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

Predictive Simulations of ITER Including Neutral Beam Driven Toroidal Rotation

International Nuclear Information System (INIS)

Halpern, Federico D.; Kritz, Arnold H.; Bateman, G.; Pankin, Alexei Y.; Budny, Robert V.; McCune, Douglas C.

2008-01-01

Predictive simulations of ITER [R. Aymar et al., Plasma Phys. Control. Fusion 44, 519 2002] discharges are carried out for the 15 MA high confinement mode (H-mode) scenario using PTRANSP, the predictive version of the TRANSP code. The thermal and toroidal momentum transport equations are evolved using turbulent and neoclassical transport models. A predictive model is used to compute the temperature and width of the H-mode pedestal. The ITER simulations are carried out for neutral beam injection (NBI) heated plasmas, for ion cyclotron resonant frequency (ICRF) heated plasmas, and for plasmas heated with a mix of NBI and ICRF. It is shown that neutral beam injection drives toroidal rotation that improves the confinement and fusion power production in ITER. The scaling of fusion power with respect to the input power and to the pedestal temperature is studied. It is observed that, in simulations carried out using the momentum transport diffusivity computed using the GLF23 model [R.Waltz et al., Phys. Plasmas 4, 2482 (1997)], the fusion power increases with increasing injected beam power and central rotation frequency. It is found that the ITER target fusion power of 500 MW is produced with 20 MW of NBI power when the pedesta temperature is 3.5 keV. 2008 American Institute of Physics. [DOI: 10.1063/1.2931037

ECRH-assisted plasma start-up with toroidally inclined launch: multi-machine comparison and perspectives for ITER

International Nuclear Information System (INIS)

Stober, J.; Hobirk, J.; Lunt, T.; Jackson, G.L.; Hyatt, A.W.; Luce, T.; Ascasibar, E.; Cappa, A.; Bae, Y.-S.; Joung, M.; Bucalossi, J.; Casper, T.; Gribov, Y.; Cho, M.-H.; Jeong, J.-H.; Namkung, W.; Park, S.-I.; Granucci, G.; Hanada, K.; Ide, S.

2011-01-01

Electron cyclotron resonance heating (ECRH)-assisted plasma breakdown is foreseen with full and half magnetic field in ITER. As reported earlier, the corresponding O1- and X2-schemes have been successfully used to assist pre-ionization and breakdown in present-day devices. This contribution reports on common experiments studying the effect of toroidal inclination of the ECR beam, which is ≥20 0 in ITER. All devices could demonstrate successful breakdown assistance for this case also, although in some experiments the necessary power was almost a factor of 2 higher compared with perpendicular launch. Differences between the devices with regard to the required power and vertical field are discussed and analysed. In contrast to most of these experiments, ITER will build up loop voltage prior to the formation of the field null due to the strong shielding by the vessel. Possible consequences of this difference are discussed.

Project status of manufacturing of European toroidal coils ITER. Validation tests; Estado del proyecto de fabricacion de las bobinas toroidales european para el ITER. Ensayos de validacion

Energy Technology Data Exchange (ETDEWEB)

Pando, F.; Felipe, A.; Madorran, A.; Pallisa, J.; Dormicch, O.; Valle, N.; D' Urzo, C.; Marin, M.; Pesenti, P.; Lucas, J.; Moreno, N.; Bonito-Oliva, A.; Harrison, R.; Bellesia, B.; Cornelis, M.; Cornella, J.

2015-07-01

The toroidal field coils are the ITER magnets responsible for confining the plasma inside the vacuum vessel. The consortium formed by IBERDROLA Ingenieria y Construccion, ASG Superconductors y ELYTT Energy is the responsible for the supply of 10 coils that the european agency F4E has to supply for the ITER project. At present, the coils are been manufactured in La Spezia (Italy), after the qualification of all the manufacturing process and the sucessfull manufacturing of a full scale prototype. (Author)

Performance of cable-in-conduit conductors in ITER [International Thermonuclear Experimental Reactor] toroidal field coils with varying heat loads

International Nuclear Information System (INIS)

Kerns, J.A.; Wong, R.L.

1989-01-01

The toroidal field (TF) coils in the International Thermonuclear Experimental Reactor (ITER) will operate with varying heat loads generated by ac losses and nuclear heating. The total heat load is estimated to be 2 kW per TF coil under normal operation and can be higher for different operating scenarios. Ac losses are caused by ramping the poloidal field (PF) for plasma initiation, burn, and shutdown; nuclear heating results from neutrons that penetrate into the coil past the shield. Present methods to reduce or eliminate these losses lead to larger and more expensive machines, which are unacceptable with today's budget constraints. A suitable solution is to design superconductors that operate with high heat loads. The cable-in-conduit conductor (CICC) can operate with high heat loads. One CICC design is analyzed for its thermal performance using two computer codes developed at LLNL. One code calculates the steady state flow conditions along the flow path, while the other calculates the transient conditions in the flow. We have used these codes to analyze the superconductor performance during the burn phase of the ITER plasma. The results of these analyses give insight to the choice of flow rate on superconductor performance. 4 refs., 5 figs

Analysis of quench-vent pressures for present design of ITER [International Thermonuclear Experimental Reactor] TF [toroidal field] coils

International Nuclear Information System (INIS)

Slack, D.S.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Union of the Soviet Union, and the United States. This paper examines the effects of a quench within the toroidal field (TF) coils based on current ITER design. It is a preliminary, rough analysis. Its intent is to assist ITER designers while more accurate computer codes are being developed and to provide a check against these more rigorous solutions. Rigorous solutions to the quench problem are very complex involving three-dimensional heat transfer, extreme changes in heat capacities and copper resistivity, and varying flow dynamics within the conductors. This analysis addresses all these factors in an approximate way. The result is much less accurate than a rigorous analysis. Results here could be in error as much as 30 to 40 percent. However, it is believed that this paper can still be very useful to the coil designer. Coil pressures and temperatures vs time into a quench are presented. Rate of helium vent, energy deposition in the coil, and depletion of magnetic stored energy are also presented. Peak pressures are high (about 43 MPa). This is due to the very long vent path length (446 m), small hydraulic diameters, and high current densities associated with ITER's cable-in-conduit design. The effects of these pressures as well as the ability of the coil to be self protecting during a quench are discussed. 3 refs., 3 figs., 1 tab

ITER toroidal field model coil (TFMC). Test and analysis summary report (testing handbook) chapter 3 TOSKA FACILITY

International Nuclear Information System (INIS)

Ulbricht, A.

2001-05-01

In the frame of a contract between the ITER (International Thermonuclear Experimental Reactor) Director and the European Home Team Director was concluded the extension of the TOSKA facility of the Forschungszentrum Karlsruhe as test bed for the ITER toroidal field model coil (TFMC), one of the 7 large research and development projects of the ITER EDA (Engineering Design Activity). The report describes the work and development, which were performed together with industry to extend the existing components and add new components. In this frame a new 2 kW refrigerator was added to the TOSKA facility including the cold lines to the Helium dewar in the TOSKA experimental area. The measuring and control system as well as data acquisition was renewed according to the state-of-the-art. Two power supplies (30 kA, 50 kA) were switched in parallel across an Al bus bar system and combined with an 80 kA dump circuit. For the test of the TFMC in the background field of the EURATOM LCT coil a new 20 kA power supply was taken into operation with the existing 20 kA discharge circuit. Two forced flow cooled 80 kA current leads for the TFMC were developed. The total lifting capacity for loads in the TOSKA building was increased by an ordered new 80 t crane with a suitable cross head (125 t lifting capacity +5 t net mass) to 130 t for assembling and installation of the test arrangement. Numerous pre-tests and development and adaptation work was required to make the components suitable for application. The 1.8 K test of the EURATOM LCT coil and the test of the W 7-X prototype coil count to these tests as overall pre-tests. (orig.)

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

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)

Results of ITER toroidal field coil cover plate welding test

International Nuclear Information System (INIS)

Koizumi, Norikiyo; Matsui, Kunihiro; Shimizu, Tatsuya; Nakajima, Hideo; Iijima, Ami; Makino, Yoshinobu

2012-01-01

In ITER Toroidal Field (TF) coils, cover plates (CP) are welded to the teeth of the radial plate (RP) to fix conductors in the grooves of the RP. Though the total length of the welds is approximately 1.5 km and the height and width of the RP are 14 and 9 m, respectively, welding deformation of smaller than 1 mm for local out-of-plane distortion and smaller than several millimeters for in-plane deformation is required. Therefore, laser welding is used for CP welding to reduce welding deformation as much as possible. However, the gap in welding joints is expected to be a maximum of 0.5 mm. Thus, a laser welding technique to enable welding of joints with a gap of 0.5 mm in width has been developed. Applying this technology, a CP welding trial using an RP mock-up was successfully performed. The achieved local flatness, that is, the flatness of the cross-section of the RP mock-up, is 0.6 mm. The analysis using inherent strains, which are derived from the welding test using flat plates, also indicates that better local flatness can be achieved if the initial distortion is zero. In addition, the welding deformation of a full-scale RP is evaluated via analysis using the inherent strain. The analytical results show that in-plane deformation is approximately 5 mm and large out-of-plane deformation, consisting of approximately 5 mm-long wave distortion and a twist of approximately 1.5 mm in the RP cross-section, is generated. It is expected that the required profile can be achieved by determining the original geometry of an RP by simulating deformation during welding. It is also expected that the required local flatness of a DP can be achieved, since out-of-plane deformation can be reduced by increasing the number of RPs turned over during CP welding. A more detailed study is required. (author)

Exploring the limits of a very large Nb3Sn conductor: the 80 kA conductor of the ITER toroidal field model coil

International Nuclear Information System (INIS)

Duchateau, J.L.; Ciazynski, D.; Guerber, O.; Park, S.H.; Zani, L.

2003-01-01

In Phase II experiment of the International Thermonuclear Experimental Reactor (ITER) Toroidal Field Model Coil (TFMC) the operation limits of its 80 kA Nb 3 Sn conductor were explored. To increase the magnetic field on the conductor, the TFMC was tested in presence of another large coil: the EURATOM-LCT coil. Under these conditions the maximum field reached on the conductor, was around 10 tesla. This exploration has been performed at constant current, by progressively increasing the coil temperature and monitoring the coil voltage drop in the current sharing regime. Such an operation was made possible thanks to the very high stability of the conductor. The aim of these tests was to compare the critical properties of the conductor with expectations and assess the ITER TF conductor design. These expectations are based on the documented critical field and temperature dependent properties of the 720 superconducting strands which compose the conductor. In addition the conductor properties are highly dependent on the strain, due to the compression appearing on Nb 3 Sn during the heat treatment of the pancakes and related to the differential thermal compression between Nb 3 Sn and the stainless steel jacket. No precise model exists to predict this strain, which is therefore the main information, which is expected from these tests. The method to deduce this strain from the different tests is presented, including a thermalhydraulic analysis to identify the temperature of the critical point and a careful estimation of the field map across the conductor. The measured strain has been estimated in the range -0.75% to -0.79 %. This information will be taken into account for ITER design and some adjustment of the ITER conductor design is under examination. (authors)

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

Basic analysis of weldability and machinability of structural materials for ITER Toroidal Field coils

Energy Technology Data Exchange (ETDEWEB)

Onozuka, Masanori [Mitsubishi Heavy Industries Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan)], E-mail: masanori_onozuka@mnes-us.com; Shimizu, Katsusuke; Urata, Kazuhiro; Kimura, Masahiro; Kadowaki, Hirokazu; Okamoto, Mamoru [Mitsubishi Heavy Industries Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Nakajima, Hideo; Hamada, Kazuya; Okuno, Kiyoshi [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-shi, Ibaraki 311-0193 (Japan)

2007-10-15

A manufacturing study has been conducted to examine the welding and machining capabilities for strengthened austenitic stainless steels with a high nitrogen content, JJ1 and ST-SS316LN, to be employed for ITER Toroidal Field (TF) coil structural components. It was found that the applicable EB welding condition for JJ1 was limited to up to 40 mm thick plates. A wider range of welding conditions was found in the vertical upward direction. Based on those results, a verification test up to 900 mm in length was successfully conducted. With respect to TIG welding, an average deposition rate of 26 g/min (i.e. the filler wire supplying speed of 3000 mm/min) was achieved. In addition to the welding tests, a series of machining tests has been conducted to examine the machinability of JJ1 and ST-SS316LN. Various types of machining tools were examined. In practical application, the cutting speed should be low to extend the tool life. At a cutting speed of 40 m/min, a tool life of more than 2 h (at a traveling distance of up to 9 m) was attained. The degree of cutter wear after 30 min of operation, at a cutting speed of 40 m/min, was found to be around 0.1 mm, which is within an acceptable range.

Development of optimum manufacturing technologies of radial plates for the ITER toroidal field coils

International Nuclear Information System (INIS)

Nakajima, H.; Hamada, K.; Okuno, K.; Abe, K.; Shimizu, T.; Kakui, H.; Yamaoka, H.; Maruyama, N.; Takayanagi, T.

2007-01-01

Japan Atomic Energy Agency is studying rational manufacturing method and developing the optimum manufacturing technologies of the radial plates used in the toroidal field coils for the International Thermonuclear Experimental Reactor (ITER) in collaboration with the Japanese industries. Three sector form pieces were cut by plasma cutting machine from a hot rolled plate without any difficulties and one of them was machined to a 1.32-m long curved segment of the radial plate having the same size as the actual one. However, unacceptable large deformation about 5 mm flatness, which was not observed in 1-m long straight radial plate, was found after intermediate machining. Since it would be caused by groove direction against the hot rolled direction and/or curved shape of grooves, two trial manufactures of 0.4-m long straight radial plates have been performed to clarify the cause of the large deformation. Detailed investigation showed that the large deformation could be avoided if the groove direction would have been parallel to a rolling direction of the plate. Welding trials by using fiber laser technique was also performed and penetration of 15 mm could be obtained in a welding speed of 0.1 m/min at 5 kW laser power. An optimum manufacturing method has been proposed based on the development of manufacturing technologies

Review of design principles for ITER VV remote inspection in magnetic field

International Nuclear Information System (INIS)

Izard, Jean-Baptiste; Perrot, Yann; Friconneau, Jean-Pierre

2009-01-01

Because ITER magnet system has a limited number of mechanical and thermal stress cycles, shut down number of the toroidal field is limited during lifetime of ITER. Any inspection device able to withstand the toroidal field between two plasma shots will enhance the inspection frequency capacity of ITER during operation phase. In addition to the high magnetic field the system should also cope with high temperature, ultra-high vacuum and high radiation, in order to keep the reactor availability high. Radiation, ultra-high vacuum and temperature constraints already addressed by on going R and D activities within Europe-considering the required level of radiation is to date the highest encountered in remote handling, and that facing all these constraints at once is an additional issue to overcome. Whereas, operating remote handling systems in high magnetic field is quite new field of investigation. This paper aims to be a guideline for future designers to help them choose among options the adequate solution for an ITER relevant inspection device. It provides the designer an objective view of the different effects that stem from technical choices and help them deciding whether a technology is relevant or not depending on the task's requirements. We have selected a set of technologies and products available for structural design, actuation, sensing and data transmission in order to design inspection remote handling equipment for ITER in the given constraints. These different solutions are commented with specific considerations and directions to have them fit in the specifications. Different design strategies to cope with magnetic field are then discussed, which imply either insensitive design or using the magnetic field as a potential energy source and as a positioning help. This analysis is the first result of one of the projects in the PREFIT partnership, part of the European Fusion Training Scheme.

Engineering, Manufacture and Preliminary Testing of the ITER Toroidal Field (TF) Magnet Helium Cold Circulator

Science.gov (United States)

Rista, P. E. C.; Shull, J.; Sargent, S.

2015-12-01

The ITER cryodistribution system provides the supercritical Helium (SHe) forced flow cooling to the magnet system using cold circulators. The cold circulators are located in each of five separate auxiliary cold boxes planned for use in the facility. Barber-Nichols Inc. has been awarded a contract from ITER-India for engineering, manufacture and testing of the Toroidal Field (TF) Magnet Helium Cold Circulator. The cold circulator will be extensively tested at Barber-Nichols’ facility prior to delivery for qualification testing at the Japan Atomic Energy Agency's (JAEA) test facility at Naka, Japan. The TF Cold Circulator integrates features and technical requirements which Barber-Nichols has utilized when supplying helium cold circulators worldwide over a period of 35 years. Features include a vacuum-jacketed hermetically sealed design with a very low helium leak rate, a heat shield for use with both nitrogen & helium cold sources, a broad operating range with a guaranteed isentropic efficiency over 70%, and impeller design features for high efficiency. The cold circulator will be designed to meet MTBM of 17,500 hours and MTBF of 36,000 hours. Vibration and speed monitoring are integrated into a compact package on the rotating assembly with operation and health monitoring in a multi-drop PROFIBUS communication environment using an electrical cabinet with critical features and full local and network PLC interface and control. For the testing in Japan and eventual installation in Europe, the cold circulator must be certified to the Japanese High Pressure Gas Safety Act (JHPGSA) and CE marked in compliance with the European Pressure Equipment Directive (PED) including Essential Safety Requirements (ESR). The test methodology utilized at Barber-Nichols’ facility and the resulting test data, validating the high efficiency of the TF Cold Circulator across a broad operating range, are important features of this paper.

Engineering, Manufacture and Preliminary Testing of the ITER Toroidal Field (TF) Magnet Helium Cold Circulator

International Nuclear Information System (INIS)

C Rista, P E; Shull, J; Sargent, S

2015-01-01

The ITER cryodistribution system provides the supercritical Helium (SHe) forced flow cooling to the magnet system using cold circulators. The cold circulators are located in each of five separate auxiliary cold boxes planned for use in the facility. Barber-Nichols Inc. has been awarded a contract from ITER-India for engineering, manufacture and testing of the Toroidal Field (TF) Magnet Helium Cold Circulator. The cold circulator will be extensively tested at Barber-Nichols’ facility prior to delivery for qualification testing at the Japan Atomic Energy Agency's (JAEA) test facility at Naka, Japan. The TF Cold Circulator integrates features and technical requirements which Barber-Nichols has utilized when supplying helium cold circulators worldwide over a period of 35 years. Features include a vacuum-jacketed hermetically sealed design with a very low helium leak rate, a heat shield for use with both nitrogen and helium cold sources, a broad operating range with a guaranteed isentropic efficiency over 70%, and impeller design features for high efficiency. The cold circulator will be designed to meet MTBM of 17,500 hours and MTBF of 36,000 hours. Vibration and speed monitoring are integrated into a compact package on the rotating assembly with operation and health monitoring in a multi-drop PROFIBUS communication environment using an electrical cabinet with critical features and full local and network PLC interface and control. For the testing in Japan and eventual installation in Europe, the cold circulator must be certified to the Japanese High Pressure Gas Safety Act (JHPGSA) and CE marked in compliance with the European Pressure Equipment Directive (PED) including Essential Safety Requirements (ESR). The test methodology utilized at Barber-Nichols’ facility and the resulting test data, validating the high efficiency of the TF Cold Circulator across a broad operating range, are important features of this paper. (paper)

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

ITER CTA newsletter. No. 2

International Nuclear Information System (INIS)

2001-10-01

This ITER CTA newsletter contains results of the ITER toroidal field model coil project presented by ITER EU Home Team (Garching) and an article in commemoration of the late Dr. Charles Maisonnier, one of the former leaders of ITER who made significant contributions to its development

Rotation and neoclassical ripple transport in ITER

Science.gov (United States)

Paul, E. J.; Landreman, M.; Poli, F. M.; Spong, D. A.; Smith, H. M.; Dorland, W.

2017-11-01

Neoclassical transport in the presence of non-axisymmetric magnetic fields causes a toroidal torque known as neoclassical toroidal viscosity (NTV). The toroidal symmetry of ITER will be broken by the finite number of toroidal field coils and by test blanket modules (TBMs). The addition of ferritic inserts (FIs) will decrease the magnitude of the toroidal field ripple. 3D magnetic equilibria in the presence of toroidal field ripple and ferromagnetic structures are calculated for an ITER steady-state scenario using the variational moments equilibrium code (VMEC). Neoclassical transport quantities in the presence of these error fields are calculated using the stellarator Fokker-Planck iterative neoclassical conservative solver (SFINCS). These calculations fully account for E r , flux surface shaping, multiple species, magnitude of ripple, and collisionality rather than applying approximate analytic NTV formulae. As NTV is a complicated nonlinear function of E r , we study its behavior over a plausible range of E r . We estimate the toroidal flow, and hence E r , using a semi-analytic turbulent intrinsic rotation model and NUBEAM calculations of neutral beam torque. The NTV from the \\vert{n}\\vert = 18 ripple dominates that from lower n perturbations of the TBMs. With the inclusion of FIs, the magnitude of NTV torque is reduced by about 75% near the edge. We present comparisons of several models of tangential magnetic drifts, finding appreciable differences only for superbanana-plateau transport at small E r . We find the scaling of calculated NTV torque with ripple magnitude to indicate that ripple-trapping may be a significant mechanism for NTV in ITER. The computed NTV torque without ferritic components is comparable in magnitude to the NBI and intrinsic turbulent torques and will likely damp rotation, but the NTV torque is significantly reduced by the planned ferritic inserts.

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

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.

ITER jako živý

Czech Academy of Sciences Publication Activity Database

Řípa, Milan

2010-01-01

Roč. 10, č. 6 (2010), s. 18-19 Institutional research plan: CEZ:AV0Z20430508 Keywords : Fusion * ITER * magnetic field * ELMs * cryo pumps * central solenoid * correction coils * superconducting coils * toroidal field coils * poloidal field coils * divertor * Cadarache Subject RIV: BL - Plasma and Gas Discharge Physics http://www.tretipol.cz/900-iter-jako-zivy

Toroid field coil shear key installation study, DOE task No. 22

International Nuclear Information System (INIS)

Jones, C.E.; Meier, R.W.; Yuen, J.L.

1995-01-01

Concepts for fitting and installation of the scissor keys, triangular keys, and truss keys in the ITER Toroidal Field (TF) Coil Assembly were developed and evaluated. In addition, the process of remote removal and replacement of a failed TF coil was considered. Two concepts were addressed: central solenoid installed last (Naka Option 1) and central solenoid installed first (Naka Option 2). In addition, a third concept was developed which utilized the favorable features of both concepts. A time line for installation was estimated for the Naka Option 1 concept

Poloidal field system for the ITER hard design option

International Nuclear Information System (INIS)

Schultz, J.H.; Pillsbury, R.D.

1992-01-01

This paper reports on ITER, the International Thermonuclear Experimental Reactor, a collaborative design by the US, EC, Japan, and the USSR of a tokamak fusion reactor that will demonstrate the physics and test the technology needed for commercial fusion reactors. In 1990, the ITER team completed a Conceptual Design Activity (CDA) in which a candidate design was shown to meet the specified goals of the ITER activity at a conceptual level. The four parties have agreed to an Engineering Design Activity (EDA) that includes the necessary additional design and analysis, along with the R and D needed to construct ITER with confidence. The CDA design includes a toroidal field (TF) magnet system that provides the main containment field and a poloidal field (PF) system used to control plasma current and position. The PF system is also used as transformer primary to induce and sustain current in the plasma. Since the volt-seconds available for full-current plasma burn are less than 10% of the total available volt-seconds from the PF system, an area of concern in the CDA design is that unfavorable plasma conditions could compromise the ability of the physics base case design to achieve long pulse burns. A High Aspect Ratio Design (HARD) was conceived as an alternative design option with a much larger bore in the central solenoid to enhance ITER's capabilities for long-burn operation

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)

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.

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.

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

The importance of the toroidal magnetic field for the feasibility of a tokamak burning plasma experiment

International Nuclear Information System (INIS)

Mazzucato, E.

2000-01-01

The next step in the demonstration of the scientific feasibility of a tokamak fusion reactor is a DT burning plasma experiment for the study and control of self-heated plasmas. In this paper, the authors examine the role of the toroidal magnetic field on the confinement of a tokamak plasma in the ELMy H-mode regime--the operational regime foreseen for ITER

Comparison of edge plasma perturbation during ELM control using one vs. two toroidal rows of RMP coils in ITER similar shaped plasmas on DIII-D

Energy Technology Data Exchange (ETDEWEB)

Fenstermacher, M.E., E-mail: fenstermacher@fusion.gat.co [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551 (United States); Evans, T.E.; Osborne, T.H.; Schaffer, M.J.; DeGrassie, J.S.; Gohil, P.; Groebner, R.J. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Moyer, R.A. [University of California, San Diego, 9500 Gilman Dr., La Jolla, California 92093 (United States)

2009-06-15

Large Type-I edge localized modes (ELMs) were suppressed by n = 3 resonant magnetic perturbations (RMPs) from a set of internal coils in plasmas with an ITER similar shape at the ITER pedestal collisionality, nu{sub e}*approx0.1 and low edge safety factor (q{sub 95} approx 3.6), with either a single toroidal row of the internal RMP coils or two poloidally separated rows of coils. ELM suppression with a single row of internal coils was achieved at approximately the same q{sub 95} surface-averaged perturbation field as with two rows of coils, but required higher current per coil. Maintaining complete suppression of ELMs using n = 3 RMPs from a single toroidal row of internal coils was less robust to variations in input neutral beam injection torque than previous ELM suppression cases using both rows of internal coils. With either configuration of RMP coils, maximum ELM size is correlated with the width of the edge region having good overlap of the magnetic islands from vacuum field calculations.

Development of the ITER Continuous External Rogowski: From conceptual design to final design

Energy Technology Data Exchange (ETDEWEB)

Moreau, Philippe, E-mail: philippe.jacques.moreau@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Spuig, Pascal; Le-luyer, Alain; Malard, Philippe; Cantone, Bruno; Pastor, Patrick; Saint-Laurent, François [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Vayakis, George; Delhom, Dominique [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Arshad, Shakeib [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Lister, Jonathan; Toussaint, Matthieu; Marmillod, Philippe; Testa, Duccio; Schlatter, Christian [Ecole polytechnique fédérale de Lausanne, Centre de Recherches en Physique des Plasmas, 1015 Lausanne (Switzerland); Peruzzo, Simone [Consorzio RFX, C.so Stati Uniti 4, 35127 Padova (Italy)

2015-10-15

Highlights: • ITER Continuous External Rogowskis are designed for plasma current measurement. • CER are located in the casing of Toroidal Field Coils and will operate at 4.5 K. • The design of the sensors has been completed and validated through prototypes. • Detailed assembly procedure inside the toroidal field coil casing has been defined. • The CER has passed all the ITER and F4E design review procedures. - Abstract: In ITER, an accurate measurement of plasma current, with high reliability, is mandatory as this parameter is used to demonstrate licensing compliance with regulatory limits. For that purpose, several independent measurements based on magnetic diagnostics have been proposed. Rogowski coils are standard inductive sensors for current measurement in many applications. In ITER, three continuous external Rogowski coils are to be installed in the casing of the toroidal field coils. These sensors are remarkable from several points of view: overall length is about 40 m, high sensitivity needed, located in the toroidal field coil casing at 4.5 K and complex 3D routing with tight bending radius of 50 mm. Since 2005 an extensive work has been carried out to develop and analyze several design options complying with ITER specifications. Prototypes of a selected continuous external Rogowski design were built and tested successfully in terms of electrical, thermal, mechanical and vacuum characteristics. Finally a detailed assembly procedure inside the toroidal field coil casing has been defined according to the coil manufacturing and assembly constraints.

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

Comparative investigation of ELM control based on toroidal modelling of plasma response to RMP fields

Science.gov (United States)

Liu, Yueqiang

2016-10-01

The type-I edge localized mode (ELM), bursting at low frequency and with large amplitude, can channel a substantial amount of the plasma thermal energy into the surrounding plasma-facing components in tokamak devices operating at the high-confinement mode, potentially causing severe material damages. Learning effective ways of controlling this instability is thus an urgent issue in fusion research, in particular in view of the next generation large devices such as ITER and DEMO. Among other means, externally applied, three-dimensional resonant magnetic perturbation (RMP) fields have been experimentally demonstrated to be successful in mitigating or suppressing the type-I ELM, in multiple existing devices. In this work, we shall report results of a comparative study of ELM control using RMPs. Comparison is made between the modelled plasma response to the 3D external fields and the observed change of the ELM behaviour on multiple devices, including MAST, ASDEX Upgrade, EAST, DIII-D, JET, and KSTAR. We show that toroidal modelling of the plasma response, based on linear and quasi-linear magnetohydrodynamic (MHD) models, provides essential insights that are useful in interpreting and guiding the ELM control experiments. In particular, linear toroidal modelling results, using the MARS-F code, reveal the crucial role of the edge localized peeling-tearing mode response during ELM mitigation/suppression on all these devices. Such response often leads to strong peaking of the plasma surface displacement near the region of weak equilibrium poloidal field (e.g. the X-point), and this provides an alternative practical criterion for ELM control, as opposed to the vacuum field based Chirikov criteria. Quasi-linear modelling using MARS-Q provides quantitative interpretation of the side effects due to the ELM control coils, on the plasma toroidal momentum and particle confinements. The particular role of the momentum and particle fluxes, associated with the neoclassical toroidal

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

Effects of 3D magnetic perturbations on toroidal plasmas

International Nuclear Information System (INIS)

Callen, J.D.

2011-01-01

stochasticity and increase plasma transport in the edge of H-mode plasmas. These various effects of 3D fields can be used to modify directly the plasma toroidal rotation (and possibly transport via multiple RMPs for controlling edge localized modes) and indirectly anomalous plasma transport. The present understanding and modelling of these various 3D magnetic field perturbation effects including for test blanket modules in ITER are summarized. Finally, implications of the present understanding and key open issues for developing a predictive capability of them for ITER are discussed. (topical review)

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

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

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

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

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)

Feedback control of resistive wall modes in toroidal devices

International Nuclear Information System (INIS)

Liu, Y.Q.

2002-01-01

Active feedback of resistive wall modes is investigated using cylindrical theory and toroidal calculations. For tokamaks, good performance is obtained by using active coils with one set of coils in the poloidal direction and sensors detecting the poloidal field inside the first wall, located at the outboard mid-plane. With suitable width of the feedback coil such a system can give robust control with respect to variations in plasma current, pressure and rotation. Calculations are shown for ITER-like geometry with a double wall. The voltages and currents in the active coils are well within the design limits for ITER. Calculations for RFP's are presented for a finite number of coils both in the poloidal and toroidal directions. With 4 coils in the poloidal and 24 coils in the toroidal direction, all non-resonant modes can be stabilized both at high and low theta. Several types of sensors, including radial and internal poloidal or toroidal sensors, can stabilize the RWM, but poloidal sensors give the most robust performance. (author)

Basic study on weldability and machinability of structural materials for ITER toroidal field coils

International Nuclear Information System (INIS)

Onozuka, M.; Shimizu, K.; Urata, K.; Kimura, M.; Kadowaki, H.; Okamoto, M.; Nakajima, H.; Hamada, K.; Okuno, K.

2006-01-01

The toroidal field (TF) coils for ITER are very large components. The main structural component of the coil is the coil case, which requires a massive complex geometry with high fabrication accuracy to attain the required magnetic performance for plasma operations. To provide high mechanical strength and toughness at cryogenic temperature, the structural components employ high-strength austenite stainless steels that have been specially developed for ITER. However, one of the main drawbacks of using those materials is the difficulty of manufacturing capabilities. A manufacturing study has been conducted to examine welding and machining capabilities for JJ1 and ST-SS316LN, to be employed for TF coil structural components. Both materials include a high nitrogen content up to around 0.2%, which makes welding and machining difficult compared with conventional stainless steels. Electron beam welding conditions were studied for the JJ1 material. The applicable welding condition was found for a bead length of up to about 300 mm in the case of 40 mm thick plates. No optimal condition was found for plates thicker than 40 mm. An additional experimental study was also conducted to explore suitable welding conditions for different welding positions and directions. It was found that the appearance of defects depends on the welding positions and directions. A wider range of welding conditions was found for cases in the vertical upward direction, as opposed to those in the vertical downward and horizontal directions. Based on those results, a verification test up to 900 mm in length was conducted. The test results showed that vertical upward EB welding should be used for the coil case wherever possible. With respect to TIG welding, an average deposition rate as high as 26 g/min (i.e. the filler wire supplying speed of 3,000 mm/min) was achieved. A series of tests have been conducted to examine machinability of JJ1 and ST-SS316LN. Various types of milling tools, including face

Error Field Assessment from Driven Mode Rotation: Results from Extrap-T2R Reversed-Field-Pinch and Perspectives for ITER

Science.gov (United States)

Volpe, F. A.; Frassinetti, L.; Brunsell, P. R.; Drake, J. R.; Olofsson, K. E. J.

2012-10-01

A new ITER-relevant non-disruptive error field (EF) assessment technique not restricted to low density and thus low beta was demonstrated at the Extrap-T2R reversed field pinch. Resistive Wall Modes (RWMs) were generated and their rotation sustained by rotating magnetic perturbations. In particular, stable modes of toroidal mode number n=8 and 10 and unstable modes of n=1 were used in this experiment. Due to finite EFs, and in spite of the applied perturbations rotating uniformly and having constant amplitude, the RWMs were observed to rotate non-uniformly and be modulated in amplitude (in the case of unstable modes, the observed oscillation was superimposed to the mode growth). This behavior was used to infer the amplitude and toroidal phase of n=1, 8 and 10 EFs. The method was first tested against known, deliberately applied EFs, and then against actual intrinsic EFs. Applying equal and opposite corrections resulted in longer discharges and more uniform mode rotation, indicating good EF compensation. The results agree with a simple theoretical model. Extensions to tearing modes, to the non-uniform plasma response to rotating perturbations, and to tokamaks, including ITER, will be discussed.

Magnet design technical report---ITER definition phase

International Nuclear Information System (INIS)

Henning, C.

1989-01-01

This report contains papers on the following topics: conceptual design; radiation damage of ITER magnet systems; insulation system of the magnets; critical current density and strain sensitivity; toroidal field coil structural analysis; stress analysis for the ITER central solenoid; and volt-second capabilities and PF magnet configurations

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)

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

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.

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

Structural characteristics of proposed ITER [International Thermonuclear Experimental Reactor] TF [toroidal field] coil conductor

International Nuclear Information System (INIS)

Gibson, C.R.; Miller, J.R.

1988-01-01

This paper analyzes the effect of transverse loading on a cable-in-conduit conductor which has been proposed for the toroidal field coils of the International Thermonuclear Experimental Reactor. The primary components of this conductor are a loose cable of superconducting wires, a thin-wall tube for helium containment, and a U-shaped structural channel. A method is given where the geometry of this conductor can be optimized for a given set of operating conditions. It is shown, using finite-element modeling, that the structural channel is effective in supporting loads due to transverse forces and internal pressure. In addition, it is shown that the superconducting cable is effectively shielded from external transverse loads that might otherwise degrade its current carrying capacity. 10 refs., 10 figs., 3 tabs

Compact toroid refueling of reactors

International Nuclear Information System (INIS)

Gouge, M.J.; Hogan, J.T.; Milora, S.L.; Thomas, C.E.

1988-04-01

The feasibility of refueling fusion reactors and devices such as the International Thermonuclear Engineering Reactor (ITER) with high-velocity compact toroids is investigated. For reactors with reasonable limits on recirculating power, it is concluded that the concept is not economically feasible. For typical ITER designs, the compact toroid fueling requires about 15 MW of electrical power, with about 5 MW of thermal power deposited in the plasma. At these power levels, ideal ignition (Q = ∞) is not possible, even for short-pulse burns. The pulsed power requirements for this technology are substantial. 6 ref., 1 figs

Investigation of intrinsic toroidal rotation scaling in KSTAR

Science.gov (United States)

Yoo, J. W.; Lee, S. G.; Ko, S. H.; Seol, J.; Lee, H. H.; Kim, J. H.

2017-07-01

The behaviors of an intrinsic toroidal rotation without any external momentum sources are investigated in KSTAR. In these experiments, pure ohmic discharges with a wide range of plasma parameters are carefully selected and analyzed to speculate an unrevealed origin of toroidal rotation excluding any unnecessary heating sources, magnetic perturbations, and strong magneto-hydrodynamic activities. The measured core toroidal rotation in KSTAR is mostly in the counter-current direction and its magnitude strongly depends on the ion temperature divided by plasma current (Ti/IP). Especially the core toroidal rotation in the steady-state is well fitted by Ti/IP scaling with a slope of ˜-23, and the possible explanation of the scaling is compared with various candidates. As a result, the calculated offset rotation could not explain the measured core toroidal rotation since KSTAR has an extremely low intrinsic error field. For the stability conditions for ion and electron turbulences, it is hard to determine a dominant turbulence mode in this study. In addition, the intrinsic toroidal rotation level in ITER is estimated based on the KSTAR scaling since the intrinsic rotation plays an important role in stabilizing resistive wall modes for future reference.

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.

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.

Analytical study of cover plate welding deformation of the radial plate of the ITER toroidal field coil

International Nuclear Information System (INIS)

Ohmori, Junji; Koizumi, Norikiyo; Shimizu, Tatsuya; Okuno, Kiyoshi; Hasegawa, Mitsuru

2009-09-01

The winding pack (WP) of the Toroidal Field (TF) coil of ITER consists of 7 double-pancakes (DPs). In the DP, the conductor is embedded in a groove of a radial plate (RP), and cover plates (CP) are welded to the RP teeth to fix the conductors in the RP groove. The dimensions of the DP are 15 m in height and 9 m in width while the tolerances of the DP are very severe, such as a flatness of 2 mm and an in-plane deviation of a few millimeters. It is therefore required to reduce the deformation of the DP by CP welding. In order to estimate welding deformation, the authors apply an analytical method in which the CP welding deformation of the DP can be calculated using inherent strain evaluated from welding deformation measured using a RP mock-up. Calculated results indicate that out-of-plane distortion can be kept to within required tolerances, but in-plane deformation is larger than allowed when welding thickness is 2.5 mm. The in-plane deformation is mainly caused by the bending of the curved RP region. Therefore, reducing the welding thickness at the curved region emerges as the most promising solution of this issue. Calculated results assuming a welding thickness of 1 mm at the curved region show that the in-plane deformation conforms to required tolerances. Furthermore, since the maximum out-of-plane deformation is within tolerances but marginal, an alternative design in which the number of welding lines is half that of the reference design, is proposed not only to improve the out-of-plane distortion but also to simplify the manufacture of the DP. It is found that the alternative design is effective in reducing welding distortion. (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.

Development of Optimum Manufacturing Technologies of Radial Plates for the ITER Toroidal Field Coils

International Nuclear Information System (INIS)

Nakajima, H.; Hamada, K.; Okuno, K.; Abe, K.; Kakui, H.; Yamaoka, H.; Maruyama, N.

2006-01-01

A stainless steel structure called a radial plate is used in the toroidal field (TF) coils of the International Thermonuclear Experimental Reactor (ITER) in order to support large electromagnetic force generated in the conductors. It is a 13.7 m x 8.7 m D-shaped plate having 11 grooves on each side in which conductors are wound. Although severe dimensional accuracy, for example flatness within 2 mm, and tight schedule that all radial plates for 9 TF coils (63 plates) have to be manufactured in about 4 years are required in manufacture of the radial plates, there are no industries in the world who have manufactured a large complicated structure like the radial plate with high accuracy. Japan Atomic Energy Agency (JAEA) has been studying rational manufacturing method and developing the optimum manufacturing technologies of the radial plates in order to satisfy the above requirements in collaboration with the Ishikawajima-Harima Heavy Industries Co., Ltd. (IHI). Several trial manufactures of radial plates have been performed to clarify the following key points: · Effect of nitrogen content in material on machinability · Effect of cutting direction of a piece on deformation caused by machining · Effect of machining shape (curve or straight) on machining condition · Effect of laser welding technique on penetration and welding deformation Three different 316LN materials having nitrogen content of 0.12 %, 0.17%, and 0.20% were used to investigate nitrogen content effect on machinability. Machinability of lower nitrogen content material was slightly better than that of higher nitrogen content material. Three sectoral pieces were cut by plasma cutting technique from a hot rolled plate without any difficulties and one of them was machined to a curved segment of the radial plate having the same size as actual one. However, unacceptable large deformation over 5 mm flatness was found during machining which would be caused by curved shape of grooves and/or cutting direction

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 of the outer poloidal field coils for ITER

International Nuclear Information System (INIS)

Sborchia, C.; Mitchell, N.; Yoshida, K.

1995-01-01

The ITER poloidal field (PF) system consists of a central solenoid (CS or PF-1), which is not subject of this paper, and six ring coils using a 40 kA forced flow cooled superconductor. The coils, placed around the toroidal field (TF) system, are used to start-up the plasma with typical ramp-up times of 100 s and burn duration of 1000 s. They also provide control and shaping of the plasma, with small, frequent current variations on a 1-5 s time scale. The magnetic field produced by the coils ranges from about 4.5 to 8 T and the AC losses in the conductor are significant: the largest coils require cooling path lengths up to 1000 m as well as the use of 2 in-hand winding. The field level and high thermal loads make the use of Nb 3 Sn strand attractive. This paper describes the basic design of the six ring (outer) coils developed by the ITER Joint Central Team in collaboration with the four Home Teams. The coil structural material is provided by a thick conductor jacket and by a bonded insulation system. The forces acting on the coils during typical operational scenarios and plasma disruption/vertical instabilities have been evaluated: radial forces are self-reacted by hoop stresses in the ring coil, with tensile stresses up to 300 MPa in the conductor jacket, and the vertical forces are resisted by a discrete support system, with shear stresses up to 10 MPa in the insulation. (orig./WL)

Impact of error fields on equilibrium configurations in ITER

Energy Technology Data Exchange (ETDEWEB)

Barbato, Lucio [DIEI, Università di Cassino and Lazio Meridionale, Cassino (Italy); Formisano, Alessandro, E-mail: alessandro.formisano@unina2.it [Department of Industrial and Information Engineering, Seconda Univ. di Napoli, Aversa (Italy); Martone, Raffaele [Department of Industrial and Information Engineering, Seconda Univ. di Napoli, Aversa (Italy); Villone, Fabio [DIEI, Università di Cassino and Lazio Meridionale, Cassino (Italy)

2015-10-15

Highlights: • Error fields (EF) are discrepancies from nominal magnetic field, and may alter plasma behaviour. • They are due to, e.g., coils manufacturing and assembly errors. • EF impact in ITER equilibria is analyzed using numerical simulations. • A high accuracy 3D field computation module and a Grad-Shafranov solver are used. • Deformations size allow using a linearized model, and performing a sensitivity analysis. - Abstract: Discrepancies between design and actual magnetic field maps in tokamaks are unavoidable, and are associated to a number of causes, e.g. manufacturing and assembly tolerances on magnets, presence of feeders and joints, non-symmetric iron parts. Such error fields may drive plasma to loss of stability, and must be carefully controlled using suitable correction coils. Anyway, even when kept below safety threshold, error fields may alter the behavior of plasma. The present paper, using as example the error fields induced by tolerances in toroidal field coils, quantifies their effect on the plasma boundary shape in equilibrium configurations. In particular, a procedure able to compute the shape perturbations due to given deformations of the coils has been set up and used to carry out a thorough statistical analysis of the error field-shape perturbations relationship.

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

ITER EDA newsletter. V. 6, no. 4

International Nuclear Information System (INIS)

1997-04-01

This issue of the ITER EDA (Engineering Design Activities) Newsletter reports on the Toroidal Field Model Coil Project (Coil System, Objectives, Design, Project Management, Testing); contains a report of A Combined Workshop of Confinement Modeling and Database and Confinement and Transport Expert Groups held at the San Diego ITER Joint Work Site from April 14. to 18. Progress and status on implementing the ITER Confinement R and D needs as specified at the last Workshops of the Expert Groups in Montreal (Oct. 1996) were reported. 7 figs, 1 tab

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.)

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

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

Numerical simulation of liquid-metal-flows in radial-toroidal-radial bends

International Nuclear Information System (INIS)

Molokov, S.; Buehler, L.

1993-09-01

Magnetohydrodynamic flows in a U-bend and right-angle bend are considered with reference to the radial-toroidal-radial concept of a self-cooled liquid-metal blanket. The ducts composing bends have rectangular cross-section. The applied magnetic field is aligned with the toroidal duct and perpendicular to the radial ones. At high Hartmann number the flow region is divided into cores and boundary layers of different types. The magnetohydrodynamic equations are reduced to a system of partial differential equations governing wall electric potentials and the core pressure. The system is solved numerically by two different methods. The first method is iterative with iteration between wall potential and the core pressure. The second method is a general one for the solution of the core flow equations in curvilinear coordinates generated by channel geometry and magnetic field orientation. Results obtained are in good agreement. They show, that the 3D-pressure drop of MHD flows in a U-bend is not a critical issue for blanket applications. (orig./HP) [de

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.

The High Aspect Ratio Design (HARD): A candidate ITER concept with improved technology phase performance

International Nuclear Information System (INIS)

Nevins, W.M.; Perkins, L.J.; Wesley, J.C.

1992-10-01

The High Aspect Ratio Design (HARD) International Thermonuclear Experimental Reactor (ITER) concept developed by the US ITER team is an alternate to the low-aspect-ratio ITER design developed by the ITER participants during the Conceptual Design Activity (CDA). The CDA design, referred to hereafter as ITER CDA, has an aspect ratio, A, of 2.79, a toroidal magnetic field, B T , of 4.85 T, and a plasma current, I p , of 22 MA for operation with an ignited plasma. In contrast, HARD employs higher aspect ratio, A = 4.0, higher toroidal field, B T = 7.11 T, and lower plasma current, I p = 14.8 MA for ignition operation. The cross sections of the two designs are compared in. The parameters and performance of HARD and ITER CDA for inductively driven ignition operation are compared in. The HARD parameters provide the same ignition performance (ignition margin evaluated against ITER-89P confinement scaling) as ITER CDA in a device with comparable size and cost. However, the reason for advancing HARD rather than ITER CDA as the ITER design concept is not inductively driven ignition performance but HARD's significantly enhanced potential to achieve the technology testing and steady-state operation goals of the ITER objectives with non-inductive current drive

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.

Minimum dimension of an ITER like Tokamak with a given Q

Energy Technology Data Exchange (ETDEWEB)

Johner, J

2004-07-01

The minimum dimension of an ITER like tokamak with a given amplification factor Q is calculated for two values of the maximum magnetic field in the superconducting toroidal field coils. For ITERH-98P(y,2) scaling of the energy confinement time, it is shown that for a sufficiently large tokamak, the maximum Q is obtained for the operating point situated both at the maximum density and at the minimum margin with respect to the H-L transition. We have shown that increasing the maximum magnetic field in the toroidal field coils from the present 11.8 T to 16 T would result in a strong reduction of the machine size but has practically no effect on the fusion power. Values obtained for {beta}{sub N} are found to be below 2. Peak fluxes on the divertor plates with an ITER like divertor and a multi-machine expression for the power radiated in the plasma mantle, are below 10 MW/m{sup 2}.

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

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.

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)

Situation of the project of manufacture of 10 european toroidal coils for ITER; Situacion del proyecto de fabricacion de 10 bobinas toroidales europeas para el ITER

Energy Technology Data Exchange (ETDEWEB)

Felipe, A.; Mrenio, A.; Pando, F.; Pallisa, J.; Merino, O.; Condado, J. P.; Madorran, A.; Dormicchi, O.; Valle, N.; Presenti, P.; Durzo, C.; Pittaluga, S.; Lucas, J.; Ruiz de Villa, E.; Harrison, R.; Cornelis, M.; Cornella, J.; Poncet, L.; Bonito-Oliva, A.

2013-07-01

The toroidal coils are part of the magnetic confinement system, of tool of plasma ITER being them making a significant technological challenge since there is no previous experience of manufacture of similar dimensions superconducting coils (14 m X 9 m). F4E, is the agency responsible for making 10 of these coils, having awarded to the consortium of Iberdrola Ingenieria, ASG Superconductors and Elytt Energy making them. This project is now in the process of manufacture of the first Double Pancake prototype that will serve as a qualification of the manufacturing process.

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

Tokamak equilibria with non-parallel flow in a triangularity-deformed axisymmetric toroidal coordinate system

Directory of Open Access Journals (Sweden)

Ap Kuiroukidis

2018-01-01

Full Text Available We consider a generalized Grad–Shafranov equation (GGSE in a triangularity-deformed axisymmetric toroidal coordinate system and solve it numerically for the generic case of ITER-like and JET-like equilibria with non-parallel flow. It turns out that increase of the triangularity improves confinement by leading to larger values of the toroidal beta and the safety factor. This result is supported by the application of a criterion for linear stability valid for equilibria with flow parallel to the magnetic field. Also, the parallel flow has a weaker stabilizing effect.

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)

Effects of 3D Magnetic Perturbations on Toroidal Plasmas

International Nuclear Information System (INIS)

Callen, J.D.

2010-01-01

Full text: To lowest order tokamaks are two-dimensional (2D) axisymmetric magnetic systems. But small 3D magnetic perturbations (both externally applied and from plasma instabilities) have many interesting and useful effects on tokamak (and quasi-symmetric stellarator) plasmas. Plasma transport equations that include these effects, especially on diamagnetic-level toroidal plasma rotation, have recently been developed. The 3D magnetic perturbations and their plasma effects can be classified according to their toroidal mode number n: low n (1 to 5) resonant (q = m/n in plasma) and non-resonant fields, medium n (due to toroidal field ripple), and high n (due to microturbulence). This paper concentrates on low and medium n perturbations. Low n non-resonant magnetic fields induce a neoclassical toroidal viscosity (NTV) that damps toroidal plasma rotation throughout the plasma toward an offset flow in the counter-I p direction; recent tokamak experiments have confirmed and exploited these predictions by applying external low n non-resonant magnetic perturbations. Medium n perturbations have similar effects plus possible ripple trapping and resultant edge ion losses. A low n resonant magnetic field induces a toroidal plasma torque in the vicinity of the rational surface; when large enough it can stop plasma rotation there and lead to a locked mode, which often causes a plasma disruption. Externally applied 3D magnetic perturbations usually have many components; in the plasma their lowest n components are amplified by plasma responses, particularly at high beta. Low n plasma instabilities (e.g., NTMs, RWMs) cause additional 3D magnetic perturbations in tokamak plasmas; tearing modes can bifurcate the topology and form magnetic islands. Finally, multiple resonant magnetic perturbations (RMPs) can cause local magnetic stochasticity and influence H-mode edge pedestal transport. These various effects of 3D magnetic perturbations can be used to control the toroidal plasma

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

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)

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.

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)

Qualification of the US Made Conductors for ITER TF Magnet System

International Nuclear Information System (INIS)

Martovetsky, Nicolai N.; Hatfield, Daniel R.; Miller, John R.; Bruzzone, P.; Stepanov, B.; Seber, B.

2010-01-01

The US Domestic Agency (USDA) is one of the six suppliers of the Toroidal Field (TF) conductor for the International Thermonuclear Experimental Reactor (ITER). In order to qualify conductors according to ITER requirements we prepared several lengths of the CICC and short samples for testing in the SULTAN facility in CRPP, Switzerland. We also fully characterized the strands that were used in these SULTAN samples. Fabrication experience and test results are presented and discussed.

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.

Numerical determination of axisymmetric toroidal magnetohydrodynamic equilibria

International Nuclear Information System (INIS)

Johnson, J.L.; Dalhed, H.E.; Greene, J.M.

1978-07-01

Numerical schemes for the determination of stationary axisymmetric toroidal equilibria appropriate for modeling real experimental devices are given. Iterative schemes are used to solve the elliptic nonlinear partial differential equation for the poloidal flux function psi. The principal emphasis is on solving the free boundary (plasma-vacuum interface) equilibrium problem where external current-carrying toroidal coils support the plasma column, but fixed boundary (e.g., conducting shell) cases are also included. The toroidal current distribution is given by specifying the pressure and either the poloidal current or the safety factor profiles as functions of psi. Examples of the application of the codes to tokamak design at PPPL are given

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

Design considerations for ITER magnet systems

International Nuclear Information System (INIS)

Henning, C.D.; Miller, J.R.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is now completing a definition phase as a beginning of a three-year design effort. Preliminary parameters for the superconducting magnet system have been established to guide further and more detailed design work. Radiation tolerance of the superconductors and insulators has been of prime importance, since it sets requirements for the neutron-shield dimension and sensitively influences reactor size. The major levels of mechanical stress in the structure appear in the cases of the inboard legs of the toroidal-field (TF) coils. The cases of the poloidal-field (PF) coils must be made thin or segmented to minimize eddy current heating during inductive plasma operation. As a result, the winding packs of both the TF and PF coils includes significant fractions of steel. The authors present here preliminary ITER magnet systems design parameters taken from trade studies, design, and analyses performed by the Home Teams of the four ITER participants, by the ITER Magnet Design Unit in Garching, and by other participants at workshops organized by the Magnet Design Unit

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

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

European technology activities to prepare for ITER component procurement

International Nuclear Information System (INIS)

Gasparotto, M.

2006-01-01

phase. In particular the manufacturing feasibility of the superconducting strand for the toroidal and poloidal field coils and of the toroidal field coils has been demonstrated, Studies of manufacturing techniques for the vacuum vessel, blanket modules and divertor are also in progress, and a number of EU industries have been prepared to successfully participate to the ITER construction. The paper will report on the main R(and)D activities performed in the EU and the major achievements in preparation for ITER component procurement. (author)

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.)

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...

ITER tokamak device

International Nuclear Information System (INIS)

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

1991-01-01

The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

Wall conditioning for ITER: Current experimental and modeling activities

Energy Technology Data Exchange (ETDEWEB)

Douai, D., E-mail: david.douai@cea.fr [CEA, IRFM, Association Euratom-CEA, 13108 St. Paul lez Durance (France); Kogut, D. [CEA, IRFM, Association Euratom-CEA, 13108 St. Paul lez Durance (France); Wauters, T. [LPP-ERM/KMS, Association Belgian State, 1000 Brussels (Belgium); Brezinsek, S. [FZJ, Institut für Energie- und Klimaforschung Plasmaphysik, 52441 Jülich (Germany); Hagelaar, G.J.M. [Laboratoire Plasma et Conversion d’Energie, UMR5213, Toulouse (France); Hong, S.H. [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Lomas, P.J. [CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Lyssoivan, A. [LPP-ERM/KMS, Association Belgian State, 1000 Brussels (Belgium); Nunes, I. [Associação EURATOM-IST, Instituto de Plasmas e Fusão Nuclear, 1049-001 Lisboa (Portugal); Pitts, R.A. [ITER International Organization, F-13067 St. Paul lez Durance (France); Rohde, V. [Max-Planck-Institut für Plasmaphysik, 85748 Garching (Germany); Vries, P.C. de [ITER International Organization, F-13067 St. Paul lez Durance (France)

2015-08-15

Wall conditioning will be required in ITER to control fuel and impurity recycling, as well as tritium (T) inventory. Analysis of conditioning cycle on the JET, with its ITER-Like Wall is presented, evidencing reduced need for wall cleaning in ITER compared to JET–CFC. Using a novel 2D multi-fluid model, current density during Glow Discharge Conditioning (GDC) on the in-vessel plasma-facing components (PFC) of ITER is predicted to approach the simple expectation of total anode current divided by wall surface area. Baking of the divertor to 350 °C should desorb the majority of the co-deposited T. ITER foresees the use of low temperature plasma based techniques compatible with the permanent toroidal magnetic field, such as Ion (ICWC) or Electron Cyclotron Wall Conditioning (ECWC), for tritium removal between ITER plasma pulses. Extrapolation of JET ICWC results to ITER indicates removal comparable to estimated T-retention in nominal ITER D:T shots, whereas GDC may be unattractive for that purpose.

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 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

Conceptual design of SC magnet system for ITER, (6)

International Nuclear Information System (INIS)

Yoshida, Kiyoshi; Sugimoto, Makoto; Tsuji, Hiroshi

1991-08-01

The International Thermonuclear Experimental Reactor (ITER) is an experimental thermonuclear tokamak reactor in order to test the basic physics performance and technologies. The conceptual design activity (CDA) of ITER required the joint work at a technical site at the Max Plank Institute for Plasma Physics in the Garching, Germany from 1988 to 1990. The technical proposals from Japan were summarized by the Fusion Experimental Reactor (FER) Team and the Superconducting Magnet Laboratory of the Japan Atomic Energy Research Institute (JAERI). This paper describes the Japanese contributions of the R and D proposals to the magnet system for the ITER. These proposals were discussed in ITER CDA design team and summarized in ITER Technical report No. 20. The development program of Toroidal Field Coil is basically proposed from Japan with the design and analysis reports. The Japanese proposals are almost adopted in the ITER Long-Term R and D program. (author)

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

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.

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)

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

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)

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...

Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

International Nuclear Information System (INIS)

Abiteboul, J.

2012-10-01

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

Toroidal current asymmetry and boundary conditions in disruptions

Science.gov (United States)

Strauss, Henry

2014-10-01

It was discovered on JET that disruptions were accompanied by toroidal asymmetry of the plasma current. The toroidal current asymmetry ΔIϕ is proportional to the vertical current moment ΔMIZ , with positive sign for an upward vertical displacement event (VDE) and negative sign for a downward VDE. It was claimed that this could only be explained by Hiro current. It is shown that instead it is essentially a kinematic effect produced by the VDE displacement of a 3D magnetic perturbation. This is verified by M3D simulations. The simulation results do not require penetration of plasma into the boundary, as in the Hiro current model. It is shown that the normal velocity perpendicular to the magnetic field vanishes at the wall, in the small Larmor radius limit of electromagnetic sheath boundary conditions. Plasma is absorbed into the wall only via the parallel velocity, which is small, penetrates only an infinitesimal distance into the wall, and does not affect forces exerted by the plasma on the wall. Supported by USDOE and ITER.

The First Benchmarking of ITER BR Nb3Sn Strand of CNDA

International Nuclear Information System (INIS)

Long Feng; Liu Fang; Wu Yu; Ni Zhipeng

2012-01-01

According to the International Thermonuclear Experimental Reactor (ITER) Procurement Arrangement (PA) of Cable-In-Conduit Conductor (CICC) unit lengths for the Toroidal Field (TF) and Poloidal Field (PF) magnet systems of ITER, at the start of process qualification, the Domestic Agency (DA) shall be required to conduct a benchmarking of the room and low temperature acceptance tests carried out at the Strand Suppliers and/or at its Reference Laboratories designated by the ITER Organization (IO). The first benchmarking was carried out successfully in 2009. Nineteen participants from six DAs (China, European Union, Japan, South Korea, Russia, and the United States) participated in the first benchmarking. Bronze-route (BR) Nb 3 Sn strand and samples prepared by the ITER reference lab (CERN) were sent out to each participant by CERN. In this paper, the test facility and test results of the first benchmarking by the Chinese DA (CNDA) are presented.

Nb3Sn conductor development for the ITER magnets

International Nuclear Information System (INIS)

Mitchell, N.

1997-01-01

The ITER magnet system consists of Toroidal Field (TF) coils, Poloidal Field (PF) coils, the Central Solenoid (CS) and error field correction coils (CC). The conductors for the coils are Nb 3 Sn or NbTi cable in conduit type, forced flow cooled with supercritical helium having a maximum operating current in the range 40-60 kA. To qualify the Nb 3 Sn conductor, two large model coils (energy up to 640 MJ) are being wound by the Home Teams of the Parties to the ITER EDA Agreement. A total of 24 t of strand has been completed for the CS model coil and 4 t for the TF model coil, and fabricated into 7 km of conductor in unit lengths up to 210 m, by an international collaboration involving 12 companies in Europe, Japan, Russia and the USA

Status and issues of the European contribution to ITER

International Nuclear Information System (INIS)

Bindslev, H.

2015-01-01

Highlights: • We describe the technical status of F4E's contributions to the ITER International Fusion Energy Project. • The foundations of the ITER Tokamak Complex have been completed. • We describe the production of the Toroidal Field coils and the achieved accuracy. • The first stage of ITER's pre-qualification programme for the ITER first wall panels was completed. • Technical developments for several other ITER components are described. - Abstract: Fusion for Energy (F4E), on behalf of Europe, is responsible for the procurement of most of the high-technology items for the ITER device. This paper provides an overview of the technical status of Europe's contributions to ITER and the related challenges. In particular, we report on progress in the construction of the buildings at the Cadarache site, the fabrication of the superconducting magnets and the vacuum vessel and the testing and qualification of the in-vessel components (first wall and divertor). The status of the design and development of the additional heating systems and the test blanket modules will also be described.

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.

Study on ECH-assisted start-up using trapped particle configuration in KSTAR and application to ITER

Science.gov (United States)

Lee, Jeongwon; Kim, Jayhyun; An, YoungHwa; Yoo, Min-Gu; Hwang, Y. S.; Na, Yong-Su

2017-12-01

ECH-assisted start-up using trapped particle configuration (TPC) is firstly studied in a superconducting, conventional tokamak, KSTAR. First, improved and efficient start-up using TPC than conventional field null configuration (FNC) is achieved by enhanced pre-ionization plasma quality. TPC shows the broader operation window in terms of the poloidal field quality and the deuterium prefill pressure than that of FNC. Surprisingly the particle trapping enhances the plasma start-up performance even with much lower particle trapping ratio than that of spherical torus. Reliability of TPC with low trapping ratio is investigated by 0D plasma evolution code, TECHP0D. Second, the characteristics of TPC start-up are explored with experiments and modellings. Two kind of start-up failure conditions are identified by the magnetic pitch and prefill pressure scan experiments: (i) low ionization rate; and (ii) low density condition. These experimental observation has consistency with TECHP0D modeling result. Finally, reliable ITER-relevant low toroidal electric field start-up using TPC is achieved with drastically improved success rate of start-up in KSTAR. In this experiments, validity of the ITER-like toroidally inclined ECH/ECCD injection is also expected. Time dependent ITER start-up scenario using TPC is proposed and the superiority of it than FNC is discussed.

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.)

Fast ion power loads on ITER first wall structures in the presence of NTMs and microturbulence

International Nuclear Information System (INIS)

Kurki-Suonio, T.; Asunta, O.; Hirvijoki, E.; Koskela, T.; Snicker, A.; Sipilae, S.; Hauff, T.; Jenko, F.; Poli, E.

2011-01-01

The level and distribution of the wall power flux of energetic ions in ITER have to be known accurately in order to ensure the integrity of the first wall. Until now, most quantitative estimates have been based on the assumption that fast ion transport is dictated by neoclassical effects only. However, in ITER, the fast ion distribution is likely to be affected by various MHD effects and probably also by microturbulence. We have now upgraded our orbit-following Monte Carlo code ASCOT so that it has simple, theory-based models for neoclassical tearing mode (NTM)-type islands as well as for turbulent diffusion. ASCOT also allows for full-orbit following, which is important close to the material surfaces and, possibly, also when strong toroidal inhomogeneities are present in the magnetic field. Here we introduce the new models, preliminary results obtained with them, and how these models could be made more realistic in the future. The simulations are carried out for thermonuclear alpha particles in ITER scenario 2 plasma, because we consider this combination to be most critical for the successful operation of ITER. Neither the turbulent transport nor NTM-type islands are found to introduce alarming changes in the wall loads. However, at this stage it was not possible to combine the island structures with the non-axisymmetric magnetic field of ITER, and it remains to be seen what the combined effect of drift islands together with the toroidal ripple and local field aberrations, such as those due to test blanket modules and resonant magnetic perturbations will be.

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

Disruption, vertical displacement event and halo current characterization for ITER

International Nuclear Information System (INIS)

Wesley, J.; Fujisawa, N.; Ortolani, S.; Putvinski, S.; Rosenbluth, M.N.

1997-01-01

Characteristics, in ITER, of plasma disruptions, vertical displacement events (VDEs) and the conversion of plasma current to runaway electron current in a disruption are presented. In addition to the well known potential of disruptions to produce rapid thermal energy and plasma current quenches and theoretical predictions that show the likelihood of ∼ 50% runaway conversion, an assessment of VDE and halo current characteristics in vertically elongated tokamaks shows that disruptions in ITER will result in VDEs with peak in-vessel halo currents of up to 50% of the predisruption plasma current and with toroidal peaking factors (peak/average current density) of up to 4:1. However, the assessment also shows an inverse correlation between the halo current magnitude and the toroidal peaking factor; hence, ITER VDEs can be expected to have a product of normalized halo current magnitude times toroidal peaking factor of ≤ 75%. (author). 3 refs, 2 figs, 3 tabs

Magnetohydrodynamic Stability of a Toroidal Plasma's Separatrix

International Nuclear Information System (INIS)

Webster, A. J.; Gimblett, C. G.

2009-01-01

Large tokamaks capable of fusion power production such as ITER, should avoid large edge localized modes (ELMs), thought to be triggered by an ideal magnetohydrodynamic instability due to current at the plasma's separatrix boundary. Unlike analytical work in a cylindrical approximation, numerical work finds the modes are stable. The plasma's separatrix might stabilize modes, but makes analytical and numerical work difficult. We generalize a cylindrical model to toroidal separatrix geometry, finding one parameter Δ ' determines stability. The conformal transformation method is generalized to allow nonzero derivatives of a function on a boundary, and calculation of the equilibrium vacuum field allows Δ ' to be found analytically. As a boundary more closely approximates a separatrix, we find the energy principle indicates instability, but the growth rate asymptotes to zero

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

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.

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.

Integration test of ITER full-scale vacuum vessel sector

International Nuclear Information System (INIS)

Nakahira, M.; Koizumi, K.; Oka, K.

1999-01-01

The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven ITER R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of - 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (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).

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

Finite toroidal flow generated by unstable tearing mode in a toroidal plasma

Energy Technology Data Exchange (ETDEWEB)

Hao, G. Z., E-mail: haogz@swip.ac.cn; Wang, A. K.; Xu, Y. H.; He, H. D.; Xu, M.; Qu, H. P.; Peng, X. D.; Xu, J. Q.; Qiu, X. M. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Liu, Y. Q. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Sun, Y. [Institute of Plasma Physics, Chinese Academic of Sciences, P.O. Box 1126, Hefei 230031 (China); Cui, S. Y. [School of Mathematics and Statistics Science, Ludong University, Yantai 264025 (China)

2014-12-15

The neoclassical toroidal plasma viscosity torque and electromagnetic torque, generated by tearing mode (TM) in a toroidal plasma, are numerically investigated using the MARS-Q code [Liu et al., Phys. Plasmas 20, 042503 (2013)]. It is found that an initially unstable tearing mode can intrinsically drive a toroidal plasma flow resulting in a steady state solution, in the absence of the external momentum input and external magnetic field perturbation. The saturated flow is in the order of 0.5%ω{sub A} at the q=2 rational surface in the considered case, with q and ω{sub A} being the safety factor and the Alfven frequency at the magnetic axis, respectively. The generation of the toroidal flow is robust, being insensitive to the given amplitude of the perturbation at initial state. On the other hand, the flow amplitude increases with increasing the plasma resistivity. Furthermore, the initially unstable tearing mode is fully stabilized by non-linear interaction with the self-generated toroidal flow.

Field load and displacement boundary condition computer program used for the finite element analysis and design of toroidal field coils in a tokamak

International Nuclear Information System (INIS)

Smith, R.A.

1975-06-01

The design evaluation of toroidal field coils on the Princeton Large Torus (PLT), the Poloidal Diverter Experiment (PDX) and the Tokamak Fusion Test Reactor (TFTR) has been performed by structural analysis with the finite element method. The technique employed has been simplified with supplementary computer programs that are used to generate the input data for the finite element computer program. Significant automation has been provided by computer codes in three areas of data input. These are the definition of coil geometry by a mesh of node points, the definition of finite elements via the node points and the definition of the node point force/displacement boundary conditions. The computer programs by name that have been used to perform the above functions are PDXNODE, ELEMENT and PDXFORC. The geometric finite element modeling options for toroidal field coils provided by PDXNODE include one-fourth or one-half symmetric sections of circular coils, oval shaped coils or dee-shaped coils with or without a beveled wedging surface. The program ELEMENT which defines the finite elements for input to the finite element computer code can provide considerable time and labor savings when defining the model of coils of non-uniform cross-section or when defining the model of coils whose material properties are different in the R and THETA directions due to the laminations of alternate epoxy and copper windings. The modeling features provided by the program ELEMENT have been used to analyze the PLT and the TFTR toroidal field coils with integral support structures. The computer program named PDXFORC is described. It computes the node point forces in a model of a toroidal field coil from the vector crossproduct of the coil current and the magnetic field. The model can be of one-half or one-fourth symmetry to be consistent with the node model defined by PDXNODE, and the magnetic field is computed from toroidal or poloidal coils

Self-similar solutions for toroidal magnetic fields in a turbulent jet

International Nuclear Information System (INIS)

Komissarov, S.S.; Ovchinnikov, I.L.

1989-01-01

Self-similar solutions for weak toroidal magnetic fields transported by a turbulent jet of incompressible fluid are obtained. It is shown that radial profiles of the self-similar solutions form a discrete spectrum of eigenfunctions of a linear differential operator. The strong depatures from the magnetic flux conservation law, used frequently in turbulent jet models for extragalactic radio sources, are found

Design and development of ITER high-frequency magnetic sensor

NARCIS (Netherlands)

Ma, Y.; Vayakis, G.; Begrambekov, L. B.; Cooper, J.J.; Duran, I.; Hirsch, M.; Laqua, H.P.; Moreau, Ph.; Oosterbeek, J.W.; Spuig, P.; Stange, T.; Walsh, M.

2016-01-01

High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in

Toroidal equilibrium of a non-neutral plasma with toroidal current, inertia and pressure

International Nuclear Information System (INIS)

Bhattacharyya, S.N.; Avinash, K.

1992-01-01

Equilibrium of non-neutral clouds in a toroidal vessel with toroidal magnetic field is demonstrated in the presence of a toroidal current, finite mass and finite pressure. With a toroidal current, it is shown that in a large-aspect-ratio conducting torus the equilibrium is governed by competition between forces produced by image charges and image currents. When μ 0 ε 0 E r 2 >B θ 2 (whe re E r and B θ are the self electrostatic and self magnetic fields of the cloud), the confinement is electrostatic and plasma shifts inwards; when μ 0 ε 0 E r 2 θ 2 , the confinement is magnetic and plasma shifts outwards. For μ 0 ε 0 E r 2 = B θ 2 there is no equilibrium. With finite mass or finite pressure, it is shown, in a large-aspect-ratio approximation, that the fluid drift surfaces and equipotential surfaces are displaced with respect to each other. In both cases the fluid drift surfaces are shifted inwards from the equipotential surfaces. (author)

TIBER-II TF [toroidal-field] winding pack design

International Nuclear Information System (INIS)

Kerns, J.A.; Miller, J.R.; Slack, D.S.; Summers, L.T.

1987-01-01

The superconducting, toroidal-field (TF) coils in the Tokamak Ignition/Burn Engineering Reactor (TIBER II) are designed with cable-in-conduit conductor (CICC) using Nb 3 Sn composite strands. To design the CICC winding pack, we used an optimization technique that maximizes the conductor stability without violating the constraints imposed by the structure, electrical insulation, quench protection, and fabrication technique. Detailed helium-properties codes calculate the heat removal along a flow path, and detailed field calculations determine the temperature, current, and stability margins. The conductor sheath is designed as distributed structure to partially support the combined in-plane and out-of-plane loads generated within the winding pack. Pancakes of the coil are wound, reacted, and insulated before being potted in the case. This design is aggressive but fully consistent with good engineering practice. 5 refs., 4 figs., 2 tabs

Resistive demountable toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Jassby, D.L.; Jacobsen, R.A.; Kalnavarns, J.; Masson, L.S.; Sekot, J.P.

1981-07-01

Readily demountable TF (toroidal-field) coils allow complete access to the internal components of a tokamak reactor for maintenance of replacement. The requirement of readily demountable joints dictates the use of water-cooled resistive coils, which have a host of decisive advantages over superconducting coils. Previous papers have shown that resistive TF coils for tokamak reactors can operate in the steady state with acceptable power dissipation (typically, 175 to 300 MW). This paper summarizes results of parametric studies of size optimization of rectangular TF coils and of a finite-element stress analysis, and examines several candidate methods of implementing demountable joints for rectangular coils constructed of plate segments

Form factor of some types of toroidal solenoids

International Nuclear Information System (INIS)

Koryavko, V.I.; Litvinenko, Yu.A.

1979-01-01

Obtained were the type of dependence between consumed power and formed field for toroidal helical-wound solenoids and the expression for the form factor analogous to the Fabry coefficient for cylindrical solenoids. Determined were optimum dimensions of the helical winding of ''forceless'' toroidal solenoids satisfying the condition of the formation of maximum field at minimum consumed power. Investigations also covered some types of conventional toroidal solenoids. Presented in the paper diagrams permitted to chose dimensions of the considered toroidal solenoids according to their consumed power and winding material volume

Steady state compact toroidal plasma production

Science.gov (United States)

Turner, William C.

1986-01-01

Apparatus and method for maintaining steady state compact toroidal plasmas. A compact toroidal plasma is formed by a magnetized coaxial plasma gun and held in close proximity to the gun electrodes by applied magnetic fields or magnetic fields produced by image currents in conducting walls. Voltage supply means maintains a constant potential across the electrodes producing an increasing magnetic helicity which drives the plasma away from a minimum energy state. The plasma globally relaxes to a new minimum energy state, conserving helicity according to Taylor's relaxation hypothesis, and injecting net helicity into the core of the compact toroidal plasma. Controlling the voltage so as to inject net helicity at a predetermined rate based on dissipative processes maintains or increases the compact toroidal plasma in a time averaged steady state mode.

Test of ITER conductors in SULTAN: An update

International Nuclear Information System (INIS)

Bruzzone, Pierluigi; Stepanov, Boris; Wesche, Rainer; Herzog, Robert; Calzolaio, Ciro; Vogel, Martin

2011-01-01

The ITER Toroidal Field (TF) conductor qualification phase has been carried out by testing short sample prototype conductors in the SULTAN test facility. This phase, started in 2007, has been substantially completed after minor adjustment of the conductor specification and test procedures. All the parties involved in the TF conductor procurement passed the qualification phase. Starting 2010, the samples for TF process qualification phase are tested in SULTAN. A summary of the results for all the ITER Qualification samples and an updated statistics are presented for the V-I and V-T characteristics of the cable-in-conduit conductors (CICC), including Nb 3 Sn and NbTi samples assembled with either a 'bottom joint' or a 'U-bend'. The technical improvements of the test facility are reported, including the enhanced cyclic loading rate and the calibration of the current meter. An outlook of the ITER conductor tests in the coming years is also presented.

The ITER poloidal field system

Energy Technology Data Exchange (ETDEWEB)

Wesley, J [General Atomics, San Diego, CA (USA); Beljakov, V; Kavin, A; Korshakov, V; Kostenko, A; Roshal, A; Zakharov, L [Kurchatov Inst. of Atomic Energy, Moscow (USSR); Bulmer, R; Kaiser, T; Miller, J R; Pearlstein, L D [Lawrence Livermore National Lab., CA (USA); Hogan, J [Oak Ridge National Lab., TN (USA); Kurihara, K; Shimomura, Y; Sugihara, M; Yoshino, R [Japan Atomic Energy Resea

1990-12-15

The ITER poloidal field (PF) system uses superconducting coils to provide the plasma equilibrium fields, slow equilibrium control and plasma flux linkage (V-s) needed for the ITER Operations and Research Program. Double-null (DN) divertor plasmas and operation scenarios for 22 MA Physics (high-Q/ignition) and 15 MA Technology (high-fluence testing) phases are provided. For 22 MA plasmas, total PF flux swing is 333 V-s. This provides inductive current drive (CD) for start-up with 66 V-s of resistive loss and 440-s (330-s minimum) sustained burn. The PF system also allows plasma start-up and shutdown scenarios, and can maintain the plasma configuration during burn over a range of current and pressure profiles. Other capabilities include increased plasma current (25 MA with inductive CD; 28 MA with non-inductive CD assist), divertor separatrix sweeping, and semi-DN and single-null plasmas.

The vacuum-arc plasma motion in a toroidal magnetic field

International Nuclear Information System (INIS)

Timoshenko, A.I.; Gnybida, M.V.; Taran, V.S.; Tereshin, V.I.; Chechelnitskij, O.G.

2005-01-01

The separation of the vacuum-arc plasma from macro-particles in the curvilinear plasma filters allows obtaining coatings with especially high characteristics. However, inside such filters the significant plasma losses also have been occurred. At the same time, increasing in the filter's efficiency is a difficult task without an effective mathematical model that really would describe the vacuum-arc plasma motion in a toroidal magnetic field. The description based on the flax-tube model was in fact only the first approximation in the decision of this problem. According to detailed flax-tube analysis of ions passage through the quarter torus plasma guide, the efficiency of the filter should grow up to 85% as the positive potential U, applied to the body of the plasma guide, is on the increase. However, the experiment showed that maximum of transparency reach up to ∼ 12%, at potential about of +18 Volts, and comes down under the further increase in potential. Such big digression from experiment does not justify the use of flux-tube model for designing of curvilinear plasma filters. We offer the new approach to the description of the vacuum-arc plasma motion in a toroidal magnetic field based on the solutions of steady-state (∂/∂t=0) Vlasov-Maxwell equations for the long plasma column aligned parallel to a constant axial magnetic field. The relations for the self-consistent electric polarization fields, which appear due to displacement of the electron component from ionic one on the curvilinear part of motion, were derived within a framework of the drift approximation. The dynamics of the central part of the plasma flow in the electric polarization fields was considered in detail. The displacement of the plasma flow at the output of the plasma guide was calculated for the carbon and titanium plasmas. The good agreement with the experimental data was obtained. (author)

A steady-state axisymmetric toroidal system

International Nuclear Information System (INIS)

Hirano, K.

1984-01-01

Conditions for achieving a steady state in an axisymmetric toroidal system are studied with emphasis on a very-high-beta field-reversed configuration. The analysis is carried out for the electromotive force produced by the Ohkawa current that is induced by neutral-beam injection. It turns out that, since the perpendicular component of the current j-vectorsub(perpendicular) to the magnetic field can be generated automatically by the diamagnetic effect, only the parallel component j-vectorsub(parallel) must be driven by the electromotive force. The drive of j-vectorsub(parallel) generates shear in the field line so that the pure toroidal field on the magnetic axis is rotated towards the plasma boundary and matched to the external field lines. This matching condition determines the necessary amount of injection beam current and power. It is demonstrated that a very-high-beta field-reversed configuration requires only a small amount of current-driving beam power because almost all the toroidal current except that close to the magnetic axis is carried by the diamagnetic current due to high beta. A low-beta tokamak, on the other hand, needs very high current-driving power since most of the toroidal current is composed of j-vectorsub(parallel) which must be driven by the beam. (author)

Toroidal asymmetries in divertor impurity influxes in NSTX

Directory of Open Access Journals (Sweden)

F. Scotti

2017-08-01

Full Text Available Toroidal asymmetries in divertor carbon and lithium influxes were observed in NSTX, due to toroidal differences in surface composition, tile leading edges, externally-applied three-dimensional (3D fields and toroidally-localized edge plasma modifications due to radio frequency heating. Understanding toroidal asymmetries in impurity influxes is critical for the evaluation of total impurity sources, often inferred from measurements with a limited toroidal coverage. The toroidally-asymmetric lithium deposition induced asymmetries in divertor lithium influxes. Enhanced impurity influxes at the leading edge of divertor tiles were the main cause of carbon toroidal asymmetries and were enhanced during edge localized modes. Externally-applied 3D fields led to strike point splitting and helical lobes observed in divertor impurity emission, but marginal changes to the toroidally-averaged impurity influxes. Power coupled to the scrape-off layer SOL plasma during radio frequency (RF heating of H-mode discharges enhanced impurity influxes along the non-axisymmetric divertor footprint of flux tubes connecting to plasma in front of the RF antenna.

Computational model for superconducting toroidal-field magnets for a tokamak reactor

International Nuclear Information System (INIS)

Turner, L.R.; Abdou, M.A.

1978-01-01

A computational model for predicting the performance characteristics and cost of superconducting toroidal-field (TF) magnets in tokamak reactors is presented. The model can be used to compare the technical and economic merits of different approaches to the design of TF magnets for a reactor system. The model has been integrated into the ANL Systems Analysis Program. Samples of results obtainable with the model are presented

Stress analysis of the conceptual design configurations of constant tension D-shaped superconducting toroidal field coils for TNS

International Nuclear Information System (INIS)

Fernades, R.; Smith, R.A.

1977-01-01

Conceptual design configurations of D-shaped toroidal field coils applicable to the TNS program are studied under the action of the toroidal field loading condition and the vertical field loading condition, but not the fault condition. Although the analysis is specific to an 8 Tesla design using a niobium titanium superconductor, the results can be extended to a coil with a different conductor material and subjected to a field of different magnitude provided the condition of linear elasticity is not violated. The analysis technique used is the finite element method, with three dimensional finite elements defined in the ANSYS computer code, and supplemented by closed form analytical solutions

Experimental studies of plasma confinement in toroidal systems

International Nuclear Information System (INIS)

Bodin, H.A.B.; Keen, B.E.

1977-01-01

In this article the closed-line magnetic field approach to the plasma isolation and confinement problem in toroidal systems is reviewed. The theoretical aspects of closed-line magnetic field systems, indicating that topologically such systems are toroidal, are surveyed under the headings; topology of closed-line systems, equilibrium in different configurations and classification of toroidal devices, MHD stability, non-ideal effects in MHD stability, microscopic stability, and plasma energy loss. A section covering the experimental results of plasma confinement in toroidal geometry considers Stellerators, Tokamaks, toroidal pinch -the reversed-field pinch, screw pinches and high-β Tokamaks, Levitrons and multipoles (internal-ring devices), and miscellaneous toroidal containment devices. Recent achievements and the present position are discussed with reference to the status of Tokamak research, low-β stellerator research and high-β research. It is concluded from the continuing progress made in this research that the criteria for the magnetic containment of plasmas can be met. Further, it is concluded that the construction of a successful and economic fusion reactor is within the scope of advancing science and technology. 250 references. (U.K.)

Experimental studies of plasma confinement in toroidal systems

Energy Technology Data Exchange (ETDEWEB)

Bodin, H A.B.; Keen, B E [UKAEA, Abingdon. Culham Lab.

1977-12-01

In this article the closed-line magnetic field approach to the plasma isolation and confinement problem in toroidal systems is reviewed. The theoretical aspects of closed-line magnetic field systems, indicating that topologically such systems are toroidal, are surveyed under the headings; topology of closed-line systems, equilibrium in different configurations and classification of toroidal devices, MHD stability, non-ideal effects in MHD stability, microscopic stability, and plasma energy loss. A section covering the experimental results of plasma confinement in toroidal geometry considers Stellerators, Tokamaks, toroidal pinch -the reversed-field pinch, screw pinches and high-..beta.. Tokamaks, Levitrons and multipoles (internal-ring devices), and miscellaneous toroidal containment devices. Recent achievements and the present position are discussed with reference to the status of Tokamak research, low-..beta.. stellerator research and high-..beta.. research. It is concluded from the continuing progress made in this research that the criteria for the magnetic containment of plasmas can be met. Further, it is concluded that the construction of a successful and economic fusion reactor is within the scope of advancing science and technology. 250 references.

Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

International Nuclear Information System (INIS)

Sannazzaro, G.; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H.

2001-01-01

In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses

Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Sannazzaro, G. E-mail: sannazg@itereu.de; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H

2001-11-01

In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses.

Magnetic field profiles during turbulent heating in a toroidal hydrogen plasma

International Nuclear Information System (INIS)

Kalfsbeek, H.W.

1978-12-01

A description is given of the measurements of both poloidal and toroidal magnetic field components as functions of radius and time in a small turbulently heated tokamak. These measurements have been carried out with an array of miniature pick-up coils, enclosed in a quartz tube which is inserted into the plasma. The electric fields inside the plasma, as well as the parallel resistivity profiles are deduced from the measured magnetic fields. The ohmically dissipated energy is determined from the field distributions and compared with the total input energy. The experimental results are compared with the outcome of a numerical model. The consistency with information obtained from other diagnostic measurements is checked. (Auth.)

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

OCLATOR (One Coil Low Aspect Toroidal Reactor)

International Nuclear Information System (INIS)

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 β 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

Effect of neoclassical toroidal viscosity on error-field penetration thresholds in tokamak plasmas.

Science.gov (United States)

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

2007-08-10

A model for field-error penetration is developed that includes nonresonant as well as the usual resonant field-error effects. The nonresonant components cause a neoclassical toroidal viscous torque that keeps the plasma rotating at a rate comparable to the ion diamagnetic frequency. The new theory is used to examine resonant error-field penetration threshold scaling in Ohmic tokamak plasmas. Compared to previous theoretical results, we find the plasma is less susceptible to error-field penetration and locking, by a factor that depends on the nonresonant error-field amplitude.

Design and issues of the ITER in-vessel components: ITER Joint central team and home teams

International Nuclear Information System (INIS)

Parker, R.R.

1998-01-01

This paper surveys the status of the design of the in-vessel components for ITER, in particular the major components, namely the vacuum vessel, blanket and first wall, and divertor, and the interface of selected ancillary systems such as those used for RF heating and current drive, and for diagnostics. The vacuum vessel is a double-walled structure constructed from two toroidal shells joined by ribs. The space between the skins is filled with shield plates directly cooled by water. The structural material is 316 LN IG (ITER grade). Toroidal supports joining the vessel midplane ports with the TF structure limit possible differential toroidal displacements, as might occur due to seismic or vertical displacement events (VDEs). A variety of load conditions corresponding to normal and off-normal loads have been considered and in all cases peak vessel stresses are within allowables. The blanket system consists of approximately 700 modules, each weighing ∝4 t. The integrated first wall consists of a beryllium-tiled copper mat bonded to the water-cooled SS shield block. The copper mat functions as a heat sink and has imbedded in it an array of SS tubes providing water cooling. The modules are mechanically attached to a toroidal backplate. Loads due to centered disruptions are reacted via hoop stress in the backplate, whereas net vertical and horizontal loads such as those arising from VDEs are transferred through the backplate and divertor supports to the vessel. (orig.)

Characterization of compact-toroid injection during formation, translation, and field penetration

Science.gov (United States)

Matsumoto, T.; Roche, T.; Allfrey, I.; Sekiguchi, J.; Asai, T.; Gota, H.; Cordero, M.; Garate, E.; Kinley, J.; Valentine, T.; Waggoner, W.; Binderbauer, M.; Tajima, T.

2016-11-01

We have developed a compact toroid (CT) injector system for particle refueling of the advanced beam-driven C-2U field-reversed configuration (FRC) plasma. The CT injector is a magnetized coaxial plasma gun (MCPG), and the produced CT must cross the perpendicular magnetic field surrounding the FRC for the refueling of C-2U. To simulate this environment, an experimental test stand has been constructed. A transverse magnetic field of ˜1 kG is established, which is comparable to the C-2U axial magnetic field in the confinement section, and CTs are fired across it. On the test stand we have been characterizing and studying CT formation, ejection/translation from the MCPG, and penetration into transverse magnetic fields.

Characterization of compact-toroid injection during formation, translation, and field penetration

Energy Technology Data Exchange (ETDEWEB)

Matsumoto, T., E-mail: cstd14003@g.nihon-u.ac.jp; Sekiguchi, J.; Asai, T. [Nihon University, Chiyoda-ku, Tokyo 101-8308 (Japan); Roche, T.; Allfrey, I.; Gota, H.; Cordero, M.; Garate, E.; Kinley, J.; Valentine, T.; Waggoner, W.; Binderbauer, M. [Tri Alpha Energy, Inc., P.O. Box 7010, Rancho Santa Margarita, California 92688 (United States); Tajima, T. [Tri Alpha Energy, Inc., P.O. Box 7010, Rancho Santa Margarita, California 92688 (United States); Department of Physics and Astronomy, University of California, Irvine, California 92697 (United States)

2016-11-15

We have developed a compact toroid (CT) injector system for particle refueling of the advanced beam-driven C-2U field-reversed configuration (FRC) plasma. The CT injector is a magnetized coaxial plasma gun (MCPG), and the produced CT must cross the perpendicular magnetic field surrounding the FRC for the refueling of C-2U. To simulate this environment, an experimental test stand has been constructed. A transverse magnetic field of ∼1 kG is established, which is comparable to the C-2U axial magnetic field in the confinement section, and CTs are fired across it. On the test stand we have been characterizing and studying CT formation, ejection/translation from the MCPG, and penetration into transverse magnetic fields.

Alternatives of ITER vacuum vessel support system

International Nuclear Information System (INIS)

Ohmori, Junji; Kitamura, Kazunori; Araki, Masanori; Ohno, Isamu; Shoji, Teruaki

2002-07-01

Optional designs of vacuum vessel (VV) support have been performed for the International Thermonuclear Experimental Reactor (ITER) to reduce stresses and buckling concern of the flexible plate structure in ITER-FDR. One of the optional designs is hanging type VV support concept that consists of top hanging supports at the top of VV and middle radial stoppers in the middle of outboard VV. The hanging supports are located at the toroidal field (TF) coil inboard top region (R∼5400 mm) using the narrow window space surrounded by a poloidal field coil (PF1) and TF coil. The radial stoppers are located inside TF coil cases in the TF coil outboard middle region (R∼9300 mm). The upper flange connection of the radial stoppers should slide in vertical direction to eliminate thermal stress produced by relative thermal displacement between VV wall and TF coil case. Both supports consist of flexible plates and are mounted on 18 locations in toroidal direction. The radial and toroidal reaction forces are shared with the hanging supports and the radial stoppers. However, the vertical force is sustained by only the hanging supports. The others are compressive type support concept that consists of nine VV supports located in alternate divertor port regions in toroidal direction. Two designs have been performed for the VV support concept. One is mounted on TF inter-coil structures (OIS) and the other is on cryostat ring. The compressive support on TF coil OIS is dependent on TF coil movement but that on cryostat is independent. In the optional designs, the bending stress due to the relative thermal displacement between TF coil and VV is classified to primary stress according to ASME Sec. III NF. The stress due to TF coil displacement is also considered as primary stress. The stress due to non-uniform temperature distribution of the flexible plate is classified to secondary stress. The preliminary structural assessments for the optional designs have been performed for all load

Evaluation of the toroidal torque driven by external non-resonant non-axisymmetric magnetic field perturbations in a tokamak

Energy Technology Data Exchange (ETDEWEB)

Kasilov, Sergei V. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, Technische Universität Graz Petersgasse 16, A–8010 Graz (Austria); Institute of Plasma Physics National Science Center “Kharkov Institute of Physics and Technology” ul. Akademicheskaya 1, 61108 Kharkov (Ukraine); Kernbichler, Winfried; Martitsch, Andreas F.; Heyn, Martin F. [Fusion@ÖAW, Institut für Theoretische Physik—Computational Physics, Technische Universität Graz Petersgasse 16, A–8010 Graz (Austria); Maassberg, Henning [Max-Planck Institut für Plasmaphysik, D-17491 Greifswald (Germany)

2014-09-15

The toroidal torque driven by external non-resonant magnetic perturbations (neoclassical toroidal viscosity) is an important momentum source affecting the toroidal plasma rotation in tokamaks. The well-known force-flux relation directly links this torque to the non-ambipolar neoclassical particle fluxes arising due to the violation of the toroidal symmetry of the magnetic field. Here, a quasilinear approach for the numerical computation of these fluxes is described, which reduces the dimension of a standard neoclassical transport problem by one without model simplifications of the linearized drift kinetic equation. The only limiting condition is that the non-axisymmetric perturbation field is small enough such that the effect of the perturbation field on particle motion within the flux surface is negligible. Therefore, in addition to most of the transport regimes described by the banana (bounce averaged) kinetic equation also such regimes as, e.g., ripple-plateau and resonant diffusion regimes are naturally included in this approach. Based on this approach, a quasilinear version of the code NEO-2 [W. Kernbichler et al., Plasma Fusion Res. 3, S1061 (2008).] has been developed and benchmarked against a few analytical and numerical models. Results from NEO-2 stay in good agreement with results from these models in their pertinent range of validity.

Prototyping and testing of the Continuous External Rogowski ITER magnetic sensor

Energy Technology Data Exchange (ETDEWEB)

Moreau, Ph., E-mail: philippe.jacques.moreau@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Le-Luyer, A.; Malard, P.; Pastor, P.; Saint-Laurent, F.; Spuig, P. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Lister, J.; Toussaint, M.; Marmillod, P.; Testa, D. [Centre de Recherches en Physique des Plasmas, EPFL (Switzerland); Peruzzo, S. [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Knaster, J. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); IFMIF EVEDA, Rokkasho (Japan); Vayakis, G.; Hughes, S.; Patel, K.M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2013-10-15

Highlights: ► ITER Continuous External Rogowski (CER) are designed for plasma and vacuum vessel current measurement. ► CER are located in the casing of Toroidal Field Coils and thus will operate at 4 K. ► The design of the sensors has been completed. ► CER prototypes have been manufactured by 2 suppliers. ► The R and D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER. -- Abstract: The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R and D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R and D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed.

Prototyping and testing of the Continuous External Rogowski ITER magnetic sensor

International Nuclear Information System (INIS)

Moreau, Ph.; Le-Luyer, A.; Malard, P.; Pastor, P.; Saint-Laurent, F.; Spuig, P.; Lister, J.; Toussaint, M.; Marmillod, P.; Testa, D.; Peruzzo, S.; Knaster, J.; Vayakis, G.; Hughes, S.; Patel, K.M.

2013-01-01

Highlights: ► ITER Continuous External Rogowski (CER) are designed for plasma and vacuum vessel current measurement. ► CER are located in the casing of Toroidal Field Coils and thus will operate at 4 K. ► The design of the sensors has been completed. ► CER prototypes have been manufactured by 2 suppliers. ► The R and D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER. -- Abstract: The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R and D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R and D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed

Challenges and status of ITER conductor production

International Nuclear Information System (INIS)

Devred, A; Backbier, I; Bessette, D; Bevillard, G; Gardner, M; Jong, C; Lillaz, F; Mitchell, N; Romano, G; Vostner, A

2014-01-01

Taking the relay of the large Hadron collider (LHC) at CERN, ITER has become the largest project in applied superconductivity. In addition to its technical complexity, ITER is also a management challenge as it relies on an unprecedented collaboration of seven partners, representing more than half of the world population, who provide 90% of the components as in-kind contributions. The ITER magnet system is one of the most sophisticated superconducting magnet systems ever designed, with an enormous stored energy of 51 GJ. It involves six of the ITER partners. The coils are wound from cable-in-conduit conductors (CICCs) made up of superconducting and copper strands assembled into a multistage cable, inserted into a conduit of butt-welded austenitic steel tubes. The conductors for the toroidal field (TF) and central solenoid (CS) coils require about 600 t of Nb 3 Sn strands while the poloidal field (PF) and correction coil (CC) and busbar conductors need around 275 t of Nb–Ti strands. The required amount of Nb 3 Sn strands far exceeds pre-existing industrial capacity and has called for a significant worldwide production scale up. The TF conductors are the first ITER components to be mass produced and are more than 50% complete. During its life time, the CS coil will have to sustain several tens of thousands of electromagnetic (EM) cycles to high current and field conditions, way beyond anything a large Nb 3 Sn coil has ever experienced. Following a comprehensive R and D program, a technical solution has been found for the CS conductor, which ensures stable performance versus EM and thermal cycling. Productions of PF, CC and busbar conductors are also underway. After an introduction to the ITER project and magnet system, we describe the ITER conductor procurements and the quality assurance/quality control programs that have been implemented to ensure production uniformity across numerous suppliers. Then, we provide examples of technical challenges that have been

Challenges and status of ITER conductor production

Science.gov (United States)

Devred, A.; Backbier, I.; Bessette, D.; Bevillard, G.; Gardner, M.; Jong, C.; Lillaz, F.; Mitchell, N.; Romano, G.; Vostner, A.

2014-04-01

Taking the relay of the large Hadron collider (LHC) at CERN, ITER has become the largest project in applied superconductivity. In addition to its technical complexity, ITER is also a management challenge as it relies on an unprecedented collaboration of seven partners, representing more than half of the world population, who provide 90% of the components as in-kind contributions. The ITER magnet system is one of the most sophisticated superconducting magnet systems ever designed, with an enormous stored energy of 51 GJ. It involves six of the ITER partners. The coils are wound from cable-in-conduit conductors (CICCs) made up of superconducting and copper strands assembled into a multistage cable, inserted into a conduit of butt-welded austenitic steel tubes. The conductors for the toroidal field (TF) and central solenoid (CS) coils require about 600 t of Nb3Sn strands while the poloidal field (PF) and correction coil (CC) and busbar conductors need around 275 t of Nb-Ti strands. The required amount of Nb3Sn strands far exceeds pre-existing industrial capacity and has called for a significant worldwide production scale up. The TF conductors are the first ITER components to be mass produced and are more than 50% complete. During its life time, the CS coil will have to sustain several tens of thousands of electromagnetic (EM) cycles to high current and field conditions, way beyond anything a large Nb3Sn coil has ever experienced. Following a comprehensive R&D program, a technical solution has been found for the CS conductor, which ensures stable performance versus EM and thermal cycling. Productions of PF, CC and busbar conductors are also underway. After an introduction to the ITER project and magnet system, we describe the ITER conductor procurements and the quality assurance/quality control programs that have been implemented to ensure production uniformity across numerous suppliers. Then, we provide examples of technical challenges that have been encountered and

Integration test of ITER full-scale vacuum vessel sector

International Nuclear Information System (INIS)

Nakahira, M.; Koizumi, K.; Oka, K.

2001-01-01

The full-scale Sector Model Project, which was initiated in 1995 as one of the Large Seven R and D Projects, completed all R and D activities planned in the ITER-EDA period with the joint effort of the ITER Joint Central Team (JCT), the Japanese, the Russian Federation (RF) and the United States (US) Home Teams. The fabrication of a full-scale 18 toroidal sector, which is composed of two 9 sectors spliced at the port center, was successfully completed in September 1997 with the dimensional accuracy of ± 3 mm for the total height and total width. Both sectors were shipped to the test site in JAERI and the integration test was begun in October 1997. The integration test involves the adjustment of field joints, automatic Narrow Gap Tungsten Inert Gas (NG-TIG) welding of field joints with splice plates, and inspection of the joint by ultrasonic testing (UT), which are required for the initial assembly of ITER vacuum vessel. This first demonstration of field joint welding and performance test on the mechanical characteristics were completed in May 1998 and the all results obtained have satisfied the ITER design. In addition to these tests, the integration with the mid plane port extension fabricated by the Russian Home Team, and the cutting and re-welding test of field joints by using full-remotized welding and cutting system developed by the US Home Team, are planned as post EDA activities. (author)

Iberdrola project engineering in the manufacture of the ITER superconducting coils

International Nuclear Information System (INIS)

Felipe, A.; Merino, A.

2012-01-01

ITER in a large-scale project that aims to demonstrate that it is possible to produce commercial energy from fusion. During its operational lifetime, ITER will test key technologies necessary for the next step: the demonstration fusion power plant that will prove that it is possible to capture fusion energy for commercial use. IBERDROLA Ingenieria y Construccion is the leader of a Consortium with ASG superconductors (Italy) and Elytt Energy (Spain) that is in charge of the manufacturing of one of the most relevant component: 10 Toroidal Field Coils. the development of this project presents significant technological challenges, where the main processes are the one related to high accuracy required during all manufacturing processes. (Author)

Status of Design and Manufacturing of ITER 1st batch Assembly Tools

Energy Technology Data Exchange (ETDEWEB)

Bae, Jin Ho; Nam, Kyoun Go; Chung, Si Kun; Ha, Min Su [ITER Korea National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Geun Hong [ITER Organization, St Paul lez Durance (France)

2016-05-15

The ITER tokamak assembly tools are purpose-built and specially designed to complete the ITER tokamak machine which includes; Vacuum Vessel (VV), VV Thermal Shield (VVTS), Toroidal Field Coil (TFC) and other components contained in the cryostat. KODA has carried out the preliminary and final design of these assembly tools. This paper shows that the current status, first quarter of the 2016, including manufacturing of ITER 1st batch assembly tools and briefly summarized the design process through design work of Sector Sub-assembly Tool (SSAT) that is most important tool representing ITER 1st batch assembly tools. KODA (Korea Domestic Agency) should provide 128 kinds of the purpose-built assembly tools for ITER Tokamak machine, and the ITER 1st batch assembly tools are split into 3 groups. The FDR for Group A was performed in December 2014, and design of SSAT has been verified by FE analysis and engineering calculation using EN cords. The SSAT is now under manufacturing phase to meet the ITER milestone. After factory acceptance test of SSAT on end of 2016, the 1st SSAT will be delivered and arrived in ITER site on second quarter of the 2017.

Compact toroid formation, compression, and acceleration

International Nuclear Information System (INIS)

Degnan, J.H.; Peterkin, R.E. Jr.; Baca, G.P.; Beason, J.D.; Bell, D.E.; Dearborn, M.E.; Dietz, D.; Douglas, M.R.; Englert, S.E.; Englert, T.J.; Hackett, K.E.; Holmes, J.H.; Hussey, T.W.; Kiuttu, G.F.; Lehr, F.M.; Marklin, G.J.; Mullins, B.W.; Price, D.W.; Roderick, N.F.; Ruden, E.L.; Sovinec, C.R.; Turchi, P.J.; Bird, G.; Coffey, S.K.; Seiler, S.W.; Chen, Y.G.; Gale, D.; Graham, J.D.; Scott, M.; Sommars, W.

1993-01-01

Research on forming, compressing, and accelerating milligram-range compact toroids using a meter diameter, two-stage, puffed gas, magnetic field embedded coaxial plasma gun is described. The compact toroids that are studied are similar to spheromaks, but they are threaded by an inner conductor. This research effort, named MARAUDER (Magnetically Accelerated Ring to Achieve Ultra-high Directed Energy and Radiation), is not a magnetic confinement fusion program like most spheromak efforts. Rather, the ultimate goal of the present program is to compress toroids to high mass density and magnetic field intensity, and to accelerate the toroids to high speed. There are a variety of applications for compressed, accelerated toroids including fast opening switches, x-radiation production, radio frequency (rf) compression, as well as charge-neutral ion beam and inertial confinement fusion studies. Experiments performed to date to form and accelerate toroids have been diagnosed with magnetic probe arrays, laser interferometry, time and space resolved optical spectroscopy, and fast photography. Parts of the experiment have been designed by, and experimental results are interpreted with, the help of two-dimensional (2-D), time-dependent magnetohydrodynamic (MHD) numerical simulations. When not driven by a second discharge, the toroids relax to a Woltjer--Taylor equilibrium state that compares favorably to the results of 2-D equilibrium calculations and to 2-D time-dependent MHD simulations. Current, voltage, and magnetic probe data from toroids that are driven by an acceleration discharge are compared to 2-D MHD and to circuit solver/slug model predictions. Results suggest that compact toroids are formed in 7--15 μsec, and can be accelerated intact with material species the same as injected gas species and entrained mass ≥1/2 the injected mass

ITER operational space for full plasma current H-mode operation

Energy Technology Data Exchange (ETDEWEB)

Mattei, M. [Assoc. Euratom-ENEA-CREATE, Seconda University di Napoli, Aversa (Italy)], E-mail: massimiliano.mattei@unirc.it; Cavinato, M.; Saibene, G.; Portone, A. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain); Albanese, R.; Ambrosino, G. [Assoc. Euratom-ENEA-CREATE, University Napoli Federico II, Napoli (Italy); Horton, L.D. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Kessel, C. [Princeton Plasma Physics Laboratory, Princeton University (United States); Koechl, F. [Assoc. EURATOM-OAW/ATI, Vienna (Austria); Lomas, P.J. [Euratom/UKAEA Fusion Assoc., Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Nunes, I. [Assoc. EURATOM/IST, Centro de Fusao Nuclear, Lisbon (Portugal); Parail, V. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Sartori, R. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain); Sips, A.C.C. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Thomas, P.R. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain)

2009-06-15

Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I{sub p}{sup 4}. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current I{sub p} of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7-0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.

Halo current and resistive wall simulations of ITER

International Nuclear Information System (INIS)

Strauss, H.R.; Zheng Linjin; Kotschenreuther, M.; Park, W.; Jardin, S.; Breslau, J.; Pletzer, A.; Paccagnella, R.; Sugiyama, L.; Chu, M.; Chance, M.; Turnbull, A.

2005-01-01

A number of ITER relevant problems in resistive MHD concern the effects of a resistive wall: vertical displacement events (VDE), halo currents caused by disruptions, and resistive wall modes. Simulations of these events have been carried out using the M3D code. We have verified the growth rate scaling of VDEs, which is proportional to the wall resistivity. Simulations have been done of disruptions caused by large inversion radius internal kink modes, as well as by nonlinear growth of resistive wall modes. Halo current flowing during the disruption has asymmetries with toroidal peaking factor up to about 3. VDEs have larger growth rates during disruption simulations, which may account for the loss of vertical feedback control during disruptions in experiments. Further simulations have been made of disruptions caused by resistive wall modes in ITER equilibria. For these modes the toroidal peaking factor is close to 1. Resistive wall modes in ITER and reactors have also been investigated utilizing the newly developed AEGIS (Adaptive EiGenfunction Independent Solution) linear full MHD code, for realistically shaped, fully toroidal equilibria. The AEGIS code uses an adaptive mesh in the radial direction which allows thin inertial layers to be accurately resolved, such as those responsible for the stabilization of resistive wall modes (RWM) by plasma rotation. Stabilization of resistive wall modes by rotation and wall thickness effects are examined. (author)

Kinetic Modifications to MHD Phenomena in Toroidal Plasmas

International Nuclear Information System (INIS)

Cheng, C.Z.; Gorelenkov, N.N.; Kramer, G.J.; Fredrickson, E.

2004-01-01

Particle kinetic effects involving small spatial and fast temporal scales can strongly affect MHD phenomena and the long time behavior of plasmas. In particular, kinetic effects such as finite ion gyroradii, trapped particle dynamics, and wave-particle resonances have been shown to greatly modify the stability of MHD modes. Here, the kinetic effects of trapped electron dynamics and finite ion gyroradii are shown to have a large stabilizing effect on kinetic ballooning modes in low aspect ratio toroidal plasmas such as NSTX [National Spherical Torus Experiment]. We also present the analysis of Toroidicity-induced Alfven Eigenmodes (TAEs) destabilized by fast neutral-beam injected ions in NSTX experiments and TAE stability in ITER due to alpha-particles and MeV negatively charged neutral beam injected ions

Modeling of divertor particle and heat loads during application of resonant magnetic perturbation fields for ELM control in ITER

Energy Technology Data Exchange (ETDEWEB)

Schmitz, O., E-mail: o.schmitz@fz-juelich.de [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Becoulet, M. [CEA/IRFM, Cadarache, 13108 St. Paul-lez-Durance Cedex (France); Cahyna, P. [IPP AS CR, Za Slovankou 3, 18200 Prague 8 (Czech Republic); Evans, T.E. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Feng, Y. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Frerichs, H.; Kirschner, A. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Kukushkin, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Laengner, R. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Lunt, T. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Reiser, D.; Reiter, D. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany); Saibene, G. [Fusion for Energy Joint Undertaking, Barcelona (Spain); Samm, U. [Forschungszentrum Jülich, IEK-4, Association EURATOM-FZJ, Jülich (Germany)

2013-07-15

First results from three-dimensional modeling of the divertor heat and particle flux pattern during application of resonant magnetic perturbation fields as ELM control scheme in ITER with the EMC3-Eirene fluid plasma and kinetic neutral transport code are discussed. The formation of a helical magnetic footprint breaks the toroidal symmetry of the heat and particle fluxes. Expansion of the flux pattern as far as 60 cm away from the unperturbed strike line is seen with vacuum RMP fields, resulting in a preferable heat flux spreading. Inclusion of plasma response reduces the radial extension of the heat and particle fluxes and results in a heat flux peaking closer to the unperturbed level. A strong reduction of the particle confinement is found. 3D flow channels are identified as a consistent reason due to direct parallel outflow from inside of the separatrix. Their radial inward expansion and hence the level of particle pump out is shown to be dependent on the perturbation level.

Commercial tokamak reactors with resistive toroidal field magnets

International Nuclear Information System (INIS)

Bombery, L.; Cohn, D.R.; Jassby, D.L.

1984-01-01

Scaling relations and design concepts are developed for commercial tokamak reactors that use watercooled copper toroidal field (TF) magnets. Illustrative parameters are developed for reactors that are scaled up in size from LITE test reactor designs, which use quasi-continuous copper plate magnets. Acceptably low magnet power requirements may be attainable in a moderate beta (β = 0.065) commercial reactor with a major radius of 6.2 m. The shielding thickness and magnet size are substantially reduced relative to values in commercial reactors with superconducting magnets. Operation at high beta (β = 0.14) leads to a reduction in reactor size, magnet-stored energy, and recirculating power. Reactors using resistive TF magnets could provide advantages of physically smaller devices, improved maintenance features, and increased ruggedness and reliability

Design of the ITER Tokamak Assembly Tools

International Nuclear Information System (INIS)

Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

2006-01-01

ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

The Linear Stability Properties of Medium- to High- n TAEs in ITER

Energy Technology Data Exchange (ETDEWEB)

Gorelenkov, N N; Budny, R V; Kessel, C E; Kramer, G J; McCune, D; Manickam, J; Nazikian, R

2008-02-14

This document provides a detailed report on the successful completion of the DOE OFES Theory Milestone for FY2007: Improve the simulation resolution of linear stability properties of Toroidal Alfvén Eigenmodes (TAE) driven by energetic particles and neutral beams in ITER by increasing the numbers of toroidal modes used to 15.

The forming of a superconductor cable during the winding of a large toroidal field coil

International Nuclear Information System (INIS)

Messemer, G.; Zehlein, H.

1984-01-01

The feasible range for the tension force which acts on a superconductor cable during the winding of a large D-shaped toroidal field coil depends strongly on the mechanical properties of the cable, on the geometry of the winding pack and on the arrangement of the equipment. The upper limit is imposed by possible damage within the cable. The lower limit is set by the need to assure enough compaction and to overcome the friction forces between the layers. Within this 'corridor' optimal control of elastic prestresses is desirable: this may be chosen with regard to the residual stresses and/or the elastic springback after removal of the coil former. This paper presents a simplified elastica conductor model built by a finite chain of intervals with constant bending moment and curvature. This paper describes the discrete model as well as the iterative shooting method, which finds the equilibrium shape of the conductor. The distributions of bending moment and shear forces around the D-shaped contour, as well as along the conductor, are given. Desirable improvements are outlined. In particular, the possibility of mitigating the stress concentration effect by supporting rollers suitably placed along the 'free' conductor near the bobbin is discussed. (author)

ITER technology R and D during the EDA

International Nuclear Information System (INIS)

Mizoguchi, T.

2001-01-01

A short overview of the ITER technology R and D achievements is presented. It includes R and D programme in the area of superconducting magnets, L-1 central solenoid model coil, L-2 toroidal field model coil, L-3 vacuum vessel sector, L-4 blanket module, L-5 divertor cassette, L-6 blanket and L-7 divertor remote handling systems. In addition to the seven large R and D projects, development of components for fuelling, pumping, tritium processing, heating/current drive, power supplies and plasma diagnostics, as well as safety-related R and D have significantly progressed

World's largest DC flywheel generator for the toroidal field power supply of JAERI's JFT-2M Tokamak nuclear fusion reactor

International Nuclear Information System (INIS)

Tani, Takashi; Nakanishi, Yuji; Horita, Tsuyoshi; Kawase, Chiharu; Oyabu, Isao; Kishimoto, Takeshi.

1996-01-01

Mitsubishi Electric has delivered the world's largest DC generator for the toroidal field coil power supply of the JFT-2M Tokamak at the Japan Atomic Energy Research Institute. The unit rotates at 225 or 460 rpm, providing a maximum rated output of 2,700 V, 19,000 A and 51.3 MW. The toroidal field is a DC field, so use of a DC generator permits a simpler design consuming less floor space than an AC drive system. The generator was manufactured following extensive studies on commutation, mechanical strength and insulation. (author)

Heating of toroidal plasmas by neutral injection

International Nuclear Information System (INIS)

Stix, T.H.

1971-08-01

This paper presents a brief review of the physics of ion acceleration, charge exchange and ionization, trajectories for fast ions in toroidal magnetic fields, and fast-ion thermalization. The injection of fast atoms is found to be a highly competitive method both for heating present-day experimental toroidal plasmas and for bringing full-scale toroidal CTR plasmas to low-density ignition. 13 refs., 9 figs

Plasma burn-through simulations using the DYON code and predictions for ITER

International Nuclear Information System (INIS)

Kim, Hyun-Tae; Sips, A C C; De Vries, P C

2013-01-01

This paper will discuss simulations of the full ionization process (i.e. plasma burn-through), fundamental to creating high temperature plasma. By means of an applied electric field, the gas is partially ionized by the electron avalanche process. In order for the electron temperature to increase, the remaining neutrals need to be fully ionized in the plasma burn-through phase, as radiation is the main contribution to the electron power loss. The radiated power loss can be significantly affected by impurities resulting from interaction with the plasma facing components. The DYON code is a plasma burn-through simulator developed at Joint European Torus (JET) (Kim et al and EFDA-JET Contributors 2012 Nucl. Fusion 52 103016, Kim, Sips and EFDA-JET Contributors 2013 Nucl. Fusion 53 083024). The dynamic evolution of the plasma temperature and plasma densities including the impurity content is calculated in a self-consistent way using plasma wall interaction models. The recent installation of a beryllium wall at JET enabled validation of the plasma burn-through model in the presence of new, metallic plasma facing components. The simulation results of the plasma burn-through phase show a consistent good agreement against experiments at JET, and explain differences observed during plasma initiation with the old carbon plasma facing components. In the International Thermonuclear Experimental Reactor (ITER), the allowable toroidal electric field is restricted to 0.35 (V m −1 ), which is significantly lower compared to the typical value (∼1 (V m −1 )) used in the present devices. The limitation on toroidal electric field also reduces the range of other operation parameters during plasma formation in ITER. Thus, predictive simulations of plasma burn-through in ITER using validated model is of crucial importance. This paper provides an overview of the DYON code and the validation, together with new predictive simulations for ITER using the DYON code. (paper)

Design and technical status of the EU contribution to ITER

International Nuclear Information System (INIS)

Gasparotto, Maurizio; Federici, Gianfranco; Casci, Federico Riccardo

2009-01-01

Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R and D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost. This paper provides an update of the design and technical status of EU contributions to ITER.

Electromagnetic torque on the toroidal plasma and the error-field induced torque

International Nuclear Information System (INIS)

Pustovitov, V. D.

2007-01-01

The electromagnetic torque on the toroidal plasma is calculated assuming a linear plasma response to the applied perturbation, which may be the error field or the field created by the correction coils, or both. The result is compared with recently published expressions for the error field induced torque (Zheng et al 2006 Nucl. Fusion 46 L9, Zheng and Kotschenreuther 2006 Phys. Rev. Lett. 97 165001), and the conclusions of these papers are revised. We resolve the problem of the torque resonance raised there. It is shown that the strong increase in the torque due to the static error field must occur at the resistive wall mode stability limit and not at the no-wall stability limit

Low-n shear Alfven spectra in axisymmetric toroidal plasmas

International Nuclear Information System (INIS)

Cheng, C.Z.; Chance, M.S.

1985-11-01

In toroidal plasmas, the toroidal magnetic field is nonuniform over a magnetic surface and causes coupling of different poloidal harmonics. It is shown both analytically and numerically that the toroidicity not only breaks up the shear Alfven continuous spectrum, but also creates new, discrete, toroidicity-induced shear Alfven eigenmodes with frequencies inside the continuum gaps. Potential applications of the low-n toroidicity-induced shear Alfven eigenmodes on plasma heating and instabilities are addressed. 17 refs., 4 figs

Dynamic ELM and divertor control using resonant toroidal multi-mode magnetic fields in DIII-D and EAST

Science.gov (United States)

Sun, Youwen

2017-10-01

A rotating n = 2 Resonant Magnetic Perturbation (RMP) field combined with a stationary n = 3 RMP field has validated predictions that access to ELM suppression can be improved, while divertor heat and particle flux can also be dynamically controlled in DIII-D. Recent observations in the EAST tokamak indicate that edge magnetic topology changes, due to nonlinear plasma response to magnetic perturbations, play a critical role in accessing ELM suppression. MARS-F code MHD simulations, which include the plasma response to the RMP, indicate the nonlinear transition to ELM suppression is optimized by configuring the RMP coils to drive maximal edge stochasticity. Consequently, mixed toroidal multi-mode RMP fields, which produce more densely packed islands over a range of additional rational surfaces, improve access to ELM suppression, and further spread heat loading on the divertor. Beneficial effects of this multi-harmonic spectrum on ELM suppression have been validated in DIII-D. Here, the threshold current required for ELM suppression with a mixed n spectrum, where part of the n = 3 RMP field is replaced by an n = 2 field, is smaller than the case with pure n = 3 field. An important further benefit of this multi-mode approach is that significant changes of 3D particle flux footprint profiles on the divertor are found in the experiment during the application of a rotating n = 2 RMP field superimposed on a static n = 3 RMP field. This result was predicted by modeling studies of the edge magnetic field structure using the TOP2D code which takes into account plasma response from MARS-F code. These results expand physics understanding and potential effectiveness of the technique for reliably controlling ELMs and divertor power/particle loading distributions in future burning plasma devices such as ITER. Work supported by USDOE under DE-FC02-04ER54698 and NNSF of China under 11475224.

Toroidal Trivelpiece-Gould modes

International Nuclear Information System (INIS)

Stoessel, F.P.

1979-01-01

Electron plasma waves are treated in quasi-electrostatic approximation in a toroidal cavity of rectangular cross-section in an infinitely strong azimuthal magnetic field. The differential equation for the electrostatic potential, derived from fluid equations, can be separated using cylindrical coordinates. The eigenvalue problem for the radial dependence is solved numerically by a shooting method. Eigenvalues are given for different aspect ratios. Comparison with appropriate modes of the straight geometry shows that the toroidal frequencies generally lie some percent above those for the straight case. Plots of the eigenfunctions demonstrate clearly the influence of toroidicity. The deviation from symmetry (which should appear for straight geometry) depends not only on the aspect ratio but also strongly on the mode numbers. (author)

Motion of a compact toroid inside a cylindrical flux conserver

Energy Technology Data Exchange (ETDEWEB)

Jarboe, T.R.; Henins, I.; Hoida, H.W.; Linford, R.K.; Marshall, J.; Platts, D.A.; Sherwood, A.R.

1980-10-13

Compact toroids have been generated in a cylindrical resistive flux conserver. They are observed to rotate so that their major axis is perpendicular to the axis of the flux conserver. Subsequently they remain stationary and their magnetic fields decay with a time constant of about 100 ..mu..s. This is the first observation of the predicted tipping mode and its saturation when no external fields are present. The compact toroids contain toroidal fields and are initially prolate in shape.

Motion of a compact toroid inside a cylindrical flux conserver

International Nuclear Information System (INIS)

Jarboe, T.R.; Henins, I.; Hoida, H.W.; Linford, R.K.; Marshall, J.; Platts, D.A.; Sherwood, A.R.

1980-01-01

Compact toroids have been generated in a cylindrical resistive flux conserver. They are observed to rotate so that their major axis is perpendicular to the axis of the flux conserver. Subsequently they remain stationary and their magnetic fields decay with a time constant of about 100 μs. This is the first observation of the predicted tipping mode and its saturation when no external fields are present. The compact toroids contain toroidal fields and are initially prolate in shape

Suspension of a field-cooled BiPbSrCaCuO high-T sub c superconductor under a toroidal permanent magnet

CERN Document Server

Lee, S H; Choe, W; Lee, T S

2002-01-01

Magnetic flux measurements of a toroidal magnet revealed a concave-shaped field distribution with a single minimum and a null field along the axis of the torus at the point where the field reversed. The non-linear magnetic field of the toroidal magnet perpendicular to the Ag sub 2 O-doped superconducting disc sample with trapped magnetic flux distorted the field line distribution. As a result, the interaction force between the magnet and the sample exhibited regions of repulsive, null, attractive, null and finally repulsive force. The asymmetrical concave-shaped force pattern along the axis with two null force points indicates that the force exerted on the sample changes direction, the transition from repulsive to attractive at the null force point, and the force becomes repulsive again beyond the second null force point as the distance along the axis increases. The magnetic field simulation using the Poisson numerical code for the toroidal magnet of 46 mm OD, 12 mm ID and 10 mm thickness was in close agreeme...

Current, temperature and confinement time scaling in toroidal reversed-field pinch experiments ZT-I and ZT-S

International Nuclear Information System (INIS)

Baker, D.A.; Burkhardt, L.C.; Di Marco, J.N.; Haberstich, A.; Hagenson, R.L.; Howell, R.B.; Karr, H.J.; Schofield, A.E.

1977-01-01

The scaling properties of a toroidal reversed-field Z pinch have been investigated over a limited range by comparing two experiments having conducting walls and discharge-tube minor diameters which differ by a factor of approximately 1.5. Both the confinement time of the plasma column and the electron temperature were found to increase about a factor of two with the increased minor diameter. Both the poloidal field diffusion and the decay of the toroidal reversed field were significantly reduced with the larger tube diameter. These results support the hypothesis that the loss of stability later in the discharge is caused by diffusion-induced deterioration of initially favourable plasma-field profiles to MHD unstable ones. This conclusion has been verified by stability analysis of the magnetic field profiles. Fusion reactor calculations show that small reactors are conceptually possible assuming good containment can be achieved for current densities approximately >20MAm -2 . (author)

Modeling of ITER edge plasma in the presence of resonant magnetic perturbations

Energy Technology Data Exchange (ETDEWEB)

Rozhansky, V.; Kaveeva, E.; Veselova, I.; Voskoboynikov, S. [Peter the Great St. Petersburg Polytechnic University, St. Petersburg (Russian Federation); Coster, D. [Max-Planck Institut fur Plasmaphysik, EURATOM Association, Garching (Germany)

2016-08-15

The modeling of the ITER edge is performed with the use of the code B2SOLPS5.2 in the presence of the electron conductivity caused by RMPs as well as for the reference case with the same input parameters but without RMPs. The radial electric field close to the neoclassical one is obtained without RMPs. Even the modest level of RMPs changes the direction of the electric field and causes the toroidal spin-up of the edge plasma. At the same time the pump-out effect is small. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Escape of magnetic toroids from the Sun

International Nuclear Information System (INIS)

Bieber, John W.; Rust, David M.

1996-01-01

Analysis of heliospheric magnetic fields at 1 AU shows that 10 24 Mx of net toroidal flux escapes from the Sun per solar cycle. This rate is compared with the apparent rate of flux emergence at the solar surface, and it is concluded that escaping toroids will remove at least 20% of the emerging flux, and may remove as much as 100% of emerging flux if multiple eruptions occur on the toroids. The data imply that flux escapes the Sun with an efficiency far exceeding Parker's upper limit estimate of 3%. Toroidal flux escape is almost certainly the source of the observed overwinding of the interplanetary magnetic field spiral. Two mechanisms to facilitate net flux escape are discussed: helicity charging to push open the fields and flux transport with reconnection to close them off. We estimate the Sun will shed ∼2x10 45 Mx 2 of magnetic helicity per solar cycle, leading to a mean helicity density of 100 Mx 2 cm -3 at 1 AU, which agrees well with observations

Next generation toroidal devices

International Nuclear Information System (INIS)

Yoshikawa, Shoichi

1998-10-01

A general survey of the possible approach for the next generation toroidal devices was made. Either surprisingly or obviously (depending on one's view), the technical constraints along with the scientific considerations lead to a fairly limited set of systems for the most favorable approach for the next generation devices. Specifically if the magnetic field strength of 5 T or above is to be created by superconducting coils, it imposes minimum in the aspect ratio for the tokamak which is slightly higher than contemplated now for ITER design. The similar technical constraints make the minimum linear size of a stellarator large. Scientifically, it is indicated that a tokamak of 1.5 times in the linear dimension should be able to produce economically, especially if a hybrid reactor is allowed. For the next stellarator, it is strongly suggested that some kind of helical axis is necessary both for the (almost) absolute confinement of high energy particles and high stability and equilibrium beta limits. The author still favors a heliac most. Although it may not have been clearly stated in the main text, the stability afforded by the shearless layer may be exploited fully in a stellarator. (author)

Next generation toroidal devices

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Shoichi [Princeton Plasma Physics Lab., Princeton Univ., NJ (United States)

1998-10-01

A general survey of the possible approach for the next generation toroidal devices was made. Either surprisingly or obviously (depending on one`s view), the technical constraints along with the scientific considerations lead to a fairly limited set of systems for the most favorable approach for the next generation devices. Specifically if the magnetic field strength of 5 T or above is to be created by superconducting coils, it imposes minimum in the aspect ratio for the tokamak which is slightly higher than contemplated now for ITER design. The similar technical constraints make the minimum linear size of a stellarator large. Scientifically, it is indicated that a tokamak of 1.5 times in the linear dimension should be able to produce economically, especially if a hybrid reactor is allowed. For the next stellarator, it is strongly suggested that some kind of helical axis is necessary both for the (almost) absolute confinement of high energy particles and high stability and equilibrium beta limits. The author still favors a heliac most. Although it may not have been clearly stated in the main text, the stability afforded by the shearless layer may be exploited fully in a stellarator. (author)

An iterative reconstruction of cosmological initial density fields

Science.gov (United States)

Hada, Ryuichiro; Eisenstein, Daniel J.

2018-05-01

We present an iterative method to reconstruct the linear-theory initial conditions from the late-time cosmological matter density field, with the intent of improving the recovery of the cosmic distance scale from the baryon acoustic oscillations (BAOs). We present tests using the dark matter density field in both real and redshift space generated from an N-body simulation. In redshift space at z = 0.5, we find that the reconstructed displacement field using our iterative method are more than 80% correlated with the true displacement field of the dark matter particles on scales k < 0.10h Mpc-1. Furthermore, we show that the two-point correlation function of our reconstructed density field matches that of the initial density field substantially better, especially on small scales (<40h-1 Mpc). Our redshift-space results are improved if we use an anisotropic smoothing so as to account for the reduced small-scale information along the line of sight in redshift space.

PC-based package for interactive assessment of MHD equilibrium and poloidal field coil design in axisymmetric toroidal geometry

International Nuclear Information System (INIS)

Kelleher, W.P.

1987-01-01

In the assessment of Magnetohydrodynamic (MHD) equilibrium and Poloidal Field Coil (PFC) arrangement for toroidal axisymmetric geometry, the Grad-Shafranov equation must be solved, either analytically or numerically. Existing numerical tools have been developed primarily for mainframe usage and can prove cumbersome for screening assessments and parametric evaluations. The objective of this thesis was to develop a personal computer (PC)-based calculational tool for assessing MHD/PFC problems in a highly interactive mode, well suited for scoping studies. The approach adopted involves a two-step process: first the MHD equilibrium is calculated and then the PFC arrangement, consistent with the equilibrium, is determined in an interactive design environment. The PC-based system developed consists of two programs: (1) PCEQ, which solve the MHD equilibrium problem and (2) PFDE-SIGN, which is employed to arrive at a PFC arrangement. PCEQ provides an output file including, but not limited to, the following: poloidal beta, total beta, safety factors, q, on axis and on edge. PCEQ plots the following contours and/or profiles: flux, pressure and toroidal current density, safety factor, and ratio of plasma toroidal field to vacuum field

A method for external measurement of toroidal equilibrium parameters

International Nuclear Information System (INIS)

Brunsell, P.; Hellblom, G.; Brynolf, J.

1992-01-01

A method has been developed for determining from external magnetic field measurements the horizontal shift, the vertical shift and the poloidal field asymmetry parameter (Λ) of a toroidal plasma in force equilibrium. The magnetic measurements consist of two toroidal differential flux loops, giving the average vertical magnetic field and the average radial magnetic field respectively, together with cosine-coils for obtaining the m=1 cosine harmonic of the external poloidal magnetic field component. The method is used to analyse the evolution of the toroidal equilibrium during reversed-field pinch discharges in the Extrap T1-U device. We find that good equilibrium control is needed for long plasma pulses. For non-optimized externally applied vertical fields, the diagnostic clearly shows a horizontal drift motion of the pinch resulting in earlier discharge termination. (au)

System design of toroidal field power supply of CDD tokamak

Energy Technology Data Exchange (ETDEWEB)

Liu, Zheng Zhi

1996-12-01

This report deals with system design of Toroidal Field Power Supply of CDD tokamak (CDD-TFPS). The general design philosophy and design variations are introduced. After the outline of CDD-TFPS, the short-circuit calculation, the evaluation of converter parameters, the compatibility of converter and line are carried out. the specifications of major components, semi-conductor devices and accessories are given. High attention is paid to protection system. The design of sub-control and grounding system are described too. Some more general material for power supply design are attached in appendices for reference. (author). 30 tabs., 21 figs.

System design of toroidal field power supply of CDD tokamak

International Nuclear Information System (INIS)

Liu, Zheng Zhi.

1996-12-01

This report deals with system design of Toroidal Field Power Supply of CDD tokamak (CDD-TFPS). The general design philosophy and design variations are introduced. After the outline of CDD-TFPS, the short-circuit calculation, the evaluation of converter parameters, the compatibility of converter and line are carried out. the specifications of major components, semi-conductor devices and accessories are given. High attention is paid to protection system. The design of sub-control and grounding system are described too. Some more general material for power supply design are attached in appendices for reference. (author). 30 tabs., 21 figs

European Technological Effort in Preparation of ITER Construction

International Nuclear Information System (INIS)

Andreani, Roberto

2005-01-01

Europe has started since the '80s with the preparatory work done on NET, the Next European Torus, the successor of JET, to prepare for the construction of the next generation experiment on the road to the fusion reactor. In 2000 the European Fusion Development Agreement (EFDA) has been signed by sixteen countries, including Switzerland, not a member of the Union. Now the signatory countries have increased to twenty-five. A vigorous programme of design and R and D in support of ITER construction has been conducted by EFDA through the coordinated effort of the national institutes and laboratories supported financially, in the framework of the VI European Framework Research Programme (2002-2006), by contracts of association with EURATOM. In the last three years, with the expenditure of 160 M[Euro], the accent has been particularly put on the preparation of the industrial manufacturing activities of components and systems for ITER. Prototypes and manufacturing methods have been developed in all the main critical areas of machine construction with the objective of providing sound and effective solutions: vacuum vessel, toroidal field coils, poloidal field coils, remote handling equipment, plasma facing components and divertor components, electrical power supplies, generators and power supplies for the Heating and Current Drive Systems and other minor subsystems.Europe feels to be ready to host the ITER site and to provide adequate support and guidance for the success of construction to our partners in the ITER collaboration, wherever needed

The ITER poloidal field configuration and operation scenario

International Nuclear Information System (INIS)

Gribov, Y.; Portone, A.; Mondino, P.L.

1995-01-01

The ITER Poloidal Field (PF) system must satisfy the following requirements. (1) ITER must have a well-controlled, single null divertor magnetic configuration with nominal plasma current 21MA and moderate plasma elongation k95 < 1.65. (2) For a variety of plasma scenarios the ITER PF system must provide: inductive breakdown and start-up in an expanding-aperture limiter configuration near the outboard first wall; an inductive current ramp-up to the nominal plasma current with a reasonable assumption of resistive loss during current ramp-up; a pulse length of 1,000s for ignition and inductively-sustained burn at nominal plasma current; plasma shutdown (following fusion power termination) in a similar contracting-aperture limiter configuration. The present design of the PF system can satisfy the ITER requirements within specified limitations

Progress and achievements of R&D activities for the ITER vacuum vessel

Science.gov (United States)

Nakahira, M.; Takahashi, H.; Koizumi, K.; Onozuka, M.; Ioki, K.

2001-04-01

The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R&D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18° toroidal sector, which is composed of two 9° sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of +/-3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R&D activities for the ITER vacuum vessel.

The ITER magnets: Preparation for full size construction based on the results of the model coil programme

International Nuclear Information System (INIS)

Huguet, M.

2003-01-01

The ITER magnets are long-lead time items and the preparation of their construction is the subject of a major and coordinated effort of the ITER International Team and Participant Teams. The results of the ITER model coil programme constitute the basis and the main source of data for the preparation of the technical specifications for the procurement of the ITER magnets. A review of the salient results of the ITER model coil programme is given and the significance of these results for the preparation of full size industrial production is explained. The model coil programme has confirmed the validity of the design and the manufacturer's ability to produce the coils with the required quality level. The programme has also allowed the optimisation of the conductor design and the identification of further development which would lead to cost reductions of the toroidal field coil case. (author)

Hamilton-Jacobi theory for continuation of magnetic field across a toroidal surface supporting a plasma pressure discontinuity

International Nuclear Information System (INIS)

McGann, M.; Hudson, S.R.; Dewar, R.L.; Nessi, G. von

2010-01-01

The vanishing of the divergence of the total stress tensor (magnetic plus kinetic) in a neighborhood of an equilibrium plasma containing a toroidal surface of discontinuity gives boundary and jump conditions that strongly constrain allowable continuations of the magnetic field across the surface. The boundary conditions allow the magnetic fields on either side of the discontinuity surface to be described by surface magnetic potentials, reducing the continuation problem to that of solving a Hamilton-Jacobi equation. The characteristics of this equation obey Hamiltonian equations of motion, and a necessary condition for the existence of a continued field across a general toroidal surface is that there exist invariant tori in the phase space of this Hamiltonian system. It is argued from the Birkhoff theorem that existence of such an invariant torus is also, in general, sufficient for continuation to be possible. An important corollary is that the rotational transform of the continued field on a surface of discontinuity must, generically, be irrational.

Efficiency of wave-driven rigid body rotation toroidal confinement

Science.gov (United States)

Rax, J. M.; Gueroult, R.; Fisch, N. J.

2017-03-01

The compensation of vertical drifts in toroidal magnetic fields through a wave-driven poloidal rotation is compared with compensation through the wave driven toroidal current generation to support the classical magnetic rotational transform. The advantages and drawbacks associated with the sustainment of a radial electric field are compared with those associated with the sustainment of a poloidal magnetic field both in terms of energy content and power dissipation. The energy content of a radial electric field is found to be smaller than the energy content of a poloidal magnetic field for a similar set of orbits. The wave driven radial electric field generation efficiency is similarly shown, at least in the limit of large aspect ratio, to be larger than the efficiency of wave-driven toroidal current generation.

Barrel Toroid fully charged to nominal field, and it works!

CERN Multimedia

Herman ten Kate

After a few weeks of testing up to intermediate currents, finally, on Thursday evening November 9, the current in the Barrel Toroid was pushed up to its nominal value of 20500 A and even 500 A beyond this value to prove that we have some margin. It went surprisingly well. Of course, the 8 coils forming the toroid were already tested individually at the surface but still, some surprise may have come from those parts added to the toroid in the cavern for the first time like the 8 cryoring sections linking the coils as well as the valve box at the bottom in sector 13 regulating the helium flow or the current lead cryostat on the top in sector 5. No training quenches, nothing to worry about, and the test was concluded with a fast dump triggered at 00:40 in the very early morning of November 10. (left) The toroid current during the evening and night of November 9. (right) The test crew oscillated between fear and hope while looking at the control panels as the current approached 21kA. Big relief was in the...

Rotation and toroidal magnetic field effects on the stability of two-component jets

Science.gov (United States)

Millas, Dimitrios; Keppens, Rony; Meliani, Zakaria

2017-09-01

Several observations of astrophysical jets show evidence of a structure in the direction perpendicular to the jet axis, leading to the development of 'spine and sheath' models of jets. Most studies focus on a two-component jet consisting of a highly relativistic inner jet and a slower - but still relativistic - outer jet surrounded by an unmagnetized environment. These jets are believed to be susceptible to a relativistic Rayleigh-Taylor-type instability, depending on the effective inertia ratio of the two components. We extend previous studies by taking into account the presence of a non-zero toroidal magnetic field. Different values of magnetization are examined to detect possible differences in the evolution and stability of the jet. We find that the toroidal field, above a certain level of magnetization σ, roughly equal to 0.01, can stabilize the jet against the previously mentioned instabilities and that there is a clear trend in the behaviour of the average Lorentz factor and the effective radius of the jet when we continuously increase the magnetization. The simulations are performed using the relativistic MHD module from the open source, parallel, grid adaptive, mpi-amrvac code.

Superconducting toroidal field coil current densities for the TFCX

International Nuclear Information System (INIS)

Kalsi, S.S.; Hooper, R.J.

1985-04-01

A major goal of the Tokamak Fusion Core Experiment (TFCX) study was to minimize the size of the device and achieve lowest cost. Two key factors influencing the size of the device employing superconducting magnets are toroidal field (TF) winding current density and its nuclear heat load withstand capability. Lower winding current density requires larger radial build of the winding pack. Likewise, lower allowable nuclear heating in the winding requires larger shield thickness between the plasma and coil. In order to achieve a low-cost device, it is essential to maximize the winding's current density and nuclear heating withhstand capability. To meet the above objective, the TFCX design specification adopted as goals a nominal winding current density of 3500 A/cm 2 with 10-T peak field at the winding and peak nuclear heat load limits of 1 MW/cm 3 for the nominal design and 50 MW/cm 3 for an advanced design. This study developed justification for these current density and nuclear heat load limits

Compact-Toroid fusion reactor based on the field-reversed theta pinch

International Nuclear Information System (INIS)

Hagenson, R.L.; Krakowski, R.A.

1981-03-01

Early scoping studies based on approximate, analytic models have been extended on the basis of a dynamic plasma model and an overall systems approach to examine a Compact Toroid (CTOR) reactor embodiment that uses a Field-Reversed Theta Pinch as a plasma source. The field-reversed plasmoid would be formed and compressionally heated to ignition prior to injection into and translation through a linear burn chamber, thereby removing the high-technology plasmoid source from the hostile reactor environment. Stabilization of the field-reversed plasmoid would be provided by a passive conducting shell located outside the high-temperature blanket but within the low-field superconducting magnets and associated radiation shielding. On the basis of this batch-burn but thermally steady-state approach, a reactor concept emerges with a length below approx. 40 m that generates 300 to 400 MWe of net electrical power with a recirculating power fraction less than 0.15

Fabrication of the KSTAR toroidal field coil structure

International Nuclear Information System (INIS)

Choi, C.H.; Sa, J.W.; Park, H.K.; Hong, K.H.; Shin, H.; Kim, H.T.; Bak, J.S.; Lee, G.S.; Kwak, J.H.; Moon, H.G.; Yoon, H.H.; Lee, J.W.; Lee, S.K.; Song, J.Y.; Nam, K.M.; Byun, S.E.; Kim, H.C.; Ha, E.T.; Ahn, H.J.; Kim, D.S.; Lee, J.S.; Park, K.H.; Hong, C.D.

2005-01-01

The KSTAR toroidal field (TF) coil structure is under fabrication upon completion of engineering design and prototype construction. The prototype TF coil structure has been fabricated within allowable tolerances. Encasing of the prototype TF coil (TF00) in the prototype structure has been carried out through major processes involving a coil encasing, an enclosing weld, a vacuum pressure impregnation, and an outer surface machining. During the enclosing weld of the TF00 coil structure, we have measured temperatures and stresses on the coil surface. Assembly test had been performed with the TF00 coil structure. We have chosen Type 316LN as material of the TF coil structure. We used the narrow-gap TIG welding method. Doosan Heavy Industries and Construction Company (DHI) will complete the fabrication of the TF coil structure in Feb. 2006. (author)

Toroidal Thermonuclear device

International Nuclear Information System (INIS)

Takizawa, Teruhiro; Shizuoka, Yoshihide.

1982-01-01

Purpose: To reduce the shielding capacity of a current breaker for a current transformer coil and to facilitate the manufacture and the assembly of the current transformer coil. Constitution: A first current transformer coil is provided between a vacuum container for enclosing a plasma and a toroidal magnetic field coil, and a secon current transformer coil is provided outside the toroidal magnetic field coil. The rise of the plasma current is performed by the variation in the current of the coil of the first transformer having high electromagnetic coupling with the plasma current, and the variation in the magnetic flux necessary for maintaining the plasma is performed by the variation in the current of the second transformer coil. In this manner, the current shielding capacity of the first transformer coil can be reduced to decrease the number of coil turns, thereby facilitating the manufacture and assembly. (Seki, T.)

Design analysis of the ITER divertor

International Nuclear Information System (INIS)

Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F.; Merola, M.; Riccardi, B.; Petrizzi, L.; Villari, R.

2007-01-01

The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

Design analysis of the ITER divertor

Energy Technology Data Exchange (ETDEWEB)

Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F. [L.T. Calcoli SaS, Merate (Lecco) (Italy); Merola, M. [ITER Team, Cadarache (France); Riccardi, B. [EFDA CSU Garching (Germany); Petrizzi, L.; Villari, R. [CRE ENEA sulla Fusione Frascati, Roma (Italy)

2007-07-01

The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

Numerical analyses of magnetic field and force in toroidal superconducting magnetic energy storage using unit coils (abstract)

International Nuclear Information System (INIS)

Kanamaru, Y.; Nakayama, T.; Amemiya, Y.

1997-01-01

Superconducting magnetic energy storage (SMES) is more useful than other systems of electric energy storage because of its larger amounts of stored energy and its higher efficiency. There are two types of SMES. One is the solenoid type and the other is the toroidal type. Some models of solenoid-type SMES are designed in the U.S. and in Japan. But the large scale SMES causes a high magnetic field in the living environment, and causes the erroneous operation of electronic equipment. The authors studied some suitable designs of magnetic shielding for the solenoidal-type SMES to reduce the magnetic field in the living environment. The toiroidal type SMES is studied in this article. The magnetic leakage flux of the toiroidal-type SMES is generally lower than that of the solenoid-type SMES. The toroidal-type SMES is constructed of unit coils, which are convenient for construction. The magnetic leakage flux occurs between unit coils. The electromagnetic force of the coils is very strong. Therefore analyses of the leakage flux and electromagnetic force are important to the design of SMES. The authors studied the number, radius, and length of unit coils. The storage energy is 5 G Wh. The numerical analyses of magnetic fields in the toroidal type SMES are obtained by analytical solutions. copyright 1997 American Institute of Physics

Born-Infeld magnetars: larger than classical toroidal magnetic fields and implications for gravitational-wave astronomy

Science.gov (United States)

Pereira, Jonas P.; Coelho, Jaziel G.; de Lima, Rafael C. R.

2018-05-01

Magnetars are neutron stars presenting bursts and outbursts of X- and soft-gamma rays that can be understood with the presence of very large magnetic fields. In this setting, nonlinear electrodynamics should be taken into account for a more accurate description of such compact systems. We study that in the context of ideal magnetohydrodynamics and make a realization of our analysis to the case of the well known Born-Infeld (BI) electromagnetism in order to come up with some of its astrophysical consequences. We focus here on toroidal magnetic fields as motivated by already known magnetars with low dipolar magnetic fields and their expected relevance in highly magnetized stars. We show that BI electrodynamics leads to larger toroidal magnetic fields when compared to Maxwell's electrodynamics. Hence, one should expect higher production of gravitational waves (GWs) and even more energetic giant flares from nonlinear stars. Given current constraints on BI's scale field, giant flare energetics and magnetic fields in magnetars, we also find that the maximum magnitude of magnetar ellipticities should be 10^{-6}-10^{-5}. Besides, BI electrodynamics may lead to a maximum increase of order 10-20% of the GW energy radiated from a magnetar when compared to Maxwell's, while much larger percentages may arise for other physically motivated scenarios. Thus, nonlinear theories of the electromagnetism might also be probed in the near future with the improvement of GW detectors.

Neoclassical tearing mode stabilization by ECCD in ITER

International Nuclear Information System (INIS)

Giruzzi, G.; Zabiego, M.

2001-01-01

A dynamic model, based on a 3-D Fokker-Planck code coupled to the island evolution equations, is used to evaluate the feasibility of active control of Neoclassical Tearing modes by Electron Cyclotron Current Drive (ECCD) in International Thermonuclear Experimental Reactor (ITER). The parameters of the present version of ITER, i.e., RTO/RC ITER (IAM option) are used. Both m=3, n=2 and m=2, n=1 modes are considered. It is shown that an Electron Cyclotron wave system at 140 GHz, with toroidally steerable antennas, can stabilize both modes simultaneously if a power ≥30 MW is available

Compact toroid challenge experiment with the increasing in the energy input into plasma and the level of trapped magnetic field

Energy Technology Data Exchange (ETDEWEB)

Romadanov, I.V.; Ryzhkov, S.V., E-mail: ryzhkov@power.bmstu.ru

2014-12-15

Highlights: • Compact torus formation method with high level of magnetic flux is proposed. • A compact torus is produced in a theta-pinch-coil with pulse mode of operation. • Key feature is a pulse of current in an axial direction. • We report a level of linked magnetic flux is higher than theta-pinch results. - Abstract: The present work reports on compact toroid hydrogen plasma creation by means of a specially designed discharge system and results of magnetic fields introduction. Experiments in the compact toroid challenge (CTC) device at P.N. Lebedev Physical Institute (FIAN) have been conducted since 2005. The CTC device differs from the conventional theta-pinch formation in the use of an axial current for enhanced efficiency. We have used a novel technique to maximize the flux linked to the plasma. The purpose of this method is to increase the energy input into the plasma and the level of trapped magnetic flux using an additional toroidal magnetic field. A study of compact torus formation with axial and toroidal currents was done and a new method is proposed and implemented.

A numerical solution for a toroidal plasma in equilibrium

International Nuclear Information System (INIS)

Hintz, E.; Sudano, J.P.

1982-01-01

The iterative techniques alternating direction implicit (ADI), sucessive ove-relaxation (SOR) and Gauss-Seidel are applied to a nonlinear elliptical second order differential equation (Grand-Shafranov). This equation was solve with the free boundary conditions plasma-vacuum interface over a rectangular section in cylindrical coordinates R and Z. The current density profile, plasma pressure profile, magnetic and isobaric surfaces are numerically determined for a toroidal plasma in equilibrium. (L.C.) [pt

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)

Turbulent transport of toroidal angular momentum in low flow gyrokinetics

International Nuclear Information System (INIS)

Parra, Felix I; Catto, Peter J

2010-01-01

We derive a self-consistent equation for the turbulent transport of toroidal angular momentum in tokamaks in the low flow ordering that only requires solving gyrokinetic Fokker-Planck and quasineutrality equations correct to second order in an expansion on the gyroradius over scale length. We also show that according to our orderings the long wavelength toroidal rotation and the long wavelength radial electric field satisfy the neoclassical relation that gives the toroidal rotation as a function of the radial electric field and the radial gradients of pressure and temperature. Thus, the radial electric field can be solved for once the toroidal rotation is calculated from the transport of toroidal angular momentum. Unfortunately, even though this methodology only requires a gyrokinetic model correct to second order in gyroradius over scale length, current gyrokinetic simulations are only valid to first order. To overcome this difficulty, we exploit the smallish ratio B p /B, where B is the total magnetic field and B p is its poloidal component. When B p /B is small, the usual first order gyrokinetic equation provides solutions that are accurate enough to employ for our expression for the transport of toroidal angular momentum. We show that current δf and full f simulations only need small corrections to achieve this accuracy. Full f simulations, however, are still unable to determine the long wavelength, radial electric field from the quasineutrality equation.

Progress and achievements of R and D activities for the ITER vacuum vessel

International Nuclear Information System (INIS)

Nakahira, M.; Takahashi, H.; Koizumi, K.; Onozuka, M.; Ioki, K.

2001-01-01

The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R and D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18 deg. toroidal sector, which is composed of two 9 deg. sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of ±3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R and D activities for the ITER vacuum vessel. (author)

Finite element and node point generation computer programs used for the design of toroidal field coils in tokamak fusion devices

International Nuclear Information System (INIS)

Smith, R.A.

1975-06-01

The structural analysis of toroidal field coils in Tokamak fusion machines can be performed with the finite element method. This technique has been employed for design evaluations of toroidal field coils on the Princeton Large Torus (PLT), the Poloidal Diverter Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The application of the finite element method can be simplified with computer programs that are used to generate the input data for the finite element code. There are three areas of data input where significant automation can be provided by supplementary computer codes. These concern the definition of geometry by a node point mesh, the definition of the finite elements from the geometric node points, and the definition of the node point force/displacement boundary conditions. The node point forces in a model of a toroidal field coil are computed from the vector cross product of the coil current and the magnetic field. The computer programs named PDXNODE and ELEMENT are described. The program PDXNODE generates the geometric node points of a finite element model for a toroidal field coil. The program ELEMENT defines the finite elements of the model from the node points and from material property considerations. The program descriptions include input requirements, the output, the program logic, the methods of generating complex geometries with multiple runs, computational time and computer compatibility. The output format of PDXNODE and ELEMENT make them compatible with PDXFORC and two general purpose finite element computer codes: (ANSYS) the Engineering Analysis System written by the Swanson Analysis Systems, Inc., and (WECAN) the Westinghouse Electric Computer Analysis general purpose finite element program. The Fortran listings of PDXNODE and ELEMENT are provided

Superconducting magnets for toroidal fusion reactors

International Nuclear Information System (INIS)

Haubenreich, P.N.

1980-01-01

Fusion reactors will soon be employing superconducting magnets to confine plasma in which deuterium and tritium (D-T) are fused to produce usable energy. At present there is one small confinement experiment with superconducting toroidal field (TF) coils: Tokamak 7 (T-7), in the USSR, which operates at 4 T. By 1983, six different 2.5 x 3.5-m D-shaped coils from six manufacturers in four countries will be assembled in a toroidal array in the Large Coil Test Facility (LCTF) at Oak Ridge National Laboratory (ORNL) for testing at fields up to 8 T. Soon afterwards ELMO Bumpy Torus (EBT-P) will begin operation at Oak Ridge with superconducting TF coils. At the same time there will be tokamaks with superconducting TF coils 2 to 3 m in diameter in the USSR and France. Toroidal field strength in these machines will range from 6 to 9 T. NbTi and Nb 3 Sn, bath cooling and forced flow, cryostable and metastable - various designs are being tried in this period when this new application of superconductivity is growing and maturing

Toroidal Plasma Thruster for Interplanetary and Interstellar Space Flights

International Nuclear Information System (INIS)

Gorelenkov, N.N.; Zakharov, L.E.; Gorelenkova, M.V.

2001-01-01

This work involves a conceptual assessment for using the toroidal fusion reactor for deep space interplanetary and interstellar missions. Toroidal thermonuclear fusion reactors, such as tokamaks and stellarators, are unique for space propulsion, allowing for a design with the magnetic configuration localized inside toroidal magnetic field coils. Plasma energetic ions, including charged fusion products, can escape such a closed configuration at certain conditions, a result of the vertical drift in toroidal rippled magnetic field. Escaping particles can be used for direct propulsion (since toroidal drift is directed one way vertically) or to create and heat externally confined plasma, so that the latter can be used for propulsion. Deuterium-tritium fusion neutrons with an energy of 14.1 MeV also can be used for direct propulsion. A special design allows neutrons to escape the shield and the blanket of the tokamak. This provides a direct (partial) conversion of the fusion energy into the directed motion of the propellant. In contrast to other fusion concepts proposed for space propulsion, this concept utilizes the natural drift motion of charged particles out of the closed magnetic field configuration

Status of the ITER magnets

International Nuclear Information System (INIS)

Mitchell, N.; Bauer, P.; Bessette, D.; Devred, A.; Gallix, R.; Jong, C.; Knaster, J.; Libeyre, P.; Lim, B.; Sahu, A.; Simon, F.

2009-01-01

The first 2 years of the ITER IO has seen substantial progress towards the construction of the magnets, in three main areas. Firstly, the design has been developed under the conflicting constraints to minimise construction costs and to maximise plasma physics performance. Building construction momentum while updating the design to take account of new physics assessments of the coil requirements has been challenging. Secondly, with a stabilising design, it has been possible for the Domestic Agencies to launch the first industrial procurement contracts. And thirdly, critical R and D to confirm the performance of the Nb3Sn cable in conduit design is proceeding successfully. The design consolidation has been accompanied by design reviews involving the international community. The reviews conducted by magnet experts have enabled a consensus to be built on choosing between some of the design options in the original ITER basic design in 2001. The major design decisions were to maintain the circular Nb 3 Sn conductor embedded in radial plates for the toroidal field (TF) coils and to maintain NbTi-based conductors for the PF coils. Cold testing, at low current, is also being introduced for quality control purposes for all coils.

Highlights from the assembly of the helical field coils for the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Benson, R.D.

1985-01-01

The helical field (HF) coils in the Advanced Toroidal Facility (ATF) device consist of a set of 24 identical segments connected to form a continuous pair of helical coils wrapped around a toroidal vacuum vessel. Each segment weighs approximately 1364 kg (3000 lb) and is composed of 14 water-cooled copper plate conductors bolted to a cast stainless steel structural support member with a T-shape cross section (known as the structural tee). The segment components are electrically insulated with Kapton adhesive tape, G-10, Tefzel, and rubber to withstand 2.5 kV. As a final insulator and structural support, the entire segment is vacuum impregnated with epoxy. This paper offers a brief overview of the processes used to assemble the component parts into a completed segment, including identification of items that required special attention. 4 figs

Topological symmetry breaking of self-interacting fractional Klein-Gordon field theories on toroidal spacetime

International Nuclear Information System (INIS)

Lim, S C; Teo, L P

2008-01-01

Quartic self-interacting fractional Klein-Gordon scalar massive and massless field theories on toroidal spacetime are studied. The effective potential and topologically generated mass are determined using zeta-function regularization technique. Renormalization of these quantities are derived. Conditions for symmetry breaking are obtained analytically. Simulations are carried out to illustrate regions or values of compactified dimensions where symmetry-breaking mechanisms appear

Equilibrium and stability of theta-pinch plasma in modified toroidal multiple mirror field

International Nuclear Information System (INIS)

Shiina, S.; Saito, K.; Osanai, Y.; Itagaki, T.; Karakizawa, T.; Gesso, H.; Todoroki, J.; Kawakami, I.; Yoshimura, H.

1976-01-01

To confine a high-beta plasma a new toroidal magnetic configuration with closed lines of force has been proposed [1]. The configuration is an appropriate superposition of l = 0, l = +- 1, l = +- 2,sup(...), helical fields. In this experiment, it is generated by modifying the multiple mirror field by enclosing the discharge tube in a copper shell which has longitudinal gap. This configuration is preferred for the wall stabilizing effect to that with the separated helical windings. The characteristics of the equilibrium conditions are examined based on the near-axis approximation theory and compared with the experimental results. The stability of plasma in the configurations with l = 0 field and with superposition of l = 0, l = +- 2 fields is investigated in linear geometry. (author)

Discussion of discrete D shape toroidal coil

International Nuclear Information System (INIS)

Kaiho, Katsuyuki; Ohara, Takeshi; Agatsuma, Ko; Onishi, Toshitada

1988-01-01

A novel design for a toroidal coil, called the D shape coil, was reported by J. File. The coil conductors are in pure tension and then subject to no bending moment. This leads to a smaller number of emf supports in a simpler configuration than that with the conventional toroidal coil of circular cross-section. The contours of the D shape are given as solutions of a differential equation. This equation includes the function of the magnetic field distribution in the conductor region which is inversely proportional to the winding radius. It is therefore important to use the exact magnetic field distribution. However the magnetic field distribution becomes complicated when the D shape toroidal coil is comprised of discrete coils and also depends on the D shape configuration. A theory and a computer program for designing the practical pure-tension toroidal coil are developed. Using this computer code, D shape conductors are calculated for various numbers of discrete coils and the results are compared. Electromagnetic forces in the coils are also calculated. It is shown that the hoop stress in the conductors depends only on the total ampere-turns of the coil when the contours of the D shape are similar. (author)

Procurement of the ITER magnet components supplied by F4E

International Nuclear Information System (INIS)

Sborchia, Carlo; Oliva, Alessandro Bonito; Rajainmaeki, Hannu; Vostner, Alexander

2009-01-01

Fusion for Energy (F4E) will manage the in-kind contribution from Europe (EU) to ITER of about one quarter of the machine magnet components. This contribution will consist in 10 Toroidal Field (TF) coils, 5 (out of 6) Poloidal Field (PF) coils, 20% of the Nb3Sn conductor used in the TF coils and 15% of the PF coil NbTi conductor. In addition, 12 pre-compression fibreglass rings for the TF system will be procured. This paper presents the complex organization and status of the procurement activities of the F4E magnet components. The technical and managerial challenges, in terms of fabrication, quality assurance and schedule, are also highlighted.

Composite coils for toroidal field coils and method of using same

International Nuclear Information System (INIS)

Perkins, R. G.; Trujillo, S. M.

1985-01-01

A composite toroidal field (TF) generating means consisting of segmented magnetic coil windings is disclosed. Each coil winding of the TF generating means consists of a copper or copper alloy conductor segment and an aluminum or aluminum alloy conductor segment. The conductor segments are joined at a high strength, low electrical resistance joint and the joint may either be a mechanical or metallurgical one. The use of the aluminum or aluminum alloy conductor segments improves the neutron economy of the reactor with which the TF coil is associated and reduces TF coil nuclear heating and heating gradients, and activation in the TF coils

Generation of toroidal pre-heat plasma

International Nuclear Information System (INIS)

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

1979-01-01

The characteristics of toroidal plasma in the initial stage of electric discharge were investigated. A small toroidal-pinch system was used for the present work. A magnetic probe was used to measure the magnetic field. The time of beginning of discharge was determined by observing the variation of the magnetic field. The initial gas pressure dependence of the induced electric field regions, in which electric discharge can be caused, was studied. It is necessary to increase the initial induced electric field for starting discharge. The delay time of large current discharge was measured, and it was about 2 microsecond. Dependences of the electric fields at the beginning of discharge on the charging voltage of capacitors, on the initial gas pressure, and on the discharge frequency were studied. The formation mechanism of plasma column was analyzed. (Kato, T.)

Realization of toroidal field power supply control system for J-TEXT tokamak

International Nuclear Information System (INIS)

Qiu Shengshun; Zhuang Ge; Zhang Ming; Feng Jianming

2009-01-01

Based on the integrated development environment provided by QNX real-time operation system, the control system of toroidal field power supply is designed and developed. The system is proved to be reliable, stable and in real-time. It can control the power supply successfully to produce a constant current up to 92.5kA lasting for 1s and 1.74T at the magnetic axis. In conclusion, the control system can meet the requirements of the J-TEXT routine operation at present. (authors)

Development of high-mechanical strength electrical insulations for tokamak toroidal field coils

International Nuclear Information System (INIS)

Burke, C.

1977-01-01

The electrical insulation for the TF (Toroidal Field) coils is subjected to a high interlaminar shear, tensile and compressive stresses. Two candidate epoxy/glass fiber systems using prepreg and vacuum impregnation techniques were evaluated. Specimens were prepared and processed under controlled conditions to simulate specification manufacturing procedures. The strengths of the insulation were measured in interlaminar shear, tension, compression, and combined shear and compression statically. Shear modulus determinations were also made. Various techniques of surface treatments to increase bond strengths with three resin primers were tested

Commissioning Test of ATLAS End-Cap Toroidal Magnets

CERN Document Server

Dudarev, A; Foussat, A; Benoit, P; Jeckel, M; Olyunin, A; Kopeykin, N; Stepanov, V; Deront, L; Olesen, G; Ponts, X; Ravat, S; Sbrissa, K; Barth, J; Bremer, J; Delruelle, J; Metselaar, J; Pengo, R; Pirotte, O; Buskop, J; Baynham, D E; Carr, F S; Holtom, E

2009-01-01

The system of superconducting toroids in the ATLAS experiment at CERN consists of three magnets. The Barrel Toroid was assembled and successfully tested in 2006. Next, two End-Cap Toroids have been tested on surface at 77 K and installed in the cavern, 100-m underground. The End Cap Toroids are based on Al stabilized Nb-Ti/Cu Rutherford cables, arranged in double pancake coils and conduction cooled at 4.6 K. The nominal current is 20.5 kA at 4.1 T peak field in the windings and the stored energy is 250 MJ per toroid. Prior to final testing of the entire ATLAS Toroidal system, each End Cap Toroid passed a commissioning test up to 21 kA to guarantee a reliable performance in the final assembly. In this paper the test results are described. It includes the stages of test preparation, isolation vacuum pumping and leak testing, cooling down, step-by-step charging to full current, training quenches and quench recovery. By fast discharges the quench detection and protection system was checked to demonstrate a safe e...

Toroidal vortices in resistive magnetohydrodynamic equilibria

International Nuclear Information System (INIS)

Montgomery, D.; Bates, J.W.; Li, S.

1997-01-01

When a time-independent electric current flows toroidally in a uniform ring of electrically conducting fluid, a Lorentz force results, jxB, where j is the local electric current density, and B is the magnetic field it generates. Because of purely geometric effects, the curl of jxB is nonvanishing, and so jxB cannot be balanced by the gradient of any scalar pressure. Taking the curl of the fluid close-quote s equation of motion shows that the net effect of the jxB force is to generate toroidal vorticity. Allowed steady states necessarily contain toroidal vortices, with flows in the poloidal directions. The flow pattern is a characteristic open-quotes double smoke ringclose quotes configuration. The effect seems quite general, although it is analytically simple only in special limits. One limit described here is that of high viscosity (low Reynolds number), with stress-free wall boundary conditions on the velocity field, although it is apparent that similar mechanical motions will result for no-slip boundaries and higher Reynolds numbers. A rather ubiquitous connection between current-carrying toroids and vortex rings seems to be implied, one that disappears in the open-quotes straight cylinderclose quotes limit. copyright 1997 American Institute of Physics

Device for supporting a toroidal coil in a toroidal type nuclear fusion device

International Nuclear Information System (INIS)

Kitazawa, Hakaru; Sato, Hiroshi.

1975-01-01

Object: To easily manufacture a center block having a strength sufficient to withstand an electromagnetic force exerted on the center of toroidal of a toroidal coil and to increase its reliability. Structure: In a device for supporting toroidal coils wherein the electromagnetic force exerted on the center of toroidal of a plurality of toroidal coils arranged in toroidal fashion, the contact surface between the toroidal coil and the center block is arranged parallel to the center axis of toroidal so as to receive the electromagnetic force exerted on the center of toroidal of the toroidal coil as the component of force in a radial direction. (Taniai, N.)

Advanced Toroidal Facility

International Nuclear Information System (INIS)

Johnson, R.L.

1985-01-01

The Advanced Toroidal Facility (ATF) is a new magnetic confinement plasma device under construction at the Oak Ridge National Laboratory (ORNL) that will lead to improvements in toroidal magnetic fusion reactors. The ATF is a type of stellerator, known as a ''torsatron'' which theoretically has the capability to operate at greater than or equal to8% beta in steady state. The ATF plasma has a major radius of 2.1 m, an average minor radius of 0.3 m, and a field of 2 T for a 2 s duration or 1 T steady state. The ATF device consists of a helical field (HF) coil set, a set of poloidal field (PF) coils, an exterior shell structure to support the coils, and a thin, helically contoured vacuum vessel inside the coils. The ATF replaces the Impurities Studies Experiment (ISX-B) tokamak at ORNL and will use the ISX-B auxiliary systems including 4 MW of electron cyclotron heating. The ATF is scheduled to start operation in late 1986. An overview of the ATF device is presented, including details of the construction process envisioned. 9 refs., 7 figs., 3 tabs

Investigation of compact toroid penetration for fuelling spherical tokamak plasmas on CPD

International Nuclear Information System (INIS)

Fukumoto, N.; Hanada, K.; Kawakami, S.

2008-10-01

In previous Compact Toroid (CT) injection experiments on several tokamaks, although CT fuelling had been successfully demonstrated, the CT fuelling process has been not clear yet. We have thus conducted CT injection into simple toroidal or vertical vacuum magnetic fields to investigate quantitatively dynamics of CT plasmoid in the penetration process on a spherical tokamak (ST) device. Understanding the process allows us to address appropriately one of the critical issues for practical application of CT injection on reactor-grade tokamaks. In the experiment, the CT shift amount of about 0.26 m in a vertical magnetic field has been observed by using a fast camera. In addition to toroidal magnetic field, vertical one appears to affect CT trajectory in not conventional tokamak but ST devices operated at rather low toroidal fields. We have also observed CT attacks on the target plate with an IR camera. The IR image has indicated that CT shifts 39 mm at the toroidal field of 261 G. From the calorimetric measurement, an input energy due to CT impact in vacuum without magnetic fields is also estimated to be 530 J, which agrees with the initial CT kinetic energy. (author)

Video Toroid Cavity Imager

Energy Technology Data Exchange (ETDEWEB)

Gerald, Rex E. II; Sanchez, Jairo; Rathke, Jerome W.

2004-08-10

A video toroid cavity imager for in situ measurement of electrochemical properties of an electrolytic material sample includes a cylindrical toroid cavity resonator containing the sample and employs NMR and video imaging for providing high-resolution spectral and visual information of molecular characteristics of the sample on a real-time basis. A large magnetic field is applied to the sample under controlled temperature and pressure conditions to simultaneously provide NMR spectroscopy and video imaging capabilities for investigating electrochemical transformations of materials or the evolution of long-range molecular aggregation during cooling of hydrocarbon melts. The video toroid cavity imager includes a miniature commercial video camera with an adjustable lens, a modified compression coin cell imager with a fiat circular principal detector element, and a sample mounted on a transparent circular glass disk, and provides NMR information as well as a video image of a sample, such as a polymer film, with micrometer resolution.

Edge pedestal characteristics in JET and JT-60U tokamaks under variable toroidal field ripple

NARCIS (Netherlands)

Urano, H.; Saibene, G.; Oyama, N.; Parail, V.; P. de Vries,; Sartori, R.; Kamada, Y.; Kamiya, K.; Loarte, A.; Lonnroth, J.; Sakamoto, Y.; Salmi, A.; Shinohara, K.; Takenaga, H.; Yoshida, M.

2011-01-01

The effects of toroidal field (TF) ripple on the edge pedestal characteristics were examined in the TF ripple scan experiments at the plasma current I(p) of 1.1 MA in JET and JT-60U. The TF ripple amplitude delta(R) was defined as a value averaged over the existing ripple wells at the separatrix on

Global Hybrid Simulations of Energetic Particle-driven Modes in Toroidal Plasmas

International Nuclear Information System (INIS)

Fu, G.Y.; Breslau, J.; Fredrickson, E.; Park, W.; Strauss, H.R.

2004-01-01

Global hybrid simulations of energetic particle-driven MHD modes have been carried out for tokamaks and spherical tokamaks using the hybrid code M3D. The numerical results for the National Spherical Tokamak Experiments (NSTX) show that Toroidal Alfven Eigenmodes are excited by beam ions with their frequencies consistent with the experimental observations. Nonlinear simulations indicate that the n=2 mode frequency chirps down as the mode moves out radially. For ITER, it is shown that the alpha-particle effects are strongly stabilizing for internal kink mode when central safety factor q(0) is sufficiently close to unity. However, the elongation of ITER plasma shape reduces the stabilization significantly

Preparation of the ITER Poloidal Field Conductor Insert (PFCI) test

International Nuclear Information System (INIS)

Zanino, R.; Egorov, S.; Kim, K.; Martovetsky, N.; Nunoya, Y.; Okuno, K.; Salpietro, E.; Sborchia, C.; Takahashi, Y.; Weng, P.; Bangasco, M.; Savoldi Richard, L.; Polak, M.; Formisano, A.; Zapretilina, E.; Shikov, A.; Vedernikov, G.; Ciazynski, D.; Zani, L.; Muzzi, L.; Ricci, M.; Deela Corte, A.; Sugimoto, M.; Hamada, K.; Portone, A.; Hurd, F.; Mitchell, N.; Nijhuis, A.; Ilyin, Y.

2004-01-01

The Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in Europe and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, about 45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of about 6 T, and it will be cooled by supercritical He at around 4.5 K and 0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pick-up coils, temperature sensors, pressure taps, strain and displacement sensors. The test program shall be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermalhydraulic characterization. Here we give an overview of the preparatory work towards the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far. (authors)

Mechanical Commissioning of the ATLAS Barrel Toroid Magnet

CERN Document Server

Foussat, A; Dudarev, A; Bajas, H; Védrine, P; Berriaud, C; Sun, Z; Sorbi, M

2008-01-01

ATLAS is a general-purpose detector designed to run at the highest luminosity at the CERN Large Hadron Collider. Its features include the 4 T Barrel Toroid magnet, the largest superconducting magnet (25 m long, 20 m diameter) that provides the magnetic field for the ATLAS muon spectrometer. The coils integrated at CERN, were tested individually at maximum current of 22 kA in 2005. Following the mechanical assembly of the Barrel Toroid in the ATLAS underground cavern, the test of the full Barrel Toroid was performed in October 2006. Further tests are foreseen at the end 2007 when the system will include the two End Cap Toroids (ECT). The paper gives an overview of the good mechanical test results achieved in comparison with model predictions and the experience gained in the mechanical behavior of the ATLAS Toroidal coils is discussed.

Quiescent H-mode operation using torque from non-axisymmetric, non-resonant magnetic fields

International Nuclear Information System (INIS)

Burrell, K.H.; Garofalo, A.M.; Osborne, T.H.; Snyder, P.B.; Solomon, W.M.; Park, J.-K.; Fenstermacher, M.E.; Orlov, D.M.

2013-01-01

Quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas including ITER. Using magnetic torque from n = 3 fields to replace counter-I p torque from neutral beam injection, we have achieved long duration, counter-rotating QH-mode operation with neutral beam injection (NBI) torque ranging continuously from counter-I p up to co-I p values of about 1 N m. This co-I p torque is about 3 times the scaled torque that ITER will have. This range also includes operation at zero net NBI torque, applicable to rf wave heated plasmas. These n = 3 fields have been created using coils either inside or, most recently, outside the toroidal coils. Experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values ν ped * ∼0.08, β T ped ∼ 1%$ and β N = 2. Discharges have confinement quality H 98y2 = 1.3, exceeding the value required for ITER. Initial work with low q 95 = 3.4 QH-mode plasmas transiently reached fusion gain values of G = β N H 89 /q 95 2 =0.4, which is the desired value for ITER; the limits on G have not yet been established. This paper also includes the most recent results on QH-mode plasmas run without n = 3 fields and with co-I p NBI; these shots exhibit co-I p plasma rotation and require NBI torque ⩾2 N m. The QH-mode work to date has made significant contact with theory. The importance of edge rotational shear is consistent with peeling–ballooning mode theory. We have seen qualitative and quantitative agreement with the predicted torque from neoclassical toroidal viscosity. (paper)

ATLAS: Full power for the toroid magnet

CERN Multimedia

2006-01-01

The 9th of November was a memorable day for ATLAS. Just before midnight, the gigantic Barrel toroid magnet reached its nominal field of 4 teslas in the coil windings, with an electrical current of 21000 amperes (21 kA) passing through the eight superconducting coils (as seen on the graph). This achievement was obtained after several weeks of commissioning. The ATLAS Barrel Toroid was first cooled down for about six weeks in July-August to -269°C (4.8 K) and then powered up step-by-step in successive test sessions to 21 kA. This is 0.5 kA above the current required to produce the nominal magnetic field. Afterwards, the current was safely switched off and the stored magnetic energy of 1.1 gigajoules was dissipated in the cold mass, raising its temperature to a safe -218°C (55 K). 'We can now say that the ATLAS Barrel Toroid is ready for physics,' said Herman ten Kate, project leader for the ATLAS magnet system. The ATLAS barrel toroid magnet is the result of a close collaboration between the magnet la...

Steady-state resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1979-12-01

If spatially-averaged values of the beta ratio can reach 5 to 10% in tokamaks, as now seems likely, resistive toroidal-field coils may be advantageous for use in reactors intended for fusion-neutron applications. The present investigation has parameterized the design of steady-state water-cooled copper coils of rectangular cross section in order to maximize figures of merit such as the ratio of fusion neutron wall loading to coil power dissipation. Four design variations distinguished by different ohmic-heating coil configurations have been examined. For a wall loading of 0.5 MW/m 2 , minimum TF-coil lifetime costs (including capital and electricity costs) are found to occur with coil masses in the range 2400 to 4400 tons, giving 200 to 250 MW of resistive dissipation, which is comparable with the total power drain of the other reactor subsystems

Radiological dose rate calculations for the International Thermonuclear Experimental Reactor (ITER)

International Nuclear Information System (INIS)

Khater, H.Y.; Santoro, R.T.

1996-01-01

Two-dimensional biological dose rates were calculated at different locations outside the International Thermonuclear Experimental Reactor (ITER) design. An 18 degree sector of the reactor was modeled in r-θ geometry. The calculations were performed for three different pulsing scenarios. This included a single pulse of 1000 s duration, 10 pulses of 1000 s duration with a 50% duty factor, and 9470 pulses of 1000 s duration with a 50% duty factor for a total fluence of 0.3 MW.a/m 2 . The dose rates were calculated as a function of toroidal angle at locations in the space between the toroidal field (TF) coils and cryostat, and in the space between the cryostat and the biological shield. The two-dimensional results clearly showed the toroidal effect, which is dominated by contribution from the activation of the cryostat and the biological shield. After one pulse, full access to the machine is possible within a few hours following shutdown. After 10 pulses, full access is also possible within the first day following shutdown. At the end of the Basic Performance Phase (BPP), full access is possible at any of the locations considered after one week following shutdown. 5 refs., 5 figs., 2 tabs

Bi-2223 HTS winding in toroidal configuration for SMES coil

International Nuclear Information System (INIS)

Kondratowicz-Kucewicz, B; Kozak, S; Kozak, J; Wojtasiewicz, G; Majka, M; Janowski, T

2010-01-01

Energy can be stored in the magnetic field of a coil. Superconducting Magnetic Energy Storage (SMES) is very promising as a power storage system for load levelling or power stabilizer. However, the strong electromagnetic force caused by high magnetic field and large coil current is a problem in SMES systems. A toroidal configuration would have a much less extensive external magnetic field and electromagnetic forces in winding. The paper describes the design of HTS winding for SMES coil in modular toroid configuration consist of seven Bi-2223 double-pancakes as well as numerical analysis of SMES magnet model using FLUX 3D package. As the results of analysis the paper presents the optimal coil configuration and the parameters such as radius of toroidal magnet, energy stored in magnet and magnetic field distribution.

Confinement time exceeding one second for a toroidal electron plasma.

Science.gov (United States)

Marler, J P; Stoneking, M R

2008-04-18

Nearly steady-state electron plasmas are trapped in a toroidal magnetic field for the first time. We report the first results from a new toroidal electron plasma experiment, the Lawrence Non-neutral Torus II, in which electron densities on the order of 10(7) cm(-3) are trapped in a 270-degree toroidal arc (670 G toroidal magnetic field) by application of trapping potentials to segments of a conducting shell. The total charge inferred from measurements of the frequency of the m=1 diocotron mode is observed to decay on a 3 s time scale, a time scale that approaches the predicted limit due to magnetic pumping transport. Three seconds represents approximately equal to 10(5) periods of the lowest frequency plasma mode, indicating that nearly steady-state conditions are achieved.

Error field assessment from driven rotation of stable external kinks at EXTRAP-T2R reversed field pinch

Science.gov (United States)

Volpe, F. A.; Frassinetti, L.; Brunsell, P. R.; Drake, J. R.; Olofsson, K. E. J.

2013-04-01

A new non-disruptive error field (EF) assessment technique not restricted to low density and thus low beta was demonstrated at the EXTRAP-T2R reversed field pinch. Stable and marginally stable external kink modes of toroidal mode number n = 10 and n = 8, respectively, were generated, and their rotation sustained, by means of rotating magnetic perturbations of the same n. Due to finite EFs, and in spite of the applied perturbations rotating uniformly and having constant amplitude, the kink modes were observed to rotate non-uniformly and be modulated in amplitude. This behaviour was used to precisely infer the amplitude and approximately estimate the toroidal phase of the EF. A subsequent scan permitted to optimize the toroidal phase. The technique was tested against deliberately applied as well as intrinsic EFs of n = 8 and 10. Corrections equal and opposite to the estimated error fields were applied. The efficacy of the error compensation was indicated by the increased discharge duration and more uniform mode rotation in response to a uniformly rotating perturbation. The results are in good agreement with theory, and the extension to lower n, to tearing modes and to tokamaks, including ITER, is discussed.

Distortion of magnetic field lines caused by radial displacements of ITER toroidal field coils

Energy Technology Data Exchange (ETDEWEB)

Amoskov, V.M., E-mail: sytch@niiefa.spb.su [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, St. Petersburg (Russian Federation); Gribov, Y.V. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Lamzin, E.A.; Sytchevsky, S.E. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, St. Petersburg (Russian Federation)

2017-05-15

An assessment of distortions of ideal (circle) field lines caused by random radial displacements of the TF coils by |∆R| ≤ 5 mm has been performed from the statistical analysis assuming a uniform probability density function for displacements.

Topological currents in neutron stars: kicks, precession, toroidal fields, and magnetic helicity

International Nuclear Information System (INIS)

Charbonneau, James; Zhitnitsky, Ariel

2010-01-01

The effects of anomalies in high density QCD are striking. We consider a direct application of one of these effects, namely topological currents, on the physics of neutron stars. All the elements required for topological currents are present in neutron stars: degenerate matter, large magnetic fields, and parity violating processes. These conditions lead to the creation of vector currents capable of carrying momentum and inducing magnetic fields. We estimate the size of these currents for many representative states of dense matter in the neutron star and argue that they could be responsible for the large proper motion of neutron stars (kicks), the toroidal magnetic field and finite magnetic helicity needed for stability of the poloidal field, and the resolution of the conflict between type-II superconductivity and precession. Though these observational effects appear unrelated, they likely originate from the same physics — they are all P-odd phenomena that stem from a topological current generated by parity violation

Toroidal drift magnetic pumping

International Nuclear Information System (INIS)

Canobbio, E.

1977-01-01

A set of azimuthal coils which carry properly dephased rf-currents in the KHz frequency range can be used to heat toroidal plasmas by perpendicular Landau damping of subsonic Alfven waves. The heating mechanism and the rf-field structure are discussed in some detail

Room Temperature Magnetic Determination of the Current Center Line for the ITER TF Coils

CERN Document Server

Lerch, Philippe; Buzio, Marco; Negrazus, Marco; Baynham, Elwyn; Sanfilippo, Stephane; Foussat, Arnaud

2014-01-01

The ITER tokamak includes 18 superconducting D-shaped toroidal field (IT) coils. Unavoidable shape deformations as well as assembly errors will lead to field errors, which can be modeled with the knowledge of the current center line (CCL). Accurate survey during the entire manufacturing and assembly process, including transfer of survey points, is complex. In order to increase the level of confidence, a room temperature magnetic measurement of the CCL on assembled and closed winding packs is foreseen, prior to insertion into their cold case. In this contribution, we discuss the principle of the CCL determination and present a low frequency ac measurement system under development at PSI, within an ITER framework contract. The largest current allowed to flow in the TF coil at room temperature and the precision requirements for the determination of the CCL loci of the coil are hard boundaries. Eddy currents in the radial plates, the winding pack enclosures, and possibly from iron in the reinforced concrete floor...

Mirror theory applied to toroidal systems

International Nuclear Information System (INIS)

Cohen, R.H.

1987-01-01

Central features of a mirror plasma are strong departures from Maxwellian distribution functions, ambipolar potentials and densities which vary along a field line, and losses, and the mirror field itself. To examine these features, mirror theorists have developed analytical and numerical techniques to solve the Fokker-Planck equation, evaluate the potentials consistent with the resulting distribution functions, and assess the microstability of these distributions. Various combinations of mirror-plasma fetures are present and important in toroidal plasmas as well, particularly in the edge region and in plasmas with strong r.f. heating. In this paper we survey problems in toroidal plasmas where mirror theory and computational techniques are applicable, and discuss in more detail three specific examples: calculation of the toroidal generalization of the Spitzer-Haerm distribution function (from which trapped-particle effects on current drive can be calculated), evaluation of the nonuniform potential and density set up by pulsed electron-cyclotron heating, and calculation of steady-state distribution functions in the presence of strong r.f. heating and collisions. 37 refs., 3 figs

Mirror theory applied to toroidal systems

International Nuclear Information System (INIS)

Cohen, R.H.

1987-01-01

Central features of a mirror plasma are strong departures from Maxwellian distribution functions, ambipolar potentials and densities which vary along a field line, end losses, and the mirror field itself. To examine these features, mirror theorists have developed analytical and numerical techniques to solve the Fokker-Planck equation, evaluate the potentials consistent with the resulting distribution functions, and assess the microstability of these distributions. Various combinations of mirror-plasma features are present and important in toroidal plasmas as well, particularly in the edge region and in plasmas with strong rf heating. In this paper we survey problems in toroidal plasmas where mirror theory and computational techniques are applicable, and discuss in more detail three specific examples: calculation of the toroidal generalization of the Spitzer-Haerm distribution function (from which trapped-particle effects on current drive can be calculated), evaluation of the nonuniform potential and density set up by pulsed electron-cyclotron heating, and calculation of steady-state distribution functions in the presence of strong rf heating and collisions. 37 refs

HL-2A experiment and ITER-related activity at SWIP

International Nuclear Information System (INIS)

Duan Xuru

2007-01-01

In this overview the recent progress on HL-2A tokamak experiment and ITER-related activity at SWIP is summarized. Experiment on HL-2A is one of the important research activities at SWIP. In the last two years, some new hardware had been developed, these include four sets of ECRH system with a total power up to 2 MW, new diagnostics such as 8-channel laser interferometer. The studied subjects were focused on plasma auxiliary heating, fuelling, transport, edge plasma physics and turbulence, etc. Progress in these fields has been obtained. For example, the toroidal symmetry of the geodesic acoustic mode (GAM), the oscillating branch of zonal flows has been demonstrated for the first time using a novel 3-step Langmuir Probe, and the poloidal and radial structure of the low frequency electric potential and field were simultaneously observed. During ECRH experiments under different discharge conditions, the MHD instability excited by high energetic electrons was investigated. Besides, non-local heat transport due to SMBI during ECRH was studied. Another important fusion activity at SWIP is the ITER relevant technology. The R and D of four ITER procurements (first wall and shielding blanket, magnet gravity support, gas injection and glow discharge cleaning system, neutron flux measurement) has been undertaken. Progress has been made, e.g. the technology for manufacturing high purity (>99%) ITER specified Be plate and CuCrZr alloy is obtained, their major mechanical and physical properties were measured. For ITER-TBM, a structural material named as CLF-1, a type of reduced activation ferritic/martenstic steel, was developed. Besides, some progress in fusion reactor design and related technology was achieved. (authors)

Compact toroid development: activity plan for field reversed configurations

International Nuclear Information System (INIS)

1984-06-01

This document contains the description, goals, status, plans, and approach for the investigation of the properties of a magnetic configuration for plasma confinement identified as the field reversed configuration (FRC). This component of the magnetic fusion development program has been characterized by its potential for physical compactness and a flexible range of output power. The included material represents the second phase of FRC program planning. The first was completed in February 1983, and was reported in DOE/ER-0160; Compact Toroid Development. This planning builds on that previous report and concentrates on the detailed plans for the next several years of the current DOE sponsored program. It has been deliberately restricted to the experimental and theoretical efforts possible within the present scale of effort. A third phase of this planning exercise will examine the subsequent effort and resources needed to achieve near term (1987 to 1990) FRC technical objectives

Investigations of toroidal wave numbers of the kink instabilities in a toroidal pinch plasma

International Nuclear Information System (INIS)

Hamajima, Takataro; Irisawa, Juichi; Tsukada, Tokuaki; Sugito, Osamu; Maruyama, Hideaki

1979-01-01

The axial toroidal wave numbers of the kink instability of toroidal pinch plasma were measured and investigated with a specially designed coil, and the results were compared with the MHD theory. The schematic figure and the particulars of the experimental apparatus are briefly illustrated in the first part. The method of generating theta-Z pinch plasma, the wave form of the magnetic flux density in Z-direction and the plasma current are also explained. The 360 deg stereoscopic framing photographs were taken with an image converter camera at the intervals of 0.5 μs after the initiation of the main electric discharge in Z-circuit. From these photographs, the growth of the kink instability was observed. The measured magnetic field distribution at t = 2 μs is presented. In the second part, the radial displacement of plasma and toroidal wave number were measured from the above framing photographs. Then the spectra of plasma displacement were analyzed by the Fourier analysis. The measured results of toroidal wave number was analyzed by both the skin current model and the diffuse current model. Many new results obtained from the present study were mainly derived from the observation of the framing photographs, and they are summarized in the final part of this paper. (Aoki, K.)

Mechanical stress calculations for toroidal field coils by the finite element method

International Nuclear Information System (INIS)

Soell, M.; Jandl, O.; Gorenflo, H.

1976-09-01

After discussing fundamental relationships of the finite element method, this report describes the calculation steps worked out for mechanical stress calculations in the case of magnetic forces and forces produced by thermal expansion or compression of toroidal field coils using the SOLID SAP IV computer program. The displacement and stress analysis are based on the 20-node isoparametric solid element. The calculation of the nodal forces produced by magnetic body forces are discussed in detail. The computer programs, which can be used generally for mesh generation and determination of the nodal forces, are published elsewhere. (orig.) [de

HTMR: an experimental tokamak reactor with hybrid copper/superconductor toroidal field magnet

International Nuclear Information System (INIS)

Avanzini, P.G.; Raia, G.; Rosatelli, F.; Zampaglione, V.

1985-01-01

The feasibility of a hybrid configuration superconducting coils/copper coils for a next generation tokamak TF magnet has been investigated. On the basis of this hybrid solution, the conceptual design has been developed for a medium-high toroidal field tokamak reactor (HTMR). The results of this study show the possibility of designing a tokamak reactor with reduced size in comparison with other INTOR like devices, still gaining some margins in front of the uncertainties in the scaling laws for plasma physics parameters and retaining the presence of a blanket with a tritium breeding ratio of about 1

Overview of the divertor design and its integration into RTO/RC-ITER

International Nuclear Information System (INIS)

Janeschitz, G.; Tivey, R.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Heidl, H.; Ibbott, C.; Martin, E.

2000-01-01

The design of the divertor and its integration into the reduced technical objectives/reduced cost-international thermonuclear energy reactor (RTO/RC-ITER) is based on the experience gained from the 1998 design of international thermonuclear energy reactor (ITER) and on the research and development performed throughout the engineering design activities (EDA). This paper gives an overview of the layout and functional design of the RTO/RC-ITER divertor, including the integration into the machine and the remote replacement of the divertor cassettes. Design guidelines are presented which have allowed quick preparation of divertor layouts suitable for further study using the B2-EIRENE edge plasma code. As in the 1998 design, the divertor is segmented into cassettes, and the segmentation, which is three per sector, is driven by access through the divertor level ports. Maintaining this access and avoiding interference with poloidal field coils means that the divertor level ports need to be inclined (7 deg.). This opens up the possibility of incorporating inboard and outboard baffles into the divertor cassettes. The cassettes are transported in-vessel by making use of the toroidal rails onto which the cassettes are finally clamped in position. Significant reduction of the space available between the X-point and the vacuum vessel results in re-positioning of the toroidal rails in order to retain sufficient depth for the inner and outer divertor legs. This, in turn, requires some changes to the remote handling (RH) concept. Remote handling (RH) is now based on using a cantilevered articulated gripper during the radial movement of the cassettes inside the RH ports. However, the principle to use a cassette toroidal mover (CTM) for in vessel handling is unchanged, hence maintaining the validity of previous EDA research and development. The space previously left below the cassettes for RH was also used for pumping. Elimination of this space has led to re-siting of the pumping

Upgrade of DIII-D toroidal magnetic field power supply controls

International Nuclear Information System (INIS)

Petrach, P.M.; Rouleau, A.R.; McNulty, R.D.; Patrick, D.B.; Walin, J.L.

1993-11-01

The toroidal magnetic field power supply for the DIII-D tokamak is of the 12 pulse line commutated variety. It consists of four individual modules and a main system control cabinet which are combined to deliver 127,000 A and 1000 V to the toroidal field coil. The modules are connected in a series-parallel configuration but can be run alone or two at a time as well. Normally on DIII-D experiments, the series-parallel configuration is required. The original design provided each individual module with its own voltage and current control loop and a main control loop. A problem with this design was that the individual control loops would cause a current sharing imbalance in the parallel modules if the calibrated loops drifted by the slightest amount. It was determined that individual control loops were not needed and a single phase lock firing circuit was employed in the system cabinet with fiber optic links to the modules for gate drive signals. Since all four modules have to be on line for DIII-D to operate, a problem in any of the five E ampersand I control loops resulted in the supply, and, therefore, the tokamak, being idled. By reducing the number of control loops to one, the sharing problem was eliminated, as well as 4 out of 5 potential control failures. The original supply employed relay logic for sequence control and fault monitoring. There were over 130 relays in each module plus an additional 100 in the system cabinet. The combination of the number of relays with the required interconnecting wiring, the age of the supply, the vibrations of the cabinets and the harsh environment, resulted in a continuously escalating number of phantom, and often intermittent, faults. The fault and sequence logic relays were replaced by a new Programmable Logic Controller (PLC). All existing interconnect wire was removed and replaced with multiconductor cables that connect directly from fault sensors and input devices to the PLC

Structural analysis of the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Sannazzaro, G.; Ioki, K.; Johnson, G.; Onozuka, M.; Utin, Y. [ITER Joint Work Site, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States); Swanson, J. [USHT, Raytheon, Princeton (United States)

1998-07-01

The ITER Vacuum Vessel (VV) must withstand a large number of loading conditions including electromagnetic, seismic, operational and upset pressure, thermal and test loads. All of the loading conditions and load combinations have been categorized and classified to permit the allowable stress to be defined in accordance with the recommendations of the ASME code. The most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and the load combination of seismic and electromagnetic loads due to a plasma vertical instability. The areas of high stress are the regions around the VV and the blanket supports, and the attachment of the ports to the main shell. In all of the loading conditions and load combinations the calculated stresses are below the allowable values. (authors)

Thermal convection in a toroidal duct of a liquid metal blanket. Part I. Effect of poloidal magnetic field

International Nuclear Information System (INIS)

Zhang, Xuan; Zikanov, Oleg

2017-01-01

Highlights: • 2D convection flow develops with internal heating and strong axial magnetic field. • Poloidal magnetic field suppresses turbulence at high Hartmann number. • Flow structure is dominated by large-scale counter-rotation vortices. • Effective heat transfer is maintained by surviving convection structures. - Abstract: We explore the effect of poloidal magnetic field on the thermal convection flow in a toroidal duct of a generic liquid metal blanket. Non-uniform strong heating (the Grashof number up to 10 11 ) arising from the interaction of high-speed neutrons with the liquid breeder, and strong magnetic field (the Hartmann number up to 10 4 ) corresponding to the realistic reactor conditions are considered. The study continues our earlier work , where the problem was solved for a purely toroidal magnetic field and the convection was found to result in two-dimensional turbulence and strong mixing within the duct. Here, we find that the poloidal component of the magnetic field suppresses turbulence, reduces the flow's kinetic energy and high-amplitude temperature fluctuations, and, at high values of Hartmann number, leads to a steady-state flow. At the same time, the intense mixing by the surviving convection structures remains able to maintain effective heat transfer between the liquid metal and the walls.

Simulation of the flow-reversal effect in dual channel CICC for ITER

CERN Document Server

Bottura, L; Calvi, M; Herzog, R; Marinucci, C

2007-01-01

The discovery of an upward counter flow of helium in the outer annulus of the vertically oriented and top-to-bottom cooled ITER PFFSJS (Poloidal Field Coil-Full Size Joint Sample) in 2002 led to closer investigations of the effect because it may lead to a reduction of the operational margin of the superconductor used in the ITER environment. Recently, further thermo-hydraulic experiments were carried out on the TFAS2 sample (Toroidal Field Advanced Strand sample 2) with the intent to asses the effect in detail. First investigations confirmed the initial assumption that the origin of the effect lies in the buoyancy of the heated, and thus less dense, helium in the outer annulus of the cable. The helium there is in good contact with the superconducting strands heated by neutron irradiation, ac losses or heat influx, but is thermally and hydraulically less well coupled to the downward flowing helium in the central channel. This paper presents an analysis of the TFAS2 experiments using the simulation program THEA...

Advanced fuelling system for ITER

Energy Technology Data Exchange (ETDEWEB)

Raman, Roger [University of Washington, Seattle, WA (United States)], E-mail: raman@aa.washington.edu

2008-12-15

Steady-state high-performance discharges in reactors, such as the Advanced Tokamak (AT) scenarios would rely on optimized density and pressure profiles that must be maintained. This maximizes the bootstrap current fraction, reduces reactor recycling power and reduces thermal stresses. Other than a system for the balance of current drive not provided by bootstrap current drive, no other sources of input power, such as from neutral beams, are allowed. For these systems, a precision fuelling system would be the ideal way to control the fusion burn by controlling and maintaining the required pressure profile. This requires a fuelling system that is capable of depositing fuel at any radial location within the plasma while at the same time not altering the density profile to a level that degrades the required pressure profile. Present fuelling systems are incapable of meeting these requirements. An advanced fuelling system based on Compact Toroid injection has the potential to meet these needs while simultaneously providing a source of toroidal momentum input. Description of a conceptual Compact Toroid fueller for ITER is presented in conjunction with a plan for developing this much needed technology.

Reactor-relevant quiescent H-mode operation using torque from non-axisymmetric, non-resonant magnetic fields

International Nuclear Information System (INIS)

Burrell, K. H.; Garofalo, A. M; Osborne, T. H.; Schaffer, M. J.; Snyder, P. B.; Solomon, W. M.; Park, J.-K.; Fenstermacher, M. E.

2012-01-01

Results from recent experiments demonstrate that quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas. Using magnetic torque from n=3 fields to replace counter-I p torque from neutral beam injection (NBI), we have achieved long duration, counter-rotating QH-mode operation with NBI torque ranging from counter-I p to up to co-I p values of 1-1.3 Nm. This co-I p torque is 3 to 4 times the scaled torque that ITER will have. These experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values of ν ped * and β N ped . These discharges exhibited confinement quality H 98y2 =1.3, in the range required for ITER. In preliminary experiments using n=3 fields only from a coil outside the toroidal coil, QH-mode plasmas with low q 95 =3.4 have reached fusion gain values of G=β N H 89 /q 95 2 =0.4, which is the desired value for ITER. Shots with the same coil configuration also operated with net zero NBI torque. The limits on G and co-I p torque have not yet been established for this coil configuration. QH-mode work to has made significant contact with theory. The importance of edge rotational shear is consistent with peeling-ballooning mode theory. Qualitative and quantitative agreements with the predicted neoclassical toroidal viscosity torque is seen.

Integrated core-SOL simulations of L-mode plasma in ITER and Indian demo

International Nuclear Information System (INIS)

Wisitsorasak, Apiwat; Onjun, Thawatchai; Kanjanaput, Wittawat

2015-01-01

Core-SOL simulations are carried out using 1.5D BALDUR integrated predictive modeling code to investigate tokamak plasma in ITER and Indian DEMO reactors operating in low confinement mode (L-Mode). In each simulation, the plasma current, temperature, and density profiles in both core and SOL region are evolved self-consistency. The SOL is simulated by integrating the fluid equations, including sources, along the field lines. The solutions in SOL subsequently provide as the boundary conditions of core plasma region on low-confinement mode. The core plasma transport model is described using a combination of anomalous transport by Multi-Mode-Model version 2001 (MMM2001) and neoclassical transport calculated by NCLASS module together with the toroidal velocity based on the torque due to Neoclassical Toroidal Viscosity (NTV). In addition, a sensitivity analysis is explored by varying plasma parameters, such as plasma density and auxiliary heating power. Furthermore, the ignition tests are conducted to observed plasma response in each design after shutting down an auxiliary heating. (author)

The Superconducting Toroid for the New International AXion Observatory (IAXO)

CERN Document Server

Shilon, I.; Silva, H.; Wagner, U.; ten Kate, H.H.J.

2013-01-01

IAXO, the new International AXion Observatory, will feature the most ambitious detector for solar axions to date. Axions are hypothetical particles which were postulated to solve one of the puzzles arising in the standard model of particle physics, namely the strong CP (Charge conjugation and Parity) problem. This detector aims at achieving a sensitivity to the coupling between axions and photons of one order of magnitude beyond the limits of the current detector, the CERN Axion Solar Telescope (CAST). The IAXO detector relies on a high-magnetic field distributed over a very large volume to convert solar axions to detectable X-ray photons. Inspired by the ATLAS barrel and end-cap toroids, a large superconducting toroid is being designed. The toroid comprises eight, one meter wide and twenty one meters long racetrack coils. The assembled toroid is sized 5.2 m in diameter and 25 m in length and its mass is about 250 tons. The useful field in the bores is 2.5 T while the peak magnetic field in the windings is 5....

Design considerations for ITER [International Thermonuclear Experimental Reactor] magnet systems

International Nuclear Information System (INIS)

Henning, C.D.; Miller, J.R.

1988-01-01

The International Thermonuclear Experimental Reactor (ITER) is now completing a definition phase as a beginning of a three-year design effort. Preliminary parameters for the superconducting magnet system have been established to guide further and more detailed design work. Radiation tolerance of the superconductors and insulators has been of prime importance, since it sets requirements for the neutron-shield dimension and sensitively influences reactor size. The major levels of mechanical stress in the structure appear in the cases of the inboard legs of the toroidal-field (TF) coils. The cases of the poloidal-field (PF) coils must be made thin or segmented to minimize eddy current heating during inductive plasma operation. As a result, the winding packs of both the TF and PF coils includes significant fractions of steel. The TF winding pack provides support against in-plane separating loads but offers little support against out-of-plane loads, unless shear-bonding of the conductors can be maintained. The removal of heat due to nuclear and ac loads has not been a fundamental limit to design, but certainly has non-negligible economic consequences. We present here preliminary ITER magnetic systems design parameters taken from trade studies, designs, and analyses performed by the Home Teams of the four ITER participants, by the ITER Magnet Design Unit in Garching, and by other participants at workshops organized by the Magnet Design Unit. The work presented here reflects the efforts of many, but the responsibility for the opinions expressed is the authors'. 4 refs., 3 figs., 4 tabs

Current control necessary for toroidal plasma equilibrium

International Nuclear Information System (INIS)

Nagao, S.

1987-01-01

It is shown that a significant amount of dipole current is necessary for the plasma equilibrium of toroidal configurations in general. Through the vector product with the poloidal field, this dipole current force has to balance with the hoop force of plasma pressure itself of the annular shape. The measurement of such a current of dipole type may be interesting for the confirmation of the plasma equilibrium in the toroidal system. Moreover it is certained that there is a new mode of a tokamak operation with such a dipole current component and with smaller vertical field than that based on the classical tokamak theory. (author) [pt

Some analytical results for toroidal magnetic field coils with elongated minor cross-sections

International Nuclear Information System (INIS)

Raeder, J.

1976-09-01

The problem of determining the shape of a flexible current filament forming part of an ideal toroidal magnetic field coil is solved in a virtually analytical form. Analytical formulae for characteristic coil dimensions, stored magnetic energies, inductances and forces are derived for the so-called D-coils. The analytically calculated inductances of ideal D-coils are compared with numerically calculated ones for the case of finite numbers of D-shaped current filaments. Finally, the magnetic energies stored in ideal rectangular, elliptic and D-coils are compared. (orig.) [de

Extending SIESTA capabilities: removing field-periodic and stellarator symmetric limitations

Science.gov (United States)

Cook, C. R.; Hirshman, S. P.; Sanchez, R.; Anderson, D. T.

2011-10-01

SIESTA is a three-dimensional magnetohydrodynamics equilibrium code capable of resolving magnetic islands in toroidal plasma confinement devices. Currently SIESTA assumes that plasma perturbations, and thus also magnetic islands, are field-periodic. This limitation is being removed from the code by allowing the displacement toroidal mode number to not be restricted to multiples of the number of field periods. Extending SIESTA in this manner will allow larger, lower-order resonant islands to form in devices such as CTH. An example of a non-field-periodic perturbation in CTH will be demonstrated. Currently the code also operates in a stellarator-symmetric fashion in which an ``up-down'' symmetry is present at some toroidal angle. Nearly all of the current tokamaks (and ITER in the future) operate with a divertor and as such do not possess stellarator symmetry. Removal of this symmetry restriction requires including both sine and cosine terms in the Fourier expansion for the geometry of the device and the fields contained within. The current status of this extension of the code will be discussed, along with the method of implementation. U.S. DOE Contract No. DE-AC05-00OR22725 with UT-Battelle, LLC.

Advanced Toroidal Facility (ATF)

International Nuclear Information System (INIS)

Thompson, P.B.

1985-01-01

The Advanced Toroidal Facility (ATF) is a new magnetic plasma confinement device, under construction at Oak Ridge National Laboratory (ORNL), which will lead to improvements in toroidal magnetic fusion reactors. ATF is a type of stellarator known as a torsatron which theoretically has the capability at greater than or equal to8% beta in steady state. The ATF plasma has a major radius of 2.1 m, an average minor radius of 0.3 m, and a field of 2 T for a 5-s duration or 1 T steady state. The ATF device consists of a helical field (HF) coil set, a set of poloidal field (PF) coils, an exterior shell structure to support the coils, and a thin helically contoured vacuum vessel inside the coils. The ATF replaces the ISX-B tokamak at ORNL and will use the ISX-B auxiliary systems including 4 MW of neutral injection heating and 0.2 MW of electron cyclotron heating. ATF device is scheduled to start operation in the fall of 1986. An overview of the ATF device is presented including details of the construction process envisioned

Characterization of high temperature superconductor cables for magnet toroidal field coils of the DEMO fusion power plant

CERN Document Server

Bayer, Christoph M

2017-01-01

Nuclear fusion is a key technology to satisfy the basic demand for electric energy sustainably. The official EUROfusion schedule foresees a first industrial DEMOnstration Fusion Power Plant for 2050. In this work several high temperature superconductor sub-size cables are investigated for their applicability in large scale DEMO toroidal field coils. Main focus lies on the electromechanical stability under the influence of high Lorentz forces at peak magnetic fields of up to 12 T.

Characterization of high temperature superconductor cables for magnet toroidal field coils of the DEMO fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Bayer, Christoph M.

2017-05-01

Nuclear fusion is a key technology to satisfy the basic demand for electric energy sustainably. The official EUROfusion schedule foresees a first industrial DEMOnstration Fusion Power Plant for 2050. In this work several high temperature superconductor sub-size cables are investigated for their applicability in large scale DEMO toroidal field coils. Main focus lies on the electromechanical stability under the influence of high Lorentz forces at peak magnetic fields of up to 12 T.

Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

International Nuclear Information System (INIS)

Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

1976-01-01

This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended

Celebration for the ATLAS Barrel Toroid magnet

CERN Multimedia

2007-01-01

Representatives from Funding Agencies and Barrel Toroid Magnet Laboratories during the ceremony. From left to right: Jean Zinn-Justin (Head of DAPNIA/CEA/Saclay), CERN Director-General Robert Aymar, and Roberto Petronzio (President INFN).Allan Clark (DPNC University Geneva) and Enrique Fernandez (IFAE Barcelona) were among the guests visiting the ATLAS cavern. The barrel toroid is visible in the background. A celebration took place at Point 1 on 13 December to toast the recent powering-up of the ATLAS barrel toroid magnet to full field (Bulletin No. 47-48/06). About 70 guests were invited to attend, mainly composed of representatives from funding partners and key members of the laboratory management teams of the barrel toroid magnet, representing CEA France, INFN Italy, BMBF Germany, Spain, Sweden, Switzerland, Russia, JINR Dubna and CERN. An introductory speech by ATLAS spokesperson Peter Jenni the scene for evening. This was followed by the ATLAS magnet system project leader Herman Ten Kate's account of the...

ITER vacuum vessel, in vessel components and plasma facing materials

International Nuclear Information System (INIS)

Ioki, Kimihiro; Enoeda, M.; Federici, G.

2007-01-01

Design of the NB ports including duct liners under heat loads of the neutral beams has been developed. Design of the in-wall shielding has been developed in more details considering the supporting structure and the assembly method. The ferromagnetic inserts have previously not been installed in the outboard midplane region due to irregularity caused by the tangential ports for NB injection. Due to this configuration, the maximum ripple is relatively large (∝1 %) in a limited region of the plasma and the toroidal field flux lines fluctuate ∝10 mm in the FW region. To avoid these problems, additional ferromagnetic inserts are to be installed in the equatorial port region. Detailed studies were carried out on the ITER vacuum vessel to define appropriate codes and standards in the context of the ITER licensing in France. A set of draft documents regarding the ITER vacuum vessel structural code were prepared including an RCC-MR Addendum for the ITER VV with justified exceptions or modifications. The main deviation from the base Code is the extensive use of UT in lieu of radiography for the volumetric examination of all one-side access welds of the outer shell and field joint. The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. Development of new beryllium materials is progressing in China and Russia. The ITER limiters will be installed in equatorial ports at two toroidal locations. The limiter plasma-facing surface protrudes ∝8 cm from the FW during the start-up and shutdown phase. In the new limiter concept, the limiters are retracted by ∝8 cm during the plasma flat top phase. This concept gives important advantages; (i) mitigation of the particle and heat loads due to disruptions, ELMs and blobs, (ii) improvement of the power coupling with the ICRH antenna

Evaluation of mechanical strength of the joints in JT-60 toroidal field coil conductors

International Nuclear Information System (INIS)

Nishio, Satoshi; Ohkubo, Monoru; Sasajima, Hiroshi

1980-04-01

Toroidal field (TF) coils of JT-60 produce a toroidal field of 45 kG at a plasma axis, they have an inner bore of 3.90 m and a weight of about 80 metric tons per coil. Eighteen TF coils are located around a torus axis at regular intervals. TF coil conductors are mostly jointed by high frequency induction brazing, the rest jointed by welding. In deciding the details of the jointing procedures, the conductor size and the requested mechanical strength are mainly taken into consideration. Described are non-destructive inspection methods for the brazed joints, strength evaluation, and the inspection criteria. Ultrasonic testing method is found to be the most effective in evaluation of mechanical properties of the brazed joints especially in terms of fatigue strength. In section 1, specifications of the TF coils are given. In section 2, the ultrasonic inspection method and the detectability of this apparatus are described in detail, the defects of known size are compared with the indication values and display figures. The apparatus developed for JT-60 is operated automatically also recording the inspectionresults. In section 3, mechanical strength of the brazed joints with initial defects is discussed on the basis of Fracture Mechanics theory and results of the fatigue crack growth test. The inspection criteria in accordance with the descriptions of section 2 and 3 are given in section 4. (author)

Curvature-induced electrostatic drift modes in a toroidal plasma

International Nuclear Information System (INIS)

Venema, M.

1985-01-01

This thesis deals with a number of problems in the theory of linear stability of a hot, fully ionized plasma immersed in a strong magnetic field. The most widely used system to magnetically confine a plasma is the tokamak. This is a toroidal, current carrying device with a strong, externally imposed, magnetic field. The author discusses the linear theory of unstable, low-frequency waves in the gradient region, restricted to electrostatic waves. In that case the resulting radial fluxes of particles and energy are due to electric cross-field drifts. In the presence of magnetic fluctuations and small-scale reconnection phenomena, radial transport could also be predominantly along field lines. At present, it is not clear which of the two mechanisms is the dominant feature of the observed anomalous transport. First, the author introduces the theory of drift waves in toroidal geometry. Next, the electrostratic drift modes in toroidal geometry (weakly collisional regime), the equations for low-frequency waves in the strongly collisional regime and the electrostatic drift modes (strongly collisional regime) are discussed. (Auth.)

Divertor cassette movers prototypes for ITER

International Nuclear Information System (INIS)

Bogusch, E.; Batz, R.; Bieber, O.; Gottfried, R.; Cerdan, G.

1998-01-01

Following competitive tendering, in October 1996 Siemens was contracted by the European Commission to design and supply an assembly of four Divertor Cassette Movers Prototypes including the control and command systems for the movers proper. The assembly consisting of one Cassette Toroidal Mover (CTM), one Radial Mover Tractor (TRC), one Second Cassette Carrier (SCC), and one Radial Cassette Carrier (RCC) represents key components of the Divertor Test Platform at Brasimone, one of the seven large R+D projects for ITER. By detailed design, high-precision manufacturing and testing of these devices, Siemens contributed to the verification of an important task within the European R and D program towards ITER construction. Replacement of the divertor cassettes is a scheduled maintenance operation throughout the life of ITER. The successful fabrication and testing of the Divertor Cassette Movers Prototypes is all important milestone to verify this delicate operation. (authors)

Upgrade of DC power supply system in ITER CS model coil test facility

International Nuclear Information System (INIS)

Shimono, Mitsugu; Uno, Yasuhiro; Yamazaki, Keita; Kawano, Katsumi; Isono, Takaaki

2014-03-01

Objective of the ITER CS Model Coil Test Facility is to evaluate a large scale superconducting conductor for fusion using the Central Solenoid (CS) Model Coil, which can generate a 13T magnetic field in the inner bore with a 1.5 m diameter. The facility is composed of a helium refrigerator / liquefier system, a DC power supply system, a vacuum system and a data acquisition system. The DC power supply system supplies currents to two superconducting coils, the CS Model Coil and an insert coil. A 50-kA DC power supply is installed for the CS Model Coil and two 30 kA DC power supplies are installed for an insert coil. In order to evaluate superconducting performance of a conductor used for ITER Toroidal Field (TF) coils whose operating current is 68 kA, the line for an insert coil is upgraded. A 10 kA DC power supply was added, DC circuit breakers were upgraded, bus bars and current measuring instrument were replaced. In accordance to the upgrade, operation manual was revised. (author)

Effects of toroidal field ripple on injected deuterons in the FED device

International Nuclear Information System (INIS)

Fowler, R.H.; Rome, J.A.

1981-07-01

A Monte Carlo beam deposition and thermalization code is used to assess the effects of toroidal field (TF) ripple on injected fast deuterons in the Fusion Engineering Device (FED). The code uses realistic geometry for the beam, plasma equilibrium, TF ripple, and vacuum chamber. For injection at an angle of 35 0 (co) from perpendicular, no particles were ripple trapped and less than 1% of the injected power went to the wall and the limiter. However, due to the large amounts of computer time required by these programs, only 100 particles were followed in the rippled case and the results must be regarded as preliminary

Force-free field inside a toroidal magnetic cloud

Czech Academy of Sciences Publication Activity Database

Romashets, E. P.; Vandas, Marek

2003-01-01

Roč. 30, č. 20 (2003), s. 2065, /SSC 8-1 - SSC 8-4/ ISSN 0094-8276 R&D Projects: GA AV ČR IBS1003006; GA ČR GA205/03/0953 Institutional research plan: CEZ:AV0Z1003909 Keywords : magnetic clouds * toroid al flux rope * analytical solution Subject RIV: BN - Astronomy, Celestial Mechanics, Astrophysics Impact factor: 2.422, year: 2003

Theory for neoclassical toroidal plasma viscosity in tokamaks

International Nuclear Information System (INIS)

Shaing, K C; Chu, M S; Hsu, C T; Sabbagh, S A; Seol, Jae Chun; Sun, Y

2012-01-01

Error fields and magnetohydrodynamic modes break toroidal symmetry in tokamaks. The broken symmetry enhances the toroidal plasma viscosity, which results in a steady-state toroidal plasma flow. A theory for neoclassical toroidal plasma viscosity in the low-collisionality regimes is developed. It extends stellarator transport theory to include multiple modes and to allow for |m − nq| ∼ 1. Here, m is the poloidal mode number, n is the toroidal mode number and q is the safety factor. The bounce averaged drift kinetic equation is solved in several asymptotic limits to obtain transport fluxes. These fluxes depend non-linearly on the radial electric field except for those in the 1/ν regime. Here, ν is the collision frequency. The theory is refined to include the effects of the superbanana plateau resonance at the phase space boundary and the finite ∇B drift on the collisional boundary layer fluxes. Analytical expressions that connect all asymptotic limits are constructed and are in good agreement with the numerical results. The flux–force relations that relate transport fluxes to forces are used to illustrate the roles of transport fluxes in the momentum equation. It is shown that the ambipolar state is reached when the momentum equation is relaxed. It is also shown that the origin of the momentum for plasma flow generated without momentum sources is the local unbalance of particles' momenta and is diamagnetic in nature regardless of the details of the theory. (paper)

Low-frequency fluctuations in a pure toroidal magnetized plasma

Indian Academy of Sciences (India)

A magnetized, low- plasma in pure toroidal configuration is formed and extensively studied with ion mass as control parameter. Xenon, krypton and argon plasmas are formed at a fixed toroidal magnetic field of 0.024 T, with a peak density of ∼ 1011 cm-3, ∼ 4 × 1010 cm-3 and ∼ 2 × 1010 cm−3 respectively.

TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

International Nuclear Information System (INIS)

Hatch, A.J.

1975-01-01

1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

Self-consistent perturbed equilibrium with neoclassical toroidal torque in tokamaks

International Nuclear Information System (INIS)

Park, Jong-Kyu; Logan, Nikolas C.

2017-01-01

Toroidal torque is one of the most important consequences of non-axisymmetric fields in tokamaks. The well-known neoclassical toroidal viscosity (NTV) is due to the second-order toroidal force from anisotropic pressure tensor in the presence of these asymmetries. This work shows that the first-order toroidal force originating from the same anisotropic pressure tensor, despite having no flux surface average, can significantly modify the local perturbed force balance and thus must be included in perturbed equilibrium self-consistent with NTV. The force operator with an anisotropic pressure tensor is not self-adjoint when the NTV torque is finite and thus is solved directly for each component. This approach yields a modified, non-self-adjoint Euler-Lagrange equation that can be solved using a variety of common drift-kinetic models in generalized tokamak geometry. The resulting energy and torque integral provides a unique way to construct a torque response matrix, which contains all the information of self-consistent NTV torque profiles obtainable by applying non-axisymmetric fields to the plasma. This torque response matrix can then be used to systematically optimize non-axisymmetric field distributions for desired NTV profiles. Published by AIP Publishing.

ITER alpha particle diagnostics using knock-on ion tails

International Nuclear Information System (INIS)

Fisher, R.K.; Parks, P.B.; McChesney, J.M.

1995-09-01

Alpha particles will play a critical role in the physics and successful operation of ITER. Achieving fusion ignition requires that the α particles created by deuterium-tritium (D-T) reactions deposit a large fraction of their energy in the reacting plasma before they are lost. Toroidal field ripple can localize any alpha particle losses and cause first wall damage. We have proposed a new method of measuring the fast confined α-particle distribution in a reacting plasma. The same elastic collisions that transfer the alpha energy to the D-T plasma ions and allow fusion ignition will also create a high energy tail on the deuterium and tritium ion energy distributions. Some of these energetic tail ions will undergo fusion reactions with the background plasma producing neutrons whose energy is increased significantly above 14 MeV due to the kinetic energy of the reacting ions. Measurement of this high energy tail on the D-T neutron distribution as a function of plasma minor radius would provide information on the alpha density profile with a time response equal to the ion slowing-down time. Although this technique may provide only limited information on the α-particle energy distribution, experimental studies of fast ions on existing tokamaks have shown that the observed slowing-down is essentially classical. Hence the α-energy distribution is expected to be classical except in situations where the α-confinement is poor. The confinement of α's can be affected by ripple losses and a number of instabilities. Toroidal field ripple can cause both prompt orbit losses and stochastic ripple diffusion losses. Magnetohydrodynamic activity, including fishbone instabilities, toroidal Alfven eigenmodes, and sawtooth oscillations, may also affect alpha confinement. The diagnostic proposed here, by monitoring the confined alpha population, can provide valuable information on the confinement of fast alphas in a reacting plasma

Moving toroidal limiter

International Nuclear Information System (INIS)

Ikuta, Kazunari; Miyahara, Akira.

1983-06-01

The concept of the limiter-divertor proposed by Mirnov is extended to a toroidal limiter-divertor (which we call moving toroidal limiter) using the stream of ferromagnetic balls coated with a low Z materials such as plastics, graphite and ceramics. An important advantage of the use of the ferromagnetic materials would be possible soft landing of the balls on a catcher, provided that the temperature of the balls is below Curie point. Moreover, moving toroidal limiter would work as a protector of the first wall not only against the vertical movement of plasma ring but also against the violent inward motion driven by major disruption because the orbit of the ball in the case of moving toroidal limiter distributes over the small major radius side of the toroidal plasma. (author)

Application of plasma focus device to compression of toroidal plasma

International Nuclear Information System (INIS)

Ikuta, Kazunari

1980-01-01

A new concept of compressing a toroidal plasma using a plasma focus device is considered. Maximum compression ratio of toroidal plasma is determined merely by the initial density ratio of the toroidal plasma to a sheet plasma in a focus device because of the Rayleigh-Taylor instability. An initiation senario of plasma-linear is also proposed with a possible application of this concepts to the creation of a burning plasma in reversed field configurations, i.e., burning plasma vortex. (author)

Complete suppression of Pfirsch-Schlueter current in a toroidal l=3 stellarator

International Nuclear Information System (INIS)

Sato, Yasuhiko; Wakatani, Masahiro; Yokoyama, Masayuki; Pustovitov, V.D.

1999-10-01

Pfirsch-Schlueter (P-S) current is an inherent property of a finite pressure toroidal equilibrium of tokamak and stellarator. However, it was pointed out recently (V.D. Pustovitov, Nuclear Fusion 36 (1996) 583) that the P-S current would be suppressed completely if the external vertical field could be adjusted to satisfy the condition Ω= in an l=3 stellarator. Here Ω= 2 >/B 0 2 -2ε cosθ, l is a pole number, |B tilde| the vacuum helical magnetic field, B 0 the toroidal field, ε the inverse aspect ratio, θ the poloidal angle and denotes the average over the toroidal angle. An example of such a stellarator equilibrium is presented in this paper. For this stellarator equilibrium, behavior of rotational transform and Boozer magnetic spectrum is clarified when the pressure is increased. Both formation of helical magnetic axis and reduction of toroidal curvature are important ingredients to reduce the P-S current. However, the collisionless particle confinement is not improved in this example. (author)

2-D Reflectometer Modeling for Optimizing the ITER Low-field Side Reflectometer System

International Nuclear Information System (INIS)

Kramer, G.J.; Nazikian, R.; Valeo, E.J.; Budny, R.V.; Kessel, C.; Johnson, D.

2005-01-01

The response of a low-field side reflectometer system for ITER is simulated with a 2?D reflectometer code using a realistic plasma equilibrium. It is found that the reflected beam will often miss its launch point by as much as 40 cm and that a vertical array of receiving antennas is essential in order to observe a reflection on the low-field side of ITER

Impurity accumulation and performance of ITER and DEMO plasmas in the presence of transport barriers

International Nuclear Information System (INIS)

Chatthong, B; Promping, J; Onjun, T

2017-01-01

In this work, the impurity accumulations and their performance in the presence of both ITB and ETB in ITER and DEMO plasmas are investigated using a BALDUR integrated predictive modelling code. In these simulations, a combination of a neoclassical transport model NCLASS and an anomalous transport model Mixed Bohm/gyro-Bohm is used. The boundary condition is described at the top of the pedestal, which is calculated theoretically based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The toroidal flow is calculated based on the NTV (neoclassical toroidal viscosity) toroidal velocity model. The time evolution of plasma temperature and density profiles of ITER and DEMO (Korean K-DEMO and Japanese DEMO models A, B and C) plasmas are simulated in H -mode scenario with and without ITB formation. It is found that Japanese DEMO model C yields highest plasma temperature; while Korean DEMO yields the best plasma performance among those designs considered. Impurity accumulation is found to be highest in Japanese DEMO model B. (paper)

Saddle-splay screening and chiral symmetry breaking in toroidal nematics

OpenAIRE

Koning, Vinzenz; van Zuiden, Benjamin C.; Kamien, Randall D.; Vitelli, Vincenzo

2013-01-01

We present a theoretical study of director fields in toroidal geometries with degenerate planar boundary conditions. We find spontaneous chirality: despite the achiral nature of nematics the director configuration show a handedness if the toroid is thick enough. In the chiral state the director field displays a double twist, whereas in the achiral state there is only bend deformation. The critical thickness increases as the difference between the twist and saddle-splay moduli grows. A positiv...

The Direct Effect of Toroidal Magnetic Fields on Stellar Oscillations: An Analytical Expression for the General Matrix Element

Energy Technology Data Exchange (ETDEWEB)

Kiefer, René; Schad, Ariane; Roth, Markus [Kiepenheuer-Institut für Sonnenphysik, Schöneckstraße 6, D-79104 Freiburg (Germany)

2017-09-10

Where is the solar dynamo located and what is its modus operandi? These are still open questions in solar physics. Helio- and asteroseismology can help answer them by enabling us to study solar and stellar internal structures through global oscillations. The properties of solar and stellar acoustic modes are changing with the level of magnetic activity. However, until now, the inference on subsurface magnetic fields with seismic measures has been very limited. The aim of this paper is to develop a formalism to calculate the effect of large-scale toroidal magnetic fields on solar and stellar global oscillation eigenfunctions and eigenfrequencies. If the Lorentz force is added to the equilibrium equation of motion, stellar eigenmodes can couple. In quasi-degenerate perturbation theory, this coupling, also known as the direct effect, can be quantified by the general matrix element. We present the analytical expression of the matrix element for a superposition of subsurface zonal toroidal magnetic field configurations. The matrix element is important for forward calculations of perturbed solar and stellar eigenfunctions and frequency perturbations. The results presented here will help to ascertain solar and stellar large-scale subsurface magnetic fields, and their geometric configuration, strength, and change over the course of activity cycles.

Ferritic insertion for reduction of toroidal magnetic field ripple on JT-60U

International Nuclear Information System (INIS)

Shinohara, K.; Sakurai, S.; Ishikawa, M.; Tsuzuki, K.; Suzuki, Y.; Masaki, K.; Naito, O.; Kurihara, K.; Suzuki, T.; Koide, Y.; Fujita, T.; Miura, Y.

2007-01-01

Ferritic steel tiles (FSTs) have been installed to improve the energetic ion confinement by reducing a toroidal magnetic field ripple. Aiming at cost-effective installation, orbit-following calculations of energetic ions were carried out for a design of the installation of ferritic steel on the JT-60U by using the fully three dimensional magnetic field orbit-following Monte-Carlo (F3D OFMC) code, which had been developed for ferritic insert experiments on the JFT-2M and can treat the complex magnetic field structure produced by ferritic inserts. The installed FSTs add a non-linear magnetic field on magnetic sensors for plasma control and an equilibrium calculation. The code for real-time control has been modified to take into account the magnetic field by the FSTs. The plasma operation was successfully resumed after usual conditioning processes and real-time plasma control was successfully carried out. The heat load measurement indicates the improved confinement of energetic ions. These results are important for practical application of the ferritic steel, which is a leading candidate of a structural material on a DEMO reactor

Third-order transfer matrices calculated for an electrostatic toroidal sector condenser including fringing-field effects

CERN Document Server

Mordik, S N

2002-01-01

The third-order transfer matrices are calculated for an electrostatic toroidal sector condenser using a rigorously conserved matrix method that implies the conservation of the beam phase volume at each step in the calculations. The transfer matrices (matrizants) obtained, include the fringing-field effect due to the stray fields. In the case of a rectangular distribution of the field components along the optical axis, the analytical expressions for all aberration coefficients, including the dispersion ones, are derived accurate to the third-order terms. In simulations of real fields with the stray field width other than zero, a smooth distribution of the field components is used for which similar aberration coefficients were calculated by means of the conserved numerical method . It has been found that for a smooth model, as the stray field width tends to zero, the aberration coefficients approach the corresponding aberration values in the rectangular model.

Third-order transfer matrices calculated for an electrostatic toroidal sector condenser including fringing-field effects

International Nuclear Information System (INIS)

Mordik, S.N.; Ponomarev, A.G.

2002-01-01

The third-order transfer matrices are calculated for an electrostatic toroidal sector condenser using a rigorously conserved matrix method that implies the conservation of the beam phase volume at each step in the calculations. The transfer matrices (matrizants) obtained, include the fringing-field effect due to the stray fields. In the case of a rectangular distribution of the field components along the optical axis, the analytical expressions for all aberration coefficients, including the dispersion ones, are derived accurate to the third-order terms. In simulations of real fields with the stray field width other than zero, a smooth distribution of the field components is used for which similar aberration coefficients were calculated by means of the conserved numerical method . It has been found that for a smooth model, as the stray field width tends to zero, the aberration coefficients approach the corresponding aberration values in the rectangular model

Structural analysis of the ITER vacuum vessel from disruption loading with halo asymmetry

International Nuclear Information System (INIS)

Riemer, B.W.; Sayer, R.O.

1996-01-01

Static structural analyses of the ITER vacuum vessel were performed with toroidally asymmetric disruption loads. Asymmetric halo current conditions were assumed to modify symmetric disruption loads which resulted in net lateral loading on the vacuum vessel torus. Structural analyses with the asymmetric loading indicated significantly higher vessel stress and blanket support forces than with symmetric disruption loads. A recent change in the vessel support design which provided toroidal constraints at each mid port was found to be effective in reducing torus lateral movement and vessel stress

New Toroid shielding design

CERN Multimedia

Hedberg V

On the 15th of June 2001 the EB approved a new conceptual design for the toroid shield. In the old design, shown in the left part of the figure above, the moderator part of the shielding (JTV) was situated both in the warm and cold areas of the forward toroid. It consisted both of rings of polyethylene and hundreds of blocks of polyethylene (or an epoxy resin) inside the toroid vacuum vessel. In the new design, shown to the right in the figure above, only the rings remain inside the toroid. To compensate for the loss of moderator in the toroid, the copper plug (JTT) has been reduced in radius so that a layer of borated polyethylene can be placed around it (see figure below). The new design gives significant cost-savings and is easier to produce in the tight time schedule of the forward toroid. Since the amount of copper is reduced the weight that has to be carried by the toroid is also reduced. Outgassing into the toroid vacuum was a potential problem in the old design and this is now avoided. The main ...

Studies on limiter confined toroidal plasma in BETA

International Nuclear Information System (INIS)

Bera, D.; Reddy, C.; Jayakumar, R.; Kaw, P.K.

1984-01-01

Plasma equilibrium and stability in the presence of a toroidal magnetic field and a poloidal limiter is being studied experimentally in the BETA experiment. In a simple toroidal magnetic field, plasma cannot be in equilibrium because of the effect of the magnetic field curvature, which tends to expand the plasma. The electric field, which causes this expansion, is short circuited if a poloidal conducting limiter is placed and this brings about a quasi-equilibrium. In the model the charge separation current flows on the surface of the plasma and closes the path by flowing parallel to the magnetic field away from the limiter and transverse to the field at the limiter. For such an equilibrium, the vertical pressure profile is expected to be uniform, while the radial pressure profile is determined by transport. Such a profile is unstable to Rayleigh-Taylor instabilities, if the magnetic field gradient and the pressure gradient have the same sense

Remote replacement of TF [toroidal field] and PF [poloidal field] coils for the compact ignition tokamak

International Nuclear Information System (INIS)

Macdonald, D.; Watkin, D.C.; Hollis, M.J.; DePew, R.E.; Kuban, D.P.

1990-01-01

The use of deuterium-tritium fuel in the Compact Ignition Tokamak will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion devices auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. In-vessel remote maintenance included replacement of divertor and first wall hardware, faraday shields, and for an in-vessel inspection system. Provision for remote replacement of a vacuum vessel sector, toroidal field coil or poloidal field ring coil was not included in the project baseline. As a result of recent coil failures experienced at a number of facilities, the CIT project decided to reconsider the question of remote recovery from a coil failure and, in January of 1990, initiated a coil replacement study. This study focused on the technical requirements and impact on fusion machine design associated with remote recovery from any coil failure

Samus Toroid Installation Fixture

Energy Technology Data Exchange (ETDEWEB)

Stredde, H.; /Fermilab

1990-06-27

The SAMUS (Small Angle Muon System) toroids have been designed and fabricated in the USSR and delivered to D0 ready for installation into the D0 detector. These toroids will be installed into the aperture of the EF's (End Toroids). The aperture in the EF's is 72-inch vertically and 66-inch horizontally. The Samus toroid is 70-inch vertically by 64-inch horizontally by 66-inch long and weighs approximately 38 tons. The Samus toroid has a 20-inch by 20-inch aperture in the center and it is through this aperture that the lift fixture must fit. The toroid must be 'threaded' through the EF aperture. Further, the Samus toroid coils are wound about the vertical portion of the aperture and thus limit the area where a lift fixture can make contact and not damage the coils. The fixture is designed to lift along a surface adjacent to the coils, but with clearance to the coil and with contact to the upper steel block of the toroid. The lift and installation will be done with the 50 ton crane at DO. The fixture was tested by lifting the Samus Toroid 2-inch off the floor and holding the weight for 10 minutes. Deflection was as predicted by the design calculations. Enclosed are sketches of the fixture and it relation to both Toroids (Samus and EF), along with hand calculations and an Finite Element Analysis. The PEA work was done by Kay Weber of the Accelerator Engineering Department.

Activation of the concrete in the bio shield of ITER

International Nuclear Information System (INIS)

Kalcheva, S.

2005-02-01

Calculations of neutron spectra in different parts of the tokamak building of ITER are performed. A computational geometry model of the tokamak building is prepared using MCNP-4C. The model includes adequate material composition and geometry description of the main parts of the tokamak for PPCS plant model A: toroidal field coils, vacuum vessel, shield, blanket structure, first wall, divertor, 14.1 MeV neutron source. The design and the dimensions of the bio shield are taken from the current ITER design. MCNP calculations of the neutron spectra in the bio shield (concrete) of ITER are performed, using the neutron spectra in TF coils calculated at UKAEA as external neutron source. The neutron spectra in the concrete calculated by MCNP are used as input data in the code EASY99 for estimations of the activation of the concrete in the bio shield around the tokamak. The time evolutions of the maximum (in the bio shield floor) and minimum (in the bio shield side walls) specific activity (Bq/kg) and dose rate (Sv/h.) of the main dominant nuclides in the concrete are evaluated and compared for 3 different concrete types, used as biological shield in the PWR and BR3 reactors. (author)

Eigenvalues of relaxed toroidal plasmas of arbitrary sharp edged cross sections. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

Khalil, Sh M [Plasma Physics and Nuclear Fusion Department, Nuclear Research Center, Atomic Energy Authority, Cairo, (Egypt)

1996-03-01

Relaxed (force-free) toroidal plasmas described by the equations cur 1 B={mu}B, and grad {mu}=O (B is the magnetic field) induces interest in nuclear fusion. Its solution is perceived to describe the gross of the reversed field pinch (RFP), spheromak configuration, current limitation in toroidal plasmas, and others. The parameter {mu} plays an important roll in relaxed states. It cannot exceed the smallest eigenvalue {mu} (min), and that for a toroidal discharge there is a maximum toroidal current which is connected to this value. The values of{mu} were calculated numerically, using the methods of collocation points, for toroids of arbitrary aspect ratio {alpha} ({alpha} = R/a, ratio of major/minor radii of tokamak) and arbitrary curved cross-sections (circle, ellipse, cassini, and D-shaped). The aim of present work is to prove the applicability of the numerical methods for calculating the eigenvalues for toroidal plasmas having sharp edged cross sections and arbitrary aspect ratio. The lowest eigenvalue {mu} (min), satisfy the boundary condition {beta} .n = O (or RB. = O) for which the toroidal flux are calculated. These are the zero field eigenvalues of the equation cur 1 b={mu}B. The poloidal magnetic field lines corresponding to different cross sections are shown by plotting the boundary condition B.n=O. The plots showed good fulfillment of the boundary condition along the whole boundaries of different cross sections. Dependence of eigenvalues {mu}a on aspect ratio {alpha} is also obtained. Several runs of the programme with various wave numbers K showed that {mu}a is very insensitive to the choice of K. 8 figs.

Confinement margins for ignition and driven operation in Iter Eda ID

International Nuclear Information System (INIS)

Johner, J.

1995-09-01

Preliminary calculations for ITER EDA ID have been performed using the 1/2D thermal equilibrium code HELIOS. It is found that: - The maximum ignition margin for ITER ID (29%) is 6% less than for ITER OD (35%) and 5% less than for ITER CDA (34%). - Decreasing the ration τ * He /τ E from the nominal value 10 to a value of 5 gives a 12% gain in the maximum ignition margin. Increasing the ration from 10 to 15 causes a 22% loss in the margin. Furthermore, ignited equilibria non longer exist for τ * He /τ E ≥ 17.6. - Operation in driven mode with 50 MW of external power increases the confinement capability by 13%. With 100 MW, the improvement is 24%. - Lowering the fusion power from 1500 to 1000 MW slightly improves the maximum ignition margin (+5%) and allows operation below the Greenwald density limit. - A 10% reduction of the toroidal magnetic field with a correlative diminution of the plasma current for constant safety factor operation, causes a dramatic reduction (-18%) of the maximum ignition margin. - A fraction of neon of 0.68% would completely suppress the ignition margin. Furthermore, ignited equilibria, with the nominal fusion power and τ * He /τ E , no longer exist when the neon fraction exceeds 0.75%. (Author). 2 refs., 10 figs

Design and development of the ITER vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Koizumi, K.; Nakahira, M.; Itou, Y.; Tada, E. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan); Johnson, G.; Ioki, K.; Elio, F.; Iizuka, T.; Sannazzaro, G.; Takahashi, K.; Utin, Y.; Onozuka, M. [ITER Joint Central Team (JCT), Garching (Germany); Nelson, B. [US Home Team, Oak Ridge National Laboratory (United States); Vallone, C. [EU Home Team, NET Team, Garching (Germany); Kuzmin, E. [RF Home Team, Efremov Institute, City (Russian Federation)

1998-09-01

In ITER, the vacuum vessel (VV) is designed to be a water cooled, double-walled toroidal structure made of 316LN stainless steel with a D-shaped cross section approximately 9 m wide and 15 m high. The design work which began at the beginning of the ITER-EDA is nearing completion by resolving the technical issues. In parallel with the design activities, the R and D program, full-scale VV sector model project, was initiated in 1995 to resolve the design and fabrication issues. The full-scale sector model corresponds to an 18 sector (9 sub-sector x 2) and is being fabricated on schedule. To date, 60% of the fabrication had been completed. The fabrication of full-scale model including sector-to-sector connection will be completed by the end of 1997 and performance tests are scheduled until the end of ITER-EDA. This paper describes the latest status of the ITER VV design and the full-scale sector model project. (orig.) 3 refs.

Verification of SuperMC with ITER C-Lite neutronic model

Energy Technology Data Exchange (ETDEWEB)

Zhang, Shu [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Yu, Shengpeng [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); He, Peng, E-mail: peng.he@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2016-12-15

Highlights: • Verification of the SuperMC Monte Carlo transport code with ITER C-Lite model. • The modeling of the ITER C-Lite model using the latest SuperMC/MCAM. • All the calculated quantities are consistent with MCNP well. • Efficient variance reduction methods are adopted to accelerate the calculation. - Abstract: In pursit of accurate and high fidelity simulation, the reference model of ITER is becoming more and more detailed and complicated. Due to the complexity in geometry and the thick shielding of the reference model, the accurate modeling and precise simulaion of fusion neutronics are very challenging. Facing these difficulties, SuperMC, the Monte Carlo simulation software system developed by the FDS Team, has optimized its CAD interface for the automatic converting of more complicated models and increased its calculation efficiency with advanced variance reduction methods To demonstrate its capabilites of automatic modeling, neutron/photon coupled simulation and visual analysis for the ITER facility, numerical benchmarks using the ITER C-Lite neutronic model were performed. The nuclear heating in divertor and inboard toroidal field (TF) coils and a global neutron flux map were evaluated. All the calculated nuclear heating is compared with the results of the MCNP code and good consistencies between the two codes is shown. Using the global variance reduction methods in SuperMC, the average speed-up is 292 times for the calculation of inboard TF coils nuclear heating, and 91 times for the calculation of global flux map, compared with the analog run. These tests have shown that SuperMC is suitable for the design and analysis of ITER facility.

Effects of magnetic geometry, fluctuations, and electric fields on confinement in the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Isler, R.C.; Aceto, S.; Baylor, L.R.; Bigelow, T.S.; Bell, G.L.; Bell, J.D.; Carreras, B.A.; Colchin, R.J.; Crume, E.C.; Dominguez, N.; Dory, R.A.; Dunlap, J.L.; Dyer, G.R.; England, A.C.; Gandy, R.F.; Glowienka, J.C.; Hanson, G.R.; Harris, J.H.; Hiroe, S.; Horton, L.D.; Jernigan, T.C.; Ji, H.; Langley, R.A.; Lee, D.K.; Likin, K.M.; Lyon, J.F.; Ma, C.H.; Morimoto, S.; Murakami, M.; Okada, H.; Qualls, A.L.; Rasmussen, D.A.; Rome, J.A.; Sato, M.; Schwelberger, J.G.; Shats, M.G.; Simpkins, J.E.; Thomas, C.E.; Uckan, T.; Wade, M.R.; Wilgen, J.B.; Wing, W.R.; Yamada, H.; Zielinski, J.J.

1992-01-01

Recent experiments in the Advanced Toroidal Facility (ATF) [Fusion Technol. 10, 179 (1986)] have been directed toward investigations of the basic physics mechanisms that control confinement in this device. Measurements of the density fluctuations throughout the plasma volume have provided indications for the existence of theoretically predicted dissipative trapped electron and resistive interchange instabilities. These identifications are supported by results of dynamic configuration scans of the magnetic fields during which the magnetic well volume, shear, and fraction of confined trapped particles are changed continuously. The influence of magnetic islands on the global confinement has been studied by deliberately applying error fields which strongly perturb the nested flux-surface geometry, and the effects of electric fields have been investigated by means of biased limiter experiments

Pure tension superconducting toroidal-field coil system design studies for the Argonne Experimental Power Reactor

International Nuclear Information System (INIS)

Wang, S.T.; Purcell, J.R.; Demichele, D.W.; Turner, L.R.

1975-11-01

As part of the Argonne Tokamak Experimental Power Reactor (TEPR) design studies, a toroidal field (TF) coil system has been designed. NbTi was chosen as the most suitable superconductor and 8T was regarded as a practical peak field level in this study. The 16-coil design was chosen as a reasonable compromise between 2 percent field ripple and 3 m access gap. To minimize the coil structure and the bending moments on the conductor, a pure tension coil shape is necessary. A correct approach for determining the pure tension coil profile in a bumpy TF coil system is given. Verification of the pure tension coil by a three-dimensional stress analysis is presented. For coil quench protection, a series-connected scheme is proposed

Formation of a compact torus using a toroidal plasma gun

International Nuclear Information System (INIS)

Levine, M.A.; Pincosy, P.A.

1981-01-01

Myers, Levine and Pincosy earlier reported results using a toroidal plasma gun. The device differs from the usual coaxial plasma gun in the use of a strong toroidal bias current for enhanced efficiency, a pair of disk-like accelerating electrodes for reduced viscosity and a fast pulsed toroidal gas valve for more effective use of the injected gas sample. In addition, a technique is used for generating a toroidal current in the plasma ring. The combination offers an opportunity to deliver a plasma with a large amount of energy and to vary the density and relative toroidal and poloidal magnetic field intensities over a range of values. It is the purpose of this paper to report further experimental results, to project the gun's applications to the formation of a compact torus, and to propose a simple modification of the present apparatus as a test

Design considerations for ITER [International Thermonuclear Experimental Reactor] magnet systems: Revision 1

International Nuclear Information System (INIS)

Henning, C.D.; Miller, J.R.

1988-01-01

The International Thermonuclear Experimental Reactor (ITER) is now completing a definition phase as a beginning of a three-year design effort. Preliminary parameters for the superconducting magnet system have been established to guide further and more detailed design work. Radiation tolerance of the superconductors and insulators has been of prime importance, since it sets requirements for the neutron-shield dimension and sensitively influences reactor size. The major levels of mechanical stress in the structure appear in the cases of the inboard legs of the toroidal-field (TF) coils. The cases of the poloidal-field (PF) coils must be made thin or segmented to minimize eddy current heating during inductive plasma operation. As a result, the winding packs of both the TF and PF coils includes significant fractions of steel. The TF winding pack provides support against in-plane separating loads but offers little support against out-of-plane loads, unless shear-bonding of the conductors can be maintained. The removal of heat due to nuclear and ac loads has not been a fundamental limit to design, but certainly has non-negligible economic consequences. We present here preliminary ITER magnet systems design parameters taken from trade studies, designs, and analyses performed by the Home Teams of the four ITER participants, by the ITER Magnet Design Unit in Garching, and by other participants at workshops organized by the Magnet Design Unit. The work presented here reflects the efforts of many, but the responsibility for the opinions expressed is the authors'. 4 refs., 3 figs., 4 tabs

Alfven Eigenmode Stability with Beams in ITER-like Plasma

International Nuclear Information System (INIS)

Gorelenkov, N.N.; Berk, H.L.; Budny, R.V.

2004-01-01

Toroidicity Alfven Eigenmodes (TAE) in ITER can be driven unstable by two groups of energetic particles, the 3.5 MeV α-particle fusion products and the tangentially injected 1MeV beam ions. Stability conditions are established using the perturbative NOVA/NOVA-K codes. A quasi-linear diffusion model is then used to assess the induced redistribution of energetic particles

Effect of alpha particles on Toroidal Alfven Eigenmodes

International Nuclear Information System (INIS)

Berk, H.L.

1992-11-01

An overview is given of the analytic structure for the linear theory of the Toroidal Alfven Eigenmode (TAE), where multiple gap structures occur. A discussion is given of the alpha particle drive and the various dissipation mechanisms that can stabilize the system. A self-consistent calculation of the TAE mode, for a low-beta high-aspect-ratio plasma, indicates that though the alpha particle drive is comparable to the dissipation mechanisms, overall stability is still achieved for ignited ITER-like plasma. A brief discussion is given of the nonlinear theory for the TAE mode and how nonlinear alpha particle dynamics can be treated by mapping methods

Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 21, Materials and processes selection. Volume 2

International Nuclear Information System (INIS)

Smith, B.R.

1995-01-01

This document identifies the candidate materials and manufacturing processes selected for development of the TPX Toroidal Field (TF) Magnet. Supporting rationale and selection criteria are provided for justification and the materials properties database report is included for completeness. Specific properties for each material selection are included in this document

Guiding Center Equations in Toroidal Equilibria

International Nuclear Information System (INIS)

White, Roscoe; Zakharov, Leonid

2002-01-01

Guiding center equations for particle motion in a general toroidal magnetic equilibrium configuration are derived using magnetic coordinates. Previous derivations made use of Boozer coordinates, in which the poloidal and toroidal angle variables are chosen so that the Jacobian is inversely proportional to the square of the magnetic field. It is shown that the equations for guiding center motion in any equilibrium possessing nested flux surfaces have exactly the same simple form as those derived in this special case. This allows the use of more spatially uniform coordinates instead of the Boozer coordinates, greatly increasing the accuracy of calculations in large beta and strongly shaped equilibria

Compact toroids with Alfvenic flows

International Nuclear Information System (INIS)

Wang Zhehui; Tang, X.Z.

2004-01-01

The Chandrasekhar equilibria form a class of stationary ideal magnetohydrodynamics equilibria stabilized by magnetic-field-aligned Alfvenic flows. Analytic solutions of the Chandrasekhar equilibria are explicitly constructed for both field-reversed configurations and spheromaks. Favorable confinement property of nested closed flux surfaces and the ideal magnetohydrodynamic stability of the compact toroids are of interest for both magnetic trapping of high energy electrons in astrophysics and confinement of high temperature plasmas in laboratory

Conceptual design of SC magnet system for ITER, (2)

International Nuclear Information System (INIS)

Koizumi, Koichi; Hasegawa, Mitsuru; Yoshida, Kiyoshi

1991-08-01

The International Thermonuclear Experimental Reactor (ITER) is an experimental tokamak machine testing the basic plasma performance and technologies required for future tokamak reactor. The design proposals for the Superconducting (SC) Magnet System from Japan were summarized by the Fusion Experimental Reactor (FER) Design Team and the Superconducting Magnet Laboratory of the Japan Atomic Energy Research Institute (JAERI). This report is one of the series reports on 'Conceptual design of superconducting magnet system for ITER', and describes the major results of the stress analysis regarding the Toroidal Field (TF) coil, the Center Solenoid (CS) coil and the Equilibrium Field (EF) coil and their support structures. Among the design issues, the mechanical design of the coil system was one of the most critical items, not only because of the huge electromagnetic loads due to large size and high magnetic field, but also because of the demand of high reliability under neutron irradiation. In order to satisfy both the coil performance and the mechanical reliability, different types of conductors were employed for each coils. The mechanical behaviors and the safety margin of each coil were analyzed by using finite element method (FEM) of MSC/NASTRAN. The procedure to obtain the equivalent winding stiffness employed for the each FEM analysis is also described in this report. The details on the coil specifications, conductor design and mechanical design for each coils are described in other report of the series reports. (J.P.N.)

Detailed electromagnetic numerical evaluation of eddy currents induced by toroidal and poloidal magnetic field variation and halo currents

International Nuclear Information System (INIS)

Roccella, M.; Marin, A.; Lucca, F.; Merola, M.

2008-01-01

A detailed evaluation of the EM loads in the ITER divertor during plasma disruptions is mandatory for the correct dimensioning of the divertor component. The EM loads during plasma disruptions are mainly produced by: (1) toroidal flux variation (TFV) during the thermal quench (TQ) and current quench (CQ); (2) halo currents (HC); and (3) poloidal flux variation (PFV) during TQ and CQ phase. The new ITER reference disruption and the last changes in the divertor design have been considered in the EM models created to calculate all the EM loads due to TFV, HC and PFV. All the analyses have been performed for the three different main design options of the divertor plasma facing units (PFU). The effects of PFV have been analyzed using an EM-zooming procedure that has allowed a good detail of the component model, while new numerical approaches have been developed for the evaluation of the effects due to TFV and HC maintaining the same detail for the divertor model. Separate models have been developed to evaluate the equivalent electrical resistivities of the various PFU options; this allows in the full 3D model a strong simplification of a geometry which would otherwise be very complex. The effect of an electrical surface bridging of the PFU castellation has also been taken into account

Detailed electromagnetic numerical evaluation of eddy currents induced by toroidal and poloidal magnetic field variation and halo currents

Energy Technology Data Exchange (ETDEWEB)

Roccella, M. [L.T. Calcoli S.a.S. Piazza Prinetti, 26/B, Merate (Lecco) (Italy)], E-mail: massimo.roccella@ltcalcoli.it; Marin, A.; Lucca, F. [L.T. Calcoli S.a.S. Piazza Prinetti, 26/B, Merate (Lecco) (Italy); Merola, M. [ITER Team, Cadarache (France)

2008-12-15

A detailed evaluation of the EM loads in the ITER divertor during plasma disruptions is mandatory for the correct dimensioning of the divertor component. The EM loads during plasma disruptions are mainly produced by: (1) toroidal flux variation (TFV) during the thermal quench (TQ) and current quench (CQ); (2) halo currents (HC); and (3) poloidal flux variation (PFV) during TQ and CQ phase. The new ITER reference disruption and the last changes in the divertor design have been considered in the EM models created to calculate all the EM loads due to TFV, HC and PFV. All the analyses have been performed for the three different main design options of the divertor plasma facing units (PFU). The effects of PFV have been analyzed using an EM-zooming procedure that has allowed a good detail of the component model, while new numerical approaches have been developed for the evaluation of the effects due to TFV and HC maintaining the same detail for the divertor model. Separate models have been developed to evaluate the equivalent electrical resistivities of the various PFU options; this allows in the full 3D model a strong simplification of a geometry which would otherwise be very complex. The effect of an electrical surface bridging of the PFU castellation has also been taken into account.

Magnetic analysis of the magnetic field reduction system of the ITER neutral beam injector

Energy Technology Data Exchange (ETDEWEB)

Barrera, Germán, E-mail: german.barrera@ciemat.es [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Ahedo, Begoña; Alonso, Javier; Ríos, Luis [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Chareyre, Julien; El-Ouazzani, Anass [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 07/08, 08019 Barcelona (Spain)

2015-10-15

The neutral beam system for ITER consists of two heating and current drive neutral beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. For the correct operation of the beam, the ion source and the ion path until it is neutralized must operate under a very low magnetic field environment. To prevent the stray ITER field from penetrating inside those mentioned critical areas, a magnetic field reduction system (MFRS) will envelop the beam vessels and the high voltage transmission lines to ion source. This system comprises the passive magnetic shield (PMS), a box like assembly of thick low carbon steel plates, and the Active Correction and Compensation Coils (ACCC), a set of coils carrying a current which depends on the tokamak stray field. This paper describes the magnetic model and analysis results presented at the PMS and ACCC preliminary design review held in ITER organization in April 2013. The paper focuses on the magnetic model description and on the description of the analysis results. The iterative process for obtaining optimized currents in the coils is presented. The set of coils currents chosen among the many possible solutions, the magnetic field results in the interest regions and the fulfillment of the magnetic field requirements are described.

Numerical stress analysis of toroidal coil by three-dimensional finite element method

International Nuclear Information System (INIS)

Nishimura, Hidetomo; Shimamoto, Susumu

1977-10-01

A structure analysis program based on finite element method for toroidal coils, developed in JAERI, and its example application to a medium-size tokamak are described. In this application, the effects of material anisotropy, poloidal field and spring constant value were studied, and also the influence of toroidal coil failure on the peak stress. The following were revealed. The effect of anisotropy on the peak stress in reinforcement must be considered. The effect of poloidal field on the peak stress is small compared with that of toroidal field. The spring constant value between coil and support does not much influence the peak stress value, The peak stress in reinforcement rises with increasing number of failed coils. In the case of 2000 nodes on the structure, CPU time with the program is about 40 min. (auth.)

Measurement of toroidal plasma current in RF heated helical plasmas

International Nuclear Information System (INIS)

Besshou, Sakae

1993-01-01

This report describes the measurement of toroidal plasma current by a semiflexible Rogowski coil in a helical vacuum chamber. A Rogowski coil measures the toroidal plasma current with a resolution of 0.1 kA, frequency range of up to 1 kHz and sensitivity of 6.5 x 10 -9 V · s/A. We measured the spontaneous toroidal plasma current (from -1.2 to +1.2 kA) under electron cyclotron resonance heating at 0.94 T toroidal field in the Heliotron-E device. We found that the measured direction of toroidal plasma current changes its sign as in the predicted behavior of a neoclassical diffusion-driven bootstrap current, depending on the horizontal position of the plasma column. We explain the observed plasma currents in terms of the compound phenomenon of an ohmic current and a neoclassical diffusion-driven current. The magnitude of the neoclassical current component is smaller than the value predicted by a collisionless neoclassical theory. (author)

Experiments and Simulations of ITER-like Plasmas in Alcator C-Mod

International Nuclear Information System (INIS)

Wilson, R.; Kessel, C.E.; Wolfe, S.; Hutchinson, I.H.; Bonoli, P.; Fiore, C.; Hubbard, A.E.; Hughes, J.; Lin, Y.; Ma, Y.; Mikkelsen, D.; Reinke, M.; Scott, S.; Sips, A.C.C.; Wukitch, S.

2010-01-01

Alcator C-Mod is performing ITER-like experiments to benchmark and verify projections to 15 MA ELMy H-mode Inductive ITER discharges. The main focus has been on the transient ramp phases. The plasma current in C-Mod is 1.3 MA and toroidal field is 5.4 T. Both Ohmic and ion cyclotron (ICRF) heated discharges are examined. Plasma current rampup experiments have demonstrated that (ICRF and LH) heating in the rise phase can save voltseconds (V-s), as was predicted for ITER by simulations, but showed that the ICRF had no effect on the current profile versus Ohmic discharges. Rampdown experiments show an overcurrent in the Ohmic coil (OH) at the H to L transition, which can be mitigated by remaining in H-mode into the rampdown. Experiments have shown that when the EDA H-mode is preserved well into the rampdown phase, the density and temperature pedestal heights decrease during the plasma current rampdown. Simulations of the full C-Mod discharges have been done with the Tokamak Simulation Code (TSC) and the Coppi-Tang energy transport model is used with modified settings to provide the best fit to the experimental electron temperature profile. Other transport models have been examined also.

Loads specification and embedded plate definition for the ITER cryoline system

Science.gov (United States)

Badgujar, S.; Benkheira, L.; Chalifour, M.; Forgeas, A.; Shah, N.; Vaghela, H.; Sarkar, B.

2015-12-01

ITER cryolines (CLs) are complex network of vacuum-insulated multi and single process pipe lines, distributed in three different areas at ITER site. The CLs will support different operating loads during the machine life-time; either considered as nominal, occasional or exceptional. The major loads, which form the design basis are inertial, pressure, temperature, assembly, magnetic, snow, wind, enforced relative displacement and are put together in loads specification. Based on the defined load combinations, conceptual estimation of reaction loads have been carried out for the lines located inside the Tokamak building. Adequate numbers of embedded plates (EPs) per line have been defined and integrated in the building design. The finalization of building EPs to support the lines, before the detailed design, is one of the major design challenges as the usual logic of the design may alter. At the ITER project level, it was important to finalize EPs to allow adequate design and timely availability of the Tokamak building. The paper describes the single loads, load combinations considered in load specification and the approach for conceptual load estimation and selection of EPs for Toroidal Field (TF) Cryoline as an example by converting the load combinations in two main load categories; pressure and seismic.

Toroidal simulation magnet tests

International Nuclear Information System (INIS)

Walstrom, P.L.; Domm, T.C.

1975-01-01

A number of different schemes for testing superconducting coils in a simulated tokamak environment are analyzed for their merits relative to a set of test criteria. Two of the concepts are examined in more detail: the so-called cluster test scheme, which employs two large background field coils, one on either side of the test coil, and the compact torus, a low-aspect ratio toroidal array of a small number of coils in which all of the coils are essentially test coils. Simulation of the pulsed fields of the tokamak is discussed briefly

A Conceptual Design and Structural Analysis for ITER Mid-plane Brace Tools

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Kihak; Robert, Shaw [ITER Organization, St Paul lez Durance Cedex (France)

2010-10-15

The ITER, International Thermonuclear Experimental Reactor, Tokamak machine is mainly composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The ITER Tokamak assembly tools are purpose-built tools to assemble the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the IO (ITER international organization), Korea has carried out the conceptual design of assembly tools with IO cooperation. The 40 .deg. sector assemblies attached mid-plane brace tools sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. The sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and mid-plane brace tools. The mid-plane brace tool is assembled to inner surface of VV and TFCs in phase of sector sub-assembly after completion of all sector components. VV, TFC and VVTS are separated fully before completion of 9 sectors at Tokamak in-pit. In this paper the mid-plane brace tools is introduced about function, structure and status of research and development are also described

Propulsion using the electron spiral toroid

International Nuclear Information System (INIS)

Seward, Clint

1998-01-01

A new propulsion method is proposed which could potentially reduce propellant needed for space travel by three orders of magnitude. It uses the newly patented electron spiral toroid (EST), which stores energy as magnetic field energy. The EST is a hollow toroid of electrons, all spiraling in parallel paths in a thin outer shell. The electrons satisfy the coupling condition, forming an electron matrix. Stability is assured as long as the coupling condition is satisfied. The EST is held in place with a small external electric field; without an external magnetic field. The EST system is contained in a vacuum chamber. The EST can be thought of as an energetic entity, with electrons at 10,000 electron volts. Propulsion would not use combustion, but would heat propellant through elastic collisions with the EST surface and eject them for thrust. Chemical rocket combustion heats propellant to 4000 deg. C; an EST will potentially heat the propellant 29,000 times as much, reducing propellant needs accordingly. The thrust can be turned ON and OFF. The EST can be recharged as needed

Effects of toroidal field ripple on suprathermal ions in tokamak plasmas

International Nuclear Information System (INIS)

Goldston, R.J.; Towner, H.H.

1980-02-01

Analytic calculations of three important effects of toroidal field ripple on suprathermal ions in tokamak plasmas are presented. In the first process, collisional ripple-trapping, beam ions become trapped in local magnetic wells near their banana tips due to pitch-angle scattering as they traverse the ripple on barely unripple-trapped orbits. In the second process, collisionless ripple-trapping, near-perpendicular untrapped ions are captured (again near a banana tip) due to their finite orbits, which carry them out into regions of higher ripple. In the third process, banana-drift diffusion, fast-ion banana orbits fail to close precisely, due to a ripple-induced variable lingering period near the banana tips. These three mechanisms lead to substantial radial transport of banana-trapped, neutral-beam-injected ions when the quantity α* identical with epsilon/sin theta/Nqdelta is of order unity or smaller

Effects of toroidal field ripple on suprathermal ions in tokamak plasmas

International Nuclear Information System (INIS)

Goldston, R.J.; Towner, H.H.

1981-01-01

Analytic calculations of three important effects of toroidal field ripple on suprathermal ions in tokamak plasmas are presented. In the first process, collisional ripple-trapping, ions become trapped in local magnetic wells near their banana tips owing to pitch-angle scattering as they traverse the ripple on barely unripple-trapped orbits. In the second process, collisionless ripple-trapping, ions are captured (again near a banana tip) owing to their finite orbits, which carry them out into regions of higher ripple. In the third process, banana-drift diffusion, fast-ion banana orbits fail to close precisely, due to a ripple-induced 'variable lingering period' near the banana tips. These three mechanisms lead to substantial radial transport of banana-trapped, neutral-beam-injected ions when the quantity α* is identical with epsilonsinthetaNqdelta is of order unity or smaller. (author)

Mechanical design and construction qualification program on ITER correction coils structures

Energy Technology Data Exchange (ETDEWEB)

Foussat, A., E-mail: arnaud.foussat@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Weiyue, Wu; Jing, Wei; Shuangsong, Du [Academy of Science Institute of Plasma Physics, PO 1126, Hefei, Anhui 230031 (China); Sgobba, S. [European Center for Nuclear Research, CH-1211 Geneva 23 (Switzerland); Hongwei, Li [China International Nuclear Fusion Energy Program Execution Center, Ministry of Science and Technology, 15B Fuxing Rd., Beijing 100862 (China); Libeyre, Paul; Jong, Cornelis; Klofac, Kamil; Mitchell, Neil [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2014-04-01

The ITER Magnet system consists of 4 main coils sub-systems, i.e. 18 toroidal field coils (TFC), a central solenoid (CS), 6 poloidal field coils (PF) and 3 sets of correction coils (CC). The ITER fusion project has selected the stainless steel 316LN as main material for the magnet structure. The CC contribute to reducing the range of magnetic error fields created by imperfections in the location and geometry of the other coils used to confine, heat, and shape the plasma. During plasma operation, a large number of loading condition scenarios have been considered and structural analysis performed on key items like Cable-In-Conduit Conductor and the coil case. The results obtained are used for both static and fatigue structural assessment defining the present baseline design. For the construction of the structural cases, welding techniques such as GTAW (Gas Tungsten Arc Welding) and techniques resulting in low distortion and shrinkage like EBW (Electron Beam Welding) or Laser Beam Welding (LBW) with filler metal wire have been selected. Those methods are considered for future qualifications to guarantee proper weld parameters and specified weld properties. In order to determine the strength and fracture toughness of 316LN stainless steel welds with respect to design criteria, some mechanical tests have been carried out at 7 K (or 77 K), and room temperature.

Supporting structures of the toroidal field coils of intor-net

International Nuclear Information System (INIS)

Farfaletti-Casali, F.; Biggio, M.; Macco, A.; Perfumo, A.; Reale, M.

1984-01-01

The design of the toroidal field (TF) coil supporting structures for INTOR-NET Phase 2A (European Configuration) is described. In order to identify the proposed design several preliminary evaluations were performed. These evaluations indicated that suitable supporting structures are those shown in the annexed figures, where each coil is guided and centered by a separate reinforcing structure, on which the intercoil structures are attached. A simplified structural analysis was carried out considering only the cyclic out-of-plane loads due to the poloidal field, acting on the coils, since this constitutes the most critical loading condition due to its impact on the fatigue behavior of the material. This analysis was performed with finite element method for displacements and stresses calculations by using the SAP IV-code. The calcaulation model was made with the aid of the GIFTS system. The results show that the maximum equivalent stress does not exceed the stress limit assumed in the INTOR specifications for stainless steel (AISI 316 type) at cryogenic temperature and under cyclic loading, for the operating conditions of INTOR. Consequently the proposed concept for the support of the TF coils can be considered a workable scheme. Further detailed analysis must be done, so as to demonstrate the complete feasibility of the system. (orig.)

Design features of HTMR-hybrid toroidal magnet tokamak reactor

International Nuclear Information System (INIS)

Rosatelli, F.; Avanzini, P.G.; Derchi, D.; Magnasco, M.; Grattarola, M.; Peluffo, M.; Raia, G.; Brunelli, B.; Zampaglione, V.

1984-01-01

The HTMR (Hybrid Toroidal Magnet Tokamak Reactor) conceptual design is aimed to demonstrate the feasibility of a Tokamak reactor which could fulfil the scientific and technological objectives expected from next generation devices with size and costs as small as possible. A hybrid toroidal field magnet, made up by copper and superconducting coils, seems to be a promising solution, allowing a considerable flexibility in machine performances, so as to gain useful margins in front of the uncertainties in confinement time scaling laws and beta and plasma density limits. The optimization procedure for the hybrid magnet, configuration, the main design features of HTMR and the preliminary mechanical calculations of the superconducting toroidal coils are described. (author)

Modal analysis of a stiffened toroidal shell sector

International Nuclear Information System (INIS)

Cerreta, R.; Di Pietro, E.; Pizzuto, A.

1987-01-01

This paper presents the results of the modal analysis of a sector of the toroidal vacuum vessel of a new experimental machine for research in the field of controlled thermonuclear fusion (FTU - Frascati Tokamak Upgrade). The vacuum vessel, one of the most critical components of the experimental device, consist of 12 stainless steel toroidal sectors, and it is designed to withstand pulsed electromagnetic loads during operation. Results of the modal analysis of the stiffened toroidal shell sector are compared and discussed with regard to the experimental data. Theoretical eigenvalues and eigenvectors have been predicted by means of ABAQUS finite element code. Experimental analysis has been carried out on a full scale model and natural frequencies have been measured. Satisfactory agreement between experimental and theoretical eigenvalues has been found

Structural analysis of the NET toroidal field coils and conductor

International Nuclear Information System (INIS)

Mitchell, N.; Collier, D.; Gori, R.

1989-01-01

The NET toroidal field coils will utilise A15-type superconductor at 4.2 K to generate fields up to 11.5 T. The superconductor strands themselves are sensitive to strain, which causes degradation of their current carrying capacity, and thus the detailed behaviour of the coil conductor must be analysied so that the strian can be minimised. This analysis must include the manufacturing processes of the conductor as well as the normal and abnormal loperational loads. The conductor will be insulated and bonded by glass fibre reinforced epoxy resin, with limited bonding shear strength, and the overall support of the complete coil system must be designed to reduce these shear stresses. The coils will be subjected to pulse loads form the poloidal field coils, and analysis of the slip between the various coil components, such as conductors and the coil case, giving rise to frictional heating and possible loss of superconducting properties is another important factor, which has been investigated by a number of stress analyses. The manufacturing, thermal and normal magnetic loads on the coils and the analysis leading to the proposed structural design are described. In addition to the normal operating conditions, there is a range of abnormal load conditions which could result from electrical or mechanical faults on the coils. The effect of these potential faults has been analysed and the coil design modified to prevent catastrophic structural failure. (author). 13 refs.; 8 figs.; 1 tab

Proceedings of the third symposium on the physics and technology of compact toroids in the magnetic fusion energy program

International Nuclear Information System (INIS)

Siemon, R.E.

1981-03-01

This document contains papers contributed by the participants of the Third Symposium on Physics and Technology of Compact Toroids in the Magnetic Fusion Energy Program. Subjects include reactor aspects of compact toroids, energetic particle rings, spheromak configurations (a mixture of toroidal and poloidal fields), and field-reversed configurations

Stationary magnetohydrodynamic equilibrium of toroidal plasma in rotation

International Nuclear Information System (INIS)

Missiato, O.

1986-01-01

The stationary equations of classical magnetohydrodynamics are utilized to study the toroidal motion of a thermonuclear magnetically - confined plasma with toroidal symmetry (Tokamak). In the present work, we considered a purely toroidal stationary rotation and te problem is reduced to studing a second order partial differencial equation of eliptic type Maschke-Perrin. Assuming that the temperature remains constant on the magnetic surfaces, an analitic solution, valid for low Mach numbers (M ≤ 0 .4), was obtained for the above-mentioned equation by means of a technique developed by Pantuso Sudano. From the solution found, we traced graphs for the quantities which described the equilibrium state of the plasma, namely: mass density, pressure, temperature, electric current density and toroidal magnetic field. Finally we compare this analitical model with others works which utilized differents analitical models and numerical simulations. We conclude that the solutions obtained are in good agreement with the previos results. In addition, however, our model contains the results of Sudano-Goes with the additional advantage of employing much simple analitical expressions. (author) [pt

Structural design of the toroidal configuration of the HTS SMES cooling system

International Nuclear Information System (INIS)

Yeom, H.K.; Koh, D.Y.; Ko, J.S.; Kim, H.B.; Hong, Y.J.; Kim, S.H.; Seong, K.C.

2011-01-01

The superconducting magnetic energy storage (SMES) system is working on around 30 K, because the magnet is made of high temperature superconductor. To maintain the cryogenic temperature, the superconducting coil is cooled by cryogen, helium gas or liquid neon. But there are some weak points in the cryogen cooling system. For example periodic charge of the cryogen and size is big and so on. So, we have designed the conduction cooling system for toroidal configuration HTS SMES. The toroidal type HTS SMES has some merits, so it is very small magnetic field leakage, and magnetic field applied perpendicular to the tape surface can be reduced. Our system has 28 numbers of HTS double pancake coils and they are arrayed toroidal configuration. The toroidal inner radius is 162 mm, and outer radius is 599 mm, and height is about 162 mm. In this study, we have designed the cooling structure and analyzed temperature distribution of cooling path, thermal stress and deformation of the cooling structure.

Fatigue behavior of an insulation system for the ITER magnets

International Nuclear Information System (INIS)

Prokopec, R.; Humer, K.; Weber, H.W.

2006-01-01

The application of glass-fiber reinforced plastics as insulation materials for fusion magnet coils (e.g. the Toroidal Field Coils of ITER) requires the full characterization of their mechanical performance under ITER-relevant conditions. One of the methods of testing material's response under dynamic load is the tension-tension fatigue procedure. This test can be used to simulate the pulsed tokamak-operation of the ITER coils over a lifetime of more than 20 years. Furthermore, it provides information on the maximum tensile or shear stress in the ITER-relevant range of 10 4 -10 5 cycles. In order to simulate the operation conditions of ITER as closely as possible, several fatigue parameters can be set in the test programme, e.g., the minimum-to-peak stress ratio R and the frequency ν of the sinusoidal load function. Further, the fatigue process can be run under load or strain control. All of these parameters may influence the mechanical response of the insulation system under cyclic load. Therefore, it is highly desirable to investigate the influence of test parameter variations on the measured stress-lifetime diagrams. The investigations were performed at 77 K using an industrial glass-fiber reinforced composite impregnated with epoxy resin. For both the load and the strain controlled mode, R-values of 0.3 and 0.5 and a frequency of 10 Hz were chosen. The results showed almost no deviations in the lifetime behavior between the load and the strain controlled mode, up to the ITER specified number of pulses, i.e. 3 x 10 4 cycles. Beyond this point, the residual strength levels were lower by 5-30 % under strain control than under load control. This effect is more pronounced at higher cycle numbers and for lower R-ratios. (author)

Non-invasive diagnostics of ion beams in strong toroidal magnetic fields with standard CMOS cameras

Science.gov (United States)

Ates, Adem; Ates, Yakup; Niebuhr, Heiko; Ratzinger, Ulrich

2018-01-01

A superconducting Figure-8 stellarator type magnetostatic Storage Ring (F8SR) is under investigation at the Institute for Applied Physics (IAP) at Goethe University Frankfurt. Besides numerical simulations on an optimized design for beam transport and injection a scaled down (0.6T) experiment with two 30°toroidal magnets is set up for further investigations. A great challenge is the development of a non-destructive, magnetically insensitive and flexible detector for local investigations of an ion beam propagating through the toroidal magnetostatic field. This paper introduces a new way of beam path measurement by residual gas monitoring. It uses a single board camera connected to a standard single board computer by a camera serial interface all placed inside the vacuum chamber. First experiments with one camera were done and in a next step two under 90 degree arranged cameras were installed. With the help of the two cameras which are moveable along the beam pipe the theoretical predictions are experimentally verified successfully. Previous experimental results have been confirmed. The transport of H+ and H2+ ion beams with energies of 7 keV and at beam currents of about 1 mA is investigated successfully.

Microwave produced plasma in a Toroidal Device

Science.gov (United States)

Singh, A. K.; Edwards, W. F.; Held, E. D.

2010-11-01

A currentless toroidal plasma device exhibits a large range of interesting basic plasma physics phenomena. Such a device is not in equilibrium in a strict magneto hydrodynamic sense. There are many sources of free energy in the form of gradients in plasma density, temperature, the background magnetic field and the curvature of the magnetic field. These free energy sources excite waves and instabilities which have been the focus of studies in several devices in last two decades. A full understanding of these simple plasmas is far from complete. At Utah State University we have recently designed and installed a microwave plasma generation system on a small tokamak borrowed from the University of Saskatchewan, Saskatoon, Canada. Microwaves are generated at 2.45 GHz in a pulsed dc mode using a magnetron from a commercial kitchen microwave oven. The device is equipped with horizontal and vertical magnetic fields and a transformer to impose a toroidal electric field for current drive. Plasmas can be obtained over a wide range of pressure with and without magnetic fields. We present some preliminary measurements of plasma density and potential profiles. Measurements of plasma temperature at different operating conditions are also presented.

Qualification tests for ITER TF conductors in SULTAN

International Nuclear Information System (INIS)

Bruzzone, P.; Stepanov, B.; Wesche, R.

2009-01-01

From February 2007 to May 2008, 18 short length conductor sections have been tested in SULTAN for design verification and manufacturer qualification of the ITER Toroidal Field (TF) conductor. The test program is focussed on the current sharing temperature, T cs , at the nominal operating conditions, 68 kA current and 11.15 T effective field, which can be fully reproduced in the SULTAN test facility. A broad range of results was observed, with over 2 K difference among the T cs of the conductors. In average, the results are poorer compared to the potential performance estimated from the strand scaling law. The key parameters to mitigate the degradation are not yet clearly identified. The experimental challenges to test conductors with performance degradation are highlighted, including enhanced instrumentation sets, the application of gas flow calorimetry to sense the current sharing power and the post-processing of voltage data to cancel the transverse potential across the cable. The updated schedule of the tests in SULTAN is presented with the short-term action plan for conductor test.

Basic performance tests on vibration of support structure with flexible plates for ITER tokamak device

International Nuclear Information System (INIS)

Takeda, Nobukazu; Kakudate, Satoshi; Shibanuma, Kiyoshi

2005-01-01

The vibration experiments of the support structures with flexible plates for the ITER major components such as toroidal field coil (TF coil) and vacuum vessel (VV) were performed using small-sized flexible plates aiming to obtain its basic mechanical characteristics such as dependence of the stiffness on the loading angle. The experimental results obtained by the hammering and frequency sweep tests were agreed each other, so that the experimental method is found to be reliable. In addition, the experimental results were compared with the analytical ones in order to estimate an adequate analytical model for ITER support structure with flexible plates. As a result, the bolt connection of the flexible plates on the base plate strongly affected on the stiffness of the flexible plates. After studies of modeling the bolts, it is found that the analytical results modeling the bolts with finite stiffness only in the axial direction and infinite stiffness in the other directions agree well with the experimental ones. Using this adequate model, the stiffness of the support structure with flexible plates for the ITER major components can be calculated precisely in order to estimate the dynamic behaviors such as eigen modes and amplitude of deformation of the major components of the ITER tokamak device. (author)

Application of calorimetry to the assessment of the performance of ITER Nb3Sn TF conductor samples in SULTAN tests

International Nuclear Information System (INIS)

Richard, L Savoldi; Zanino, R

2008-01-01

In the frame of the International Thermonuclear Experimental Reactor (ITER), several short full-size Nb 3 Sn samples of candidate toroidal field (TF) conductors were tested in 2007 at the SULTAN facility, PSI Villigen, Switzerland, in conditions relevant to the ITER TF (background magnetic field of 10.78 T and transport current of 68 kA). The performance of a SULTAN sample is determined by the current sharing temperature T CS . This can be obtained in principle from voltage measurements along the conductor sample, but the procedure is not free of issues and ambiguities. Here a complementary approach, based on the calorimetric assessment of the Joule heating due to current sharing, is critically discussed. Suitable algorithms are defined and the respective error bars are estimated, also based on numerical thermal-hydraulic modeling. The calorimetric approach is then applied to assess the performance of the samples tested in 2007 and compared with the results of the standard (electrical) approach

Superconducting endcap toroid design report

Energy Technology Data Exchange (ETDEWEB)

Walters, C.R.; Baynham, D.E.; Holtom, E.; Coombs, R.C.

1992-10-01

The Atlas Experiment proposed for the LHC machine will use toroidal magnet systems to achieve high muon momentum resolutions. One of the options under consideration is an air cored superconducting toroidal magnet system consisting of a long barrel toroid with small and cap toroids inserted in it to provide high resolution at high pseudorapidity. The design of the barrel toroid has been studied over the past two years and the design outline is given in a Saclay Report. More recently consideration has been given to an end cap toroid system which is based on air cored superconducting coils. This report presents the basic engineering design of such a system, the proposals for fabrication, assembly and installation, and an outline cost estimate for one end cap is presented in Appendix 1.

Drifts, currents, and power scrape-off width in SOLPS-ITER modeling of DIII-D

International Nuclear Information System (INIS)

Meier, E. T.; Goldston, R. J.; Kaveeva, E. G.; Makowski, M. A.; Mordijck, S.

2016-01-01

The effects of drifts and associated flows and currents on the width of the parallel heat flux channel (λ q ) in the tokamak scrape-off layer (SOL) are analyzed using the SOLPS-ITER 2D fluid transport code. Motivation is supplied by Goldston’s heuristic drift (HD) model for λ q , which yields the same approximately inverse poloidal magnetic field dependence seen in multi-machine regression. The analysis, focusing on a DIII-D H-mode discharge, reveals HD-like features, including comparable density and temperature fall-off lengths in the SOL, and up-down ion pressure asymmetry that allows net cross-separatrix ion magnetic drift flux to exceed net anomalous ion flux. In experimentally relevant high-recycling cases, scans of both toroidal and poloidal magnetic field (B tor and B pol ) are conducted, showing minimal λ q dependence on either component of the field. Insensitivity to B tor is expected, and suggests that SOLPS-ITER is effectively capturing some aspects of HD physics. Absence of λ q dependence on B pol , however, is inconsistent with both the HD model and experimental results. As a result, the inconsistency is attributed to strong variation in the parallel Mach number, which violates one of the premises of the HD model.

Low-loss metamaterial electromagnetically induced transparency based on electric toroidal dipolar response

Energy Technology Data Exchange (ETDEWEB)

Li, Hai-ming; Liu, Shao-bin, E-mail: lsb@nuaa.edu.cn; Liu, Si-yuan; Ding, Guo-wen; Yang, Hua; Yu, Zhi-yang; Zhang, Hai-feng [Key Laboratory of Radar Imaging and Microwave Photonics, Nanjing University of Aeronautics and Astronautics, Nanjing, 210016 (China); Wang, Shen-yun [Research Center of Applied Electromagnetic, Nanjing University of Information Science and Technology, Nanjing, 210044 (China)

2015-02-23

In this paper, a low-loss and high transmission analogy of electromagnetically induced transparency based on electric toroidal dipolar response is numerically and experimentally demonstrated. It is obtained by the excitation of the low-loss electric toroidal dipolar response, which confines the magnetic field inside a dielectric substrate with toroidal geometry. The metamaterial electromagnetically induced transparency (EIT) structure is composed of the cut wire and asymmetric split-ring resonators. The transmission level is as high as 0.88, and the radiation loss is greatly suppressed, which can be proved by the surface currents distributions, the magnetic field distributions, and the imaginary parts of the effective permeability and permittivity. It offers an effective way to produce low-loss and high transmission metamaterial EIT.

Measurement of tokamak error fields using plasma response and its applicability to ITER

International Nuclear Information System (INIS)

Strait, E.J.; Buttery, R.J.; Chu, M.S.; Garofalo, A.M.; La Haye, R.J.; Schaffer, M.J.; Casper, T.A.; Gribov, Y.; Hanson, J.M.; Reimerdes, H.; Volpe, F.A.

2014-01-01

The nonlinear response of a low-beta tokamak plasma to non-axisymmetric fields offers an alternative to direct measurement of the non-axisymmetric part of the vacuum magnetic fields, often termed ‘error fields’. Possible approaches are discussed for determination of error fields and the required current in non-axisymmetric correction coils, with an emphasis on two relatively new methods: measurement of the torque balance on a saturated magnetic island, and measurement of the braking of plasma rotation in the absence of an island. The former is well suited to ohmically heated discharges, while the latter is more appropriate for discharges with a modest amount of neutral beam heating to drive rotation. Both can potentially provide continuous measurements during a discharge, subject to the limitation of a minimum averaging time. The applicability of these methods to ITER is discussed, and an estimate is made of their uncertainties in light of the specifications of ITER's diagnostic systems. The use of plasma response-based techniques in normal ITER operational scenarios may allow identification of the error field contributions by individual central solenoid coils, but identification of the individual contributions by the outer poloidal field coils or other sources is less likely to be feasible. (paper)

Design features of HTMR-Hybrid Toroidal Magnet Tokamak Reactor

International Nuclear Information System (INIS)

Rosatelli, F.; Avanzini, P.G.; Brunelli, B.; Derchi, D.; Magnasco, M.; Grattarola, M.; Peluffo, M.; Raia, G.; Zampaglione, V.

1985-01-01

The HTMR (Hybrid Toroidal Magnet Tokamak Reactor) conceptual design is aimed to demonstrate the feasibility of a Tokamak reactor which could fulfill the scientific and technological objectives expected from next generation devices (e.g. INTOR-NET) with size and costs as small as possible. An hybrid toroidal field magnet, made up by copper and superconducting coils, seems to be a promising solution, allowing a considerable flexibility in machine performances, so as to gain useful margins in front of the uncertainties in confinement time scaling laws and beta and plasma density limits. In this paper the authors describe the optimization procedure for the hybrid magnet configuration, the main design features of HTMR and the preliminary mechanical calculations of the superconducting toroidal coils

Design point selection for an ignited ITER

International Nuclear Information System (INIS)

Spears, W.R.; Mizoguchi, T.; Perkins, J.; Putvinskij, S.; Flanagan, C.

1989-01-01

The parameter choice for the ITER design is dominated by the need to achieve ignition. Different energy confinement time scalings extrapolate to different device parameters, and the final design chosen must have a reasonable chance of success under the presently favoured scalings. This study investigates the device performance over the plasma current - aspect ratio design space, using sets of simplified equations and more detailed parametric analysis codes, with a view to clarifying the acceptable parameter region. Alterations to the initial assumptions of plasma elongation, plasma safety factor and peak toroidal field, and the effect of limiting the wall load, are also investigated. The study concludes that a device with aspect ratio around 3 and plasma current about 20 MA would be an appropriate choice. Under the assumptions of the study such a device would have a major radius of about 5.8m and would require the attainment of enhancement factors around 1.5 over L-mode scaling. (author). 6 refs, 11 figs, 3 tabs

Alfven continuum with toroidicity

International Nuclear Information System (INIS)

Riyopoulos, S.; Mahajan, S.M.

1985-06-01

The symmetry property of the MHD wave propagation operator is utilized to express the toroidal eigenmodes as a superposition of the mutually orthogonal cylindrical modes. Because of the degeneracy among cylindrical modes with the same frequency but resonant surfaces of different helicity the toroidal perturbation produces a zeroth order mixing of the above modes. The toroidal eigenmodes of frequency ω 0 2 have multiple resonant surfaces, with each surface shifted relative to its cylindrical position and carrying a multispectral content. Thus a single helicity toroidal antenna of frequency ω 0 couples strongly to all different helicity resonant surfaces with matching local Alfven frequency. Zeroth order coupling between modes in the continuum and global Alfven modes also results from toroidicity and degeneracy. Our perturbation technique is the MHD counterpart of the quantum mechanical methods and is applicable through the entire range of the MHD spectrum

Toroidally Resolved Structure of Divertor Heat Flux in RMP H-mode Discharges on DIII-D

International Nuclear Information System (INIS)

Jakubowski, M.W.; Evans, T.E.; Fenstermacher, M.E.; Lasnier, C.J.; Wolf, R.C.; Baylor, Larry R.; Boedo, J.A.; Burrell, K.H.; DeGrassie, J.S.; Gohil, P.; Mordijck, S.; Laengner, R.; Leonard, A.W.; Moyer, R.A.; Petrie, T.W.; Petty, C.C.; Pinsker, R.I.; Rhodes, T.L.; Schaffer, M.J.; Schmitz, O.; Snyder, P.B.; Stoschus, H.; Osborne, T.H.; Orlov, D.M.; Unterberg, Ezekial A.; Watkins, J.G.

2011-01-01

As shown on DIII-D edge localized modes (ELMs) can be either completely eliminated or mitigated with resonant magnetic perturbation (RMP) fields. Two infrared cameras, separated 105 degrees toroidally, were used to make simultaneous measurements of ELM heat loads with high frame rates. Without the RMP fields ELMs display a variety of different heat load dynamics and a range of toroidal variability that is characteristic of their 3D structure. Comparing radial averages there is no asymmetry between two toroidal locations. With RMP-mitigated ELMs, the variability in the radially averaged power loads is significantly reduced and toroidal asymmetries in power loads are introduced. In addition to RMP ELM suppression scenarios an RMP scenario with only very small ELMs and very good confinement has been achieved.

Calculations of toroidal EXTRAP equilibria for different toroidal ring current configurations

International Nuclear Information System (INIS)

Drake, J.R.; Scheffel, J.

1985-12-01

EXTRAP is a concept in which a pure Z-pinch is generated along the axis of an octupole field. Experiments in a linear as well as in a sector geometry have demonstrated that the pinch becomes stable against instabilities for many Alfven times. The octupole field in EXTRAP is produced by four, external, current-carrying rings. In the toroidal geometry these rings must be supplemented by additional rings to compensate for the plasma loop force and transformer core leakage flux. Equilibrium studies are carried out for two basically ring designs. The studies are based on numerical equilibrium copmputations using the GOYA code. Sensitivity of the equilibrium to technical imperfections is also analyzed. (author)

Models for large superconducting toroidal magnet systems

International Nuclear Information System (INIS)

Arendt, F.; Brechna, H.; Erb, J.; Komarek, P.; Krauth, H.; Maurer, W.

1976-01-01

Prior to the design of large GJ toroidal magnet systems it is appropriate to procure small scale models, which can simulate their pertinent properties and allow to investigate their relevant phenomena. The important feature of the model is to show under which circumstances the system performance can be extrapolated to large magnets. Based on parameters such as the maximum magnetic field and the current density, the maximum tolerable magneto-mechanical stresses, a simple method of designing model magnets is presented. It is shown how pertinent design parameters are changed when the toroidal dimensions are altered. In addition some conductor cost estimations are given based on reactor power output and wall loading

Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel

Science.gov (United States)

Maheshwari, A.; Pathak, H. A.; Mehta, B. K.; Phull, G. S.; Laad, R.; Shaikh, M. S.; George, S.; Joshi, K.; Khan, Z.

2017-04-01

ITER Vacuum Vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with In-Wall Shielding Blocks (IWS) and Water. The main purpose of IWS is to provide neutron shielding during ITER plasma operation and to reduce ripple of Toroidal Magnetic Field (TF). Although In-Wall Shield Blocks (IWS) will be submerged in water in between the walls of the ITER Vacuum Vessel (VV), Outgassing Rate (OGR) of IWS materials plays a significant role in leak detection of Vacuum Vessel of ITER. Thermal Outgassing Rate of a material critically depends on the Surface Roughness of material. During leak detection process using RGA equipped Leak detector and tracer gas Helium, there will be a spill over of mass 3 and mass 2 to mass 4 which creates a background reading. Helium background will have contribution of Hydrogen too. So it is necessary to ensure the low OGR of Hydrogen. To achieve an effective leak test it is required to obtain a background below 1 × 10-8 mbar 1 s-1 and hence the maximum Outgassing rate of IWS Materials should comply with the maximum Outgassing rate required for hydrogen i.e. 1 x 10-10 mbar 1 s-1 cm-2 at room temperature. As IWS Materials are special materials developed for ITER project, it is necessary to ensure the compliance of Outgassing rate with the requirement. There is a possibility of diffusing the gasses in material at the time of production. So, to validate the production process of materials as well as manufacturing of final product from this material, three coupons of each IWS material have been manufactured with the same technique which is being used in manufacturing of IWS blocks. Manufacturing records of these coupons have been approved by ITER-IO (International Organization). Outgassing rates of these coupons have been measured at room temperature and found in acceptable limit to obtain the required Helium Background. On the basis of these measurements, test reports have been generated and got

Test of piezo-ceramic motor technology in ITER relevant high magnetic fields

Energy Technology Data Exchange (ETDEWEB)

Monti, Chiara, E-mail: chiara.monti@enea.it [Associazione EURATOM-ENEA sulla Fusione, via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Besi Vetrella, Ugo; Mugnaini, Giampiero; Neri, Carlo; Rossi, Paolo; Viola, Rosario [Associazione EURATOM-ENEA sulla Fusione, via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Dubus, Gregory; Damiani, Carlo [Fusion for Energy, c/ Josep Pla, 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

2014-10-15

In the framework of a Fusion for Energy (F4E) grant, a test campaign started in 2012 in order to assess the performance of the in-vessel viewing system (IVVS) probe concept and to verify its compatibility when exposed to ITER typical working conditions. ENEA laboratories went through with several tests simulating high magnetic fields, high temperature, high vacuum, gamma radiation and neutron radiation. A customized motor has been adopted to study the performances of ultrasonic piezo motors technology in high magnetic field conditions. This paper reports on the testing activity performed on the motor in a multi Tesla magnetic field. The job was carried out in a test facility of ENEA laboratories able to achieve 14 T. A maximum field of 10 T, fully compliant with ITER requirements (8 T), was applied. A specific mechanical assembly has been designed and manufactured to hold the motor in the region with high homogeneity of the field. Results obtained so far indicate that the motor is compatible with high magnetic fields, and are presented in the paper.

Simulation of environment effects on retroreflectors in ITER

International Nuclear Information System (INIS)

Voitsenya, V.S.; Donne, A.J.H.; Bardamid, A.F.; Belyaeva, A.I.; Berezhnyj, V.L.; Galuza, A.A.; Gil, Ch.; Konovalov, V.G.; Lipa, M.; Malaquais, A.; Naidenkova, D.I.; Ryzhkov, V.I.; Schunke, B.; Solodovchenko, S.I.; Topkov, A.N.

2005-01-01

Two plasma diagnostics in ITER will use cube-corner reflectors (CCR): poloidal polarimetry and toroidal interferometry/polarimetry. The multichannel poloidal polarimetry system is planned to operate at a wavelength of 118.8 μm. The multichannel toroidal interferometry/polarimetry system is based on a CO 2 laser operating at wavelengths of 10.6 and 9.27 μm. The long term sputtering by charge exchange atoms and/or deposition of carbon-based (or beryllium-based) contaminant layers can affect the optical properties of the CCR. The role of both these potentially deleterious effects on the CCR operation is analyzed in this article, taking into account the probing beam wavelength and the CCR locations. The conclusion is that for the intended use of a CCR in the poloidal polarimetry at 118 μm neither erosion nor deposition should pose a problem. On the other hand, in the toroidal interferometry/polarimetry system operating at 10 μm, care must be taken to reduce the charge exchange atom flux and it is likely that deposition will occur and could have deleterious effects especially on the polarization properties of the reflected radiation

Observations of toroidal and poloidal rotation in the high beta tokamak Torus II

International Nuclear Information System (INIS)

Kostek, C.A.

1983-01-01

The macroscopic rotation of plasma in a toroidal containment device is an important feature of the equilibrium. Toroidal and poloidal rotation in the high beta tokamak Torus II is measured experimentally by examining the Doppler shift of the 4685.75 A He II line emitted from the plasma. The toroidal flow at an average velocity of 1.6 x 10 6 cm/sec, a small fraction of the ion thermal speed, moves in the same direction as the toroidal plasma current. The poloidal flow follows the ion diamagnetic current direction, also at an average speed of 1.6 x 10 6 cm/sec. In view of certain ordering parameters, the toroidal flow is compared with predictions from neoclassical theory in the collosional, Pfirsch-Schluter regime. The poloidal motion, however results from an E x B drift in a positive radial electric field, approaching a stable ambipolar state. This radial electric field is determined from theory by using the measured poloidal velocity. Mechanisms for the time evolution of rotation are also examined. It appears that the circulation damping is governed by a global decay of the temperature and density gradients which, in turn, may be functions of radiative cooling, loss of equilibrium due to external field decay, or the emergence of a growing instability, occasionally observed in CO 2 interferometry measurements

Form coefficient of helical toroidal solenoids

International Nuclear Information System (INIS)

Amelin, V.Z.; Kunchenko, V.B.

1982-01-01

For toroidal solenoids with continuous spiral coil, winded according to the laws of equiinclined and simple cylindrical spirals with homogeneous, linearly increasing to the coil periphery and ''Bitter'' distribution of current density, the analytical expressions for the dependence between capacity consumed and generated magnetic field, expressions for coefficients of form similar to Fabry coefficient for cylindrical solenoids are obtained and dependence of the form coefficient and relative volume of solenoid conductor on the number of revolutions of screw line per one circumvention over the large torus radius is also investigated. Analytical expressions of form coefficients and graphical material permit to select the optimum geometry as to capacity consumed both for spiral (including ''force-free'') and conventional toroidal solenoids of magnetic systems in thermonulear installations

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)

Implementation of vertically asymmetric toroidal-field ripple for beam heating of tokamak reactor plasmas

International Nuclear Information System (INIS)

Jassby, D.L.; Sheffield, G.V.; Towner, H.H.; Weissenburger, D.W.

1976-10-01

The neutral-beam energy required for adequate penetration of tokamak plasmas of high opacity can be reduced by a large factor if the beam is injected vertically into a region of large TF (toroidal-field) ripple. Energetic ions are trapped in local magnetic wells and drift vertically toward the midplane (z = 0). If the ripple is made very small on the opposite side of the midplane, drifting ions are detrapped and thermalized in the central plasma region. This paper discusses design considerations for establishing the required vertically asymmetric ripple. Examples are given of special TF-coil configurations, and of the use of auxiliary coil windings to create the prescribed ripple profiles

Supporting device for Toroidal coils

International Nuclear Information System (INIS)

Araki, Takao.

1985-01-01

Purpose: To reduce the response of a toroidal coil supporting device upon earthquakes and improve the earthquake proofness in a tokamak type thermonuclear device. Constitution: Structural materials having large longitudinal modulus and enduring great stresses, for example, stainless steels are used as the toroidal coil supporting legs and heat insulating structural materials are embedded in a nuclear reactor base mats below the supporting legs. Furthermore, heat insulating concretes are spiked around the heat insulating structural materials to prevent the intrusion of heat to the toroidal coils. The toroidal coils are kept at cryogenic state and superconductive state for the conductors. In this way, the period of proper vibrations of the toroidal coils and the toroidal coil supporting structures can be shortened thereby decreasing the seismic response. Furthermore, since the strength of the supporting legs is increased, the earthquake proofness of the coils can be improved. (Kamimura, M.)

Toroidal current asymmetry in tokamak disruptions

Science.gov (United States)

Strauss, H. R.

2014-10-01

It was discovered on JET that disruptions were accompanied by toroidal asymmetry of the toroidal plasma current I ϕ. It was found that the toroidal current asymmetry was proportional to the vertical current moment asymmetry with positive sign for an upward vertical displacement event (VDE) and negative sign for a downward VDE. It was observed that greater displacement leads to greater measured I ϕ asymmetry. Here, it is shown that this is essentially a kinematic effect produced by a VDE interacting with three dimensional MHD perturbations. The relation of toroidal current asymmetry and vertical current moment is calculated analytically and is verified by numerical simulations. It is shown analytically that the toroidal variation of the toroidal plasma current is accompanied by an equal and opposite variation of the toroidal current flowing in a thin wall surrounding the plasma. These currents are connected by 3D halo current, which is π/2 radians out of phase with the n = 1 toroidal current variations.

Ion confinement and transport in a toroidal plasma with externally imposed radial electric fields

Science.gov (United States)

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

1979-01-01

Strong electric fields were imposed along the minor radius of the toroidal plasma by biasing it with electrodes maintained at kilovolt potentials. Coherent, low-frequency disturbances characteristic of various magnetohydrodynamic instabilities were absent in the high-density, well-confined regime. High, direct-current radial electric fields with magnitudes up to 135 volts per centimeter penetrated inward to at least one-half the plasma radius. When the electric field pointed radially toward, the ion transport was inward against a strong local density gradient; and the plasma density and confinement time were significantly enhanced. The radial transport along the electric field appeared to be consistent with fluctuation-induced transport. With negative electrode polarity the particle confinement was consistent with a balance of two processes: a radial infusion of ions, in those sectors of the plasma not containing electrodes, that resulted from the radially inward fields; and ion losses to the electrodes, each of the which acted as a sink and drew ions out of the plasma. A simple model of particle confinement was proposed in which the particle confinement time is proportional to the plasma volume. The scaling predicted by this model was consistent with experimental measurements.

TOROID II

Science.gov (United States)

2009-01-01

three axis fluxgate magnetometer , CMOS sun and star sensors, and a Kalman filter. The work and tasks that have been accomplished on the TOROID... magnetometer . The problem was found to be a missing ferrite bead which connects the 12V power supply to the op-amps which are used to appropriately...establish an overall operational timeline for TOROID. Testing and calibration was performed on the three-axis magnetometer which is primary attitude

Elastic stability and vibration of toroidal magnets for fusion reactors. Final report

International Nuclear Information System (INIS)

Moon, F.C.; Swanson, C.

1975-09-01

The vibration and elastic stability of a set of discrete superconducting toroidal field magnets arranged to form a ''bumpy'' torus is examined. The mutual destabilizing magnetic forces between magnet pairs are calculated using a numerical differential inductance technique. It is shown that the mutual attractive magnetic forces can produce elastic buckling of the entire toroidal set. The vibration modes of the set are also found as functions of the coil current. The response of the set of magnets to an earthquake type motion of the toroidal base is calculated. The calculations have been incorporated in a computer code which accompanies the report. Measurements are made of the lateral stiffness of a flexible, planar, superconducting coil between two rigid coils in series. These tests show a dramatic decrease in the natural bending frequency with subsequent elastic instability or ''buckling'' at a critical value of the current in the coils. These observations support a magnetoelastic analysis which shows that proposed designs, of toroidal field coils for Tokamak fusion reactors, have insufficient lateral support for mechanical stability of the magnets

Equivelar toroids with few flag-orbits

OpenAIRE

Collins, José; Montero, Antonio

2018-01-01

An $(n+1)$-toroid is a quotient of a tessellation of the $n$-dimensional Euclidean space with a lattice group. Toroids are generalizations of maps in the torus on higher dimensions and also provide examples of abstract polytopes. Equivelar toroids are those that are induced by regular tessellations. In this paper we present a classification of equivelar $(n+1)$-toroids with at most $n$ flag-orbits; in particular, we discuss a classification of $2$-orbit toroids of arbitrary dimension.

Poloidal and toroidal heat flux distribution in the CCT tokamak

International Nuclear Information System (INIS)

Brown, M.L.; Dhir, V.K.; Taylor, R.J.

1990-01-01

Plasma heat flux to the Faraday shield panels of the UCLA Continuous Current Tokamak (CCT) has been measured calorimetrically in order to identify the dominant parameters affecting the spatial distribution of heat deposition. Three heating methods were investigated: audio frequency discharge cleaning, RF heating, and AC ohmic. Significant poloidal asymmetry is present in the heat flux distribution. On the average, the outer panels received 25-30% greater heat flux than the inner ones, with the ratio of maximum to minimum values attaining a difference of more than a factor of 2. As a diagnostic experiment the current to a selected toroidal field coil was reduced in order to locally deflect the toroidal field lines outward in a ripple-like fashion. Greatly enhanced heat deposition (up to a factor of 4) was observed at this location on the outside Faraday panels. The enhancement was greatest for conditions of low toroidal field and low neutral pressure, leading to low plasma densities, for which Coulomb collisions are the smallest. An exponential model based on a heat flux e-folding length describes the experimentally found localization of thermal energy quite adequately. (orig.)

Neoclassical transport in toroidal systems

International Nuclear Information System (INIS)

Wobig, H.

1992-01-01

The neoclassical theory of general toroidal equilibria is reformulated. The toroidal equilibrium of tokamaks and stellarators are described in Hamada coordinates. The relevant geometrical parameters are identified and it is shown how the reduction of Pfirsch-Schluter currents affects neoclassical transport and bootstrap effects. General flux-friction relations between thermodynamic forces and fluxes are derived. In drift-kinetic approximation the neoclassical transport coefficients are Onsager symmetric. Since a toroidal loop voltage is included, the theory is valid for all toroidal systems. (Author)

Compression of toroidal plasma by imploding plasma-liner

International Nuclear Information System (INIS)

Ikuta, Kazunari.

1979-07-01

A new concept of compressing a plasma in a closed magnetic configuration by a version of liner implosion flux compression technique is considered. The liner consists of a dense plasma cylinder, i.e. the plasma-liner. Maximum compression ratio of toroidal plasma is determined just by the initial density ratio of the toroidal plasma to the liner plasma because of the Rayleigh-Taylor instability. A start-up senario of plasma-liner is also proposed with a possible application of this concept to the creation of a burning plasma in reversed field configurations, i.e. burning plasma vortex. (author)

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

International Nuclear Information System (INIS)

Lee, Yi-Kang

2016-01-01

Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be

Design and characterization of a novel toroidal split-ring resonator

International Nuclear Information System (INIS)

Bobowski, J. S.; Nakahara, Hiroko

2016-01-01

The design and characterization of a novel toroidal split-ring resonator (SRR) are described in detail. In conventional cylindrical SRRs, there is a large magnetic flux within the bore of the resonator. However, there also exists a non-negligible magnetic flux in the free space surrounding the resonator. The energy losses associated with this radiated power diminish the resonator’s quality factor. In the toroidal SRR, on the other hand, the magnetic field lines are strongly confined within the bore of the resonator resulting in high intrinsic quality factors and stable resonance frequencies without requiring additional electromagnetic shielding. This paper describes the design and construction of a toroidal SRR as well as an experimental investigation of its cw response in the frequency-domain and its time-domain response to a rf pulse. Additionally, the dependence of the toroidal SRR’s resonant frequency and quality factor on the strength of inductive coupling to external circuits is investigated both theoretically and experimentally

Fast Dump of the ATLAS Toroids

CERN Document Server

Dudarev, A; Volpini, Giovanni; Dudarev, Alexey; Kate, Herman Ten

2010-01-01

The toroidal magnet system of the ATLAS Detector at CERN consists of a Barrel Toroid (BT) and two End Cap Toroids (ECT-A and ECT-C). Each toroid is built up from eight racetrack coils wound with an aluminum stabilized NbTi conductor and indirectly cooled by forced flow liquid helium. The three toroids operate in series at 20.5 kA with a total stored energy of 1.5 GJ. In order to verify the reliability and effectiveness of the quench protection system, series of fast dump tests have been performed first of the single toroids and finally of the entire toroidal magnet system. In this paper a model to simulate the fast dump of the ATLAS toroids in single mode operation and in full system configuration is presented. The model is validated through comparison with measured data extracted from the ramp-and-quench runs. The calculated energy dissipation in the various coils is in very good agreement (within 1-2\\%) with the enthalpy changes estimated from the temperature measurements of the different parts of the cold ...

Structural analysis of ITER sub-assembly tools

International Nuclear Information System (INIS)

Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Lee, J.H.; Kim, K.K.; Im, K.; Shaw, R.

2011-01-01

The ITER Tokamak assembly tools are purpose-built assembly tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. The sector sub-assembly tools descried in this paper are main assembly tools to assemble vacuum vessel, thermal shield and toroidal filed coils into a complete 40 o sector. The 40 o sector sub-assembly tools are composed of sector sub-assembly tool, including radial beam, vacuum vessel supports and mid-plane brace tools. These tools shall have sufficient strength to transport and handle heavy weight of the ITER Tokamak machine reached several hundred tons. Therefore these tools should be designed and analyzed to confirm both the strength and structural stability even in the case of conservative assumptions. To verify structural stabilities of the sector sub-assembly tools in terms of strength and deflection, ANSYS code was used for linear static analysis. The results of the analysis show that these tools are designed with sufficient strength and stiffness. The conceptual designs of these tools are briefly described in this paper also.

Construction of force-free fields which have toroidal surfaces about a given surface

International Nuclear Information System (INIS)

Bouligand, G.

1983-05-01

A study of two-fields (B vector, rotB vector) of conservative flux which admits a family of toroidal surfaces of parameter phi on a domain limited by a given surface S, suggests their construction by a Cauchy-Arzela method of step by step. Taking into account the Newcomb condition this method is consistent with force-free magnetic fields and with helical equilibria with scalar pressure. The method supposes that B vector is of class C 1 . This construction makes use of the remarkable property of the field B vector to be the surface gradient of a generating multivalued function Q on a closed surface. Consequently, the initial surface will be given with its normal metric coefficient K; that is to say, B vector admits a family F of homotopic surfaces on a infinitesimal domain about S, an element of F. From this, the periodic part of Q is a solution of a Beltrami equation for the flux conservation of which numerical resolution is envisaged. The study of these fields is made in a biorthogonal system of coordinates. The coeffficients of the two fundamental metric forms of magnetic surfaces vary with phi and are interrelated by a sixth order differential system of equations which gives their variation [fr

Ballooning instabilities in toroidally linked mirror systems

International Nuclear Information System (INIS)

Hastie, R.J.; Watson, C.J.H.

1977-01-01

This paper examines the stability against ballooning modes of plasma equilibria in toroidally linked mirror configurations consisting of a number of quadrupole minimum-B mirrors linked toroidally. On the basis of the Kruskal-Oberman energy principle, a class of displacements is identified which are potentially unstable, and a necessary criterion for stability is derived. The criterion is obtained from the eigenvalues of an ordinary differential equation, which determines the variation of the displacement along a field line. The coefficients in the equation are determined by the configuration, and by inserting various model configurations, estimates are obtained of the maximum value of β consistent with stability. In cases of interest, quite high β-values are obtained. (author)

Neoclassical poloidal and toroidal rotation in tokamaks

International Nuclear Information System (INIS)

Kim, Y.B.; Diamond, P.H.; Groebner, R.J.

1991-01-01

Explicit expressions for the neoclassical poloidal and toroidal rotation speeds of primary ion and impurity species are derived via the Hirshman and Sigmar moment approach. The rotation speeds of the primary ion can be significantly different from those of impurities in various interesting cases. The rapid increase of impurity poloidal rotation in the edge region of H-mode discharges in tokamaks can be explained by a rapid steepening of the primary ion pressure gradient. Depending on ion collisionality, the poloidal rotation speed of the primary ions at the edge can be quite small and the flow direction may be opposite to that of the impurities. This may cast considerable doubts on current L to H bifurcation models based on primary ion poloidal rotation only. Also, the difference between the toroidal rotation velocities of primary ions and impurities is not negligible in various cases. In Ohmic plasmas, the parallel electric field induces a large impurity toroidal rotation close to the magnetic axis, which seems to agree with experimental observations. In the ion banana and plateau regime, there can be non-negligible disparities between primary ion and impurity toroidal rotation velocities due to the ion density and temperature gradients. Detailed analytic expressions for the primary ion and impurity rotation speeds are presented, and the methodology for generalization to the case of several impurity species is also presented for future numerical evaluation

Neoclassical toroidal viscosity in perturbed equilibria with general tokamak geometry

Energy Technology Data Exchange (ETDEWEB)

Logan, Nikolas C.; Park, Jong-Kyu; Kim, Kimin; Wang, Zhirui [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Berkery, John W. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

2013-12-15

This paper presents a calculation of neoclassical toroidal viscous torque independent of large-aspect-ratio expansions across kinetic regimes. The Perturbed Equilibrium Nonambipolar Transport (PENT) code was developed for this purpose, and is compared to previous combined regime models as well as regime specific limits and a drift kinetic δf guiding center code. It is shown that retaining general expressions, without circular large-aspect-ratio or other orbit approximations, can be important at experimentally relevant aspect ratio and shaping. The superbanana plateau, a kinetic resonance effect recently recognized for its relevance to ITER, is recovered by the PENT calculations and shown to require highly accurate treatment of geometric effects.

Variation of the poloidal field during a disruption and consequences on the vacuum chamber, the poloidal system and the toroidal magnet (Tore II)

International Nuclear Information System (INIS)

Gatineau, F.; Leloup, C.; Pariente, M.

1977-12-01

The currents induced into the vacuum vessel and into the poloidal field coils and the overvoltages on the generators during a plasma current disruption are calculated. The subsequent applied mechanical forces and the poloidal field variations at the toroidal field conductor are deduced. The current decrease rate considered, during a disruption, ranges from d Ip/dt=0.810 9 A/s to 0.410 11 A/s [fr

Ideal magnetohydrodynamic simulations of low beta compact toroid injection into a hot strongly magnetized plasma

Energy Technology Data Exchange (ETDEWEB)

Liu, Wei [Los Alamos National Laboratory; Hsu, Scott [Los Alamos National Laboratory; Li, Hui [Los Alamos National Laboratory

2009-01-01

We present results from three-dimensional ideal magnetohydrodynamic simulations of low {beta} compact toroid (CT) injection into a hot strongly magnetized plasma, with the aim of providing insight into CT fueling of a tokamak with parameters relevant for ITER (International Thermonuclear Experimental Reactor). A regime is identified in terms of CT injection speed and CT-to-background magnetic field ratio that appears promising for precise core fueling. Shock-dominated regimes, which are probably unfavorable for tokamak fueling, are also identified. The CT penetration depth is proportional to the CT injection speed and density. The entire CT evolution can be divided into three stages: (1) initial penetration, (2) compression in the direction of propagation and reconnection, and (3) coming to rest and spreading in the direction perpendicular to injection. Tilting of the CT is not observed due to the fast transit time of the CT across the background plasma.

Initial Results of the SSPX Transient Internal Probe System for Measuring Toroidal Field Profiles

Science.gov (United States)

Holcomb, C. T.; Jarboe, T. R.; Mattick, A. T.; Hill, D. N.; McLean, H. S.; Wood, R. D.; Cellamare, V.

2000-10-01

Lawrence Livermore National Laboratory, Livermore, CA 94550, USA. The Sustained Spheromak Physics Experiment (SSPX) is using a field profile diagnostic called the Transient Internal Probe (TIP). TIP consists of a verdet-glass bullet that is used to measure the magnetic field by Faraday rotation. This probe is shot through the spheromak by a light gas gun at speeds near 2 km/s. An argon laser is aligned along the path of the probe. The light passes through the probe and is retro-reflected to an ellipsometer that measures the change in polarization angle. The measurement is spatially resolved down to the probes’ 1 cm length to within 15 Gauss. Initial testing results are given. This and future data will be used to determine the field profile for equilibrium reconstruction. TIP can also be used in conjunction with wall probes to map out toroidal mode amplitudes and phases internally. This work was performed under the auspices of US DOE by the University of California Lawrence Livermore National Laboratory under Contract No. W-7405-ENG-48.

Seismic analysis of ITER fourth PF (Poloidal Field Coil) feeder

International Nuclear Information System (INIS)

Liu, Sumei; Chen, Wei; Song, Yuntao; Ni, Xiaojun; Wang, Zhongwei; Chen, Yonghua; Gong, Chenyu

2014-01-01

The ITER feeder systems connect the ITER magnet systems located inside the main cryostat to the cryo-plant, power-supply and control system interfaces outside the cryostat. The main purpose of the feeders is to convey the cryogenic supply and electrical power to the coils as well as house the instrumentation wiring. The PF busbar which carries 52 kA current will suffer from high Lorentz force due to the background magnetic field inspired by the coils and the self-field between every pair of busbars. Except their mechanical strength and thermal insulation performance must be achieved, the dynamic mechanism on PF structure should be assessed. This paper presents the simulation and seismic analysis on ITER 4th PF feeder including the Coil Terminal Box and S-bend Box (CTB and SBB), the Cryostat Feed-through (CFT), the In-Cryostat-Feeder (ICF), especially for the ground supports and main outer-tube firstly. This analysis aims to study seismic resistance on system design under local seismograms with floor response spectrum, the structural response vibration mode and response duration results of displacement, membrane stress, and bending stress on structure under different directions actuating signals were obtained by using the single-seismic spectrum analysis and Dead Weight analysis respectively. Based on the simulative and analytical results, the system seismic resistance and the integrity of the support structure in the 4th PF feeder have been studied and the detail design confirmed

Basic toroidal Effects on Alfven Wave Current in Small Aspect Ratio Tokamaks

International Nuclear Information System (INIS)

Burma, C.; Cuperman, S.; Komoshvili, K.

1998-01-01

The Alfven wave current drive (AWCD) in small aspect ratio Tokamaks is properly calculated, with consideration of the basic toroidicity effects present in (i) the dielectric tensor-operator (involving the strongly toroidal equilibrium profiles), (ii) the structure of the r.f. fields obtained as a solution of the wave equation (through Maxwell's equations' toroidal operators as well as the conversion rate and conversion layer location, depending also on the equilibrium profiles) and (iii) the formulation of the AWCD (which, besides its dependence on the r.f. fields - affected by toroidicity as mentioned at points (i) and (ii) - also requires the equilibrium-magnetic-surface averaging of non-resonant forces involved). Thus, we consider consistent equilibrium profiles with neo-classical conductivity corresponding to an ohmic START-like discharge; use a resistive (anisotropic) MHD dielectric tensor-operator Edith practically no limitations, adequate to describe the plasma response in the pre-heated stage ; solve numerically the 2(1/2)D full- wave equation by the aid of an advanced finite element code developed in; and evaluate the AWCD by the aid of the recently proposed, quite general formulation holding in the case of strongly toroidal fusion devices and including contributions due to helicity injection, momentum transfer and plasma Bow. A general discussion of the results obtained in this work is presented

Measurement of toroidal and poloidal plasma rotation in TCA

International Nuclear Information System (INIS)

Duval, B.P.; Joye, B.; Marchal, B.

1991-01-01

With optimal observation geometry we have measured both the toroidal and poloidal rotation velocities in the edge and in the bulk of the TCA plasma. Regular calibration and correction for variations in the spectrometer temperature permitted a measurement with an error of ∼0.5 km/s which is an order of magnitude smaller than the range of measured velocities. In general, changes in the velocities are observed to be stronger and faster in the plasma edge than in the plasma bulk. With increasing density, the toroidal velocity is observed to change sign and follow the plasma density, while the poloidal velocity increases. These two effects lead to an increase in the absolute value of the radial electric field. With very strong gas puffing, the toroidal velocity is observed to again reverse and tend to zero, an effect which is stronger as the gradient of the density ramp is increased. Comparison between gas puffing and high power AWH does not show a significant difference in the radial electric field that could be responsible for the large associated density rise, which still remains unexplained. (author) 4 figs., 2 refs

Phase coexistence and electric-field control of toroidal order in oxide superlattices.

Science.gov (United States)

Damodaran, A R; Clarkson, J D; Hong, Z; Liu, H; Yadav, A K; Nelson, C T; Hsu, S-L; McCarter, M R; Park, K-D; Kravtsov, V; Farhan, A; Dong, Y; Cai, Z; Zhou, H; Aguado-Puente, P; García-Fernández, P; Íñiguez, J; Junquera, J; Scholl, A; Raschke, M B; Chen, L-Q; Fong, D D; Ramesh, R; Martin, L W

2017-10-01

Systems that exhibit phase competition, order parameter coexistence, and emergent order parameter topologies constitute a major part of modern condensed-matter physics. Here, by applying a range of characterization techniques, and simulations, we observe that in PbTiO 3 /SrTiO 3 superlattices all of these effects can be found. By exploring superlattice period-, temperature- and field-dependent evolution of these structures, we observe several new features. First, it is possible to engineer phase coexistence mediated by a first-order phase transition between an emergent, low-temperature vortex phase with electric toroidal order and a high-temperature ferroelectric a 1 /a 2 phase. At room temperature, the coexisting vortex and ferroelectric phases form a mesoscale, fibre-textured hierarchical superstructure. The vortex phase possesses an axial polarization, set by the net polarization of the surrounding ferroelectric domains, such that it possesses a multi-order-parameter state and belongs to a class of gyrotropic electrotoroidal compounds. Finally, application of electric fields to this mixed-phase system permits interconversion between the vortex and the ferroelectric phases concomitant with order-of-magnitude changes in piezoelectric and nonlinear optical responses. Our findings suggest new cross-coupled functionalities.

Phase coexistence and electric-field control of toroidal order in oxide superlattices

International Nuclear Information System (INIS)

Damodaran, A. R.; Clarkson, J. D.; Hong, Z.

2017-01-01

Systems that exhibit phase competition, order parameter coexistence, and emergent order parameter topologies constitute a major part of modern condensed-matter physics. Here, by applying a range of characterization techniques, and simulations, we observe that in PbTiO 3 /SrTiO 3 superlattices all of these effects can be found. By exploring superlattice period-, temperature- and field-dependent evolution of these structures, we observe several new features. First, it is possible to engineer phase coexistence mediated by a first-order phase transition between an emergent, low-temperature vortex phase with electric toroidal order and a high-temperature ferroelectric a 1 /a 2 phase. At room temperature, the coexisting vortex and ferroelectric phases form a mesoscale, fibre-textured hierarchical superstructure. The vortex phase possesses an axial polarization, set by the net polarization of the surrounding ferroelectric domains, such that it possesses a multi-order-parameter state and belongs to a class of gyrotropic electrotoroidal compounds. Finally, application of electric fields to this mixed-phase system permits interconversion between the vortex and the ferroelectric phases concomitant with order-of-magnitude changes in piezoelectric and nonlinear optical responses. Here, our findings suggest new cross-coupled functionalities.

RF current drive in a toroidal plasna in the banana regime

International Nuclear Information System (INIS)

Belikov, V.S.; Kolesnichenko, Ya.I.; Plotnik, I.S.

1982-01-01

The use of travelling waves for the steady-state current drive in an axisymmetric toroidal plasma in the banana regime is studied. The treatment is based on a quasi-linear equation for the electron distribution function averaged over the period of the particle motion along the small azimuth of the torus. It is show that the trapped electrons do not absorb the energy of the monochromatic (over frequency) RF field and thus only the circulating electrons contribute to the driving current and to the absorbed RF power. The current and the absorbed power are calculated by using the electron distribution function obtained for the case of narrow wave packet, both the toroidal magnetic field and the distortion of the electron distribution over transverse velocities being taken into consideration. The significant role of the barely carculating electrons is revealed. It is pointed out that the toroidal satellite resonances can affect the RF current drive by spreading and splitting the region of the wave-marticle interaction