range of q to examine the subsequent differences in the halo currents, plasma movement and other plasma behaviour during the current quench

Divertor characteristics and control on the W-shaped divertor with pump of JT-60U

International Nuclear Information System (INIS)

Hosogane, N.; Kubo, H.; Higashijima, S.

1999-01-01

Roles of the inner leg pumping and the private dome, which are special features of the W-shaped divertor of JT-60U, have been investigated. The following observations were made: The inner leg pumping functions well in attached states or partially detached states with weak X-point MARFE where the inner particle recycling is enhanced. A combination of main gas puff and inner leg pump is effective in reduction of intrinsic carbon impurity. Geometrical effects of the private dome on transport of hydrocarbons in the private flux region was confirmed by spectroscopic measurements of CD-band intensity profile and impurity transport simulation code using experimental data. (author)

Toroidal equilibrium in an iron-core reversed field pinch

International Nuclear Information System (INIS)

Miller, G.

1984-04-01

An analytical theory of toroidal equilibrium in the ZT-40M reversed field pinch is obtained, including effects of iron cores and resistive shell. The iron cores alter the form of the equilibrium condition and cause the equilibrium to be unstable on the shell resistive time scale

The Electromagnetic Field of Elementary Time-Dependent Toroidal Sources

International Nuclear Information System (INIS)

Afanas'ev, G.N.; Stepanovskij, Yu.P.

1994-01-01

The radiation field of toroidal-like time-dependent current configurations is investigated. Time-dependent charge-current sources are found outside which the electromagnetic strengths disappear but the potentials survive. This can be used to carry out time-dependent Aharonov-Bohm-like experiments and the information transfer. Using the Neumann-Helmholtz parametrization of the current density we present the time-dependent electromagnetic field in a form convenient for applications. 17 refs

Engineering study of the neutral beam and rf heating systems for DIII-D, MFTF-B, JET, JT-60 and TFTR

International Nuclear Information System (INIS)

Lindquist, W.B.; Staten, S.H.

1987-01-01

An engineering study was performed on the rf and neutral beam heating systems implemented for DIII-D, MFTF-B, JET, JT-60 and TFTR. Areas covered include: methodology used to implement the systems, technology, cost, schedule, performance, problems encountered and lessons learned. Systems are compared and contrasted in the areas studied. Summary statements were made on common problems and lessons learned. 3 refs., 6 tabs

Behaviour of tritium in the vacuum vessel of JT-60U

International Nuclear Information System (INIS)

Kobayashi, K.; Miya, N.; Ikeda, Y.; Torikai, Y.; Saito, M.; Alimov, V.

2015-01-01

The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D 2 (92.8 %) - T 2 (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily

Behaviour of tritium in the vacuum vessel of JT-60U

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, K.; Miya, N.; Ikeda, Y. [JT-60 Safety Assessment Group, JAEA, Mukoyama (Japan); Torikai, Y. [Hydrogen Isotope Research Center, University of Toyama, Gofuku (Japan); Saito, M.; Alimov, V. [ITER Project Management Group, JAEA, Mukoyama (Japan)

2015-03-15

The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D{sub 2} (92.8 %) - T{sub 2} (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily.

Pulsed-laser ablation of co-deposits on JT-60 graphite tile

International Nuclear Information System (INIS)

Sakawa, Youichi; Watanabe, Daisuke; Shibahara, Takahiro; Sugiyama, Kazuyoshi; Tanabe, Tetsuo

2007-01-01

Pulsed laser ablation of the co-deposits on a JT-60 open-divertor tile using the fourth harmonic of a 20 ps-Nd: YAG laser has been investigated. With increasing the laser intensity, three regions, non-ablation region (NAR), weak-ablation region (WAR), and strong-ablation region (SAR) were distinguished. Transition from NAR to WAR and WAR to SAR occurred at the threshold laser intensity for laser ablation and that for strong ionization of carbon atoms, respectively. The ablation accompanied desorption of H 2 and C 2 H 2 , with minor contribution of other hydrocarbons, while production of H 2 O was small. In NAR and WAR the number of the hydrogen desorbed by the laser irradiation was less than that of hydrogen retained in the ablated volume, while in SAR it was much larger, owing to thermal desorption of hydrogen gas from the region surrounding the ablated volume. For the ablative removal of hydrogen isotopes, SAR is more desirable because of higher removal efficiency and less production of hydrocarbons

Pulsed-laser ablation of co-deposits on JT-60 graphite tile

Energy Technology Data Exchange (ETDEWEB)

Sakawa, Youichi [Institute of Laser Engineering, Osaka University, Yamadaoka, Suita, Osaka 565-0871 (Japan)]. E-mail: sakawa-y@ile.osaka-u.ac.jp; Watanabe, Daisuke [Graduate School of Engineering, Nagoya University, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Shibahara, Takahiro [Graduate School of Engineering, Nagoya University, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Sugiyama, Kazuyoshi [Interdisciplinary School of Engineering Science, Kyushu University, Fukuoka, Fukuoka 812-8581 (Japan); Tanabe, Tetsuo [Interdisciplinary School of Engineering Science, Kyushu University, Fukuoka, Fukuoka 812-8581 (Japan)

2007-08-01

Pulsed laser ablation of the co-deposits on a JT-60 open-divertor tile using the fourth harmonic of a 20 ps-Nd: YAG laser has been investigated. With increasing the laser intensity, three regions, non-ablation region (NAR), weak-ablation region (WAR), and strong-ablation region (SAR) were distinguished. Transition from NAR to WAR and WAR to SAR occurred at the threshold laser intensity for laser ablation and that for strong ionization of carbon atoms, respectively. The ablation accompanied desorption of H{sub 2} and C{sub 2}H{sub 2}, with minor contribution of other hydrocarbons, while production of H{sub 2}O was small. In NAR and WAR the number of the hydrogen desorbed by the laser irradiation was less than that of hydrogen retained in the ablated volume, while in SAR it was much larger, owing to thermal desorption of hydrogen gas from the region surrounding the ablated volume. For the ablative removal of hydrogen isotopes, SAR is more desirable because of higher removal efficiency and less production of hydrocarbons.

Local transport analysis of L-mode plasmas in JT-60 tokamak

International Nuclear Information System (INIS)

Hirayama, Toshio; Kikuchi, Mitsuru; Shirai, Hiroshi; Shimizu, Katsuhiro; Yagi, Masatoshi; Koide, Yoshihiko; Ishida, Shinichi; Azumi, Masafumi.

1991-03-01

Local heat transport has been studied in auxiliary heated JT-60 plasmas with emphasis on understanding the deteriorated confinement observed in L-mode plasmas. The systematic experiment and analysis have been carried out in L-mode phase of divertor (single null, lower X-point), and limiter discharges with hydrogen neutral beam heating into hydrogen plasmas, based on sets of consistent experimental data including ion temperature profiles from CXR measurements. The deterioration in the energy confinement time with increasing the auxiliary heating power, so-called the power scaling, is mainly due to the degradation in ion energy transport. The confinement improvement as the plasma current increases is followed by both improvement in ion and electron transport properties. It is found that the ion thermal diffusivity has an approval dependence on the density. High ion temperature (T i (0) ≤ 12 keV) L-mode plasmas are attained at high β p up to 3.5. The centrally peaked ion temperature is significantly due to the improvement in ion transport property, which is reduced to the level of the electron thermal diffusivities. (author)

Detailed design studies at CEA for JT-60SA TF coils

International Nuclear Information System (INIS)

Decool, P.; Marechal, J.L.; Portafaix, C.; Lacroix, B.; Gros, G.; Verger, J.M.

2011-01-01

Following a first conceptual design activity in which the general design of the JT-60SA TF system was defined and frozen in agreement with all the participants in the project (CEA, ENEA, F4E), a second phase had to be launched to deal with the detailed design. In this paper, we present the work performed at CEA on the TF coil design during this second phase. Part of this work, concerns the determination of conductor hydraulic performances during operation as well as in factory. The thermohydraulic of the conductor was also assessed to confirm the need of helium inlets and a specific design was developed and qualified to be compatible with the available hydraulic performance of the cryoplant. The mechanical behavior is still to be assessed and qualified. Last but not least, the inner electrical joints of the coil have been modified with respect to the original twin-box design developed by CEA for the ITER coils in order to simplify the fabrication process. A dedicated qualification program for their manufacture is ongoing.

Long pulse operation of high performance plasmas in JT-60U

International Nuclear Information System (INIS)

Ide, Shunsuke

2005-01-01

Recent experimental progress in JT-60U advanced tokamak research is presented; sustainment of the normalized beta (β N ) - 3 in a normal magnetic shear plasma, the bootstrap current fraction (f BS ) - 45% in a weak shear plasma and ∼75% in a reversed magnetic shear plasma in a nearly full non-inductive current drive condition for longer than the current relaxation time. Achievement of high-density high-radiation fraction together with high-confinement in advanced plasmas was demonstrated. Achievement and foundings in long pulse operations after system modification are presented as well. A 65 s discharge of I p =0.7 MA was successfully obtained. As a result, high-β N of 2.3 was successfully sustained for a very long period of 22.3 s. In addition, a 30 s standard ELMy H-mode plasma of I p up to 1.4 MA has also been obtained. Effectiveness of divertor pumping to control particle recycling and the electron density under the wall retention was saturated was demonstrated. These achievement and issues in the development will be discussed. (author)

Compact toroid formation using barrier fields and controlled reconnection in the TRX-1 field reversed theta pinch

International Nuclear Information System (INIS)

Hoffman, A.L.; Armstrong, W.T.

1982-01-01

TRX-1 is a new 20 cm diameter, 1-m long field reversed theta pinch with a magnetic field swing of 10 kG in 3 μsec. It employs z discharge preionization and octopole barrier fields to maximize flux trapping on first half cycle operation. Cusp coils are used at the theta pinch ends to delay reconnection and fast mirror coils are used to trigger reconnection at a time designed to maximize axial heating efficiency and toroid lifetime. These controls are designed to study toroid formation methods which are claimed to be especially efficient by Russian experimenters. Studies have been conducted on flux trapping efficiency, triggered reconnection, and equilibrium and lifetime

Retention of Hydrogen Isotopes in Divertor Tiles Used in JT-60U

International Nuclear Information System (INIS)

Hirohata, Y.; Shibahara, T.; Tanabe, T.; Oya, Y.; Arai, T.; Gotoh, Y.; Masaki, K.; Yagyu, J.; Oyaidzu, M.; Okuno, K.; Nishikawa, M.; Miya, N.

2005-01-01

Retention characteristics of deuterium and hydrogen retained in graphite tiles placed in the divertor region of JT-60U were investigated by thermal desorption spectroscopy (TDS). The deuterium retained in the near surface of all graphite tiles was mostly replaced by hydrogen due to exposure to hydrogen plasma at the final stage operations, resulting in main deuterium retention in the deeper region. The dominant species desorbed from the divertor tiles were H 2 , HD, D 2 and CH 4 . The smallest retention of hydrogen isotopes (H+D) was observed in the outer divertor tile which was eroded with maximum of 20 μm depth. The amount of H+D retained in the inner divertor tiles covered by the re-deposited layers increased with the thickness of the re-deposited layers. Hydrogen isotopes concentration ((H+D)/C) in the re-deposited layers was ∼0.02, which was much smaller than those observed in JET and other devices

Dynamic response of the JT-60 vacuum vessel under the electromagnetic forces

International Nuclear Information System (INIS)

Takatsu, H.; Shimizu, M.; Ohta, M.

1982-01-01

Dynamic response analyses of the JAERI Tokamak 60 (JT-60) vacuum vessel were carried out under three kinds of saddle-like electromagnetic forces. In the analysis, the dynamic response of the bellows was obtained by dividing it into three components; the first, caused by the forced deflection due to the displacement of an adjacent rigid ring; the second, caused by inertia force; and the third, caused by a saddle-like electromagnetic force. Eigenvalue analyses showed that the 20th mode is a typical rotation mode of the rigid ring around the major radius with a natural frequency of 46.3 Hz. From the results of the dynamic response analyses, the maximum displacement response of the rigid ring was 3.1 mm and remarkable dynamic response was observed in the case of plasma disruption with a time constant of 1 ms. In cases of start-up of the plasma current and plasma disruption with a time constant of 50 ms, the rigid ring vibrates quasi-statically. It is clear that the dynamic behavior of the vacuum vessel is governed mainly by the saddle-like electromagnetic force, with a smaller effect of the inverse saddle-like electromagnetic force on the dynamic response of the vacuum vessel. (orig.)

Structural design of the superconducting toroidal field coils for ITER

International Nuclear Information System (INIS)

Wong, F.M.G.; Sborchia, C.; Thome, R.J.; Malkov, A.; Titus, P.H.

1995-01-01

Structural design issues and features of the superconducting toroidal field (TF) coils for the International Thermonuclear Experimental Reactor (ITER) will be discussed. Selected analyses of the structural and mechanical behavior of the ITER TF coils will also be presented. (orig.)

Toroidal field magnet and poloidal divertor field coil systems adapted to reactor requirements

International Nuclear Information System (INIS)

Koeppendoerfer, W.

1985-01-01

ASDEX Upgrade is a tokamak experiment with external poloidal field coils, that is now under construction at IPP Garching. It can produce elongated single-null (SN), double-null (DN) and limiter (L) configurations. The SN is the reference configuration with asymmetric load distributions in the poloidal field (PF) system and the toroidal field (TF) magnet. Plasma control and stabilization requires a rigid passive conductor close to the plasma. The design principles of the coils and support structure are described. (orig.)

Development of hard-seal gate valve and fast shutter for JT-60 neutral beam injectors

International Nuclear Information System (INIS)

Kuribayashi, S.; Minami, M.; Matsuoka, T.; Takeshita, K.; Morita, H.; Kuriyama, M.; Matsuda, S.; Shirakata, H.

1983-01-01

A 600 mm hard-seal valve and a fast shutter for the JT-60 Neutral Beam Injector were developed. The 600 mm hard-seal gate valve was fabricated and tested for 500 cycles at various temperatures of up to 250 0 C. In consequence, requirements of the endurance and vacuum tightness were satisfied. Major components of the fast shutter, i.e., swing action bellows and a high-speed pneumatic cylinder, were tested for 30,000 cycles, and their reliability was confirmed. Then the fast shutter was fabricated and tested. The test result indicated that the fast shutter fully satisfied the requirements of the molecular gas flow conductance and opening/closing speed. (author)

Toroidal rotation braking with n = 1 magnetic perturbation field on JET

DEFF Research Database (Denmark)

Sun, Y; Liang, Y; Koslowski, H R

2010-01-01

A strong toroidal rotation braking has been observed in plasmas with application of an n = 1 magnetic perturbation field on the JET tokamak. Calculation results from the momentum transport analysis show that the torque induced by the n = 1 perturbation field has a global profile. The maximal value...

Progress on large superconducting toroidal field coils

International Nuclear Information System (INIS)

Haubenreich, P.N.; Luton, J.N.; Thompson, P.B.; Beard, D.S.

1979-01-01

Large superconducting toroidal field coils of competing designs are being produced by six major industrial teams. In the US, teams headed by General Dynamics Convair, General Electric, and Westinghouse are under contract to design and fabricate one coil each to specifications established by the Large Coil Program. A facility for testing 6 coils in a toroidal array at fields to 8 to 12 tesla is under construction at Oak Ridge. Through an international agreement, EURATOM, Japan, and Switzerland will produce one coil each for testing with the US coils. Each test coil will have a 2.5 x 3.5 m D-shape winding bore and is designed to operate at a current of 10 to 18 kA at a peak field of 8T while subjected to pulsed fields of 0.14 T applied in 1.0 s. There are significant differences among the six coil designs: five use NbTi, one Nb 3 Sn; three are cooled by pool boiling helium, three by forced flow; five have welded or bolted stainless steel coil cases, one has aluminum plate structure. All are designed to be cryostable at 8T, with structural margin for extended operation. The three US coil teams are almost or completely finished with detailed design and are now procuring materials and setting up manufacturing equipment. The non-US teams are at various stages of verification testing and design. The GDC and GE coils are scheduled for delivery in the spring of 1981 and the others will be completed a year later. The 11-m diameter vessel at the test facility has been completed and major components of the test stand are being procured. Engineering and procurement to upgrade the helium liquifier-refrigerator system are under way

Toroidal fusion reactor design based on the reversed-field pinch

International Nuclear Information System (INIS)

Hagenson, R.L.

1978-07-01

The toroidal reversed-field pinch (RFP) achieves gross equilibrium and stability with a combination of high shear and wall stabilization, rather than the imposition of tokamak-like q-constraints. Consequently, confinement is provided primarily by poloidal magnetic fields, poloidal betas as large as approximately 0.58 are obtainable, the high ohmic-heating (toroidal) current densities promise a sole means of heating a D-T plasma to ignition, and the plasma aspect ratio is not limited by stability/equilibrium constraints. A reactor-like plasma model has been developed in order to quantify and to assess the general features of a power system based upon RFP confinement. An ''operating point'' has been generated on the basis of this plasma model and a relatively detailed engineering energy balance. These results are used to generate a conceptual engineering model of the reversed-field pinch reactor (RFPR) which includes a general description of a 750 MWe power plant and the preliminary consideration of vacuum/fueling, first wall, blanket, magnet coils, iron core, and the energy storage/transfer system

Development of remote pipe welding tool for divertor cassettes in JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Hayashi, Takao, E-mail: hayashi.takao@jaea.go.jp [Fusion Research and Development Directorate, Japan Atomic Energy Agency, Naka (Japan); Sakurai, Shinji; Sakasai, Akira; Shibanuma, Kiyoshi [Fusion Research and Development Directorate, Japan Atomic Energy Agency, Naka (Japan); Kono, Wataru; Ohnawa, Toshio; Matsukage, Takeshi [Toshiba Corporation, Yokohama, Kanagawa (Japan)

2015-12-15

Highlights: • Remote pipe welding tool accessing from inside of the pipe has been newly developed. • Cooling pipe with a jut on the edge expands the acceptable welding gap up to 0.5 mm. • Positioning accuracy of the laser beam is realized to be less than ±0.1 mm. • We have achieved robust welding for an angular misalignment of 0.5°. - Abstract: Remote pipe welding tool accessing from inside of the pipe has been newly developed for JT-60SA. Remote handling (RH) system is necessary for the maintenance and repair of the divertor cassette in JT-60SA. Because the space around the cooling pipe connected with the divertor cassette is very limited, the cooling pipe is to be remotely cut and welded from inside for the maintenance. A laser welding method was employed to perform circumferential welding by rotating the focusing mirror inside the pipe. However, the grooves of connection pipes are not precisely aligned for welding. The most critical issue is therefore to develop a reliable welding tool for pipe connection without a defect such as undercut weld due to a gap caused by angular and axial misalignments of grooves. In addition, an angular misalignment between two pipes due to inclination of pipe has to be taken into account for the positioning of the laser beam during welding. In this paper, the followings are proposed to solve the above issues: (1) Cooling pipe connected with the divertor is machined to have a jut on the edge so as to expand the acceptable welding gap up to 0.5 mm by filling the gap with welded jut. (2) Positioning accuracy of the laser beam for reliable welding is realized to be less than ±0.1 mm along the circumferential target for welding by a position control mechanism provided in the tool even in the case of up to angular misalignment of 0.5° between connection pipes. Based on the above proposals, we have achieved robust welding for a large gap up to 0.5 mm as well as the maximum angular misalignment of 0.5° between connection pipes

Development of a global toroidal gyrokinetic Vlasov code with new real space field solver

International Nuclear Information System (INIS)

Obrejan, Kevin; Imadera, Kenji; Li, Ji-Quan; Kishimoto, Yasuaki

2015-01-01

This work introduces a new full-f toroidal gyrokinetic (GK) Vlasov simulation code that uses a real space field solver. This solver enables us to compute the gyro-averaging operators in real space to allow proper treatment of finite Larmor radius (FLR) effects without requiring any particular hypothesis and in any magnetic field configuration (X-point, D-shaped etc). The code was well verified through benchmark tests such as toroidal Ion Temperature Gradient (ITG) instability and collisionless damping of zonal flow. (author)

Closed expressions for the magnetic field of toroidal multipole configurations

International Nuclear Information System (INIS)

Sheffield, G.V.

1983-04-01

Closed analytic expressions for the vector potential and the magnetic field for the lower order toroidal multipoles are presented. These expressions can be applied in the study of tokamak plasma cross section shaping. An example of such an application is included. These expressions also allow the vacuum fields required for plasma equilibrium to be specified in a general form independent of a particular coil configuration

Electromagnetic analysis, structural integrity and progress on mechanical design of the ITER ferromagnetic insert

Energy Technology Data Exchange (ETDEWEB)

Morimoto, M. [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: masaaki_morimoto@maia.eonet.ne.jp; Ioki, K.; Terasawa, A.; Utin, Yu.; Barabash, V.; Gribov, Y. [ITER Organization, 13108 St. Paul lez Durance (France)

2009-12-15

Ferromagnetic material is used to reduce the toroidal field ripple in JFT-2M and JT-60U . In ITER, since the ferromagnetic material is inserted in the space between the double walls of ITER Vacuum Vessel (VV), it is called 'ferromagnetic inserts'. Suitable material is selected to satisfy the design requirements of ITER. The proper location and amount of the ferromagnetic inserts are optimized with the goal of reduction of the toroidal field ripple. The ferromagnetic inserts are designed to minimize electromagnetic forces acting on them. The electromagnetic forces have been calculated with the latest disruption scenarios. Magnetization forces due to magnetic fields have also been calculated. Structural integrity has been validated by a structural analysis.

Roles of electric field on toroidal magnetic confinement

International Nuclear Information System (INIS)

Itoh, Kimitaka; Itoh, Sanae; Sanuki, Heiji; Fukuyama, Atsushi.

1992-11-01

Theoretical research on the influence of the electric field on the toroidal magnetic confinement is surveyed. The static electric field is first described. Physics pictures on the generation of the radial electric field and the influence on the confinement are shown. Neoclassical effects as well as the nonclassical processes are discussed. Emphasis is made on the connection with the improved confinement. Convective cell, i.e. the nonuniform potential on the magnetic surface is also discussed. The roles of the fluctuating electric field are then reviewed. The progress in the recent theories on the anomalous transport is addressed. Through these surveys, the impact of the experiments using the heavy ion beam probes on the modern plasma physics is illustrated. (author) 66 refs

Plasma equilibrium control during slow plasma current quench with avoidance of plasma-wall interaction in JT-60U

Science.gov (United States)

Yoshino, R.; Nakamura, Y.; Neyatani, Y.

1997-08-01

In JT-60U a vertical displacement event (VDE) is observed during slow plasma current quench (Ip quench) for a vertically elongated divertor plasma with a single null. The VDE is generated by an error in the feedback control of the vertical position of the plasma current centre (ZJ). It has been perfectly avoided by improving the accuracy of the ZJ measurement in real time. Furthermore, plasma-wall interaction has been avoided successfully during slow Ip quench owing to the good performance of the plasma equilibrium control system

Toroidal rotation braking with n = 1 magnetic perturbation field on JET

International Nuclear Information System (INIS)

Sun, Y; Liang, Y; Koslowski, H R; Harting, D; Wiegmann, C; Wiesen, S; Jachmich, S; Alfier, A; Asunta, O; Corrigan, G; Giroud, C; Gryaznevich, M P; Hender, T; Nardon, E; Parail, V; Naulin, V; Tala, T

2010-01-01

A strong toroidal rotation braking has been observed in plasmas with application of an n = 1 magnetic perturbation field on the JET tokamak. Calculation results from the momentum transport analysis show that the torque induced by the n = 1 perturbation field has a global profile. The maximal value of this torque is at the plasma core region (ρ - √ν regime in the plasma core, but it is close to the transition between the 1/ν and ν - √ν regimes. The neoclassical toroidal viscosity (NTV) torque in the 1/ν and ν - √ν regimes is calculated. The observed torque is of a magnitude in between that of the NTV torque in the 1/ν and ν - √ν regimes. The NTV torque in the ν - √ν regimes is enhanced using the Lagrangian variation of the magnetic field strength. However, it is still smaller than the observed torque by one order of magnitude.

Increase of the positive ion source power in JT-60 NBI

International Nuclear Information System (INIS)

Kawai, Mikito; Akino, Noboru; Ebisawa, Noboru

1998-09-01

Neutral Beam Injection (NBI) heating experiment in JT-60 started in 1986, and the rated injection power of 20MW at 75keV with hydrogen was achieved after several month operation. In 1991, the ion sources and power supply had been upgraded for a higher beam energy up to 120keV with deuterium, following which the ion source operation re-started aiming for an injection power of 40MW at 110keV. In the operation, the beam acceleration voltage was tried to increase by modifying the ion source structure against the break-down which occurred frequently in the ion source. The beam acceleration was, however, unstable in a beam energy range of more than 105keV because of voltage-holding deterioration in the accelerator. Therefore we changed the strategy to increase the injection power: i.e. we tried to increase the beam current with keeping the beam energy. The structure of the source has been modified to be operated in a high current regime. As a result, the deuterium neutral beam injection of 40MW at 91-96keV was achieved in July 1996. (author)

3. Laser Thomson scattering by plasmas. 3.2. Applications of incoherent Thomson scattering. 3.2.2. Incoherent Thomson scattering systems for JT-60U and JFT-2M

International Nuclear Information System (INIS)

Hatae, Takaki; Yoshida, Hidetoshi; Naito, Osamu; Yamauchi, Toshihiko

2000-01-01

Development of Thomson scattering diagnostics for the JT-60U and JFT-2M Tokamaks are described. Two Thomson scattering systems have been installed on JT-60U. The first system uses two ruby lasers (10 J, 0.25 Hz) and measures electron temperature (T e ) and density (n e ) profiles of 60 spatial points with high spatial resolution (8 mm). The second system uses a YAG laser (2 J, 30 Hz) and measures time evolution of T e and n e profiles with 15 spatial points. On JFT-2M, a TV Thomson Scattering system (TVTS) has been installed and measures at 81 spatial points with high spatial resolution (8.6 mm). These systems have provided not only profiles of all over the plasma, but successfully measured local structures to study various physics issues; e.g. H-mode edge pedestal, internal transport barrier, local MHD event. (author)

Calculation about a modification to the toroidal magnetic field of the Tokamak Novillo. Part I

International Nuclear Information System (INIS)

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

1991-07-01

The charged particles that constitute the plasma in the tokamaks are located in magnetic fields that determine its behavior. The poloidal magnetic field of the plasma current and the toroidal magnetic field of the tokamak possess relatively big gradients, which produce drifts on these particles. These drifts are largely the cause of the continuous lost of particles and of energy of the confinement region. In this work the results of numerical calculations of a modification to the 'traditional' toroidal magnetic field that one waits it diminishes the drifts by gradient and improve the confinement properties of the tokamaks. (Author)

The Pre-compression System of the Toroidal Field Coils in ITER

International Nuclear Information System (INIS)

Knaster, J.; Jong, C.; Vollmann, T.; Ferrari, M.

2006-01-01

The Toroidal Field (TF) coils of ITER will undergo out-of-plane forces caused by the machine poloidal fields required to maintain the toroidal stability of the plasma. These forces will be supported against overturning moments by links between the coils. In turn, these links consist of the Inner Intercoil Structure (IIC), which are composed by 2 sets of 4 poloidal shear keys inserted in slots between adjacent coils placed at the top and bottom part of the inboard leg, and the Outer Intercoil Structure (OIS) formed by 4 bands of shear panels at the outboard leg. The magnetic forces during energization of ITER would cause at IIC locations a toroidal gap between adjacent TF coils of 0.35 mm; during plasma operation this value could reach >1 mm causing a loosening of the keys and intensifying stress concentrations. This undesired effect will be suppressed by the application of a centripetal force of 70 MN per coil (35 MN at both the bottom and top part of the inboard leg of each of the 18 TF coils) that will be provided by 2 sets of 3 fibre-glass epoxy composite rings submitted to a toroidal hoop force of 100 MN per set. The calculated maximum stress in the rings will occur during the installation phase at room temperature, where the maximum radial elongation (∼ 25 mm) is required, and it will be less than 30% of its ultimate stress. The imposed elongation to reach that force and the lower Young modulus of the composite compared with the stainless steel one will ease component tolerances and/or settlement effects in the final assembly. (author)

Photoelastic analyses of stresses in toroidal magnetic field coils

International Nuclear Information System (INIS)

Pih, H.

1977-02-01

Several two-dimensional photoelastic stress analyses were made on models of circular and oval toroidal magnetic field coils for fusion reactors. The circumferential variation of each coil's in-plane magnetic force was simulated by applying different pressures to 16 segmented regions of the inner surface of the models. One special loading fixture was used for the model of each shape and size. Birefringence and isoclinic angles were measured in a transmission polariscope at selected points on the loaded model. Boundary stresses in the cases of known boundary conditions were determined directly from the isochromatics. Separate principal stresses were calculated using the combination of photoelastic information and isopachic data obtained by the electrical analogy method from the solution of Laplace's equation. Comparisons were made between experimental results and those computed using the finite element method. The stress distribution between theoretical and experimental agrees very well, although the finite element method yielded slightly higher stresses than the photoelastic method; further work is needed to resolve this difference. In this investigation several variations of coil geometry and methods of support were evaluated. Based on experimental results, optimum structural designs of toroidal field coils were recommended

Effect of toroidal field ripple on the formation of internal transport barriers

Energy Technology Data Exchange (ETDEWEB)

Vries, P C de; Hawkes, N C; Challis, C D; Andrew, Y; Beurskens, M; Brix, M; Giroud, C; Zastrow, K-D [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Joffrin, E [EFDA-JET CSU, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Litaudon, X [Association EURATOM-CEA, DSM/DFRC, CEA Cadarache, 13108, St Paul lez Durance (France); Brzozowski, J; Johnson, T [Association EURATOM-VR, Fusion Plasma Physics, EES, KTH, Stockholm (Sweden); Crombe, K [Department of Applied Physics, Ghent University, Ghent (Belgium); Hobirk, J [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, 85748 Garching (Germany); Loennroth, J; Salmi, A [Association Euratom-Tekes, Helsinki University of Technology, PO Box 4100, 02015 TKK (Finland); Tala, T [Association Euratom-Tekes, VTT, PO Box 1000, 02044 VTT (Finland); Yavorskij, V [Institute for Theoretical Physics, Association EURATOM-OEAW, University of Innsbruck (Austria)], E-mail: Peter.de.Vries@jet.uk

2008-06-15

The effect of a toroidal field (TF) ripple on the formation and performance of internal transport barriers (ITBs) has been studied in JET. It was found that the TF ripple had a profound effect on the toroidal plasma rotation. An increased TF ripple up to {delta} = 1% led to a lower rotation and reduced the rotational shear in the region where the ITBs were formed. ITB triggering events were observed in all cases and it is thought that the rotational shear may be less important for this process than, for example, the q-profile. However, the increase in the pressure gradient following the ITB trigger was reduced in discharges with a larger TF ripple and consequently a lower rotational shear. This suggests that toroidal rotation and its shear play a role in the growth of the ITB once it has been triggered.

Photoelastic and analytical investigation of stress in toroidal magnetic field coils

International Nuclear Information System (INIS)

Pih, H.; Gray, W.H.

1975-01-01

A series of two-dimensional photoelastic stress analyses on circular and oval toroidal magnetic field coils for fusion reactors were made. The circumferential variation of the coil's magnetic force was simulated by applying different pressures to sixteen segmented regions of the inner surface of the models. Isochromatics and isoclinics were measured at selected points on the loaded model in a transmission polariscope using a microphotometer. Separate principal stresses were obtained using the combination of photoelastic information and isopachic data measured from the solution of Laplace's equation by the electrical analog method. Analysis of the same coil geometries, loadings, and boundary conditions were made using the finite element method. General agreement between theory and experiment was realized. From this investigation several variations of coil geometry and methods of support were evaluated. Based upon this experiment, suggestions for optimum structural design of toroidal field coils are presented

Formation of a compact toroid for enhanced efficiency

Energy Technology Data Exchange (ETDEWEB)

Mozgovoy, A. G. [P.N. Lebedev Physical Institute, Moscow 119991 (Russian Federation); Romadanov, I. V.; Ryzhkov, S. V., E-mail: ryzhkov@power.bmstu.ru [Bauman Moscow State Technical University, Moscow 105005 (Russian Federation)

2014-02-15

We report here our results on the formation of a plasma configuration with the generic name of compact toroid (CT). A method of compact toroid formation to confine, heat and compress a plasma is investigated. Formation of a compact torus using an additional toroidal magnetic field helps to increase the plasma current to a maintainable level of the original magnetic field. We design the Compact Toroid Challenge (CTC) experiment in order to improve the magnetic flux trapping during field reversal in the formation of a compact toroid. The level of the magnetic field immersed in the plasma about 70% of the primary field is achieved. The CTC device and scheme of high level capturing of magnetic flux are presented.

Strength-limited magnetic field intensity of toroidal magnet systems fabricated or the base of layer-by-layer shrouded solenoids

International Nuclear Information System (INIS)

Litvinnko, Yu.A.

1982-01-01

The possibilities, as to the ultimate magnetic field strength, of tokamak magnet systems made on the base of layer-by-laeyer shrouded coils are considered numerically. The toroidal magnet system is considered which consists of N skewe, layer-by-layer shrouded, equistrong coils in the ideal torus approximation. The dependences of the ragnetic field strength on the internal- and external torus radii, pulse duration and aspect ratio for copper coils shrouded with fiberglass are calculated as an example. The analysis of the obtained results shows that using of the layer-by-layer shrouding scheme for toroidal solenoid coils leads to a considerable growth of the ultimate magnetic field strengths in a wide duration range. For example, the limiting field strength along the toroidal solenoid axis of the considered type inside the ''FT'' installation toroidal solenoid at equivalent field pulse duration of approximately 0.3 s reaches H 0 =1.3zx10 7 A/m

Stress analyses of ITER toroidal field coils under fault conditions

International Nuclear Information System (INIS)

Jong, C.T.J.

1990-02-01

The International Thermonuclear Experimental Reactor (ITER) is intended as an experimental thermonuclear tokamak reactor for testing the basic physics, performance and technologies essential to future fusion reactors. The ITER design will be based on extensive new design work, supported by new physical and technological results, and on the great body of experience built up over several years from previous national and international reactor studies. Conversely, the ITER design process should provide the fusion community with valuable insights into what key areas need further development or clarification as we move forward towards practical fusion power. As part of the design process of the ITER toroidal field coils the mechanical behaviour of the magnetic system under fault conditions has to be analysed in more detail. This paper describes the work carried out to create a detailed finite element model of two toroidal field coils as well as some results of linear elastic analyses with fault conditions. The analyses have been performed with the finite element code ANSYS. (author). 5 refs.; 8 figs.; 2 tabs

Calculation about a modification to the toroidal magnetic field of the Tokamak Novillo. Part I; Calculo sobre una modificacion al campo magnetico toroidal del Tokamak Novillo. Parte I

Energy Technology Data Exchange (ETDEWEB)

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

1991-07-15

The charged particles that constitute the plasma in the tokamaks are located in magnetic fields that determine its behavior. The poloidal magnetic field of the plasma current and the toroidal magnetic field of the tokamak possess relatively big gradients, which produce drifts on these particles. These drifts are largely the cause of the continuous lost of particles and of energy of the confinement region. In this work the results of numerical calculations of a modification to the 'traditional' toroidal magnetic field that one waits it diminishes the drifts by gradient and improve the confinement properties of the tokamaks. (Author)

LHCD current profile control experiments towards steady state improved confinement on JT-60U

International Nuclear Information System (INIS)

Ide, S.; Naito, O.; Oikawa, T.; Fujita, T.; Kondoh, T.; Seki, M.; Ushigusa, K.

2001-01-01

In JT-60U lower hybrid current drive (LHCD) experiments, a reversed magnetic shear configuration that was accompanied by the internal transport barriers was successfully maintained by means of LHCD almost in the full current drive quasi-steady state for 4.7 s. The normalized beta was kept near 1 and the neutron emission rate was almost steady as well indicating no accumulation of impurities into the plasma. Diagnostics data showed that all the profiles of the electron and ion temperatures, the electron density and the current profile were almost unchanged during the LHCD phase. Moreover, capability of LHCD in H-mode plasmas has been also investigated. It was found that the lower hybrid waves can be coupled to an H-mode edge plasma even with the plasma wall distance of about 14 cm. The maximum coupling distance was found to depend on the edge recycling. (author)

Poloidal and toroidal plasmons and fields of multilayer nanorings

OpenAIRE

Garapati, Kumar Vijay; Salhi, Marouane; Kouchekian, Sherwin; Siopsis, George; Passian, Ali

2017-01-01

Composite and janus type metallo-dielectric nanoparticles are increasingly considered as a means to control the spatial and temporal behavior of electromagnetic fields in diverse applications such as coupling to quantum emitters, achieve invisibility cloaks, and obtain quantum correlations between qubits. We investigate the surface modes of a toroidal nano-structure and obtain the canonical plasmon dispersion relations and resonance modes for arbitrarily layered nanorings. Unlike particle pla...

Reduction of toroidal magnetic field ripple in the advanced material tokamak experiment on JFT-2M

International Nuclear Information System (INIS)

Sato, M.; Miura, Y.; Kimura, H.; Yamamoto, M.; Koike, T.; Nakayama, T.; Hasegawa, M.; Urata, K.

1998-01-01

In order to reduce fast ion losses due to the toroidal field ripple, the reduction of ripple amplitude (δ) by inserting ferritic steel is studied, taking its toroidal mode number into account. The guideline of the design for reduction is wider and thicker ferritic board (FB) is located at further position from VV. The δ depends on the toroidal magnetic field. The value of B r21 /B t in the case of displacement of few cm is about 1 x 10 -5 which is one order smaller than the critical value. The offsetting of FB is not a problem for locked mode. Preliminary experiments with insertion of one or two FB's indicate no adverse effect on global plasma parameters. (author)

Reduction of toroidal magnetic field ripple in the advanced material tokamak experiment on JFT-2M

Energy Technology Data Exchange (ETDEWEB)

Sato, M.; Miura, Y.; Kimura, H.; Yamamoto, M.; Koike, T. [Japan Atomic Energy Research Inst. (Japan); Nakayama, T. [Hitachi Ltd. (Japan); Hasegawa, M. [Mitsubishi Electric Corp. (Japan); Urata, K. [Mitsubishi Heavy Industries Ltd. (Japan)

1998-07-01

In order to reduce fast ion losses due to the toroidal field ripple, the reduction of ripple amplitude ({delta}) by inserting ferritic steel is studied, taking its toroidal mode number into account. The guideline of the design for reduction is wider and thicker ferritic board (FB) is located at further position from VV. The {delta} depends on the toroidal magnetic field. The value of B{sub r21} /B{sub t} in the case of displacement of few cm is about 1 x 10{sup -5} which is one order smaller than the critical value. The offsetting of FB is not a problem for locked mode. Preliminary experiments with insertion of one or two FB's indicate no adverse effect on global plasma parameters. (author)

Toroidal Extrap Equilibria

International Nuclear Information System (INIS)

Scheffel, J.

1982-04-01

Ideal MHD-equilibria for the toroidal EXTRAP configuration have been computed with an equilibrium code. The free-boundary prob- lem is solved by using the condition that the current density is proportional to r on a flux surface. It is found that the toroidal Z-pinch, initially induced in the central zero-field region of a transverse octupole field, drifts radially outwards producing an inverse -D shaped cross-section. The plasma current of this high- beta equilibrium may be increased if the plasma is pushed back by altering the external confining magnetic field as demonstrated. (Author)

Toroidal inhomogeneity of the vertical field in a tokamak apparatus

International Nuclear Information System (INIS)

Sometani, Taro; Takashima, Hidekazu

1977-01-01

An experiment with a model device has been made on the toroidal inhomogeneity of the vertical field in a Tokamak with an iron core. The D.C. vertical field is increased near the yokes of the iron core, while the gross plasma image field (consisting of the components due to the plasma current, the primary current, and its image) is reduced there. These two vertical fields, when superposed, exert force on the plasma as a less inhomogeneous external vertical field. The vertical field can be homogenized satisfactorily by using a compensation winding wound at a proper position on the iron core even if the shielding plates, which are mounted on some Tokamaks, are dispensed with. (auth.)

Stability properties of a toroidal z-pinch in an external magnetic multipole field

International Nuclear Information System (INIS)

Eriksson, H.G.

1987-01-01

MHD stability of m=1, axisymmetric, external modes of a toroidal z-pinch immersed in an external multipole field (Extrap configuration) is studied. The description includes the effects of a weak toroidicity, a non-circular plasma cross-section and the influence of induced currents in the external conductors. It is found that the non-circularity of the plasma cross-section always has a destabilizing effect but that the m=1 mode can be stabilized by the external feedback if the non-circularity is small. (author)

Chaotic magnetic field line in toroidal plasmas

International Nuclear Information System (INIS)

Hatori, Tadatsugu; Abe, Yoshihiko; Urata, Kazuhiro; Irie, Haruyuki.

1989-05-01

This is an introductory review of chaotic magnetic field line in plasmas, together with some new results, with emphasis on the long-time tail and the fractional Brownian motion of the magnetic field line. The chaotic magnetic field line in toroidal plasmas is a typical chaotic phenomena in the Hamiltonian dynamical systems. The onset of stochasticity induced by a major magnetic perturbation is thought to cause a macroscopic rapid phenomena called the current disruption in the tokamak discharges. Numerical simulations on the basis of magnetohydrodynamics reveal in fact the disruptive phenomena. Some dynamical models which include the area-preserving mapping such as the standard mapping, and the two-wave Hamiltonian system can model the stochastic magnetic field. Theoretical results with use of the functional integral representation are given regarding the long-time tail on the basis of the radial twist mapping. It is shown that application of renormalization group technique to chaotic orbit in the two-wave Hamiltonian system proves decay of the velocity autocorrelation function with the power law. Some new numerical results are presented which supports these theoretical results. (author)

Field-reversed experiments (FRX) on compact toroids

Energy Technology Data Exchange (ETDEWEB)

Armstrong, W.T.; Linford, R.K.; Lipson, J.; Platts, D.A.; Sherwood, E.G.

1981-11-01

Equilibrium, stability, and confinement properties of compact toroids produced in field-reversed theta-pinch experiments (FRX) are reported. Two experimental facilities, FRX-A and FRX-B, have been used to study highly elongated compact toroid plasmas confined in a purely poloidal field geometry. Spatial scans and fill pressure scaling of the equilibrium plasma parameters are presented. Plasma conditions range from T/sub e/approx.150 eV, T/sub i/approx.800 eV, n/sub m/approx.1 x 10/sup 15/ cm/sup -3/ to T/sub e/approx.100 eV, T/sub i/approx.150 eV, n/sub m/approx.4 x 10/sup 15/ cm/sup -3/. Typical confined plasma dimensions are: major radius Rapprox.4 cm, minor radius aapprox.2 cm, and total length 35--50 cm. The plasma configuration remains in a stable equilibrium for up to 50 ..mu..sec followed by the destructive n = 2 rotational instability. The stable period prior to the onset of the rotational mode is up to one hundred times greater than characteristic Alfven transit times of the plasma. This stable period increases and the mode growth rate decreases with increased a/rho/sub i/ (where rho/sub i/ is the ion gyroradius). Agreement of experimental and theoretical mode frequencies for the instability is observed. Preferential particle loss has been proposed as a likely cause of rotation. The particle inventory at the onset of the instability is consistent with this hypothesis. The particle loss rate is also consistent with the predicted anomalous transport near the separatrix. Contributions to rotational instability from diffusion, end-shorting, equipartition, and compression are also discussed.

Field-reversed experiments (FRX) on compact toroids

International Nuclear Information System (INIS)

Armstrong, W.T.; Linford, R.K.; Lipson, J.; Platts, D.A.; Sherwood, E.G.

1981-01-01

Equilibrium, stability, and confinement properties of compact toroids produced in field-reversed theta-pinch experiments (FRX) are reported. Two experimental facilities, FRX-A and FRX-B, have been used to study highly elongated compact toroid plasmas confined in a purely poloidal field geometry. Spatial scans and fill pressure scaling of the equilibrium plasma parameters are presented. Plasma conditions range from T/sub e/approx.150 eV, T/sub i/approx.800 eV, n/sub m/approx.1 x 10 15 cm -3 to T/sub e/approx.100 eV, T/sub i/approx.150 eV, n/sub m/approx.4 x 10 15 cm -3 . Typical confined plasma dimensions are: major radius Rapprox.4 cm, minor radius aapprox.2 cm, and total length 35--50 cm. The plasma configuration remains in a stable equilibrium for up to 50 μsec followed by the destructive n = 2 rotational instability. The stable period prior to the onset of the rotational mode is up to one hundred times greater than characteristic Alfven transit times of the plasma. This stable period increases and the mode growth rate decreases with increased a/rho/sub i/ (where rho/sub i/ is the ion gyroradius). Agreement of experimental and theoretical mode frequencies for the instability is observed. Preferential particle loss has been proposed as a likely cause of rotation. The particle inventory at the onset of the instability is consistent with this hypothesis. The particle loss rate is also consistent with the predicted anomalous transport near the separatrix. Contributions to rotational instability from diffusion, end-shorting, equipartition, and compression are also discussed

Viscous damping of toroidal angular momentum in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Stacey, W. M. [Georgia Tech Fusion Research Center, Atlanta, Georgia 30332 (United States)

2014-09-15

The Braginskii viscous stress tensor formalism was generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry in order to provide a representation for the viscous damping of toroidal rotation in tokamaks arising from various “neoclassical toroidal viscosity” mechanisms. In the process, it was verified that the parallel viscosity contribution to damping toroidal angular momentum still vanishes even in the presence of toroidal asymmetries, unless there are 3D radial magnetic fields.

New lens system using toroidal magnetic field for intense ion beam

International Nuclear Information System (INIS)

Mohri, Akihiro; Ikuta, Kazunari; Fujita, Junji.

1976-11-01

The use of toroidal magnetic field as a lens system is proposed for producing intense ion beam. The characteristics of the lens system are obtained both analytically and numerically. Some examples of ray-trajectories are presented for different focal lengths. The system is applicable to neutral beam injection heating and micro-pellet implosion for nuclear fusion, and to the other fields such as ion beam X-ray lasers. (auth.)

Review of preliminary additional heating experiments in JT-60 (Aug. - Nov., 1986)

International Nuclear Information System (INIS)

1987-03-01

This is a prompt report on preliminary additional heating experiments in JT-60 from August to November in 1986. Neutral beam heating power was raised up to 20 MW in about a month. Plasma stored energy is about 2 MJ and energy confinement time is 0.1 ∼ 0.12 sec with the maximum heating power. The energy confinement time shows L-mode like deterioration with power, while it has little dependence on electron density. The maximum ion temperature of ∼ 7 keV and electron temperature of 4.5 keV were obtained at relatively low electron density (n-bar e = 2 - 3 x 10 19 m -3 ). Lower hybrid wave could efficiently drive plasma current up to 1.7 MA with 1.2 MW LH power. The current drive efficiency is 1 ∼ 1.7 in ohmically heated plasmas and 2 ∼ 2.8 in NB heated plasmas. Futhermore the energy confinement was improved when neutral beam was injected into entirely current driven discharges of 1 MA by LH in contrast to inductively driven target plasmas. Similar improvement in energy confinement was observed during combined heating with NB and ion cyclotron wave. (author)

Dynamic transport study of the plasmas with transport improvement in LHD and JT-60U

International Nuclear Information System (INIS)

Ida, K.; Inagaki, S.; Sakamoto, R.; Tanaka, K.; Fujita, T.; Funaba, H.; Kubo, S.; Yoshinuma, M.; Shimozuma, T.; Takeiri, Y.; Ikeda, K.; Michael, C.; Tokuzawa, T.; Sakamoto, Y.; Takenaga, H.; Isayama, A.; Matsunaga, G.; Ide, S.

2009-01-01

Transport analysis during the transient phase of heating (a dynamic transport study) applied to the plasma with internal transport barriers (ITBs) in the Large Helical Device (LHD) heliotron and the JT-60U tokamak is described. In the dynamic transport study the time of transition from the L-mode plasma to the ITB plasma is clearly determined by the onset of flattening of the temperature profile in the core region and a spontaneous phase transition from a zero curvature ITB (hyperbolic tangent shaped ITB) or a positive curvature ITB (concaved shaped ITB) to a negative curvature ITB (convex shaped ITB) and its back-transition are observed. The flattening of the core region of the ITB transition and the back-transition between a zero curvature ITB and a convex ITB suggest the strong interaction of turbulent transport in space.

Reduced transport and ER shearing in improved confinement regimes in JT-60U

International Nuclear Information System (INIS)

Shirai, H.; Kikuchi, M.; Takizuka, T.

2001-01-01

The global confinement and the local transport properties of improved core confinement plasmas in JT-60U have been studied in connection with E r shear formation. The improved core confinement mode with ITB, the internal transport barrier, is roughly classified into 'parabolic' type ITBs and 'box' type ITBs. The parabolic type ITB has the reduced thermal diffusivity, χ, in the core region; however, the E r shear, dE r /dr, is not so strong. The box type ITB has a very strong E r shear at the thin ITB layer and the χ value decreases to the level of neoclassical transport there. The estimated ExB shearing rate, ω ExB , becomes almost the same as the linear growth rate of the drift microinstability, γ L , at the ITB layer in the box type ITB. Experiments of hot ion mode plasmas during the repetitive L-H-L transition shows that the thermal diffusivity clearly depends on the E r shear and the strong E r shear contributes to the reduced thermal diffusivity. (author)

Development of the computer system for the JT-60 negative-ion based NBI

International Nuclear Information System (INIS)

Kawai, Mikito; Oohara, Hiroshi; Honda, Atsushi; Kuriyama, Masaaki; Aoyagi, Tetsuo.

1997-03-01

The negative-ion based NBI system (N-NBI) for JT-60 is the first NBI system using a negative-ion source in the world. The N-NBI is designed do deliver a neutral beam injection power of 10 MW at 500 keV. The computer for the N-NBI system is composed of UNIX workstations and VMEbus systems, and has the functions of ion source operation and data acquisition and processing. Since a real-time operating system compatible with the UNIX is adopted for the VMEbus systems, the software development environment both for the workstation and the VMEbus system is unified with the UNIX. The software has been developed with a priority to the software required for the verification tests which are performed in accordance with the progress of the N-NBI construction. The first beam injection with the N-NBI has been conducted in March using the newly developed software, and the deuterium neutral beam injection of 350 keV, 2.5 MW has achieved as of the end of October 1996. (author)

Reduced transport and Er shearing in improved confinement regimes in JT-60U

International Nuclear Information System (INIS)

Shirai, H.; Kikuchi, M.; Takizuka, T.

1999-01-01

The global confinement and the local transport properties of improved core confinement plasmas in JT-60U were studied in connection with E r shear formation. In the improved core confinement mode with internal transport barriers (ITBs), these are roughly classified into 'parabolic type' ITBs and 'box type' ITBs. The parabolic type ITB has a reduced thermal diffusivity χ in the core region; however, the E r shear, dE r /dr, is not as strong. The box type ITB has a very strong E r shear at the thin ITB layer and χ decreases to the level of neoclassical transport there. The estimated E x B shearing rate, ω ExB , becomes almost the same as the linear growth rate of the drift microinstability, γ L , at the ITB layer in the box type ITB. Experiments with hot ion mode plasmas during the repetitive L-H-L transition showed that the thermal diffusivity clearly depends on the E r shear and the strong E r shear contributes to the reduced thermal diffusivity. (author)

Hydrogen isotope distributions and retentions in the inner divertor tile of JT-60U

Energy Technology Data Exchange (ETDEWEB)

Oya, Y. [Radioisotope Center, University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-0032 (Japan)]. E-mail: yoya@ric.u-tokyo.ac.jp; Hirohata, Y. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Tanabe, T. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Shibahara, T. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Kimura, H. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Oyaidzu, M. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Arai, T. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Masaki, K. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Gotoh, Y. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Okuno, K. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Miya, N. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Hino, T. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Tanaka, S. [Graduate School of Engineering, University of Tokyo, Tokyo 113-8656 (Japan)

2005-11-15

Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature.

Hydrogen isotope distributions and retentions in the inner divertor tile of JT-60U

International Nuclear Information System (INIS)

Oya, Y.; Hirohata, Y.; Tanabe, T.; Shibahara, T.; Kimura, H.; Oyaidzu, M.; Arai, T.; Masaki, K.; Gotoh, Y.; Okuno, K.; Miya, N.; Hino, T.; Tanaka, S.

2005-01-01

Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature

Protection of toroidal field coils using multiple circuits

International Nuclear Information System (INIS)

Thome, R.J.; Langton, W.G.; Mann, W.R.; Pillsbury, R.D.; Tarrh, J.M.

1983-01-01

The protection of toroidal field (TF) coils using multiple circuits is described. The discharge of a single-circuit TF system is given for purposes of definition. Two-circuit TF systems are analyzed and the results presented analytically and graphically. Induced currents, maximum discharge voltages, and discharge time constants are compared to the single-circuit system. Three-circuit TF systems are analyzed. In addition to induced currents, maximum discharge voltages, and time constants, several different discharge scenarios are included. The impacts of having discharge rates versus final maximum coil temperatures as requirements are examined. The out-of-plane forces which occur in the three-circuit system are analyzed using an approximate model. The analysis of multiplecircuit TF systems is briefly described and results for a Toroidal Fusion Core Experiment (TFCX) scale device are given based on computer analysis. The advantages and disadvantages of using multiple-circuit systems are summarized and discussed. The primary disadvantages of multiple circuits are the increased circuit complexity and potential for out-of-plane forces. These are offset by the substantial reduction in maximum discharge voltages, as well as other design options which become available when using multiple circuits

Dynamic divertor control using resonant mixed toroidal harmonic magnetic fields during ELM suppression in DIII-D

Science.gov (United States)

Jia, M.; Sun, Y.; Paz-Soldan, C.; Nazikian, R.; Gu, S.; Liu, Y. Q.; Abrams, T.; Bykov, I.; Cui, L.; Evans, T.; Garofalo, A.; Guo, W.; Gong, X.; Lasnier, C.; Logan, N. C.; Makowski, M.; Orlov, D.; Wang, H. H.

2018-05-01

Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness of the RMP ELM suppression technique for the simultaneous control of divertor heat and particle load required in ITER.

Deuterium depth profiling in JT-60U W-shaped divertor tiles by nuclear reaction analysis

International Nuclear Information System (INIS)

Hayashi, T.; Ochiai, K.; Masaki, K.; Gotoh, Y.; Kutsukake, C.; Arai, T.; Nishitani, T.; Miya, N.

2006-01-01

Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JAERI Tokamak-60 Upgrade (JT-60U) were investigated by nuclear reaction analysis (NRA). The highest deuterium concentration of D/ 12 C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium-carbon co-deposition. In the outer and inner divertor target tiles, the D/ 12 C data were lower than 0.006. Additionally, the maximum (H + D)/ 12 C in the dome top tile was estimated to be 0.023 from the results of NRA and secondary ion mass spectroscopy (SIMS). Orbit following Monte-Carlo (OFMC) simulation showed energetic deuterons caused by neutral beam injections (NBI) were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, scanning electron microscope (SEM) observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium-carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk

Active control of internal transport barrier and confinement database in JT-60U reversed shear plasma

Energy Technology Data Exchange (ETDEWEB)

Sakamoto, Yoshiteru; Takizuka, Tomonori; Shirai, Hiroshi; Fujita, Takaaki; Kamada, Yutaka; Ide, Shunsuke; Fukuda, Takeshi; Koide, Yoshihiko [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

2001-07-01

Active control of internal transport barrier (ITB) and confinement properties of plasma with ITB have been studied in reversed shear plasmas. Modifications of the radial electric field (E{sub r}) profile by changing the combination of tangential neutral beams can control the ITB strength, where the contribution to E{sub r} from the toroidal rotation plays an important role. The ITB confinement database of reversed shear plasmas has been constructed. Stored energy is strongly correlated with poloidal magnetic field at the ITB foot. (author)

Core density fluctuations in reverse magnetic shear plasmas with internal transport barrier on JT-60U

International Nuclear Information System (INIS)

Nazikian, R.; Shinohara, K.; Yoshino, R.; Fujita, T.; Shirai, H.; Kramer, G.T.

1999-01-01

First measurements of the radial correlation length of density fluctuations in JT-60U plasmas with internal transport barrier (ITB) is reported. The measurements are obtained using a newly installed correlation reflectometer operating in the upper X-mode. Before transport barrier formation in the low beam power current ramp-up phase of the discharge, reflectometer measurements indicate density fluctuation levels n-tilde/n∼0.1-0.2% and radial correlation lengths 2-3 cm (k r p i ≤0.5) in the central plasma region (r/a r p i ∼3. However, fluctuation levels are considerably higher than measured near the magnetic axis. Reflectometer measurements obtained at the foot of the ITB also indicate high fluctuation levels compared to measurements in the central region of the discharge. (author)

The role of fluctuation-induced transport in a toroidal plasma with strong radial electric fields

Science.gov (United States)

Roth, J. R.; Krawczonek, W. M.; Powers, E. J.; Hong, J. Y.; Kim, Y. C.

1981-01-01

Previous work employing digitally implemented spectral analysis techniques is extended to demonstrate that radial fluctuation-induced transport is the dominant ion transport mechanism in an electric field dominated toroidal plasma. Such transport can be made to occur against a density gradient, and hence may have a very beneficial effect on confinement in toroidal plasmas of fusion interest. It is shown that Bohm or classical diffusion down a density gradient, the collisional Pedersen-current mechanism, and the collisionless electric field gradient mechanism described by Cole (1976) all played a minor role, if any, in the radial transport of this plasma.

Toroidal visco-resistive magnetohydrodynamic steady states contain vortices

International Nuclear Information System (INIS)

Bates, J.W.; Montgomery, D.C.

1998-01-01

Poloidal velocity fields seem to be a fundamental feature of resistive toroidal magnetohydrodynamic (MHD) steady states. They are a consequence of force balance in toroidal geometry, do not require any kind of instability, and disappear in the open-quotes straight cylinderclose quotes (infinite aspect ratio) limit. If a current density j results from an axisymmetric toroidal electric field that is irrotational inside a torus, it leads to a magnetic field B such that ∇x(jxB) is nonvanishing, so that the Lorentz force cannot be balanced by the gradient of any scalar pressure in the equation of motion. In a steady state, finite poloidal velocity fields and toroidal vorticity must exist. Their calculation is difficult, but explicit solutions can be found in the limit of low Reynolds number. Here, existing calculations are generalized to the more realistic case of no-slip boundary conditions on the velocity field and a circular toroidal cross section. The results of this paper strongly suggest that discussions of confined steady states in toroidal MHD must include flows from the outset. copyright 1998 American Institute of Physics

Recent ion cyclotron range of frequencies experiments in JT-60U

International Nuclear Information System (INIS)

Kimura, H.; Fujii, T.; Saigusa, M.; Moriyama, S.; Sato, M.; Nemoto, M.; Kondoh, T.; Hamamatsu, K.

1995-01-01

Recent results on the minority ion second harmonic heating on JT-60U are presented. Maximum coupled power reached 6.4MW. Good antenna-plasma coupling capability and a small fraction (less than 10%) of an incremental radiation loss to r.f. power are confirmed. Power absorption rate increases with increasing r.f. power and is saturated around unity at r.f. powers higher than 3MW. The sawtooth stabilization by minority ion second harmonic heating was realized over a wide parameter range, i.e. I P =0.9MA-4MA, q 95 =2.3-8.6, n e =(1.3-5)x10 19 m -3 and P IC ≥2.2MW. A figure of merit V P left angle n e right angle /P tot for efficiency of the sawtooth stabilization is about 50% higher than those in other devices where fundamental resonance minority ion heating is employed. The longest stable period reached 2.33s. Attainable sawtooth-free periods scale with the resistive diffusion time. It was found that the energy confinement is further improved by 25% during the reheating phase after the giant sawtooth crash. The electron temperature profile became more peaked at the improved confinement phase. Those phenomena were observed only in low q discharges (q 95 ≤2.9). ((orig.))

Research and development of the JAERI large tokamak (JT-60), (4)

International Nuclear Information System (INIS)

Takashima, Tetsuo; Shimizu, Masatsugu; Ohta, Mitsuru; Minaguchi, Tadayoshi; Maeda, Hideto.

1978-01-01

A pair of fast-acting movable rail limiters are to be installed in the vacuum chamber of JT-60 to suppress skin current in the plasma column. They should travel across the vacuum chamber over a stroke of about 1 m in 0.1 sec in the build-up phase of the plasma current. Each movable limiter system consists of a drive mechanism, a vacuum seal, a bearing usable at high temperatures in a vacuum, a molybdenum rail limiter head and its auxiliary members. Various engineering problems are involved in constructing such a system because the design specifications outlined above exceed the present technology. A full-scale movable limiter, therefore, was designed, constructed and then put to mechanical, electrical and vacuum-technological tests. The model features a hydraulic drive mechanism with servovalves to control the oil flow. A special vacuum seal allowing a movement at high speeds was developed. It consists of welded bellows jointed together and connected to a pantograph at the joints. It allows uniform expansion of each bellows at high speeds. Molybdenum disulphide with 20% Ta is chosen as the most suitable bearing material after conducting tests on various bearing materials. The overall test of the model showed that its specifications were met with satisfactory reliability and reproducibility. Furthermore, the endurance test demonstrated that it functioned reliably over 50,000 times of operation. (author)

The ASDEX upgrade toroidal field magnet and poloidal divertor field coil system adapted to reactor requirements

International Nuclear Information System (INIS)

Koeppendoerfer, W.; Blaumoser, M.; Ennen, K.; Gruber, J.; Gruber, O.; Jandl, O.; Kaufmann, M.; Kollotzek, H.; Kotzlowski, H.; Lackner, E.; Lackner, K.; Larcher, T. von; Noterdaeme, J.M.; Pillsticker, M.; Poehlchen, R.; Preis, H.; Schneider, H.; Seidel, U.; Sombach, B.; Speth, E.; Streibl, B.; Vernickel, H.; Werner, F.; Wesner, F.; Wieczorek, A.

1986-01-01

ASDEX Upgrade is a tokamak experiment with external poloidal field coils that is now under construction at IPP Garching. It can produce elongated single-null (SN), double-null (DN) , and limiter (L) configurations. The SN is the reference configuration with asymmetric load distributions in the poloidal field (PF) system and the toroidal field (TF) magnet. Plasma control and stabilization require a rigid passive conductor close to the plasma. The design principles of the coils and support structure are described. (orig.)

Current sustaining by RF travelling field in a collisional toroidal plasma

International Nuclear Information System (INIS)

Fukuda, Masaji; Matsuura, Kiyokata

1978-01-01

The relation between the current generated by RF travelling field and the absorbed power is studied in a collisional toroidal plasma, parameters being phase velocity and filling gap pressure or electron collision frequency. It is observed at a low magnetic field that the current is proportional to the plasma conductivity and an effective electromotive force, which is a new concept introduced on the basis of fluid model; the electromotive force is proportional to the absorbed RF power and inversely proportional to the plasma density and the phase velocity of the travelling field. (author)

Current sustaining by RF travelling field in a collisional toroidal plasma

International Nuclear Information System (INIS)

Fukuda, Masaji; Matsuura, Kiyokata.

1977-06-01

The relation between the current generation by RF travelling field and the accompanied power absorption is studied in a collisional toroidal plasma, parameters being phase velocity and filling gas pressure or electron collision frequency. It is observed at a low magnetic field that the current is proportional to the plasma conductivity and an effective electromotive force, which is a new concept introduced on the basis of fluid model; the electromotive force is proportional to the absorbed RF power and inversely proportional to the plasma density and the phase velocity of the travelling field. (auth.)

A model for the neoclassical toroidal viscosity effect on Edge plasma toroidal rotation

Energy Technology Data Exchange (ETDEWEB)

Miron, I.G. [National Institute for Laser, Plasma and Radiation Physics, Euratom-MEdC Association, Bucharest (Romania)

2013-11-15

A semianalytic expression for the edge plasma angular toroidal rotation frequency that includes the neoclassical toroidal viscosity braking influence is obtained. Based on the model presented in a previous paper [I.G. Miron, Contrib. Plasma Phys. 53, 214 (2013)], the less destabilizing error field spectrum is found in order to minimize the nonlinear effect of the NTV on the toroidal rotation of the edge of the plasma. (copyright 2013 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Spatial Variations of Poloidal and Toroidal Mode Field Line Resonances Observed by MMS

Science.gov (United States)

Le, G.; Chi, P. J.; Strangeway, R. J.; Russell, C. T.; Slavin, J. A.; Anderson, B. J.; Kepko, L.; Nakamura, R.; Plaschke, F.; Torbert, R. B.

2017-12-01

Field line resonances (FLRs) are magnetosphere's responses to solar wind forcing and internal instabilities generated by solar wind-magnetospheric interactions. They are standing waves along the Earth's magnetic field lines oscillating in either poloidal or toroidal modes. The two types of waves have their unique frequency characteristics. The eigenfrequency of FLRs is determined by the length of the field line and the plasma density, and thus gradually changes with L. For toroidal mode oscillations with magnetic field perturbations in the azimuthal direction, ideal MHD predicts that each field line oscillates independently with its own eigenfrequency. For poloidal mode waves with field lines oscillating radially, their frequency cannot change with L easily as L shells need to oscillate in sync to avoid efficient damping due to phase mixing. Observations, mainly during quiet times, indeed show that poloidal mode waves often exhibit nearly constant frequency across L shells. Our recent observations, on the other hand, reveal a clear L-dependent frequency trend for a long lasting storm-time poloidal wave event, indicating the wave can maintain its power with changing frequencies for an extended period [Le et al., 2017]. The spatial variation of the frequency shows discrete spatial structures. The frequency remains constant within each discrete structure that spans about 1 REalong L, and changes discretely. We present a follow-up study to investigate spatial variations of wave frequencies using the Wigner-Ville distribution. We examine both poloidal and toroidal waves under different geomagnetic conditions using multipoint observations from MMS, and compare their frequency and occurrence characteristics for insights into their generation mechanisms. Reference: Le, G., et al. (2017), Global observations of magnetospheric high-m poloidal waves during the 22 June 2015 magnetic storm, Geophys. Res. Lett., 44, 3456-3464, doi:10.1002/2017GL073048.

Review of JT-60U experimental results in 1998

International Nuclear Information System (INIS)

Adachi, H.; Akasaka, H.; Akino, N.

1999-09-01

Based on the high confinement regimes such as reversed shear mode, high-β p mode and H-mode, the JT-60U experiment in 1998 was devoted to expand the discharge regimes in terms of 1) achievement of high fusion gain, 2) concept optimization for long sustainment of the advanced modes for >>τ E and >τ p * with the current profile close to the steady-state solution, 3) high confinement by electron heating (T e >T i ), 4) high confinement at high electron density and/or at high divertor radiation and 5) active control of multiple parameters including both core and divertor plasmas. As for the reversed shear mode, high performance discharges satisfying Q DT eq (DT equivalent fusion gain ) >1 were obtained reproducibly and the record value of Q DT eq =1.25 was achieved in a reactor-relevant thermonuclear dominant regime due to the optimized discharge scenario using feedback control of the neutron production rate where β-values were kept in the MHD stable region during the I p ramping phase with a large radius of the internal transport barrier (ITB). The reduction of Z eff obtained after installation of W-shaped pumped divertor increased fusion reaction rate. Concerning long sustainment, the reversed shear ELMy H-mode with H 89PL -factor∼1.5-2 and β N =1.0-1.4 was kept for 5.5 s with NB heating. By off axis LH current drive, the reversed shear current profile with the ITB was kept constant for 4.7 s under full non- inductive current drive condition (LHCD=77%, bootstrap=23%) at T e -1.2T i . In the high-β p ELMy H-mode regime, benefits of the high triangularity shape were demonstrated. At a high triangularity δ X ∼0.46, β N =2.5-2.7 was sustained for 3.5 s even at the low value of q 95 =2.9-3.3. The product of β N xH-factor sustainable for >5τ E (>τ p *) increases with δ x and reaches ∼6 at δ X ∼0.46. In addition to extension of the discharge regimes, the key phenomena determining transport and stability around ITBs were studied intensively. For the

Theoretical study on the laser-driven ion-beam trace probe in toroidal devices with large poloidal magnetic field

Science.gov (United States)

Yang, X.; Xiao, C.; Chen, Y.; Xu, T.; Yu, Y.; Xu, M.; Wang, L.; Wang, X.; Lin, C.

2018-03-01

Recently, a new diagnostic method, Laser-driven Ion-beam Trace Probe (LITP), has been proposed to reconstruct 2D profiles of the poloidal magnetic field (Bp) and radial electric field (Er) in the tokamak devices. A linear assumption and test particle model were used in those reconstructions. In some toroidal devices such as the spherical tokamak and the Reversal Field Pinch (RFP), Bp is not small enough to meet the linear assumption. In those cases, the error of reconstruction increases quickly when Bp is larger than 10% of the toroidal magnetic field (Bt), and the previous test particle model may cause large error in the tomography process. Here a nonlinear reconstruction method is proposed for those cases. Preliminary numerical results show that LITP could be applied not only in tokamak devices, but also in other toroidal devices, such as the spherical tokamak, RFP, etc.

Physics issues of high bootstrap current tokamaks

International Nuclear Information System (INIS)

Ozeki, T.; Azumi, M.; Ishii, Y.

1997-01-01

Physics issues of a tokamak plasma with a hollow current profile produced by a large bootstrap current are discussed based on experiments in JT-60U. An internal transport barrier for both ions and electrons was obtained just inside the radius of zero magnetic shear in JT-60U. Analysis of the toroidal ITG microinstability by toroidal particle simulation shows that weak and negative shear reduces the toroidal coupling and suppresses the ITG mode. A hard beta limit was observed in JT-60U negative shear experiments. Ideal MHD mode analysis shows that the n = 1 pressure-driven kink mode is a plausible candidate. One of the methods to improve the beta limit against the kink mode is to widen the negative shear region, which can induce a broader pressure profile resulting in a higher beta limit. The TAE mode for the hollow current profile is less unstable than that for the monotonic current profile. The reason is that the continuum gaps near the zero shear region are not aligned when the radius of q min is close to the region of high ∇n e . Finally, a method for stable start-up for a plasma with a hollow current profile is describe, and stable sustainment of a steady-state plasma with high bootstrap current is discussed. (Author)

OCLATOR (One Coil Low Aspect Toroidal Reactor)

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, S.

1980-02-01

A new approach to construct a tokamak-type reactor(s) is presented. Basically the return conductors of toroidal field coils are eliminated and the toroidal field coil is replaced by one single large coil, around which there will be placed several tokamaks or other toroidal devices. The elimination of return conductors should, in addition to other advantages, improve the accessibility and maintainability of the tokamaks and offer a possible alternative to the search for special materials to withstand large neutron wall loading, as the frequency of changeover would be increased due to minimum downtime. It also makes it possible to have a low aspect ratio tokamak which should improve the ..beta.. limit, so that a low toroidal magnetic field strength might be acceptable, meaning that the NbTi superconducting wire could be used. This system is named OCLATOR (One Coil Low Aspect Toroidal Reactor).

Power system for tokamak fusion experiments. Motor generator with flywheel effect

International Nuclear Information System (INIS)

Miyachi, Kengo

1997-01-01

JT-60 requires an enormous electric power pulse about 1,300 MVA periodically for its plasma initiation, containment and heating. JT-60 could not receive all electric power from a commercial line for plasma experiment except about 160 MVA because the 275 kV commercial line has some limitations. Therefore JT-60 needs huge electric power sources. The power supply system of JT-60 has 3 motor generators (MG). The total capacity of MG is 1,115 MVA that consists of a toroidal MG (TMG), poloidal MG (PMG) and Heating power supply MG (HMG), and each MG has a huge flywheel effect. For example, TMG has a 4.02 GJ energy yield that consists of 6 disk flywheel. The total weight of flywheel of TMG is 650 ton. This report describes the structure, operating system, and maintenance history of 3 types of MG. (author)

Thermal properties of redeposition layers in the JT-60U divertor region

International Nuclear Information System (INIS)

Ishimoto, Y.; Gotoh, Y.; Arai, T.; Masaki, K.; Miya, N.; Oyama, N.; Asakura, N.

2006-01-01

Thermal properties of the redeposition layer on the inner plate of the W-shaped divertor of JT-60U have been measured with laser flash method so as to estimate transient heat loads onto the divertor. Morphology analysis of the redeposition layer was conducted with a scanning electron microscope. Measurement of a redeposition layer sample of more than 200 μm thick, which had been produced near the most frequent striking point, showed following results: (1) the bulk density of the redeposition layer is about half of that of carbon fiber composite material; (2) the specific heat of the layer is roughly equal to that of the isotropic graphite; (3) the thermal conductivity of the redeposition layer is two orders of magnitude smaller than that of the carbon fiber composite. This low thermal conductivity of the redeposition layer is considered to be caused by a low graphitization degree of the redeposition layer. The difference between the divertor heat loads and the loss of the plasma stored energy becomes smaller taking account of thermal properties of the redeposition layer on the inner divertor, whereas estimated heat loads due to the ELMs is still larger than the loss. This is probably caused by the poloidal distribution of the thermal properties

Physical mechanism determining the radial electric field and its radial structure in a toroidal plasma

International Nuclear Information System (INIS)

Ida, Katsumi; Miura, Yukitoshi; Itoh, Sanae

1994-10-01

Radial structures of plasma rotation and radial electric field are experimentally studied in tokamak, heliotron/torsatron and stellarator devices. The perpendicular and parallel viscosities are measured. The parallel viscosity, which is dominant in determining the toroidal velocity in heliotron/torsatron and stellarator devices, is found to be neoclassical. On the other hand, the perpendicular viscosity, which is dominant in dictating the toroidal rotation in tokamaks, is anomalous. Even without external momentum input, both a plasma rotation and a radial electric field exist in tokamaks and heliotrons/torsatrons. The observed profiles of the radial electric field do not agree with the theoretical prediction based on neoclassical transport. This is mainly due to the existence of anomalous perpendicular viscosity. The shear of the radial electric field improves particle and heat transport both in bulk and edge plasma regimes of tokamaks. (author) 95 refs